High-density U-Mo fuel for research reactors

Vatulin, A. V.; Морозов, А. В.; Suprun, V.; Petrov, Yu. I.; Trifonov, Yu. I.

doi:10.1007/s11041-005-0007-5

Cited by 18 publications

(18 citation statements)

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“…The singlephase c-alloy has also been obtained with 13 at.% Mo using an additional sample treatment (e.g. hardening and annealing [6] or annealing at around 1300 K for 24 h [7]). A sharpening of the XRD peaks after annealing certainly indicated an enhancement of the c-phase concentration by improving the molybdenum homogeneity due to diffusion [7].…”

Section: C-u Alloys Metallurgymentioning

confidence: 99%

Characterization of cubic γ-phase uranium molybdenum alloys synthesized by ultrafast cooling

Tkach

Kim-Ngan

Mašková

et al. 2012

Journal of Alloys and Compounds

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Section: C-u Alloys Metallurgymentioning

confidence: 99%

Characterization of cubic γ-phase uranium molybdenum alloys synthesized by ultrafast cooling

Tkach

Kim-Ngan

Mašková

et al. 2012

Journal of Alloys and Compounds

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“…As compensation it is suggested that fuel granules consisting of U-9%Mo (density 17.2 g/cm 3 ) in a matrix consisting of an aluminum alloy be used [3,4]. For example, if in the composition UO 2 -Al the uranium content is 3-3.5 g/cm 3 , then in the composition (U−9Mo)-Al it increases by a factor of 2.…”

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confidence: 99%

Interaction of uranium-molybdenum fuel with an aluminum matrix with deep burnup

et al. 2010

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A thermodynamic analysis is used to calculate the phase and component composition of uraniummolybdenum fuel with burnup 200 GW·days/ton. The equilibrium composition of the gas phase, consisting mainly of gaseous cesium whose pressure reaches 30 kPa, is determined more accurately. The quantitative composition of the phase of solid solutions of tellurides, whose formation degrades the structure of a fuel granule, is presented. Thermal tests of the fuel composition (U-Mo)-Al were performed. The investigation was performed in the presence of simulators of the chemically active fission products of cesium and iodine at different temperature. The interaction zone of (U-Mo)-Al is investigated by means of metallography and scanning electron microscopy. The data obtained on the composition of the indicated zone made it possible to conjecture the character of interaction between the fuel material and the aluminum matrix.The world is showing heightened interest in power reactor facilities of low and ultralow power and research nuclear reactors. These facilities use dispersion compositions in the form of UO 2 , UAl x , U 3 Si 2 , uniformly distributed in an aluminum matrix, as nuclear fuel. The 235 U enrichment often reaches >90% [1].When low (to 20%) enrichment fuel is used, the burnup decreases because of the lower volume content of 235 U [2]. As compensation it is suggested that fuel granules consisting of U-9%Mo (density 17.2 g/cm 3 ) in a matrix consisting of an aluminum alloy be used [3,4]. For example, if in the composition UO 2 -Al the uranium content is 3-3.5 g/cm 3 , then in the composition (U−9Mo)-Al it increases by a factor of 2. However, the use of the composition (U-Mo)-Al is held back by the interaction between uranium-molybdenum granules and an aluminum matrix with burnup 150-200 GW·days/ton [5]. This interaction is complicated. On the one hand, a chemical interaction is observed in the system U-Al at a temperature close to the melting temperature of aluminum. On the other hand, the effect of the fission products on the physicochemical and physical-mechanical characteristics of uranium-molybdenum alloy must be taken into account.The compatibility of the fuel with the matrix can be improved by introducing into the alloy a component, for example, silicon, that substantially lowers the interaction rate as well as by depositing a coating on the surface of the fuel particles [5][6][7][8].Even though from the technological point of view alloying with silicon is considered to be optimal, one should take note of the decrease of the stability of the γ-U-Mo phase. The introduction of silicon into an intermetallide does not exclude the focal interaction with aluminum; in addition, it can increase the temperature gradient at the boundary between a fuel granule and an aluminum matrix [6].Of special interest is the appearance of barrier coatings on the surface of uranium-molybdenum granules. Specifically, it is proposed that a layer of pure molybdenum be deposited on the fuel granules, which would then be protected from int...

