Investigating eutectic behavior and material relocation in BC-stainless steel composites using the improved MPS method

Ahmed, Z.*; Wu, S.*; Sharma, A.*; Kumar, R.*; Yamano, Hidemasa; Pellegrini, M.*; Yokoyama, Ryo*; Okamoto, Koji*

International Journal of Heat and Mass Transfer, 250, p.127343_1 - 127343_17, 2025/11

https://doi.org/10.1016/j.ijheatmasstransfer.2025.127343

Introduce of pin-indentation deformation model into fuel pin bundle deformation analysis code "BAMBOO"

Uwaba, Tomoyuki; Ito, Masahiro*; Ishitani, Ikuo*

JAEA-Technology 2025-002, 25 Pages, 2025/06

JAEA-Technology-2025-002.pdf:2.06MB

In spacer wire-type fast reactor fuel assemblies, when the bundle-duct interaction (BDI) becomes severe, contact loads applied to claddings via wires cause oval-deformation of the cladding cross-sections. The cladding oval-deformation may become significant especially in the pins at the outermost periphery of the bundle because they are subjected to a large load from the duct via their wires. In this case the cladding oval-deformation occurs in such a manner that the wire is partially sinking into the cladding, and as a result the cladding is locally dented. We developed a model to simulate such cladding local deformation and integrated the model into the BDI analysis code "BAMBOO". In the out-of-pile bundle compression tests, the gaps between pins and the duct became smaller than the wire diameter when the bundle was compressed despite the existence of a wires between the bundle outermost pins and the duct. The bundle compression test analyses by BAMBOO equipped with the model showed that the local deformation of claddings of outermost pins reproduced the similar trend to the compression test results.

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.

New filter concept for removal of fine particle generated in high level radioactive solution

Takahatake, Yoko; Watanabe, So; Watanabe, Masayuki; Sano, Yuichi; Takeuchi, Masayuki

Progress in Nuclear Science and Technology (Internet), 7, p.195 - 198, 2025/05

https://doi.org/10.15669/pnst.7.195

Extraction chromatgraphy technology for trivalent minor actinide (MA(III) ; Am(III) and Cm(III)) recovery from the solution generated by an extraction process in reprocessing of spent nuclear fuel has been developed. A fine particle is generated in the solution. The fine particle must be removed before MA recovery operation, because that leads clogging of the extraction chlomatography column. In order to prevent clogging the column, filtration system utilizing porous silica beads packed column has been designed. In this study, a fine particle trapping system was developed and particle removal performance of the system was experimentally evaluated using alumina particles as simulated fine particle. Column experiments revealed that the fine particle with the particle size from 0.12 to 15 m is cause of clogging of the filtration column. Since simulated fine particles were trapped on filtration experiments, a filtration system using the porous silica beads column is practical,

Density, surface tension, and viscosity of molten Ni-based superalloys using the maximum bubble pressure and oscillating crucible methods

Nishi, Tsuyoshi*; Matsumoto, Saori*; Yamano, Hidemasa; Hayashi, Kiichiro*; Endo, Rie*; Bell, M. R.*; Neubert, L.*; Volkova, O.*

Steel Research International, 96(5), p.2300766_1 - 2300766_6, 2025/05

https://doi.org/10.1002/srin.202300766

The density of Ni-based superalloys is measured using the maximum bubble pressure (MBP) method. The viscosity is evaluated using the oscillating crucible method. The surface tension is simultaneously measured using the MBP method.

Federated learning of creep rupture time and high temperature tensile strength prediction models

Sakurai, Junya*; Torigata, Keisuke*; Matsunaga, Manabu*; Takanashi, Naoto*; Hibino, Shinya*; Kizu, Kenichi*; Morita, Akira*; Inomoto, Masahiro*; Shimohata, Nobuaki*; Toyota, Kodai; et al.

Tetsu To Hagane, 111(5), p.246 - 262, 2025/04

https://doi.org/10.2355/tetsutohagane.TETSU-2024-124

Radiation heating effects on BC-SS eutectic melting and its relocation behaviour

Ahmed, Z.*; Sharma, A. K.*; Pellegrini, M.*; Yamano, Hidemasa; Okamoto, Koji*

Arabian Journal for Science and Engineering, 50(5), p.3361 - 3371, 2025/03

https://doi.org/10.1007/s13369-024-09412-9

Trading cards of nuclear technology; A Science communication tool developed for gaining literacy to read nuclear policy documents

Fukaya, Yuji; Yan, X.; Matsuo, Yuji*; Hayashi, Reona*

Japanese Journal of Science Communication (Internet), (36), p.45 - 56, 2025/03

https://doi.org/10.14943/112887

From the 1956 "Long-Term Program on the Development and Utilization of Atomic Energy" to the present-day "Strategic Energy Plan", documents related to relevant committees and nuclear power policies have been accumulated, and many of them may be red on the Internet. Moreover, there are diverse explanations and opinions based on these policy documents, as well as many papers on research and development that reflect them. On the other hand, 1956 was a time when Japan was transitioning from post-war reconstruction to high economic growth and was clearly aware of the possibility of war in the form of the Cold War. Under such circumstance, unless one is knowledgeable to appreciate, to some extent, the context and the source from which the benefits of a particular technology claimed in a particular policy document originate, it would be difficult to predict the actual effects of the technology when it is implemented in society. This might also mislead the general public, making it difficult to reflect the public opinion correctly in new policies. To address these dilemmas, a trading card game has been developed in which the players participate in the role-playing process as a proponent of a particular nuclear technology. The process consists of four phases, namely technology selection, understanding the current situation, policy making, and justification of the selected technology under the policy direction. The game could allow the players to predict the intent of the technology and policy communicator.

Effect of Am addition on oxygen potential in (UPuAm)O

Yokoyama, Keisuke; Watanabe, Masashi; Usui, Akane; Seki, Takayuki*; Onishi, Takashi; Kato, Masato

Nuclear Materials and Energy (Internet), 42, p.101908_1 - 101908_6, 2025/03

https://doi.org/10.1016/j.nme.2025.101908

Oxygen potential of high Am content MOX, (UPuAm)O, was measured at 1273 K, 1473 K, 1573 K, and 1623 K. by gas equilibrium method using thermogravimeter. Comparing the measured data with the literature data, it was found that the addition of 15% Am increases the oxygen potential of (U, Pu)O by 100-150 kJ/mol for the same Pu content and O/M ratio. The proportion of cations in the stoichiometric composition was determined as (UUPuAm)O, assuming the presence of Am and partial oxidation of U to U. The relationship between oxygen partial pressure and deviation x from stoichiometry in (UPuAm)O was analyzed by defect chemistry model. The equation to represent the O/M ratio was derived as a function of temperature and oxygen partial pressure. A part of this study includes the results of MEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0219214921.

None

Maruyama, Shuhei

Robutsuri No Kenkyu (Internet), (78), 7 Pages, 2025/03

no abstracts in English