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A basic method for reprocessing nuclear fuel is the Purex water process, which uses organic solvents for extractive removal, purification, and separation of uranium and plutonium from nitric acid solutions. Problems with this process include the large volume of liquid wastes, radiation instability, and the nuclear hazard of the water solutions. Today a need remains to develop other methods that do not have these problems. These include nonaqueous pyrochemical methods, such as fluoride gas distillation, chemical and electrochemical distillation, and "extraction (distribution between salt and metal melts).Recrystallization is a process in which the same material is dissolved and separated from a solution. The choice of melts for recrystallization is an important one. Analysis of the scientific and engineering literature excludes chloride, fluoride, sulfate, nitrate, and hydroxide systems in favor of molybdate systems.Motybdate melts are widely used for recrystatlizing oxides. An assembly of fundamental data on ternary state diagrams indicates it should be possible to find ternary systems (uranium and plutonium oxides-molybdenum trioxide-fusible molybdate) that have regions where a refractory oxide crystallizes at a "low" temperature ( < 1000~In order to develop a process for recrystallization from molybdate melts, we studied: 1) the interaction and properties of oxide fuel components (oxides of uranium, plutonium, and accompanying elements) with molybdate systems, i.e., the corresponding binary systems with MoO 3 and ternary systems with MoO 3 and Na2MoO4; 2) how various compounds of uranium and plutonium and their accompanying impurities behave during crystallization in molybdate systems.From the literature analysis, we selected the following binary systems: Na2MoO4-MoO 3, BaO-MoO 3, ZrO2-MoO 3, La203-MoO 3, CeO2-MoO 3, RuO2-MoO 3, UO2-MoO 3, UO3-MoO 3, PuO2-MoO 3, UO2-Na2MoO 4, U308-Na2MoO4, PUO2-Na2MoO 4, U(MoO4)2-Na2MoO 4, and Pu(MoO4)2-Na2MoO 4. We also selected specific regions [the crystallization regions of UO 2, PuO 2, U(MoO4) 2, and Pu(MoO4)2] in the ternary systems UO2-MoO3-Na2MoO4 and PuO2-MoO3-Na2MoO 4.Thermographic, x-ray, and analytic methods were used in the research. Thermography was done up to 1000-1100~ Binary Systems. Data on the interactions of the various molybdenum oxides and trioxides are shown in Table 1.No interaction of ruthenium dioxide with MoO 3 was observed. UO 2 hardly interacts with Na2MoO 4 in an inert atmosphere: the observed solubility was insignificant (-0.2% molar at 1000~and an insignificant amount of sodium diuranate forms in air, which also is only slightly soluble in Na2MoO 4 ( -0.05% molar at 1000~PuO 2 and Na2MoO 4 hardly interact (molar solubility is -0.1% at 1000~Detailed information [ 1-12] has been published on some of these systems. Analysis of the resultant state diagrams leads to the following conclusions.All the oxides are soluble in a MoO 3 melt except RuO 2. Oxides of metals which are analogs of fission products are much more soluble in a MoO 3 melt than are ura...
Spent nuclear fuel in a closed nuclear fuel cycle is reprocessed using the purex process. This technology has certain drawbacks, including the following: large amounts of radioactive wastes are produced; flammable extracting agents are used; and, radiolysis of water and organic phases with release of explosive substances occurs. In contrast to the purex process, water-free (pyrochemical or pyrometallurgical) methods are distinguished by the small size of the technological apparatus, the small volume of liquid wastes which are produced during fuel reprocessing, and radiation-resistant reagents and therefore short holding times for the fuel before reprocessing.The pyrochemical reprocessing of spent oxide fuel using molybdate melts is being studied at the A. A. Bochvar All-Russia Research Institute of Standardization in Machine Engineering [1,2]. Binary and ternary phase diagrams have been studied: oxides of the spent-fuel components -molybdenum trioxide -sodium molybdate [3][4][5]. The presence of molydenum trioxide gives rise to oxidation of the fission products which are in a metallc state; sodium molybdate dissolves the oxides obtained, which has been confirmed experimentally for the fission product 106 Ru [6]. The isothermal section of a ternary phase diagram UO 2 -MoO 3 -Na 2 MoO 4 is shown in Fig. 1. x-Ray phase analysis of melt samples obtained by heating mixtures of UO 2 , MoO 3 , and Na 2 MoO 4 to 1000°C in an inert atmosphere and then cooling at a rate of 300 deg/h was performed. The mixtures were prepared in accordance with the composition indicated at the nodal points of the diagram. A special feature of the system UO 2 -MoO 3 -Na 2 MoO 4 is crystallization in one region of the uranium dioxide diagram UO 2 (in the figure the crystallization region is indicated by a circle) and crystallization of uranium molybdate U(MoO 4 ) 2 (B) and other compounds in another region. The diagrams with PuO 2 and the oxides of fission products have a similar form. The feature mentioned above is used for developing methods for reprocessing by recrystallization purification of uranium and plutonium dioxides and uranium molybdates in molybdate melts with the appropriate compositions.
The presence of traces of water dissolved in a sodium molybdate melt and the interaction of this water with a metal surface in contact with the melt are confirmed. It is shown experimentally that U 3 O 8 can be reduced to UO 2 in sodium molybdate melt by the hydrogen that is released during such an interaction. A technological process of reprocessing spent oxide nuclear fuel using the phenomenon studied is described.There are several circumstances that make it difficult to implement pyrochemical reprocessing of spent oxide nuclear fuel using melts based on molybdate systems [1-3]:• metal holding containers corrode in melts with a high content of molybdenum trioxide, used to accelerate the dissolution of compact sintered fuel pellets; • complete dissolution of fuel during recrystallization requires a large volume of melt and a prescribed heating and cooling regime for the melt. To eliminate these difficulties in the reprocessing of spent VVÉR-1000 fuel, it is recommended that the fuel pellets be dispersed by oxidation to U 3 O 8 and reduction of the latter in a molybdate melt to UO 2 at constant temperature.It has been determined [4], and we have confirmed experimentally, that a very small amount of water is present in melts, including in molybdate melt, and that this water interacts with metals placed in the melt (material of apparatus, cladding fragments, or specially introduced fragments of nickel, iron, and other metals), hydrogen being released in the process. The hydrogen released reduces the U 3 O 8 present in the melt. The transformations named are described by the reactions Me (2) + H 2 O = MeO + H 2 ↑ and U 3 O 8 +2H 2 = 3UO 2 + 2H 2 O.The rearrangement of the structure of the fuel during such a transformation in MoO 3 -Na 2 MoO 4 melt containing enough molybdenum trioxide for molybdates of the fission products to form results in the removal of fission products from the crystal lattice of the fuel and their dissolution in the melt. The present article presents the results of a laboratory-scale check of the process resulting in the reduction of U 3 O 8 in sodium molybdate melt. U 3 O 8 was mixed with Na 2 MoO 4 in a 1:1 ratio by mass and loaded into nickel (NP-2 nickel) crucibles. The mixture was allowed to stand from 2 to 20 h at 700, 850, and 1000 °C in a helium atmosphere in a hermetically sealed furnace. After being held at the prescribed temperature, the mixture was cooled, water was used to leach out the sodium molybdate, the powder consisting of uranium oxides was dried, a chemical analysis was performed to determine the total content of uranium and tetravalent uranium, and x-ray diffraction analysis was used to determine the presence of U 3 O 8 and UO 2 phases. Standard procedures were used for the chemical analysis [5]. The x-ray diffraction analysis was performed in a RKU-86 camera using Co radiation.The content of tetravalent uranium as a function of the holding time at 850 °C in an inert atmosphere is displayed in Fig. 1. Complete reduction of 10 g U 3 O 8 to UO 2 in sodium molybdate melt und...
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