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 ( 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.

Thermal conductivities of hypostoichiometric (U, Pu, Am)O 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

The thermal conductivities of (UPuAm)O 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, = (A+BT), A(x) = 2.89x + 2.2410 (m K/W) and B(x) = (- 6.70x + 2.48) 10 (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.

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.

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.

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

Thermal physical properties of MOX Fuels, 3; Measurements of thermal diffusivity for (UPuAm)O (x=1.902.00)

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.

Enthalpy measurement and heat capacity evaluation of (U, Pu, Am)O

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

no journal, , 

no abstracts in English

Thermal conductivity of (UPuNp)O

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

no journal, , 

no abstracts in English

Enthalpy measurement and evaluation of heat capacity on PuO

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 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.

Recovery behaviors of thermal conductivities in self-irradiated MOX fuel

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

no journal, , 

In MOX fuel stored for long term, the defects induced by self-irradiation (alpha decay) increase with an increase of storage time, and consequently its thermal conductivities decrease by an accumulation of these defects. Some investigators showed that these defects were decreased or annihilated by heat treatment, and then the lattice parameter, thermal conductivity, etc of fuel were recovered. In this study, we prepared the specimens divided from MOX fuels stored for about fifteen years. We investigated a recovery behavior of the thermal conductivities of this specimen and evaluated this behavior as functions of heating temperature and heating time.