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Takashi Hirosawa - One of the best experts on this subject based on the ideXlab platform.
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Americium and plutonium release behavior from irradiated Mixed Oxide fuel during heating
Journal of Nuclear Materials, 2013Co-Authors: Isamu Sato, Takashi Hirosawa, Mitsuo Suto, Shuhei Miwa, Shin-ichi KoyamaAbstract:Abstract The release behavior of Pu and Am was investigated under the reducing atmosphere expected in sodium cooled fast reactor severe accidents. Irradiated Pu and U Mixed Oxide Fuels were heated at maximum temperatures of 2773 K and 3273 K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that Pu and Am can be released more easily than U under the reducing atmosphere. The respective release rate coefficients for Pu and Am were obtained as 3.11 × 10 −4 min −1 and 1.60 × 10 −4 min −1 at 2773 K under the reducing atmosphere with oxygen partial pressure less than 0.02 Pa. Results of thermochemical calculations indicated that the main released chemical forms would likely be PuO for Pu and Am for Am under quite low oxygen partial pressure.
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burnup dependence of melting temperature of fbr Mixed Oxide Fuels irradiated to high burnup
Journal of Nuclear Materials, 2011Co-Authors: Takashi Hirosawa, Isamu SatoAbstract:Abstract The melting temperatures of FBR MOX Fuels with Pu content of 28–30 wt.% irradiated to from 22.5 to 112.5 MWd kg−1 were measured using a rhenium inner capsule to hold the specimens. The rhenium inner capsule could prevent chemical reactions between Fuels and tungsten materials which decrease the melting temperature. The melting temperatures were about 30 K higher than the previous data using tungsten capsules. The melting temperature decreases in a linear manner with burnup due to solid solution of fission products in Fuels. However, the slopes of the lines plotting melting temperature versus burnup are almost similar to the previous data.
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melting temperature of Mixed Oxide Fuels for fast reactors
Journal of Nuclear Science and Technology, 2002Co-Authors: Koichi Konno, Takashi HirosawaAbstract:Alternation of the melting temperature of irradiated Mixed Oxide (MOX) fuel in fast reactors with progress of burnup was determined in relation to the actinide fractions contained and of the oxygen-to-metal (O/M) ratio. Based on ideal solution models of UO2-PuO2 and UO2-PuO2-Am2O3 systems and on correlations obtained of the measured melting with O/M ratio and with burnup, an equation was derived expressing the estimated melting temperature T rev (K) as function of the factors mentioned above that affect the temperature: where X 1 is the plutonium fraction (Pu/(Pu+U)), X 2 the americium fraction (Am/(Pu+U+Am)), X 3 the burnup (GWd/t), and X 4 the O/M ratio (-) (not exceeding 2.00). Toward advanced stage of burnup, the tendency of irradiated fuel melting temperature to lower with burnup progress presented a tendency to level off its rate of descent.
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melting temperature of simulated high burnup Mixed Oxide Fuels for fast reactors
Journal of Nuclear Science and Technology, 1999Co-Authors: Koichi Konno, Takashi HirosawaAbstract:The melting (solidus) temperatures were measured for fuel simulating burnups of 50, 90, 130, 170, 210 and 250 GWd/t (SIMFUEL) which were added non-radioactive soluble fission products (FPs) to unirradiated fast reactor MOX fuel. The melting temperatures for Fuels of 250 GWd/t which were blended non-radioactive soluble FPs and irradiated Fuels of 110.6 and 119.0 GWd/t were also compared to the SIMFUEL of 250 GWd/t. The melting temperature decrease of the SIMFUEL tended to saturate with increasing burnup and the melting temperatures appeared virtually constant above 170 GWd/t. An equation for melting temperature was obtained from the proposed equation in the previous report by revising the coefficient of the fifth term: where is the expected melting temperature (K), X 1 the plutonium fraction (Pu/(Pu+U)), X 2 the americium fraction (Am/(Pu+U+Am)), and X 3 the burnup (GWd/t). After the temperature measurement, the radial distribution of eight FP Oxide additives in SIMFUEL of 250 GWd/t was measured by X-ray m...
