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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 10 $^{-2}$ + 0.93C $_{Nd}$ + 1.20C $_{Sm}$ , B(m/W)=2.39 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：3 Percentile：95.99(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.

Journal Articles

Liquid phase sintering of alumina-silica co-doped cerium dioxide CeO $_{2}$ ceramics

Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Sunaoshi, Takeo*; Yamada, Tadahisa*; Nakamichi, Shinya; Murakami, Tatsutoshi

Ceramics International, 49(2), p.3058 - 3065, 2023/01

https://doi.org/10.1016/j.ceramint.2022.09.295

Times Cited Count：8 Percentile：75.06(Materials Science, Ceramics)

Journal Articles

Oxygen potential measurement of (U,Pu,Am)O $_{2 pm x}$ and (U,Pu,Am,Np)O $_{2 pm x}$

Hirooka, Shun; Matsumoto, Taku; Kato, Masato; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*

Journal of Nuclear Materials, 542, p.152424_1 - 152424_9, 2020/12

https://doi.org/10.1016/j.jnucmat.2020.152424

Times Cited Count：6 Percentile：60.71(Materials Science, Multidisciplinary)

The measurement of oxygen potential was conducted at 1,673, 1,773, and 1,873 K for (U $_{0.623}$ Pu $_{0.350}$ Am $_{0.027}$ )O $_{2}$ and at 1,873 and 1,923 K for (U $_{0.553}$ Pu $_{0.285}$ Am $_{0.015}$ Np $_{0.147}$ )O $_{2}$ by using a thermo-gravimeter and an oxygen sensor. Am inclusion in terms of substituting the U significantly increased the oxygen potential. Similarly, the inclusion of Np as a substitute for U increased the oxygen potential; however, the effect was not as large as that with the Pu or Am addition at the same rate. The results were analyzed via defect chemistry and certain defect formations were suggested in the reducing region and the near-stoichiometric region by plotting the relationship between PO $_{2}$ and the deviation from the stoichiometry. The equilibrium constants of the defect reactions were arranged to reproduce the experiment such that Am/Np contents were included in the entropy with coefficients fitting the experimental data.

Oral presentation

A Consideration of utilization of U $_{3}$ O $_{8}$ powder in MOX fuel production

Tsuchimochi, Ryota; Hirooka, Shun; Sunaoshi, Takeo*; Yamada, Tadahisa*

no journal, ,

no abstracts in English

Oral presentation

The Measurement of thermal conductivity of MOX including simulated FPs and impurities

Horii, Yuta; Hirooka, Shun; Kato, Masato; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Okumura, Kazuyuki

no journal, ,

As part of studies on physical properties of low-decontaminated fuel pellets, simulated FPs (Sm $_{2}$ O $_{3}$ and Nd $_{2}$ O $_{3}$ ) and impurities included at such as fuel fabrication process (Al $_{2}$ O $_{3}$ and SiO $_{2}$ ) were added to MOX, and their effects on thermal conductivity were evaluated. Addition of Sm $_{2}$ O $_{3}$ and Nd $_{2}$ O $_{3}$ in MOX, that can be solutionized, decreased thermal conductivity whereas addition of Al $_{2}$ O $_{3}$ and SiO $_{2}$ , that don't make a solid solution with MOX, increased thermal conductivity.

Oral presentation

Sintering behavior of recycled MOX powder

Nakamichi, Shinya; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*; Okumura, Kazuyuki

no journal, ,

no abstracts in English

Oral presentation

Thermal conductivity of MOX with simulated fission products

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

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. However, FPs could affect on thermal properties of MOX. In particular, thermal conductivity is one of the most important properties for fuel design. Many studies have been reported on the thermal conductivity of MOX that evaluated the effect of plutonium and minor actinide (Am, Np) contents. However, the number of studies on the thermal conductivity of MOX containing FPs are limited; only Nd, Eu and Zr. In this study, thermal conductivity of MOX including Nd $_{2}$ O $_{3}$ and Sm $_{2}$ O $_{3}$ , which are expected to be the main FPs remaining as solid solutions in low-decontaminated fuel, was evaluated by measuring thermal diffusivity of sintered pellet. Nd $_{2}$ O $_{3}$ powder and Sm $_{2}$ O $_{3}$ powder was added into raw MOX powder of 20% plutonium content and they were milled by ball-mill. Sintered pellets were obtained by pressing and sintering this powder. The thermal diffusivities were measured by a laser flash method from 973 K to 1673 K at every 100 K. The thermal conductivities up to 1400K are expressed by classical phonon transport model as =(A+BT) $^{-1}$ where A(mK/W)=2.010 $^{-3}$ +4.810 $^{-1}$ C $_{Nd,Sm}$ and B(m/W)=2.510 $^{-4}$ . This means the thermal conductivity decreased with enhancement of phonon scattering by adding Nd $_{2}$ O $_{3}$ /Sm $_{2}$ O $_{3}$ , similar to the previous studies.

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, ,

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*; Ohwada, 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

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*; Ohwada, 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 recombinaison 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

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