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.

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka

Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01

https://doi.org/10.1016/j.anucene.2023.110144

Computer code analysis of irradiation performance of an annular mixed oxide fuel element

Yokoyama, Keisuke; Uwaba, Tomoyuki

Journal of Nuclear Science and Technology, 60(10), p.1219 - 1227, 2023/10

https://doi.org/10.1080/00223131.2023.2188272

no abstracts in English

Oxygen potential of neodymium-doped UPuAmO uranium-plutonium-americium mixed oxides at 1573, 1773, and 1873 K

Vauchy, R.; Sunaoshi, Takeo*; Hirooka, Shun; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Materials, 580, p.154416_1 - 154416_11, 2023/07

https://doi.org/10.1016/j.jnucmat.2023.154416

Data processing and visualization of X-ray computed tomography images of a JOYO MK-III fuel assembly

Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji

Journal of Nuclear Science and Technology, 60(6), p.715 - 723, 2023/06

https://doi.org/10.1080/00223131.2022.2138603

Development of "MOX weighing and Ball-mill blending" based on experience in operation and maintenance of MOX fuel manufacturing equipment

Kawasaki, Kohei; Ono, Takanori; Shibanuma, Kimikazu; Goto, Kenta; Aita, Takahiro; Okamoto, Naritoshi; Shinada, Kenta; Ichige, Hidekazu; Takase, Tatsuya; Osaka, Yuki; et al.

JAEA-Technology 2022-031, 91 Pages, 2023/02

JAEA-Technology-2022-031.pdf:6.57MB

The document for back-end policy opened to the public in 2018 by Japan Atomic Energy Agency (hereafter, JAEA) states the decommissioning of facilities of Nuclear Fuel Cycle Engineering Laboratories and JAEA have started gathering up nuclear fuel material of the facilities into Plutonium Fuel Production Facilities (hereafter, PFPF) in order to put it long-term, stable and safe storage. Because we planned to manufacture scrap assemblies almost same with Monju fuel assembly using unsealed plutonium-uranium mixed-oxide (hereafter, MOX) powder held in PFPF and transfer them to storage facilities as part of this "concentration" task of nuclear fuel material, we obtained permission to change the use of nuclear fuel material in response to the new regulatory Requirements in Japan for that. The amount of plutonium (which is neither sintered pellets nor in a lidded powder-transport container) that could be handled in the pellet-manufacturing process was limited to 50 kg Pu or less in order to decrease the facility risk in this manufacture. Therefore, we developed and installed the "MOX weighing and blending equipment" corresponding with small batch sizes that functioned in a starting process and the equipment would decrease handling amounts of plutonium on its downstream processes. The failure data based on our operation and maintenance experiences of MOX fuel production facilities was reflected in the design of the equipment to further improve reliability and maintainability in this development. The completed equipment started its operation using MOX powder in February 2022 and the design has been validated through this half-a-year operation. This report organizes the knowledge obtained through the development of the equipment, the evaluation of the design based on the half-a-year operation results and the issues in future equipment development.

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Journal of Nuclear Science and Technology, 13 Pages, 2023/00

https://doi.org/10.1080/00223131.2023.2243949

Materials science and fuel technologies of uranium and plutonium mixed oxide

Kato, Masato; Machida, Masahiko; Hirooka, Shun; Nakamichi, Shinya; Ikusawa, Yoshihisa; Nakamura, Hiroki; Kobayashi, Keita; Ozawa, Takayuki; Maeda, Koji; Sasaki, Shinji; et al.

Materials Science and Fuel Technologies of Uranium and Plutonium mixed Oxide, 171 Pages, 2022/10

https://doi.org/10.1201/9781003298205

Innovative and advanced nuclear reactors using plutonium fuel has been developed in each country. In order to develop a new nuclear fuel, irradiation tests are indispensable, and it is necessary to demonstrate the performance and safety of nuclear fuels. If we can develop a technology that accurately simulates irradiation behavior as a technology that complements the irradiation test, the cost, time, and labor involved in nuclear fuel research and development will be greatly reduced. And safety and reliability can be significantly improved through simulation of nuclear fuel irradiation behavior. In order to evaluate the performance of nuclear fuel, it is necessary to know the physical and chemical properties of the fuel at high temperatures. And it is indispensable to develop a behavior model that describes various phenomena that occur during irradiation. In previous research and development, empirical methods with fitting parameters have been used in many parts of model development. However, empirical techniques can give very different results in areas where there is no data. Therefore, the purpose of this study is to construct a scientific descriptive model that can extrapolate the basic characteristics of fuel to the composition and temperature, and to develop an irradiation behavior analysis code to which the model is applied.

Recent studies on fuel properties and irradiation behaviors of Am/Np-bearing MOX

Hirooka, Shun; Yokoyama, Keisuke; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04

Property studies on Am/Np-bearing MOX were carried out and how the properties influences on the irradiation behaviors was discussed. Both Am and Np inclusions increase the oxygen potential of MOX. Inter-diffusion coefficients obtained by using diffusion couple technique indicate that the inter-diffusion coefficient is larger in the order of U-Am, U-Pu and U-Np. Also, the inter-diffusion coefficients were evaluated to be larger at the O/M = 2 than those of O/M 2 by several orders. The increase of oxygen potential with Am/Np leads to higher vapor pressure of UO and the acceleration of the pore migration along temperature gradient during irradiation. The redistributions of actinide elements were also considered with the relationship of the pore migration and diffusion in solid state. Thus, the obtained inter-diffusion coefficients directly influence on the redistribution rate. The obtained properties were modelled and can be installed in a fuel irradiation simulation code.

Development of fission gas release model for MOX fuel pellets with treatment of heterogeneous microstructure

Tasaki, Yudai; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03

https://doi.org/10.1080/00223131.2021.1973924