Abstract: The invention concerns a vessel for processing, transfer, accumulation and/or transfer of material containing civilian or military origin plutonium, in the form of plutonium oxide, carbide and/or nitride. The plutonium is preferably present in a concentration not higher than a guaranteed maximum between 20 and 50% wt. in said material, possibly with other actinides such as americium, neptunium, or curium. Said material is preferably in form of powder, granulate and/or tablets. The vessel comprises a volume of 20 to 70 liters for containing said material, demarcated at least by two substantially parallel walls, these two walls being separated by a distance e between 8 and 15 cm.

Abstract: The invention concerns a vessel for processing, transfer, accumulation and/or transfer of material containing civilian or military origin plutonium, in the form of plutonium oxide, carbide and/or nitride. The plutonium is preferably present in a concentration not higher than a guaranteed maximum between 20 and 50% wt. in said material, possibly with other actinides such as americium, neptunium, or curium. Said material is preferably in form of powder, granulate and/or tablets. The vessel comprises a volume of 20 to 70 liters for containing said material, demarcated at least by two substantially parallel walls, these two walls being separated by a distance e between 8 and 15 cm.

Abstract: The invention relates to a method and device for production of non-contaminated rods from MOX tablets in columns, within a confinement chamber divided into compartments, by means of a centering and aligning element for the tablets and a long pushrod with a sensitive drive.

Abstract: Process for fabrication MOX fuel from weapon originating plutonium oxide (W—PuO2), including a process for fabricating MOX fuel pellets, a dry processing of W—PuO2 powder, and a dry processing of MOX fabrication scrap.

Abstract: A method for carrying out a comprehensive quality control of MOX fuel rods, including measurement of plutonium content of pellets contained in the rod, check of internal structure and constituents in the rod, including the detection of rogue pellets, and measurement of dimensional characteristics and radioactive contamination along the rod surface, by moving the fuel rod axially, performing concurrently the measurements and checks along the fuel rod moved axially, using radiometry and radiography. An apparatus for carrying out the method.

Abstract: A method for dry process recycling of mixed (U,Pu)O2 oxide nuclear fuel waste, including a process for making fuel pellets of mixed (U,Pu)O2 oxide, including a dosage and a first mixture (1) of waste in powder form and, if required, of PuO2 and/or UO2 powders, a micronization (2) and a forced sieving (3) of said first mixture, another dosage and a second mixture (4) of the first sieved mixture, of UO2 powders and, if necessary, of the waste powder, pelletizing (6) the second mixture, and sintering (7) the resulting pellets; and a process for pre-treating the waste including pelletizing (20) and sintering (21) the powder waste to form waste pellets, and micronizing (23) the waste pellets to form the desired waste powder designed to be incorporated as waste powder, in the first (1) and/or second mixtures.

Abstract: A process for manufacturing (U,Pu)O2 mixed oxide nuclear fuel pellets. In a first blending operation, PuO2 is mixed with a UO2 powder and/or fuel manufacturing scrap, forming a first blend. The first blend is micronized and sieved to form a conditioned first blend. Non-free-flowing UO2 is selected as a second portion of UO2. The second portion of UO2 is mechanically granulated to form granulated free-flowing UO2. In a second blending operation, the conditioned first blend is further mixed with the granulated free-flowing UO2 and possibly scrap. Lubricants and/or poreformers are added. The blend is pelletized and sintered. During granulation, the non free-flowing UO2 is compressed to form tablets. The tablets are then crushed until a free-flowing crushed material has been formed. At least one portion of this free flowing crushed material is used in the second blending operation.

Abstract: A process and apparatus for solidifying radioactive waste liquid containing dissolved and/or suspended solids is disclosed. The process includes chemically treating for pH adjustement and precipitation of solids, concentrating solids with a thin-film evaporator to provide liquid concentrate containing about 50% solids, and drying the concentrate with heated mixing apparatus. The heated mixing apparatus includes a heated wall and working means for shearing dried concentrate from internal surfaces and subdividing dry concentrate into dry, powdery particles. The working means includes a rotor and helical means for positively advancing the concentrate and resulting dry particles from inlet to outlet of the mixing apparatus. The dry particles may also be encapsulated in a matrix material. Entrained particles in the vapor stream from the evaporator and mixer are removed in an integral particle separator and the vapor is subsequently condensed and may be recycled upstream of the thin-film evaporator.

Abstract: The process of drying concentrates containing boric acid and/or borates after addition of lime is improved by the addition of an oxidizing agent to the concentrate.

Abstract: There is described a method for sodium-deactivating and/or stocking irradiated nuclear fuel elements, said elements being contacted with a fluid, in which use is made as fluid, of a molten salt mixture which is uncorrosive for the element stainless steels, inert relative to uranium and plutonium oxides by the deactivating temperature and reactive with sodium metal without producing a detonating gas.

Abstract: Process for manufacturing improved nuclear fuel pellets, including compacting ceramic powder in the Compaction chamber of a pelletizing machine with only the lower punch moving on compaction, wherein the walls of the compaction chamber are widened on at least part of their height in the direction of a diameter increase towards the die-bearing table.

Abstract: A fuel assembly for gas-cooled nuclear reactor: a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed.