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Journal Articles

Thermal conductivity measurement of uranium-plutonium mixed oxide doped with Nd/Sm as simulated fission products

Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01

https://doi.org/10.1016/j.jnucmat.2023.154799
 Times Cited Count：1 Percentile：72.91(Materials Science, Multidisciplinary)

The thermal conductivities of near-stoichiometric (U,Pu,Am)O$$_{2}$$ doped with Nd$$_{2}$$O$$_{3}$$/Sm$$_{2}$$O$$_{3}$$, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT)$$^{-1}$$. The dependences of the coefficients A and B on the Nd/Sm content (C$$_{Nd}$$ and C$$_{Sm}$$, respectively) are evaluated as: A(mK/W)=1.70 $$times$$ 10$$^{-2}$$ + 0.93C$$_{Nd}$$ + 1.20C$$_{Sm}$$, B(m/W)=2.39 $$times$$ 10$$^{-4}$$.

Journal Articles

Thermal diffusivity measurement of (U, Pu)O$$_{2-x}$$ at high temperatures up to 2190 K

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*

Journal of Nuclear Materials, 443(1-3), p.286 - 290, 2013/11

https://doi.org/10.1016/j.jnucmat.2013.07.048
 Times Cited Count：5 Percentile：38.62(Materials Science, Multidisciplinary)

In this study, measurement was conducted for the sliced MOX pellets containing 30% of Pu prepared by a conventional powder metallurgy technology. Oxygen-to-metal (O/M) ratios of the samples were adjusted in the range from 1.92 to 2.00. The thermal diffusivities of these samples were measured at temperature up to 2150 K with the laser flash method. Thermal diffusivities of the near-stoichiometric samples obtained in the cooling process were greatly lower than those in the heating process unlike measurement below 1770 K. On the other hand, they were almost identical for the sample of 1.946 in O/M. It was also shown that thermal diffusivity decreased with the temperature but increased with the O/M.

Journal Articles

Thermal recovery evaluation of thermal conductivity in a self-irradiated MOX pellet

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*

Proceedings of Plutonium Futures; The Science 2010 (CD-ROM), p.339 - 340, 2010/09

Nuclear fuel pellets are stored before loading into a reactor. In some cases, the fuel pellets are left for several years. When uranium-plutonium mixed oxide (MOX) fuel pellets are stored for a long time, lattice defects induced by self-irradiation ($$alpha$$ decay) accumulate and these defects affect physical properties of the pellets, i.e. lattice parameter, electrical resistivity and thermal conductivity. The thermal conductivity of fuel pellets is one of the most important properties for fuel design and performance analyses; it is known to decrease due to the defects induced by self-irradiation, but it can be recovered by heating the pellets. In this study, the recovery behavior of thermal conductivity of a MOX fuel pellet stored for long time was investigated as a function of time and temperature, in order to make it easy to analyze the thermal performance of fuel pellets.

Journal Articles

Thermal conductivities of (U,Pu,Am)O$$_{2}$$ solid solutions

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki; Abe, Tomoyuki

Journal of Alloys and Compounds, 452(1), p.54 - 60, 2008/03

https://doi.org/10.1016/j.jallcom.2006.12.159
 Times Cited Count：30 Percentile：77.85(Chemistry, Physical)

Plutonium and uranium mixed oxide (MOX) fuel with high Pu content have been developed as a fuel of fast reactor (FR). As the storage time of Pu raw material between reprocessing and fabrication increases, americium content of the fabricated MOX fuel increases up to a few percent. In this work, the thermal conductivity of MOX fuel containing Am was investigated as a part of clarifying the effect of Am content on thermal physical properties. The pellets of (Am$$_{0.007}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$, (Am$$_{0.02}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ and (Am$$_{0.03}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ were prepared. The oxygen to metal ratio (O/M ratio) of sintered pellet was adjusted to 2.00. The thermal diffusivity measurement was carried out in the range of temperature from 900 K to 1700 K by the laser flash method, and thermal conductivity of these pellets was evaluated. The heat capacity for evaluating thermal conductivity was derived from heat capacity of UO$$_{2}$$, PuO$$_{2}$$ and AmO$$_{2}$$ by using the Kopp-Neumann rule.

Journal Articles

Thermal conductivities of hypostoichiometric (U, Pu, Am)O$$_{2-x}$$ oxide

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki

Journal of Nuclear Materials, 374(3), p.378 - 385, 2008/03

https://doi.org/10.1016/j.jnucmat.2007.09.003
 Times Cited Count：35 Percentile：89.2(Materials Science, Multidisciplinary)

The thermal conductivities of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ solid solutions (x = 0.0 - 0.1) were studied at temperatures from 900 to 1773 K. Thermal conductivities were obtained from the thermal diffusivity measured by laser flash method. The thermal conductivities obtained experimentally up to about 1400K could be expressed by a classical phonon transport model, $$lambda$$ = (A+BT)$$^{-1}$$, A(x) = 2.89$$times$$x + 2.24$$times$$10$$^{-2}$$ (m K/W) and B(x) = (- 6.70$$times$$x + 2.48) $$times$$ 10$$^{-4}$$ (m/W). The experimental values of A showed a good agreement with theoretical predictions. The experimental values of B could be fairly expressed by the theoretical prediction in the region x $$<$$ 0.04, but not deviated from the ones in the region x $$>$$ 0.04. Although this reason could not be understood well, it is most likely that the uncertainty in the measurement of melting temperature cause this difference.

Journal Articles

Measurement of thermal conductivity of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ in high temperature region

Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Ogasawara, Masahiro*; Sunaoshi, Takeo*

Transactions of the American Nuclear Society, 97(1), p.616 - 617, 2007/11

no abstracts in English

JAEA Reports

Evaluation of thermal physical properties for fast reactor fuels; Melting point and thermal conductivities

Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Sugata, Hiromasa*; et al.

JAEA-Technology 2006-049, 32 Pages, 2006/10

JAEA-Technology-2006-049.pdf:19.46MB
JAEA-Technology-2006-049(errata).pdf:0.32MB

Japan Atomic Energy Agency has developed a fast breeder reactor(FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio(O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Nuemann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content.

JAEA Reports

Development of low decontaminated MOX fuel containing MA IV; Oxygen potential and phase relation

Kato, Masato; Morimoto, Kyoichi; Kihara, Yoshiyuki; Ogasawara, Masahiro*; Tamura, Tetsuya*; Uno, Hiroki*; Sunaoshi, Takeo*

JNC TN8400 2004-022, 44 Pages, 2005/03

JNC-TN8400-2004-022.pdf:5.43MB

Japan Nuclear Development Institute has developed homogeneous mixed oxide fuel containing minor actinide as a fuel of an advanced fast reactor. Study on the sintering behavior of the fuel was carried out and the heat treatment technique for preparing homogeneous low O/M fuel had been developed. In this report, oxygen potential was measured and phase relation was evaluated, which are needed essentially for developing the new type fuel.Oxygen potential of (Npsub0.02Amsub0.02Pusub0.3Usub0.66)Osub2-X was measured by gas equilibrium method as a function of temperature and O/M ratio. The MOX with MA has slightly higher oxygen potential as compared with that of MOX without MA. And the model of oxygen potential was derived from the measurement results based on lattice defect theory.In samples with low O/M ratio, two fcc phases were observed at room temperature. The temperature of the phase separation was measured and it is observed that the addition of MA have the effect to be decreased the phase separation temperature. In the MOX containing MA and Nd simulated a low decontaminated fuel, the Pu-Am-Nd oxides were precipitated by decreasing O/M ratio in less than 1.96.

JAEA Reports

Evaluation of dissolution rate on high plutonium content MOX fuel

; Endo, Hideo; Ogasawara, Masahiro*; Shinada, Masanori*; *

JNC TN8440 2003-004, 24 Pages, 2003/05

JNC-TN8440-2003-004.pdf:5.0MB

The dissolution rate of high Pu content MOX fuel into the nitric acid was measured as a function of Pu content. MOX fuel sample which was pressed and sintered, were dissolved in boiling 7M nitric acid, and dissolution rate was measured by analysis of the Pu and U concentration in the solution. Dissolution rate of MOX fuel had the following tendencies : it decreased with increase of the Pu content and was reduced after 6 hours dissolution. These results agreed well to previous one, but dissolution rate was 3-6 times larger than that. It is estimated that the cause of this difference was due to the underestimate of surface area of MOX fuel powder and the difference of MOX O/M ratio.

JAEA Reports

None

Todokoro, Akio; Watahiki, Masatoshi; Kihara, Yoshiyuki; Ishii, Yasuhiko*; Ogasawara, Masahiro*; Otaka, Akihiro*

PNC TN8410 96-238, 86 Pages, 1996/08

PNC-TN8410-96-238.pdf:2.3MB

None

JAEA Reports

None

Todokoro, Akio; Kihara, Yoshiyuki; Watahiki, Masatoshi; Ogasawara, Masahiro*; Otaka, Akihiro*

PNC TN8410 94-403, 30 Pages, 1994/11

PNC-TN8410-94-403.pdf:1.71MB

None

JAEA Reports

None

; Ogasawara, Masahiro*; *; Nemoto, Takeshi;

PNC TN8410 93-084, 34 Pages, 1993/04

PNC-TN8410-93-084.pdf:0.67MB

None

Oral presentation

Thermal physical properties of MOX Fuels, 3; Measurements of thermal diffusivity for (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{x}$$ (x=1.90$$sim$$2.00)

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

To examine the influences of O/M ratio on the thermal conductivity of MOX fuel, the thermal diffusivities of the MOX fuel with 30% Pu content were measured, and thermal conductivities were evaluated.

Oral presentation

Thermal physical properties of MOX fuels, 2; Thermal diffusivity measurement of (U, Pu, Am)O$$_{2.00}$$

Komeno, Akira; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Enthalpy measurement and heat capacity evaluation of (U, Pu, Am)O$$_{2}$$

Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Thermal conductivity of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ at temperatures from 900 to 2200K

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Thermal conductivity of (U$$_{0.7-y}$$Pu$$_{0.3}$$Np$$_{y}$$)O$$_{2}$$

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Ogasawara, Masahiro*

no journal, , 

no abstracts in English

Oral presentation

Thermal recovery behavior of thermal conductivity in MOX pellet stored for long-term

Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Ogasawara, Masahiro*

no journal, , 

no abstracts in English

Oral presentation

Enthalpy measurement and evaluation of heat capacity on PuO$$_{2-x}$$

Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Nishi, Tsuyoshi; Ichise, Kenichi; Ito, Akinori; Ogasawara, Masahiro*

no journal, , 

Heat capacity of MOX fuel is one of the important physical properties for fuel design and for evaluation of thermal conductivity by laser flash method. In this study, the enthalpies of PuO$$_{2-x}$$ were measured as one of the evaluation of basic physical properties on MOX fuel. Additionally the temperature and O/Pu ratio dependencies on heat capacities were evaluated by analysing the enthalpy data.

Oral presentation

The Influence of Pu-content on thermal conductivities of (U,Pu)O$$_{2}$$ solid solution

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*

no journal, , 

Thermal conductivities of UO$$_{2}$$ and MOX containing Pu from 20 to 40% were measured. These data were compared with the thermal conductivities obtained in other institutes. From the results of this experiment, it was found that the thermal conductivities of MOX was lower than those of UO$$_{2}$$ and the variation of thermal conductivities of MOX hardly observed in the Pu-content range from 20% to 40%.

48 (Records 1-20 displayed on this page)