Thermal conductivity evaluation of Am-doped oxide fuels

Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Yano, Yasuhide; Tokoro, Daishiro*; Sugata, Hiromasa*; Kato, Masato*

JAEA-Research 2025-002, 18 Pages, 2025/05

JAEA-Research-2025-002.pdf:1.73MB

It is advocated as a development target of fast reactors (FRs) to allow for the of use of mixed oxide (MOX) fuels containing minor actinide (MA) separated and recovered from spent fuels with the aim of reducing the volume and toxicity of high-level radioactive waste generated from nuclear reactors. In the development of MAMOX fuels, it is important behavior to understand the thermal properties such as thermal conductivity for fuel design and analysis of the irradiation. However, there are only a few reports on the thermal properties of MA-MOX fuels, and neither the effects of MA contents nor of oxygen non-stoichiometry in MOX fuels on their thermal conductivities have been fully understood. In this study, the thermal conductivities of MOX fuels with up to 15% Am content were measured at near-stoichiometric composition and the relationship between thermal conductivity and Am content was evaluated. Moreover, the thermal conductivities of Am-doped UO fuels were also measured and evaluated by comparison with Am-MOX to evaluate the effect of Am content. The fuel samples used in this study were three types of MOX with a Pu content of 30% and different Am contents (5%, 10%, and 15%), and UO containing 15% Am. The thermal conductivities of specimens were calculated from the thermal diffusivities measured by the laser flash method, the density of the specimens and, the heat capacity at constant pressure. The oxygen partial pressure during the measurement was controlled at that of the targeted near-stoichiometric composition. The thermal conductivities of all specimens exhibited a decline with increasing temperature and Am content, with a particularly pronounced reduction observed below 1,173 K. The results of the classical phonon scattering model analysis of the measured thermal conductivities showed that the effect of lattice strain due to the Am addition was significant on the thermal resistivity change, and the effect was comparable for both MOX and UO.

Oxygen potential measurement of UAmO at 1473, 1573, and 1673 K

Watanabe, Masashi; Yokoyama, Keisuke; Vauchy, R.; Kato, Masato; Sugata, Hiromasa*; Seki, Takayuki*; Hino, Tetsushi*

Journal of Nuclear Materials, 599, p.155232_1 - 155232_5, 2024/10

https://doi.org/10.1016/j.jnucmat.2024.155232

Oxygen potential data of UAmO were measured at 1473, 1573, and 1673 K by thermogravimetry. In UAnO, where An stands for Pu or Am, and for a given value of y and Oxygen/Metal ratio, the oxygen potential of UAmO is higher than that of UPuO. The valence of cations in the hypostoichiometric region is similar to that of Nd-doped UO. At the stoichiometric composition, it is estimated to be Am, U, and U (for charge compensation of Am). The experimental data were analyzed using a defect chemistry model, and a relationship connecting the oxygen-to-metal ratio, the temperature, and the equilibrium oxygen partial pressure was proposed.

Anisotropic creep property related to non-spherical shape of mechanically alloyed powder of oxide dispersion strengthened F82H

Sakasegawa, Hideo; Nakajima, Motoki*; Kato, Taichiro*; Nozawa, Takashi*; Ando, Masami*

Materials Today Communications (Internet), 40, p.109659_1 - 109659_8, 2024/08

https://doi.org/10.1016/j.mtcomm.2024.109659

Nanometric oxide particles play an important role in improving the creep property of Oxide Dispersion Strengthened (ODS) steels. In our previous research, we examined a microstructural feature known as prior particle boundary (PPB). PPB refers to the surface of mechanically alloyed (MA) powders before consolidation. We revealed that the ODS steel with fine PPBs produced from smaller MA powders, exhibited shorter creep rupture times, compared to that with coarse PPBs produced from larger MA powders. The size of MA powders had an impact on the creep property. In this study, we examined the shape of MA powders, which were non-spherical shapes. Such shapes have the potential to induce anisotropic creep behavior. We conducted small punch creep tests on specimens with two different orientations to study the possible anisotropy. The results revealed that the creep rupture times varied depending on the orientation of specimen, thus indicating anisotropic creep property.

