Experimental evaluation of Am-and Np-bearing mixed-oxide fuel properties

Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki

Proceedings of 10th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation (CD-ROM), p.201 - 209, 2010/00

Japan Atomic Energy Agency has developed homogeneous MOX fuel containing minor actinide (MA) elements such as Np and Am. To measure physical properties of the fuel is essential for its development, because their data are needed to evaluate irradiation behavior. In this report, the physical properties, melting temperature, thermal conductivity, lattice parameter, oxygen potential and phase separation behavior, were reviewed, and effect of MA content was discussed.

The Phase state at high temperatures in the MOX-SiO system

Nakamichi, Shinya; Kato, Masato; Sunaoshi, Takeo*; Uchida, Teppei; Morimoto, Kyoichi; Kashimura, Motoaki; Kihara, Yoshiyuki

Journal of Nuclear Materials, 389(1), p.191 - 196, 2009/05

https://doi.org/10.1016/j.jnucmat.2009.01.028

Japan Atomic Energy Agency researchers have developed mixed oxide (MOX) fuels containing minor actinides (MA). These fuels were irradiated for ten minutes in the FBR Joyo in some short-term irradiation tests. The Si-condensed phases were observed at the center of the pellets in the post irradiation examination. Si impurities came to be mixed into the raw materials in the ball milling process, because Si rubber was used as the lining of the milling pot. Content of Si in the pellets was within the specification of the fuel. It is important to investigate the Si state in MOX at high temperatures like the reactor operating temperature of the fuel to evaluate irradiation behavior. In the present work, MOX specimens with mixed SiO impurity were prepared. The ratio of MOX to SiO was controlled at a mol fraction of 3 to 1. The specimens were first heated at 1973K in atmospheres of three different oxygen partial pressures to adjust the O/M ratio. Then these specimens were sealed in a tungsten capsule, and heated at 2273K or 2673K. Compounds consisting of Pu and Si were observed at grain boundaries of the MOX matrix in specimens after heat treatment. These compounds were not observed in grain interior and MOX matrix was not affected significantly by Si impurity. These compounds tended to form in specimens with low O/M ratio and in specimens heated at higher temperatures.

Thermal conductivities of (U,Pu,Am)O 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

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 Pu U)O, (Am Pu U)O and (Am Pu U)O 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, PuO and AmO by using the Kopp-Neumann rule.

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 conductivity of (U, Pu, Am)O

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki

Transactions of the American Nuclear Society, 97(1), p.618 - 619, 2007/11

Plutonium and uranium mixed oxide (MOX) fuel with high Pu-content has been developed as a fuel for fast reactors (FRs). Thermal conductivity of the oxide fuel is among the most important properties for design and performance analyses of fuel rods. Among recent reports, there have been none examining of thermal conductivity of MOX fuel containing Am except our studies. In this study, the thermal conductivities of MOX fuel with 30% Pu-content, as obtained by our group, were evaluated as functions of temperature, oxygen-to-metal (O/M) ratio and Am-content.

Evaluation of melting temperature in (PuAmU)O

Nakamichi, Shinya; Kato, Masato; Morimoto, Kyoichi; Sugata, Hiromasa*; Kashimura, Motoaki; Abe, Tomoyuki

Transactions of the American Nuclear Society, 96(1), p.191 - 192, 2007/06

JAEA has developed plutonium and uranium mixed oxide (MOX) containing 20-32%Pu content as a fuel of the fast breeder reactor. During irradiation, large temperature gradient in radial direction of a fuel pellet causes redistribution of Pu and U, and the Pu content increases to about 43% at the pellet center. The maximum temperature of the fuel pellet during irradiation is limited within the design criterion to prevent fuel melting. So, it is important to evaluate melting points of MOX containing 43%Pu. In this work, it is confirmed that the MOX with 43%Pu content is not melted by heat treatment just below the melting point which was determined by thermal arrest technique using Re inner capsule. The MOX specimen with 43%Pu content was heated at 2978K for 40s using Re inner capsule. Optical micrograph and XRD results show the specimen was heated at the temperature less than solidus temperature. So it was confirmed that (PuAmU)O was solid phase at 2978K20K.

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.

Proposal of a Nuclear Cycle Research and Development Plan in Tokai Works -The Roadmap from LWR Cycle to FBR Cycle-

Nakamura, Hirofumi; Abe, Tomiyuki; Kashimura, Takao; Nagai, Toshihisa; Maeda, Seiichiro; Yamaguchi, T.; Kuroki, Ryoichiro

JNC TN8440 2003-016, 39 Pages, 2003/07

JNC-TN8440-2003-016.pdf:0.79MB

The Generation-II Project Task Force Team has investigated a research and development plan on a future nuclear fuel cycle in Tokai works for about three months from December 19,2002. First we have discussed about the present condition of Japanese nuclear fuel cycle and have recognized it as the following. *The relation of the technology between the LWR-cycle and the FBR-cycle is not clear. *MOX Fuel Use in Light Water Reactors is important to establish technology of the FBR fuel cycle. *Radioactive waste disposal issue is urgent. Next we have proposed the three basic policies on R&D plan of nuclear fuel cycle in consideration of the F.S. on FBR-cycle. *Establishment and advancement of "the tough nuclear fuel cycle". *Early establishment of the FBR cycle technology to be able to supply energy stably for long-term. *Establishment of the radioactive waste treatment and disposal technology, and optimization of nuclear fuel cycle technology from the viewpoint of radioactive waste.And we have proposed the Japanese technical holder system to integrate all LWR and FBR cycle technology.

Thermal physical properties of MOX fuel, 1; Overall plan

Abe, Tomoyuki; Kashimura, Motoaki; Kato, Masato; Morimoto, Kyoichi

no journal, , 

no abstracts in English

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.