The allocation method of mo-99 from the metal fuel based on uranium

 

(57) Abstract:

Usage: in the technologies of production of the medical isotope Mo-99 from irradiated fuel based on uranium to reduce the amount of radioactive waste and improve process safety. The inventive irradiated target on the basis of the uranium metal is dissolved under a layer of low-melting metal in a molten chloride salt containing gloriouse agent, separating the liquid metal phase, separating the concentrate of molybdenum-99, which is then subjected to refining. 4 C.p. f-crystals, 1 table.

The invention relates to technologies of production of the medical isotope Mo-99 from irradiated fuel based on uranium.

Known methods of production of the medical isotope Mo-99, based on his selection of irradiated fuel based on uranium enriched in the isotope U-235 [U.S. Patent 4,093,697; 4,094,953; 4,701,308]. These methods include the operation of irradiation targets with uranium and dissolving them after a short extracts in aqueous solutions of acids or alkalis. The resulting solution is subjected to selection operation of Mo-99 in the form of individual fractions (by extraction or adsorption-desorption), which is subjected to refining with obtaining h is active liquid waste, containing fissile material - enriched uranium. Despite the possibility of achieving high technical indicators processes (high yield of the target product, short production cycle) they are associated with the release of large amounts of highly radioactive liquid waste (up to 35-40 l 1 CCI Mo-99), storage and processing which significantly reduces economic performance. Need a special multistage processing of these wastes for excretion of uranium and preparation of waste for disposal. In addition, to work with svezheobozhzhennym material in a large amount of the problem of protection from emissions of radionuclides of iodine, in particular - 1-131. The use of aqueous media for the dissolution of irradiated targets limits the amount of activity that is processed in one cycle (20-25 RCC).

The method for separation and collection of Mo-99 from irradiated uranium targets based on thermal chromatographic separation [U.S. Patent 4,123,4981] . The target material is subjected to oxidation, and Mo-99 is separated in the form of volatile trioxide, which is captured and treated. The method allows to avoid the formation of large volumes of waste containing fissile materials. However, the awn his full recovery. Full recovery of the isotope iodine-131 should include multistage filters high capacity, which also periodically removed to waste. This method allows to avoid the formation of liquid waste with fissile material, but leads to the formation of waste from the capture of radioactive iodine.

There is a problem of iodine in the transition to high volume production, especially in the case of accidental depressurization.

The disadvantages of this method are:

- the formation of a large volume of radioactive waste requiring special handling;

- the inability to fully capture radioactive iodine in large-scale implementation of the method (for example, up to 1000 CCI in one cycle);

- risk of releases of radioactive iodine in cases of accidental depressurization of equipment, which is a processed material of the target.

All of the above disadvantages do not allow to organize large-scale production while increasing its security.

The above-mentioned disadvantages of obtaining molybdenum-99 are eliminated by the fact that in the proposed method of extraction of molybdenum-99 from irradiated targets on the basis of uranium metal, including otdavshego metal in molten chloride salt, containing gloriouse agent, separating the liquid metal phase, then with a high-temperature distillation of the metal-solvent separated from her concentrate of molybdenum-99, which is then subjected to refining.

In the processing goes irradiated enriched uranium in the form of targets on the basis of uranium metal or its alloy, the operation is carried out in an inert atmosphere. The target is placed under a layer of fusible metal (e.g. zinc or cadmium). On the metal layer placed liquid molten chloride salt (for example, equimolar mixture of NaCl-KCl) containing gloriouse agent (for example, in the form of chloride of zinc or cadmium).

When the interaction between the metal and molten salt exchange reactions occur, leading to dissolution in the molten salt components target:

- uranium (by the reaction of the type U+ZnCl2=Zn+UCl3)

- shell components, such as zirconium;

- fission products from the group of alkali, alkaline earth and rare earth metals.

I.e. all components, equilibrium reactions are shifted to the right.

In the liquid metal phase remain fission products, chlorides are less stable than the chloride of zinc. These include noble meta is the purpose of separation of Mo-99. Is the distillation of metal (for example, zinc or cadmium) in vacuum. Then the rest with metallic Mo-99 is dissolved in an aqueous solution for further refining Mo-99 or undergoing oxidation to dissolve in the form of MoO3. While uranium and iodine in future products (and waste) are not present.

Salt phase after exposure (for decay of the isotope iodine-131) can be subjected to simple processing (e.g., electrolysis) with separation of the uranium or uranium dioxide, ready for re-irradiation. Then the molten salt mixed with some DD can be subjected to a simple precipitation cleaned and used repeatedly.

To further increase the security of the process of extracting Mo-99 emissions of radioactive iodine procedure of dissolution of the target may be held in conjunction with the sheath material, if it is made on the basis of zirconium, aluminum or other metals, chlorides which are more stable than the chloride of the metal-solvent. In addition, the material of the shell is also dissolved in the molten salt and is not present in the concentrate Mo-99 after its separation from the metal-solvent.

Thus this solution has significant differences from the handy use of uranium and reagents (salts);

to localize the iodine in the solid state, suitable for controlled storage;

- significantly reduce the probability of emission of iodine in emergency situations.

Conducted a series of experiments in a protective chamber for processing real-irradiated targets with the aim of identifying concentrate of molybdenum-99. As experimental targets were used developed in SSC RIAR target of uranium wire enriched in the uranium-235 90% diameter of about 1 mm to about 100 mm, placed in a metal matrix, zirconium shell. Was exposed to experimental targets in research reactors CM-3 during normal power on a single target up to 1 kW for 7-10 days.

Extract targets after irradiation was about two days. Recycling targets was performed in a salt melt NaCl - KCl-ZnCl2at a temperature of 720-750oC in a quartz crucible in an atmosphere of purified argon.

The Stripping of zinc from liquid metal phase after separation from the salts were carried out in a quartz crucible in an atmosphere of purified argon at a temperature of 900oC.

Table 1 shows data on the distribution of key fission products and uranium during the processing of targets N 1 and N 2. Target N 1 th branch of the zirconium sheath from the core of the target.

Thus, the invention has significant differences from the known method and allows you to achieve your goals.

All of this allows you to create a production unit capacity while increasing safety process.

1. Method of extraction of molybdenum-99 from irradiated metallic fuel based on uranium, which includes operations of the Department of molybdenum from the main mass of uranium and fission products and the concentration of molybdenum in the composition of a compact solid phase with subsequent refining, characterized in that the separation of molybdenum from uranium is carried out by dissolving the fuel from under the layer of liquid metal-solvent in the chloride melt gloriouse agent, after which the metal phase is separated from the salt and subjected to high-temperature distillation of the metal-solvent, and the residual solid compact concentrate molybdenum dissolved for further refining known methods.

2. The method according to p. 1, characterized in that the fuel dissolution in the melt is carried out by electrolysis.

3. The method according to PP.1 and 2, characterized in that conduct dissolution in chloride melt the irradiated targets complete with shell, without her PR is the Olev melt NaCl - KCl at 680 - 850oC, as the metal-solvent used zinc and gloriouse agent is zinc chloride.

5. The method according to PP.1 and 2, characterized in that the metal-solvent used cadmium, and as glorieuses agent - chloride cadmium.

 

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