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“…It is concluded that low-enrichment high-density uranium-molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium-molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used.A large number of radiation tests of new types of dispersion fuel have been conducted as part of the international program on decreasing the enrichment of fuel for research reactors (RERTR) and national programs [1][2][3][4][5]. Plate-shaped miniscale fuel elements with dispersion fuel based on various high-density alloys and uranium compounds have been tested at the first stage.…”

mentioning

confidence: 99%

“…A large number of radiation tests of new types of dispersion fuel have been conducted as part of the international program on decreasing the enrichment of fuel for research reactors (RERTR) and national programs [1][2][3][4][5]. Plate-shaped miniscale fuel elements with dispersion fuel based on various high-density alloys and uranium compounds have been tested at the first stage.…”

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confidence: 99%

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Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

et al. 2006

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The results of radiation tests are discussed and the character of the failure of fuel compositions and the operability of fuel elements is analyzed as a function of the type of fuel and the irradiation conditions. The intense interaction of the fuel with the matrix material is considered as the main factor limiting the operability of fuel elements in power-dense high-flux nuclear reasearch reactors. It is concluded that low-enrichment high-density uranium-molybdenym fuel can provide reliable operation of dispersion fuel elements in low-and medium-power research reactors. Such fuel can be used in power-dense high-flux research reactors if the fuel load is decreased below the maximum admissible amount, the compatibility of the uranium-molybdenum alloy with an aluminum matrix is radically improved, or fuel elements with a different construction, for example, monolithic, are used.A large number of radiation tests of new types of dispersion fuel have been conducted as part of the international program on decreasing the enrichment of fuel for research reactors (RERTR) and national programs [1][2][3][4][5]. Plate-shaped miniscale fuel elements with dispersion fuel based on various high-density alloys and uranium compounds have been tested at the first stage. Fuel based on uranium silicide U 3 Si 2 and uranium alloys with 7-10% molybdenum or zirconium and niobium with total content at least 10% exhibited the best radiation resistance. Consequently, mini-and full-scale fuel elements with fuel based on these alloys, primarily, uranium alloys with molybdenum (here and below, the mass fraction of the alloying element is given in %) were tested in further tests.The influence of the characteristics of the dispersion fuel and the conditions of irradiation on the radiation resistance of fuel elements were investigated in the experiments. The results obtained make it possible to draw conclusions about the regularities in the behavior of, the failure of fuel compositions of, and the operability of fuel elements as a function of the type of fuel and conditions of irradiation.Results of the First Stage of the Investigations. At the first stage of the investigations, the RERTR-1 and -2 computer programs were used to study the radiation resistance of plate-shaped miniscale fuel elements with fuel based on intermetallic compounds of uranium with iron, manganese, silicon, and aluminum and alloys of uranium with molybdenum, zirconium, and niobium. The uranium content in the composition was 4 g/cm 3 , the heat flux and temperature of the fuel-element cladding were 55 W/cm 2 and 65°C, respectively, and the burnup 40 and 70% 235 U.Some of the fuel elements successfully withstood the tests -they remain sealed, and the increase in the thickness of the plates was negligible. The microstructure of the fuel composition of such fuel elements after irradiation in the reactor was essentially identical to the initial microstructure (Fig. 1). The fuel particles retained their shape and size. The interaction of

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High-density U-Mo fuel for research reactors

Cited by 18 publications

References 1 publication

Characterization of cubic γ-phase uranium molybdenum alloys synthesized by ultrafast cooling

Characterization of cubic γ-phase uranium molybdenum alloys synthesized by ultrafast cooling

Interaction of uranium-molybdenum fuel with an aluminum matrix with deep burnup

Radiation resistance of high-density uranium—molybdenum dispersion fuel for nuclear research reactors

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