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Melting Temperature of Irradiated Fast Reactor Mixed Oxide Fuels
Journal of Nuclear Science and Technology, 1998Co-Authors: Koichi Konno, Takashi HirosawaAbstract:The melting (solidus) temperatures of irradiated Mixed Oxide Fuels were measured and the compositions of the Fuels on the temperature measurement date were calculated. The Fuels contained about 29wt% Pu initially and were irradiated up to 124GWd/t in the experimental fast reactor JOYO. A melting temperature correlation was obtained by an experimental regression analysis using 21 measurements: where [Tcirc]m is the expected melting temperature (K), X 1 the plutonium fraction (Pu/(Pu+U)) and X 2 the americium fraction (Am/(Pu+U+Am)) at the measurement, and X 3 the burnup (GWd/t). The equation shows that the melting temperature decreases by 5, 4, and 3K per 10GWd/t at 50, 100 and 150GWd/t, respectively. The effects of actinides, such as Pu and Am, on melting temperature could be explained by an ideal solution model. However, the decreases caused by soluble fission products could not be well explained in terms of the ideal solution model.
Shinsuke Yamanaka - One of the best experts on this subject based on the ideXlab platform.
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effect of americium and simulated fission products addition on oxygen potential of uranium plutonium Mixed Oxide Fuels
MRS Proceedings, 2009Co-Authors: Kosuke Tanaka, Ken Kurosaki, Masahiko Osaka, Masayoshi Uno, Hiroaki Muta, Shinsuke YamanakaAbstract:The oxygen potentials at 1273 K of Mixed Oxide (MOX) Fuels with Am and 26 kinds of fission product elements (FPs), simulating low-decontaminated MOX fuel and high burn-up of up to 250 GWd/t, have been measured by using thermogravimetric analysis (TGA). The oxygen potentials for simulated low-decontaminated MOX Fuels were higher than the Fuels without FPs and increased with increasing simulated burn-up.
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molecular dynamics studies of americium containing Mixed Oxide Fuels
Journal of Nuclear Science and Technology, 2006Co-Authors: Ken Kurosaki, Jun Adachi, M Katayama, Masahiko Osaka, Kenya Tanaka, Masayoshi Uno, Shinsuke YamanakaAbstract:The molecular dynamics (MD) calculation was performed for americium-containing Mixed Oxide Fuels, (U0-7--x Pu0.3Am x )O2 (x=0,0.016; 0.03; 0.05; 0.1; 0.15), in the temperature range from 300 to around 2,500 K to evaluate the lattice parameter, heat capacity and thermal conductivity. The MD results reveal that the calculated heat capacity and thermal conductivity are at a similar level in the entire composition range, in other words they are scarcely influenced by adding americium up to 15%. This behavior was examined from a view point of a phonon-impurity scattering mechanism.
D. Terentyev - One of the best experts on this subject based on the ideXlab platform.
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Molecular dynamics study of oxygen transport and thermal properties of Mixed Oxide Fuels
Computational Materials Science, 2007Co-Authors: D. TerentyevAbstract:Abstract Evaluation of thermal and transport properties of actinide Mixed Oxide Fuels is a problem of high importance for the development of new generation reactors. Molecular dynamics provides a powerful tool, that can be used for this goal. In this work molecular dynamic simulations have been performed with the purpose to evaluate thermal and diffusion properties of UO2, PuO2 and (Pu0.3U0.7)O2 Oxide nuclear Fuels. Specific heat, thermal conductivity and thermal expansion of these materials were estimated in the temperature range of 300–2500 K and compared with available recommendations. Migration energies of the anion defects via the interstitial and vacancy mechanisms have been estimated together with oxygen self-diffusion coefficients. The presence of a strong interaction between point defects and Pu atoms in (Pu0.3U0.7)O2 Oxide has been revealed and this interaction alters the diffusivity of point defects in comparison with the pure uranium and plutonium diOxides.
