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Journal Articles

The Effects of plutonium content and self-irradiation on thermal conductivity of mixed oxide fuel

Ikusawa, Yoshihisa; Morimoto, Kyoichi; Kato, Masato; Saito, Kosuke; Uno, Masayoshi*

Nuclear Technology, 205(3), p.474 - 485, 2019/03

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

This study evaluated the effects of plutonium content and self-irradiation on the thermal conductivity of mixed-oxide (MOX) fuel. Samples of UO$$_{2}$$ fuel and various MOX fuels were tested. The MOX fuels had a range of plutonium contents, and some samples were stored for 20 years. The thermal conductivity of these samples was determined from thermal diffusivity measurements taken via laser flash analysis. Although the thermal conductivity decreased with increasing plutonium content, this effect was slight. The effect of self-irradiation was investigated using the stored samples. The reduction in thermal conductivity caused by self-irradiation depended on the plutonium content, its isotopic composition, and storage time. The reduction in thermal conductivity over 20 years' storage can be predicted from the change of lattice parameter. In addition, the decrease in thermal conductivity caused by self-irradiation was recovered with heat treatment, and recovered almost completely at temperatures over 1200 K. From these evaluation results, we formulated an equation for thermal conductivity that is based on the classical phonon-transport model. This equation can predict the thermal conductivity of MOX fuel thermal conductivity by accounting for the influences of plutonium content and self-irradiation.

Journal Articles

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.

Journal Articles

Oxygen potential measurement of (Pu$$_{0.928}$$Am$$_{0.072}$$)O$$_{2-x}$$ at high temperatures

Matsumoto, Taku; Arima, Tatsumi*; Inagaki, Yaohiro*; Idemitsu, Kazuya*; Kato, Masato; Morimoto, Kyoichi; Sunaoshi, Takeo*

Journal of Nuclear Science and Technology, 52(10), p.1296 - 1302, 2015/10

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The oxygen potentials of (Pu$$_{0.928}$$Am$$_{0.072}$$)O$$_{2-x}$$ were measured at 1873K, 1773K and 1473K by gas equilibrium method. It was shown that following the reduction of Am at the O/M ratio above 1.96, Pu was reduced at the O/M ratio below 1.96.

Journal Articles

The Influences of Pu and Zr on the melting temperatures of the UO$$_{2}$$-PuO$$_{2}$$-ZrO$$_{2}$$ 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

 Times Cited Count:2 Percentile:71.62(Nuclear Science & Technology)

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.

Journal Articles

Development of science-based fuel technologies for Japan's Sodium-Cooled Fast Reactors

Kato, Masato; Hirooka, Shun; Ikusawa, Yoshihisa; Takeuchi, Kentaro; Akashi, Masatoshi; Maeda, Koji; Watanabe, Masashi; Komeno, Akira; Morimoto, Kyoichi

Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 12 Pages, 2014/08

Uranium and plutonium mixed oxide (MOX) fuel has been developed for Japan sodium-cooled fast reactors. Science based fuel technologies have been developed to analyse behaviours of MOX pellets in the sintering process and irradiation conditions. The technologies can provide appropriate sintering conditions, irradiation behaviour analysis results and so on using mechanistic models which are derived based on theoretical equations to represent various properties.

Journal Articles

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$$_2$$ and zircaloy.

Journal Articles

Thermal diffusivity measurement of (U, Pu)O$$_{2-x}$$ at high temperatures up to 2190 K

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

Journal of Nuclear Materials, 443(1-3), p.286 - 290, 2013/11

 Times Cited Count:4 Percentile:57.8(Materials Science, Multidisciplinary)

In this study, measurement was conducted for the sliced MOX pellets containing 30% of Pu prepared by a conventional powder metallurgy technology. Oxygen-to-metal (O/M) ratios of the samples were adjusted in the range from 1.92 to 2.00. The thermal diffusivities of these samples were measured at temperature up to 2150 K with the laser flash method. Thermal diffusivities of the near-stoichiometric samples obtained in the cooling process were greatly lower than those in the heating process unlike measurement below 1770 K. On the other hand, they were almost identical for the sample of 1.946 in O/M. It was also shown that thermal diffusivity decreased with the temperature but increased with the O/M.

