The system MoO3-UO3

Ustinov, O. A.; Andrianov, M. A.; Chebotarev, N.T.; Novoselov, G. P.

doi:10.1007/bf01154837

Cited by 16 publications

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“…Several physical properties of UO 2 MoO 4 have been investigated such as its melting point, which has been given as T % 1253 K [3] or 1283 K [2], and some of its thermodynamic functions determined [8,9]. That is, namely, Chattopadhyay et al [8] measured the Gibbs energy of formation of UO 2 MoO 4 using e.m.f.…”

Section: Introductionmentioning

confidence: 99%

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A calorimetric and thermodynamic investigation of uranyl molybdate UO2MoO4

Сулейманов

Голубев

Alekseev

et al. 2010

The Journal of Chemical Thermodynamics

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Section: Introductionmentioning

confidence: 99%

“…Uranyl molybdate, UO 2 MoO 4 , is one of a number of different uraniummolybdenum phases in the system U-Mo-O and it is important for technological reasons [1]. UO [2][3][4][5][6][7].…”

Section: Introductionmentioning

confidence: 99%

A calorimetric and thermodynamic investigation of uranyl molybdate UO2MoO4

Сулейманов

Голубев

Alekseev

et al. 2010

The Journal of Chemical Thermodynamics

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“…[1][2][3] Generally, halide-based molten salts have been used as electrolytes in electrorefining used nuclear fuel; [1-3]however, Ustinov et al have explored molybdate melts for an alternative pyrochemical processing technique, specifically for oxide fuel treatment. [4][5][6][7][8][9] We explored this alternative approach in developing a so-called modified open fuel cycle in which the actinides are recovered for recycle to an advanced reactor without extensive processing.…”

Section: Introductionmentioning

confidence: 99%

“…In this approach, recrystallization methods are used to achieve the desired separations in which the used nuclear fuel is dissolved in the molybdate melt containing a mixture of sodium molybdate (Na 2 MoO 4 ) and molybdenum trioxide (MoO 3 ) at temperatures between 1273-1473 K, and UO 2 (or transuranic oxides) recrystallize upon cooling, separating it from the fission products. [4][5][6][7][8][9][10][11][12] The process is based on the difference in solubility of the fission product metal oxides compared to the uranium (or transuranic) oxides in the molybdate melt. [4][5][6][7][8][9][10][11][12] The advantage of this process is that the fission products are more soluble in molybdate melt than the UO 2 .…”

Section: Introductionmentioning

confidence: 99%

Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

Hames

Tkáč

Paulenová

et al. 2017

Journal of Nuclear Materials

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An investigation of molybdate melts containing sodium molybdate (Na 2 MoO 4) and molybdenum trioxide (MoO 3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate the feasibility of UO 2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D>10). A melt consisting of 20 wt% MoO 3-50 wt% Na 2 MoO 4-30 wt% UO 2 heated to 1313 K and cooled to 1123 K for the physical separation of the UO 2 product from the melt, and washed once with Na 2 MoO 4 displays optimum conditions for separation of the UO 2 from the fission products.

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“…After the fuel elements are opened, the (U, Pu)O 2 fuel pellets oxidize to (U, Pu) 3 The nonvolatile or weakly volatile fission products (cesium, rubidium, strontium, barium, molybdenum, zirconium, rare-earth elements, technetium, and ruthenium) oxidize, interact with the melt forming the process molydates or other oxide salts [5][6][7][8], which dissolve and concentrate in the salt melt, additionally heating the melt and increasing solubility:…”

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confidence: 99%

Unconventional reprocessing of the oxide fuel from fast reactors

2011

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A balanced combination of thermal and fast reactors in a closed fuel cycle is currently considered to be the most promising variant of a nuclear energy complex. Fast reactors are used as the base reactors in the strategy for the development of nuclear power in Russia in the first half of the 21st century and as advanced reactors in the Gen-IV international program of the leading nuclear countries [1]. The idea of a nuclear power plant with a fast reactor is now mentioned quite often [2][3][4]. It is believed that the fuel purification factor will be low (10 3 ) recovered fuel used in the fuel assemblies.It is well known that after being removed from a reactor spent nuclear fuel is stored in cool-down ponds at the reactor facility. However, in reprocessing spent fuel from fast reactors it is irrational to let the fuel cool down initially and then use additional energy for reprocessing, keeping in mind the proclaimed orientation toward low-waste but energy-intensive pyrochemical methods. The use of heat released by a spent fuel assembly for recovery looks tempting.Concentrating on uranium-plutonium oxide fuel of fast reactors, we are proposing an unconventional approach to the solution of the problem posed: after being extracted from a reactor spent fuel assemblies should be reprocessed by the dry scheme, in which fuel assemblies are dissolved and recrystallize in pools with a low-melting salt composition as a result of their own energy release, immediately and with no cool-down. Previously studied reactions occurring in molybdate melts were used for physicochemical validation of this process.We shall now give a brief description of this proposes process. The fuel assembly removed from the core of the reactor is place in a tank with a dry salt composition. The heat emission of the assembly will promote the formation of zone of melt of the salt composition around it. It is assumed that oxygen from air is present above the composition. Here, initially, the components of the fuel-assembly jacket undergo oxidation with iron, chromium, and nickel being formed; the latter elements interact with molybdenum trioxide via well-known reactions with formation of the corresponding molydates which dissolve in the excess melt [5, 6]: 2Fe + 3/2O 2 + 3MoO 3 = Fe 2 (MoO 4 ) 3 ;Ni + 1/2O 2 + MoO 3 = NiMoO 4 ; 2Cr + 3/2O 2 + 3MoO 3 = Cr 2 (MoO 4 ) 3 .In the course of further development of the technology, the system's thermal regime which is due to exothermal oxidation reactions, endothermal reactions resulting in the formation of molybdates, and heat release from the from the assembly will be calculated in detail. Additional heating of a fuel assembly by means of an electromagnetic field can be considered as a backup technological variant.During this period, the hottest zone of the melt is located in direct proximity to the fuel assembly; away from the latter the melt becomes colder. The temperature gradient which has appeared causes the iron, chromium, and nickel molyb-

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The system MoO3-UO3

Cited by 16 publications

References 1 publication

A calorimetric and thermodynamic investigation of uranyl molybdate UO2MoO4

A calorimetric and thermodynamic investigation of uranyl molybdate UO2MoO4

Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

Unconventional reprocessing of the oxide fuel from fast reactors

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