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Journal Articles

None

Hirooka, Shun; Horii, Yuta; Hayashizaki, Kohei; Mohamad, A. B.

Kaku Nenryo, (60-1), p.17 - 20, 2025/02

no abstracts in English

Journal Articles

Uranium-plutonium-oxygen phase diagram; Investigating the solvus of fluorite's exsolution

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

https://doi.org/10.1016/j.jnucmat.2024.155233
 Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)

The fluorite exsolution/recombination in U$$_{1-y}$$Pu$$_{y}$$O$$_{2-x}$$ (y = 0.30 and 0.45) and PuO$$_{2-x}$$ 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$$sim$$50 K lower than previously reported. Finally, the systematic experimental procedure allowed us refining the locus of the solvus existing in hypostoichiometric U$$_{0.70}$$0Pu$$_{0.30}$$O$$_{2-x}$$, U$$_{0.55}$$Pu$$_{0.45}$$O$$_{2-x}$$, and PuO$$_{2-x}$$ dioxides.

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：6 Percentile：77.44(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

Sintering and microstructural behaviors of mechanically blended Nd/Sm-doped MOX

Hirooka, Shun; Horii, Yuta; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*; Vauchy, R.; Hayashizaki, Kohei; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Science and Technology, 60(11), p.1313 - 1323, 2023/11

https://doi.org/10.1080/00223131.2023.2192725
 Times Cited Count：5 Percentile：84.10(Nuclear Science & Technology)

Additive MOX pellets are fabricated by a conventional dry powder metallurgy method. Nd$$_{2}$$O$$_{3}$$ and Sm$$_{2}$$O$$_{3}$$ are chosen as the additive materials to simulate the corresponding soluble fission products dispersed in MOX. Shrinkage curves of the MOX pellets are obtained by dilatometry, which reveal that the sintering temperature is shifted toward a value higher than that of the respective regular MOX. The additives, however, promote grain growth and densification, which can be explained by the effect of oxidized uranium cations covering to a pentavalent state. Ceramography reveals large agglomerates after sintering, and Electron Probe Micro-Analysis confirms that inhomogeneous elemental distribution, whereas XRD reveals a single face-centered cubic phase. Finally, by grinding and re-sintering the specimens, the cation distribution homogeneity is significantly improved, which can simulate spent nuclear fuels with soluble fission products.

Oral presentation

Demonstration research on fast reactor recycling using low decontaminated MA-bearing MOX fuels, 9; Microstructure and thermal conductivity of MOX with simulated FPs

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

no journal, , 

The low-decontaminated fuel which contains significant amount of fission products (FPs) has been investigated as a fuel for the advanced fast reactor cycle. In this cycle, it is expected to reduce reprocessing cost and strengthen nuclear proliferation resistance of recovered plutonium accompanying high radiation dose FPs. As part of studies on physical properties of low-decontaminated fuel pellets, Nd$$_{2}$$O$$_{3}$$/Sm$$_{2}$$O$$_{3}$$ powders were added to the MOX raw powder as simulated fission products (FPs). The effects of simulated FPs on thermal conductivity were evaluated focusing on the microstructural homogeneities of simulated FPs. From the results of thermal diffusivity measurement and the EPMA mapping, the homogeneous simulated FPs decreased the thermal conductivity of the MOX.

Oral presentation

Improvement of dose rate measurement method and calculation analysis for MOX containing Americium

Okada, Toyofumi; Horii, Yuta; Tsubota, Yoichi; Komeno, Akira; Kikuno, Hiroshi

no journal, , 

In this research, purpose is getting gamma dose rate from MOX and PuO$$_{2}$$ containing Americium and evaluating exposure dose by calculation code. One of reasons of difference between measurement value and calculation value is scattering ray from inside of globe box panel because MOX is handled inside of globe box. Therefore, we improve measuring method (changing MOX position, applying shadow shield method) and get new data which difference between measurement value and calculation value (using ANISN) is smaller than before. Furthermore, we confirm that it can be available to apply PHITS for evaluating exposure dose.

Oral presentation

Phase separation/recombination in hypostoichiometric uranium-plutonium mixed oxide U$$_{1-y}$$PuyO$$_{2-x}$$ (y= 0.30; 0.45; and 1.00) using DSC

Vauchy, R.; Ogasawara, Masahiro*; Yamada, Tadahisa*; Sunaoshi, Takeo*; Tamura, Tetsuya*; Horii, Yuta; Hirooka, Shun; Murakami, Tatsutoshi

no journal, , 

The phase separation/recombination occurring in hypostoichiometric uranium-plutonium mixed oxides is important for technological applications, from manufacturing to irradiation behavior. The systematic cross-experiment procedure allowed refining the limits of the solvus existing in hypostoichiometric U$$_{1-y}$$PuyO$$_{2-x}$$ and PuO$$_{2-x}$$ dioxides to better understand the redox behavior of MOX fuel for fast reactors during manufacturing and storage.