F Hodaj - One of the best experts on this subject based on the ideXlab platform.
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ceramic processing of uranium plutonium Mixed Oxide Fuels u1 ypuy o2 with high plutonium content
Ceramics International, 2014Co-Authors: Romain Vauchy, Annecharlotte Robisson, Fabienne Audubert, F HodajAbstract:Abstract The ternary thermodynamic U–Pu–O system has been studied for decades for MOX fuel applications but the phase diagram is still not precisely described mostly in the UO 2 –PuO 2 –Pu 2 O 3 sub-system. Furthermore, uranium–plutonium Mixed Oxides containing high amounts of plutonium are now being considered within the scope of future nuclear reactors. Within this framework, obtaining homogeneous Mixed Oxides by powder metallurgy is paramount. The studied process is based on UO 2 and PuO 2 co-milling and applied to compounds with high Pu content. The objective of this study is obtaining microstructures free of local heterogeneities in the U–Pu distribution which are not suitable for research studies. Furthermore, in case of prospective irradiation application, local high Pu concentrations lead to “hot spots” in the material influencing the fission gas release behaviour such as the thermal conductivity which may raise a number of safety issues. This study describes the effect of some fabrication parameters on the powder morphology and/or, on the final microstructure ( e.g. U–Pu distribution). The co-milling, sieving and sintering steps were investigated within this scope and the resulting powders and pellets were characterised by X-ray diffraction (XRD) and optical microscopy observations, respectively.
Romain Vauchy - One of the best experts on this subject based on the ideXlab platform.
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Perspectives in using Raman spectroscopy for characterizing the microstructure of plutonium-bearing materials
2018Co-Authors: Laeticia Medyk, Patrick Simon, Aurélien Canizarès, Dario Manara, Rudy Konings, Jean-yves Colle, Romain Vauchy, Christophe Valot, Gilles Montagnac, Philippe MartinAbstract:In the frame of the development of uranium-plutonium Mixed Oxide Fuels for Sodium-cooled Fast Reactors (SFRs), characterizing nuclear materials by various techniques is paramount. These fast neutron reactors imply the use of a (U,Pu)O$_{2-x}$ ceramic fuel with a Pu/(U+Pu) content between 19 and 30 mol.%. Furthermore, the physico-chemical and microstructural properties of such Fuels, such as chemical homogeneity, oxygen stoichiometry (O/(U+Pu) ratio) and crystallographic structure, have to meet precise criteria for being introduced in the reactor core.
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ceramic processing of uranium plutonium Mixed Oxide Fuels u1 ypuy o2 with high plutonium content
Ceramics International, 2014Co-Authors: Romain Vauchy, Annecharlotte Robisson, Fabienne Audubert, F HodajAbstract:Abstract The ternary thermodynamic U–Pu–O system has been studied for decades for MOX fuel applications but the phase diagram is still not precisely described mostly in the UO 2 –PuO 2 –Pu 2 O 3 sub-system. Furthermore, uranium–plutonium Mixed Oxides containing high amounts of plutonium are now being considered within the scope of future nuclear reactors. Within this framework, obtaining homogeneous Mixed Oxides by powder metallurgy is paramount. The studied process is based on UO 2 and PuO 2 co-milling and applied to compounds with high Pu content. The objective of this study is obtaining microstructures free of local heterogeneities in the U–Pu distribution which are not suitable for research studies. Furthermore, in case of prospective irradiation application, local high Pu concentrations lead to “hot spots” in the material influencing the fission gas release behaviour such as the thermal conductivity which may raise a number of safety issues. This study describes the effect of some fabrication parameters on the powder morphology and/or, on the final microstructure ( e.g. U–Pu distribution). The co-milling, sieving and sintering steps were investigated within this scope and the resulting powders and pellets were characterised by X-ray diffraction (XRD) and optical microscopy observations, respectively.