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Kokubun, Yuji; Hosomi, Kenji; Seya, Natsumi; Nagaoka, Mika; Inoue, Kazumi; Koike, Yuko; Hasegawa, Ryo; Kubota, Tomohiro; Hirao, Moe; Iizawa, Shogo; et al.
JAEA-Review 2024-053, 116 Pages, 2025/03
Based on the regulations (the safety regulation of Tokai Reprocessing Plant, the safety regulation of nuclear fuel material usage facilities, the radiation safety rule, the regulation about prevention from radiation hazards due to radioisotopes, which are related with the nuclear regulatory acts, the local agreement concerning with safety and environment conservation around nuclear facilities, the water pollution prevention act, and by law of Ibaraki Prefecture), the effluent control of liquid waste discharged from the Nuclear Fuel Cycle Engineering Laboratories of Japan Atomic Energy Agency has been performed. This report describes the effluent control results of the liquid waste in the fiscal year 2023. In this period, the concentrations and the quantities of the radioactivity in liquid waste discharged from the reprocessing plant, the plutonium fuel fabrication facilities, and the other nuclear fuel material usage facilities were much lower than the limits authorized by the above regulations.
Vauchy, R.; Hirooka, Shun; Horii, Yuta; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Yamada, Tadahisa*; Tamura, Tetsuya*; Murakami, Tatsutoshi
Journal of Nuclear Materials, 599, p.155233_1 - 155233_11, 2024/10
Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)The fluorite exsolution/recombination in UPu
O
(y = 0.30 and 0.45) and PuO
was investigated using differential scanning calorimetry. The results are in relatively good agreement with the literature data, except for plutonia. Our values indicate that the critical temperature of the miscibility gap in Pu-O is 30
50 K lower than previously reported. Finally, the systematic experimental procedure allowed us refining the locus of the solvus existing in hypostoichiometric U
0Pu
O
, U
Pu
O
, and PuO
dioxides.
Hirooka, Shun; Morimoto, Kyoichi; Matsumoto, Taku; Ogasawara, Masahiro*; Kato, Masato; Murakami, Tatsutoshi
Journal of Nuclear Materials, 598, p.155188_1 - 155188_9, 2024/09
Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)no abstracts in English
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
Times Cited Count:8 Percentile:88.57(Materials Science, Multidisciplinary)The thermal conductivities of near-stoichiometric (U,Pu,Am)O doped with Nd
O
/Sm
O
, 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)
. The dependences of the coefficients A and B on the Nd/Sm content (C
and C
, respectively) are evaluated as: A(mK/W)=1.70
10
+ 0.93C
+ 1.20C
, B(m/W)=2.39
10
.
Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*
Journal of Nuclear Materials, 443(1-3), p.286 - 290, 2013/11
Times Cited Count:5 Percentile:36.22(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.
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.
Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki; Abe, Tomoyuki
Journal of Alloys and Compounds, 452(1), p.54 - 60, 2008/03
Times Cited Count:32 Percentile:78.03(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 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.
Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki
Journal of Nuclear Materials, 374(3), p.378 - 385, 2008/03
Times Cited Count:37 Percentile:89.08(Materials Science, Multidisciplinary)The thermal conductivities of (UPu
Am
)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) = 2.89
x + 2.24
10
(m K/W) and B(x) = (- 6.70
x + 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.
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
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
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.
Kato, Masato; Morimoto, Kyoichi; Kihara, Yoshiyuki; Ogasawara, Masahiro*; Tamura, Tetsuya*; Uno, Hiroki*; Sunaoshi, Takeo*
JNC TN8400 2004-022, 44 Pages, 2005/03
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 (NpAm
Pu
U
)O
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.
; Endo, Hideo; Ogasawara, Masahiro*; Shinada, Masanori*;
JNC TN8440 2003-004, 24 Pages, 2003/05
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.
Todokoro, Akio; Watahiki, Masatoshi; Kihara, Yoshiyuki; Ishii, Yasuhiko*; Ogasawara, Masahiro*; Otaka, Akihiro*
PNC TN8410 96-238, 86 Pages, 1996/08
None
Todokoro, Akio; Kihara, Yoshiyuki; Watahiki, Masatoshi; Ogasawara, Masahiro*; Otaka, Akihiro*
PNC TN8410 94-403, 30 Pages, 1994/11
None
Komeno, Akira; Uchida, Teppei; Ogasawara, Masahiro*; Morimoto, Kyoichi
no journal, ,
no abstracts in English
Matsumoto, Taku; Arima, Tatsumi*; Inagaki, Yaohiro*; Idemitsu, Kazuya*; Kato, Masato; Morimoto, Kyoichi; Ogasawara, Masahiro*
no journal, ,
no abstracts in English
Morimoto, Kyoichi; Ogasawara, Masahiro*
no journal, ,
Many studies discussing the influence of Pu-content on thermal conductivities of MOX fuel have been conducted by using experimental and calculation techniques. On the other hand, the disparity among these experimental results regarding the dependence of Pu-content on thermal conductivity is large and various researchers have expressed several different opinions about the influence of Pu-oxide addition on thermal conductivity. In this study, the stoichiometric MOX specimens containing from 2 to 46% of Pu were prepared and the thermal conductivities of them were measured with a laser flash apparatus. In the Pu-content range less than 46%, it was found that the influence of Pu-content on thermal conductivities was small and that the thermal conductivities decrease slightly and monotonically with an increase of Pu-content.
Morimoto, Kyoichi; Matsumoto, Taku; Ogasawara, Masahiro*
no journal, ,
Many studies discussing the influence of Pu-content on thermal conductivities of MOX fuel have been conducted using experimental and calculation techniques. On the other hand, the disparity among these experimental results regarding the dependence of Pu-content on thermal conductivity is large and various researchers have expressed several different opinions about the influence of Pu-oxide addition on thermal conductivity. In this study, the stoichiometric MOX specimens containing from 2 to 46% of Pu were prepared and the thermal conductivities of them were measured with a laser flash apparatus. In the Pu-content range less than 46%, it was found that the influence of Pu-content on thermal conductivities was small and that the thermal conductivities decrease slightly and monotonically with an increase of Pu-content.
Matsumoto, Taku; Kato, Masato; Morimoto, Kyoichi; Arima, Tatsumi*; Inagaki, Yaohiro*; Idemitsu, Kazuya*; Ogasawara, Masahiro*; Sunaoshi, Takeo*
no journal, ,
no abstracts in English