Oral presentation

Pu-content dependence of MOX oxygen potential

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Sunaoshi, Takeo*; Saito, Kosuke

no journal, , 

no abstracts in English

Oral presentation

None

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Uno, Hiroki*; Tamura, Tetsuya*; Sunaoshi, Takeo*; Saito, Kosuke

no journal, , 

no abstracts in English

Oral presentation

Manufacturing of homogeneous simulated FPs (Nd$$_{2}$$O$$_{3}$$/Sm$$_{2}$$O$$_{3}$$/Gd$$_{2}$$O$$_{3}$$/ZrO$$_{2}$$)-doped MOX for studies on MOX fuel properties

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Hayashizaki, Kohei; Uno, Hiroki*; Tamura, Tetsuya*; Sunaoshi, Takeo*; Owada, Hideaki*; Yamada, Tadahisa*; Murakami, Tatsutoshi

no journal, , 

Fission products (FPs), which are generated and stored in the fuel matrix by irradiating MOX fuels, affect the fuel properties. In previous studies, many properties of unirradiated MOX were reported. On the other hand, the number of studies on irradiated MOX properties are limited. Studying properties of irradiated materials has difficulties in handling, therefore, using simulated FP-doped (U,Pu)O$$_{2}$$ is an alternative method in studying irradiated fuel properties. In order to evaluate the effect of simulated FPs, the homogeneity is one of the important factors. In this study, two dry-processing methods, namely melting and grinding-mixing methods, respectively, are employed and evaluated the aptitude as methods to prepare the homogeneous simulated FP-doped (U,Pu)O$$_{2}$$.

Oral presentation

Thermal conductivity measurement of homogenenous/heterogeneous simulated FP-doped MOX

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

no journal, , 

no abstracts in English

Oral presentation

Phase separation and recombination in U1-yPuyO$$_{2}$$-x with y = 0.30, 0.45, and 1.00

Vauchy, R.; Ogasawara, Masahiro*; Yamada, Tadahisa*; Sunaoshi, Takeo*; Tamura, Tetsuya*; Horii, Yuta; Hirooka, Shun; Kato, Masato; Murakami, Tatsutoshi

no journal, , 

Oral presentation

Development of visual inspection technology of pellets using machine learning, 2; Demonstration experiment using MOX pellets and consideration for introduction to production lines

Goto, Kenta; Hirooka, Shun; Horii, Yuta; Nakamichi, Shinya; Murakami, Tatsutoshi; Shibanuma, Kimikazu; Ono, Takanori; Yamamoto, Kazuya; Hatanaka, Nobuhiro; Okumura, Kazuyuki

no journal, , 

no abstracts in English

Oral presentation

Manufacturing and property measurements of homogeneous simulated FP (Nd/Sm/Gd/Zr)-doped MOX

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Hayashizaki, Kohei; Uno, Hiroki*; Tamura, Tetsuya*; Sunaoshi, Takeo*; Owada, Hideaki*; Yamada, Tadahisa*; Murakami, Tatsutoshi

no journal, , 

Fission products (FPs), which are generated and stored in fuel matrix by irradiating nuclear fuels, affect thermo-physical fuel properties. To improve accuracy of computer simulation of irradiation behaviors, studies on the fuel properties containing FPs are needed. However, only a limited number of studies on the irradiated fuel properties, especially MOX fuels, have been reported in the world due to difficulties in handling of the irradiated fuels. Moreover, the effect of individual FP cannot be evaluated because many kinds of FPs are stored in the irradiated fuels. Thus, an alternative method should be suggested to easily study the effects of FPs on the fuel properties. In this study, fuel properties were measured to evaluate the effects of FPs by using simulated FP-doped MOX specimens instead of a real irradiated fuel. For the measurement, the homogeneity of FP in a specimen is also important, as well as uranium and plutonium. To obtain homogeneous specimens, re-grinding and re-sintering processes were repeated and the improvement was confirmed by EPMA and XRD at each set of the process. A specimen with suitable homogeneity for measurement was prepared by repeating the series of processes three times. Sm$$_{2}$$O$$_{3}$$, Gd$$_{2}$$O$$_{3}$$ and ZrO$$_{2}$$, which are major and soluble FPs in irradiated MOX fuels, were selected as simulated FPs. The effect of individual FP on the properties, such as thermal conductivity and thermal expansion, was evaluated on the specimens. In addition, Nd$$_{2}$$O$$_{3}$$, Sm$$_{2}$$O$$_{3}$$ and Gd$$_{2}$$O$$_{3}$$ co-doped MOX was also prepared to compare the influence of containing multiple lanthanides.

Oral presentation

Evaluation of burn-up effect on MOX fuel thermal conductivity, 1; Thermal conductivity evaluation of simulated FP-doped MOX

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Uno, Hiroki*; Ogasawara, Masahiro*; Yamada, Tadahisa*; Murakami, Tatsutoshi

no journal, , 

no abstracts in English

Oral presentation

Evaluation of burn-up effect on MOX fuel thermal conductivity, 2; Investigation of thermal conductivity model considering burn-up

Hirooka, Shun; Horii, Yuta; Murakami, Tatsutoshi

no journal, , 

no abstracts in English

Oral presentation

Measurement and modelling of properties of (U,Pu)O $$_{2}$$ doped with simulated FP elements

Hirooka, Shun; Horii, Yuta; Vauchy, R.; Hayashizaki, Kohei; Owada, Hideaki*; Furusawa, Naoya*; Yamada, Tadahisa*; Sunaoshi, Takeo*; Uno, Hiroki*; Naganuma, Masayuki; et al.

no journal, , 

Oral presentation

Measurement and Pu-content dependence of (U,Pu)O$$_{2}$$$$_{pm}$$$$_{x}$$ oxygen potential

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Sunaoshi, Takeo*; Saito, Kosuke

no journal, , 

Oral presentation

Study on adjustment technology of low O/M ratio for MOX pellets, 2; Measurement of oxygen chemical diffusion coefficient

Vauchy, R.; Hirooka, Shun; Horii, Yuta; Sunaoshi, Takeo*; Saito, Kosuke

no journal, , 

Oral presentation

Study on controlling low O/M ratio in MOX pellet, 1; Variation of O/M ratio during whole heat treatment

Hirooka, Shun; Vauchy, R.; Horii, Yuta; Akashi, Masatoshi; Sunaoshi, Takeo*; Saito, Kosuke

no journal, , 

no abstracts in English

25 (Records 1-20 displayed on this page)