High temperature nanoindentation of (U,Ce)O compounds

Frazer, D.*; Saleh, T. A.*; Matsumoto, Taku; Hirooka, Shun; Kato, Masato; McClellan, K.*; White, J. T.*

Nuclear Engineering and Design, 423, p.113136_1 - 113136_7, 2024/07

https://doi.org/10.1016/j.nucengdes.2024.113136

Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young's modulus, hardness, and creep stress exponents. In this study, (U,Ce)O solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (Ux-1,Cex)O compounds ranging from x equals 0.1 to 0.3 at up to 800 C: their Young's modulus, hardness, and creep stress exponents were evaluated. The Young's modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800 C give stress exponent values, n=4.7-6.9, that suggests dislocation motion as the deformation mechanism.

A Science-based mixed oxide property model for developing advanced oxide nuclear fuels

Kato, Masato; Oki, Takumi; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.; Ozawa, Takayuki; Uwaba, Tomoyuki; Ikusawa, Yoshihisa; Nakamura, Hiroki; Machida, Masahiko

Journal of the American Ceramic Society, 107(5), p.2998 - 3011, 2024/05

https://doi.org/10.1111/jace.19609

Model catchments for testing biosphere assessment models of geological disposal

Yamaguchi, Masaaki; Suzuki, Yuji*; Kabasawa, Satsuki; Kato, Tomoko

JAEA-Data/Code 2024-001, 21 Pages, 2024/03

JAEA-Data-Code-2024-001.pdf:3.45MB
JAEA-Data-Code-2024-001-appendix(CD-ROM).zip:8.0MB

Model catchments have developed for use in testing various assessment models that can consider specific surface environmental conditions such as topography, riverine systems, and land use in the biosphere assessment of HLW geological disposal. The model catchments consist of the topography and riverine system of the catchment area created using existing tools, as well as land use and population distribution, river discharge, sediment flux data set by algorithms from topographical data. Datasets of three types of model watersheds (Types 1 to 3, watershed area: 730 to 770 km) with different topographical characteristics have released as raster data that can be handled by geographic information systems (GIS). Since the model catchments were created virtually reflecting as much as possible the main characteristics of Japan's surface environment, they can be used as a test bed for conducting hydraulic/mass transport analysis to set the GBI and compartment model.

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

The thermal conductivities of near-stoichiometric (U,Pu,Am)O doped with NdO/SmO, 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.

Estimation of the activity median aerodynamic diameter of plutonium particles using image analysis

Takasaki, Koji; Yasumune, Takashi; Yamaguchi, Yukako; Hashimoto, Makoto; Maeda, Koji; Kato, Masato

Journal of Nuclear Science and Technology, 60(11), p.1437 - 1446, 2023/11

https://doi.org/10.1080/00223131.2023.2196969

The aerodynamic radioactive median diameter (AMAD) is necessary information to assess the internal exposure. On June 6, 2017, at a plutonium handling facility in Oarai site of Japan Atomic Energy Agency (JAEA), during the inspection work of a storage container that contains nuclear fuel materials, accidental contamination occurred and five workers inhaled radioactive materials including plutonium. Some smear papers and an air sampling filter were measured with the imaging plate, and we conservatively estimated minimum AMADs for two cases, plutonium nitrate and plutonium dioxide. As a result of AMAD estimation, even excluding a giant particle of a smear sample, the minimum AMADs of plutonium nitrate from smear papers were 4.3 - 11.3 m and those of plutonium dioxide were 5.6 - 14.1 m. Also, the minimum AMAD of plutonium nitrate from an air sampling filter was 3.0 m and that of plutonium dioxide was 3.9 m.

Theory of inverse Rashba-Edelstein effect induced by spin pumping into a two-dimensional electron gas

Yama, Masaki*; Matsuo, Mamoru; Kato, Takeo*

Physical Review B, 108(14), p.144430_1 - 144430_15, 2023/10

https://doi.org/10.1103/PhysRevB.108.144430

Validation practices of multi-physics core performance analysis in an advanced reactor design study

Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08

https://dx.doi.org/10.13182/NURETH20-40118

An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.