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.

Physical properties of non-stoichiometric (U,Pu)O

Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Morimoto, Kyoichi; Kato, Masato

2018 GIF Symposium Proceedings (Internet), p.315 - 320, 2020/05

Recently, a research group studying at Plutonium Fuel Development Facility (PFDF) in Japan Atomic Energy Agency has systematically measured vast amounts of physical properties in the non-stoichiometric (U, Pu)O. Lattice parameter, elastic modulus, thermal expansion, oxygen potential, oxygen chemical diffusion coefficient and thermal conductivity were successfully measured as function of Pu content, O/M ratio and temperature, and the effects of Pu content and O/M ratio on their physical properties were evaluated. In this work, those experimental data are reviewed, and latest experimental data set on the non-stoichiometric (U, Pu)O are presented. The data set would be available in development of a fuel performance code.

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

The Influences of Pu and Zr on the melting temperatures of the UO-PuO-ZrO pseudo-ternary system

Morimoto, Kyoichi; Hirooka, Shun; Akashi, Masatoshi; Watanabe, Masashi; Sugata, Hiromasa*

Journal of Nuclear Science and Technology, 52(10), p.1247 - 1252, 2015/10

https://doi.org/10.1080/00223131.2015.1047807

As a part of decommissioning plan of the damaged reactors at Fukushima Daiichi Nuclear Power Plant, some strategies for removing of debris from the reactors are discussed. In these considerations, it is necessary to predict a melt progression during the severe accident based on theoretical evidences. Melting temperature is one of the most important thermal characteristics to analyse a melt progression during the severe accident. In this study, the melting temperatures of specimens of U, Pu and Zr mixed oxide prepared as simulated debris were measured by the thermal arrest technique. From the results of this measurement, the influences of Pu and Zr contents on the melting temperature of the simulated debris were evaluated.

Property measurements and inner state estimation of simulated fuel debris

Hirooka, Shun; Kato, Masato; Morimoto, Kyoichi; Washiya, Tadahiro

Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 8 Pages, 2014/08

Since the severe accident at Fukushima Daiichi Nuclear Power Station, technologies to remove fuel debris from the damaged core have been developed. However, many subjects such as how to access to the core, cut the fuel debris, control criticality safety, estimate fissile materials, store removed debris and so on are still in existence. Purpose of this work is to evaluate the fuel debris properties by using analysis of simulated fuel debris and to estimate the inner state such as temperature profile and density profile which depend on elapsed time after the accident. The reported properties such as melting temperature, thermal conductivity and thermal expansion were obtained by the simulated fuel debris manufactured from UO and zircaloy.

Melting temperatures of the ZrO-MOX system

Uchida, Teppei; Hirooka, Shun; Sugata, Hiromasa*; Shibata, Katsuya*; Sato, Daisuke*; Kato, Masato; Morimoto, Kyoichi

Proceedings of International Nuclear Fuel Cycle Conference; Nuclear Energy at a Crossroads (GLOBAL 2013) (CD-ROM), p.1549 - 1553, 2013/09

Melting temperature and thermal conductivities of corium prepared from UO and zircalloy-2

Kato, Masato; Uchida, Teppei; Hirooka, Shun; Akashi, Masatoshi; Komeno, Akira; Morimoto, Kyoichi

Materials Research Society Symposium Proceedings, Vol.1444, p.91 - 96, 2012/09

https://doi.org/10.1557/opl.2012.987

Thermal recovery evaluation of thermal conductivity in a self-irradiated MOX pellet

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*

Proceedings of Plutonium Futures; The Science 2010 (CD-ROM), p.339 - 340, 2010/09

Nuclear fuel pellets are stored before loading into a reactor. In some cases, the fuel pellets are left for several years. When uranium-plutonium mixed oxide (MOX) fuel pellets are stored for a long time, lattice defects induced by self-irradiation ( decay) accumulate and these defects affect physical properties of the pellets, i.e. lattice parameter, electrical resistivity and thermal conductivity. The thermal conductivity of fuel pellets is one of the most important properties for fuel design and performance analyses; it is known to decrease due to the defects induced by self-irradiation, but it can be recovered by heating the pellets. In this study, the recovery behavior of thermal conductivity of a MOX fuel pellet stored for long time was investigated as a function of time and temperature, in order to make it easy to analyze the thermal performance of fuel pellets.

Phase separation behaviour of (UPu)O (1.92x2.00) based fuels containing actinides and/or lanthanides

Komeno, Akira; Kato, Masato; Uno, Hiroki*; Takeuchi, Kentaro; Morimoto, Kyoichi; Kashimura, Motoaki

IOP Conference Series; Materials Science and Engineering, 9, p.012016_1 - 012016_7, 2010/05

https://doi.org/10.1088/1757-899X/9/1/012016

It is expected that the important data for design of fast reactor fuel can be provided by evaluating the relationship between fuel composition and phase separation with reported and new measurement data. According to evaluation with reported data and new measured data, a relationship between fuel composition and phase separation temperature of MOX fuel was indicated. Higher minor actinides-containing MOX had a lower phase separation temperature at O/M ratio region from 1.92 to 1.96.

Oxygen chemical diffusion in hypo-stoichiometric MOX

Kato, Masato; Morimoto, Kyoichi; Tamura, Tetsuya*; Sunaoshi, Takeo*; Konashi, Kenji*; Aono, Shigenori; Kashimura, Motoaki

Journal of Nuclear Materials, 389(3), p.416 - 419, 2009/06

https://doi.org/10.1016/j.jnucmat.2009.02.018

Plutonium and uranium mixed oxide (MOX) has been developed to use as a core fuel of the fast reactor. The oxygen to metal ratio (O/M) of the MOX fuel is an important parameter to control the FCCI. The oxygen potential and the oxygen diffusion coefficient of the MOX are essential data to understand the oxygen behaviour in MOX. The oxygen potentials of the MOX were measured with accuracy as a function of O/M and temperatures in the previous work. In this work the oxygen chemical diffusion coefficient in (PuU)O and (PuU)O were investigated using thermo gravimetric technique. The kinetics of the reduction processes of (PuU)O and (PuU)O were measured by TG-DTA method. The oxygen chemical diffusion coefficients have been estimated from the reduction curves. It was concluded that the oxygen chemical diffusion coefficient in (PuU)O is a smaller than that of (PuU)O.