The invention relates to compounds which correspond to the general formula (I) below:

##STR00001##

in which: R.sup.1 and R.sup.2 represent, independently of one another, a C.sub.4 to C.sub.12 acyclic hydrocarbon group; R.sup.3 represents H; a C.sub.1 to C.sub.12 acyclic hydrocarbon group with optionally one or more heteroatoms; a C.sub.5 or C.sub.6 cyclic hydrocarbon group; or a 5- or 6-membered heterocyclic group; R.sup.4 represents H or a C.sub.1 to C.sub.12 acyclic hydrocarbon group with optionally one or more heteroatoms; R.sup.5 and R.sup.6 represent, independently of one another, H; a C.sub.1 to C.sub.12 acyclic hydrocarbon group with optionally one or more heteroatoms; a C.sub.5 or C.sub.6 cyclic hydrocarbon group; or a 5- or 6-membered heterocyclic group; on the condition however that R.sup.5 and R.sup.6 do not each represent H.

A system for extracting uranium from wet-process phosphoric acid (WPA), includes an ion exchange resin or solvent extractor for separating uranium from WPA to produce a loaded uranium solution stream and a uranium depleted WPA stream. An ion exchange resin is positioned to receive the loaded uranium solution stream and bind uranium species thereto. An anion solution stream is positioned to feed a solution comprising anions onto the ion exchange resin to form a loaded uranium eluant stream. The loaded uranium eluant stream may then be treated to provide a uranium containing product.

The proposed invention relates to processes of extraction and concentration of radio nuclides and can be used in radiochemical technologies when processing liquid radioactive wastes.

A method for extraction of americium from liquid radioactive wastes and its separation from rare-earth elements comprises simultaneous extraction of americium and rare-earth elements from radioactive nitrate solution with neutral solution of organic extracting agent in polar fluorinated organic solvent, washing of saturated with metals organic phase, selective re-extraction of americium. N,N,N,N-tetraalkyl-amide of diglycolic acid is used as an extracting agent and solution containing 5-20 g/L of complexon, 5-60 g/L of nitrogen-containing organic acid and 60-240 g/L of salting-out agent is used as a solution for re-extraction of americium.

Technical effect is the extraction of americium from acidic liquid radioactive solutions and its separation from all rare-earth elements in a single extraction cycle.

The proposed invention relates to processes of extraction and concentration of radio nuclides and can be used in radiochemical technologies when processing liquid radioactive wastes.

A method for extraction of americium from liquid radioactive wastes and its separation from rare-earth elements comprises simultaneous extraction of americium and rare-earth elements from radioactive nitrate solution with neutral solution of organic extracting agent in polar fluorinated organic solvent, washing of saturated with metals organic phase, selective re-extraction of americium. N,N,N,N-tetraalkyl-amide of diglycolic acid is used as an extracting agent and solution containing 5-20 g/L of complexon, 5-60 g/L of nitrogen-containing organic acid and 60-240 g/L of salting-out agent is used as a solution for re-extraction of americium.

Technical effect is the extraction of americium from acidic liquid radioactive solutions and its separation from all rare-earth elements in a single extraction cycle.

The invention relates to novel algae of the genus Coccomyxa, in particular the algae of a new species called C-IR3-4C, and their use for capturing metals from aqueous media, and in particular from radioactive media.

Provided is a process for recovering metals from solid radioactive waste, preferably uranium, cesium, mercury, thorium, rare earths or combinations thereof. The process comprises a leaching step and a separation step. The leaching step comprises contacting the solid radioactive waste with an aqueous inorganic acid and a leaching salt to produce a mixture of a metal-rich leachate and a metal-poor waste, which are separated in the separation step. Also provided is a process for recovering metals from solid radioactive waste comprising an attrition step, a leaching step, a washing step, a combination step and a recovery step.

This disclosure describes systems and methods for removing uranium hexafluoride (UF.sub.6) and/or other uranium fluoride (uranium fluorides identified herein generally as UF.sub.x) gases from a hydrogen fluoride (HF) gas stream.

According to one aspect of the invention, a system to separate salt from uranium. The system has a vessel, a heater, a pump, and a condenser. The vessel is adapted to receive a uranium that has a salt concentration. The heater heats the uranium for a period of time, causing the salt to turn into a salt vapor and the uranium to melt. The melted uranium releases the salt vapor. The pump circulates an inert gas that carries the salt vapor away from the melted uranium. The condenser is adapted to receive the salt vapor.

An apparatus is for the automated production of a daughter radionuclide from a parent radionuclide using a generator comprising a solid medium onto which the parent nuclide is fixed and whereby the daughter nuclide is formed by radioactive decay of the parent nuclide. The apparatus includes a fluid circuit including a chromatography column having a head port and a tail port, at least one connection port for connecting the generator to the fluid circuit, at least one inlet port for connecting fluid sources to the fluid circuit and at least one valve controlled by an electronic control unit for selectively connecting the chromatography column, the connection port and the at least one inlet port in various configurations. The various configurations include a first elution configuration for circulating an A1 solution exiting the generator and containing the daughter radionuclide, through the chromatography column from the head port to the tail port for loading the chromatography column with the daughter radionuclide; a first washing configuration for circulating an A2 washing solution from a solution inlet through the chromatography column from the head port to the tail port; and a second washing configuration for circulating an A2 washing solution from a solution inlet through the chromatography column from the tail port to the head port.

Provided are an electrokinetic device and method for in-situ leaching of uranium. The electrokinetic device for in-situ leaching of uranium includes an injection well, a pumping well, a positive electrode, a negative electrode, leaching solution, and a direct current power supply. Uranium ore is provided between the injection well and the pumping well, the negative electrode is arranged in the injection well, and the positive electrode is arranged in the pumping well. The leaching solution is injected from the injection well, flows through the uranium ore, and then is pumped from the pumping well for uranium extraction. The direct current power supply is respectively connected to the positive electrode and the negative electrode, and is configured to apply direct current between the positive electrode and the negative electrode to promote the pooling of uranium-carrying ions towards the pumping well.