Abstract: The present invention generally relates to the preparation of mixed actinide oxides, such as mixed oxides of uranium and plutonium (U, Pu) O2, by simultaneously coprecipitation and then calcinations.

Abstract: The invention relates to a heat-sensitive compound having the property of being soluble in water below a critical temperature Tc and insoluble in water above this temperature Tc, this property being thermally reversible, characterized in that it comprises a first amphiphilic and thermally reversible part corresponding to one of the following formulae (I) and (II): in which i is an integer ranging from 1 to 20 and j is an integer ranging from 3 to 30. This compound can be used to extract a chemical entity, such as uranium.

Abstract: The invention relates to a heat-sensitive compound having the property of being soluble in water below a critical temperature Tc and insoluble in water above this temperature Tc, this property being thermally reversible, characterized in that it comprises a first amphiphilic and thermally reversible part corresponding to one of the following formulae (I) and (II): 1

Abstract: Process for co-precipitating actinides in oxidation state (IV) in which:

Abstract: The present invention relates to a process for the selective separation of actinides (III) and lanthanides (III) from a reaction medium containing them. More specifically, the invention permits a selective extraction of actinides (III) by using an aqueous reaction medium containing TcO.sub.4.sup.- pertechnetate ions. The preferred extracting agents are chosen from among picolinamides and their derivatives. A particular use of the process is in the separation of actinides (III) from high activity raffinates produced during the reprocessing of nuclear fuels according to the solvent-based extraction process known as the PUREX process.

Abstract: The invention relates to a process and an installation for the destruction of organic solutes, particularly complexing agents, present in an aqueous solution such as a radioactive effluent.According to this process, the aqueous solution undergoes electrolysis in a cell (1) comprising a working electrode (5.sub.1, 5.sub.2, etc.) by applying to said electrode an adequate potential for destroying the solutes by electrolytic oxidation and periodic modification takes place to the potential applied or to the passage direction of the current passing through the cell so as to periodically carry out a regeneration of the working electrode.The solutes can be organic acids such as formic acid, dihydroxymalonic acid or oxalic acid.

Abstract: The invention relates to a process for dissolving plutonium dioxide by means of OH.sup.- radicals produced by radiolysis of water and usable for the treatment of dissolving fines and plutoniferous waste.To this end, contacting takes place in a reactor (1) of solid products containing PuO.sub.2 coming from the hopper (5) with an aqueous nitric solution (3) irradiated by radiation or charged particles for producing OH.sup.- radicals by radiolysis of said solution, in the presence of a reagent such as N.sub.2 O, able to trap the solvated electrons and the H.sup.- radicals produced in simultaneous manner and optionally a redox mediator such as silver.Irradiation can take place by an alpha or beta emitter present in the solution or by an external source such as an irradiator or an electron accelerator.

Abstract: A process for reducing and dissolving PuO.sub.2 comprises contacting the plutonium dioxide with a hydrazine-free, acid aqueous solution containing a reducing agent, whose redox potential is below +0.5 V/ENH in order to reduce and dissolve the plutonium. The reducing agent can be Cr.sup.2+ or U.sup.4+.

Abstract: The invention relates to the elimination of the chloride ions present in contaminated solid wastes, such as incineration ashes contaminated by actinides.First the chloride ions are dissolved in an aqueous solution, for example a nitric acid solution of 4 to 6 moles/l, whereafter the chloride ions thus dissolved in the aqueous solution are oxidized, for example, electrochemically, possibly in the presence of an oxidation-reduction couple such as CO.sup.3+ /CO.sup.2+, to eliminate them from the solution in the form of chlorine in a gas flow.The dissolution and oxidation can be performed in the electrolyzer (1), the gas flow being introduced at (10) and eliminated at (13). Then the actinides can be dissolved directly in the electrolyzer by electrolysis in the presence of Ag.sup.2+ ions.

Abstract: The invention relates to an ammonium plutonate uranate, its preparation process and its use for producing the mixed oxide (U,Pu)O.sub.2.The ammonium plutonate uranate of formula [Cu,Pu)O.sub.3 ].sub.2 [NH.sub.3 ].sub.x [H.sub.2 O].sub.y in which x and y have a value from 0 to 5 with x+y=5, is prepared by neutralization with ammonia of an aqueous solution of uranium VI and plutonium VI salts, e.g. uranyl and plutonyl nitrates. This solution can be prepared by oxidizing a solution of plutonium IV and uranyl nitrate. By calcining under a reducing atmosphere the ammonium plutonate uranate precipitate formed, a mixed oxide (U,Pu)O.sub.2 is obtained having a solubility in 11N HN O.sub.3 without fluoride of at least 99.9%.

Abstract: The invention relates to a process for recovering the plutonium contained in the solid waste, such as process, technological and laboratory waste.The waste is contacted with aqueous nitric acid solution having a nitric acid concentration of 2 to 8 mole/l, which contains a compound of Ag.sup.2+ soluble in the solution, for oxidizing the plutonium present in the waste and for dissolving same in the nitric solution.Preferably, contacting takes place in an electrolytic cell having an anode and a cathode, by applying a potential difference between anode and cathode which is sufficient to continuously regenerate the Ag.sup.2+ ions used for oxidizing the elements present in said waste.

Abstract: The invention relates to a process for the recovery of molybdenum-99 from an irradiated uranium alloy target.This process comprises the following stages:(a) dissolving the irradiated uranium alloy target in sulphuric acid,(b) separating the iodine and tellurium present in the thus obtained solution,(c) oxidizing the molybdenum-99 present in the Mo (VI) solution, and(d) extracting the thus oxidized molybdenum-99 with a hydroxamic acid of formula: ##STR1## in which R is a radical chosen from the group including straight or branched-chain alkyl radicals, the phenyl radical, phenyl radicals substituted by at least one alkyl radical, arylalkyl radicals and the cyclohexyl radical and R' is a hydrogen atom or an alkyl radical.The hydroxamic acid can be tri-n-butylacetohydroxamic acid and the molybdenum recovered can be used in a technetium generator for medical purposes.