Journal Articles

Melting temperatures of the ZrO$$_{2}$$-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

Journal Articles

Effect of oxygen-to-metal ratio on properties of corium prepared from UO$$_{2}$$ and zircaloy-2

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

Journal of Nuclear Materials, 437(1-3), p.130 - 134, 2013/06

 Times Cited Count:4 Percentile:57.8(Materials Science, Multidisciplinary)

Journal Articles

Melting temperature and thermal conductivities of corium prepared from UO$$_{2}$$ 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

 Times Cited Count:1 Percentile:34.17

Journal Articles

Thermal expansion of corium prepared from UO$$_2$$ and zircalloy-2

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

Materials Research Society Symposium Proceedings, Vol.1444, p.97 - 101, 2012/09

 Times Cited Count:0 Percentile:100

Journal Articles

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 ($$alpha$$ 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.

Journal Articles

Burn-up effect on MOX fuel thermal conductivity

Ikusawa, Yoshihisa; Morimoto, Kyoichi; Ozawa, Takayuki; Kato, Masato

Proceedings of Plutonium Futures; The Science 2010 (CD-ROM), p.341 - 342, 2010/09

Thermal conductivity of oxide fuel is important for fuel design and performance analyses. Uranium dioxide and uranium-plutonium mixed oxide (MOX) are used as fuels in light water reactors (LWRs), and the thermal conductivities of these oxide fuels have been measured in various laboratories. In a review of oxide fuel properties, it was reported that the thermal conductivity of oxide fuel would decrease with burn-up increase. In this study, burn-up effect on MOX fuel thermal conductivity was discussed.

Journal Articles

Phase separation behaviour of (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{2-x}$$ (1.92$$<$$x$$<$$2.00) 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

 Times Cited Count:8 Percentile:4.42

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.

Journal Articles

Experimental evaluation of Am-and Np-bearing mixed-oxide fuel properties

Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki

Proceedings of 10th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation (CD-ROM), p.201 - 209, 2010/00

Japan Atomic Energy Agency has developed homogeneous MOX fuel containing minor actinide (MA) elements such as Np and Am. To measure physical properties of the fuel is essential for its development, because their data are needed to evaluate irradiation behavior. In this report, the physical properties, melting temperature, thermal conductivity, lattice parameter, oxygen potential and phase separation behavior, were reviewed, and effect of MA content was discussed.

Journal Articles

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

 Times Cited Count:9 Percentile:40.26(Materials Science, Multidisciplinary)

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 (Pu$$_{0.2}$$U$$_{0.8}$$)O$$_{2-x}$$ and (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2-x}$$ were investigated using thermo gravimetric technique. The kinetics of the reduction processes of (Pu$$_{0.2}$$U$$_{0.8}$$)O$$_{2-x}$$ and (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2-x}$$ 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 (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2-x}$$ is a smaller than that of (Pu$$_{0.2}$$U$$_{0.8}$$)O$$_{2-x}$$.

Journal Articles

The Phase state at high temperatures in the MOX-SiO$$_{2}$$ system

Nakamichi, Shinya; Kato, Masato; Sunaoshi, Takeo*; Uchida, Teppei; Morimoto, Kyoichi; Kashimura, Motoaki; Kihara, Yoshiyuki

Journal of Nuclear Materials, 389(1), p.191 - 196, 2009/05

 Times Cited Count:2 Percentile:79.69(Materials Science, Multidisciplinary)

Japan Atomic Energy Agency researchers have developed mixed oxide (MOX) fuels containing minor actinides (MA). These fuels were irradiated for ten minutes in the FBR Joyo in some short-term irradiation tests. The Si-condensed phases were observed at the center of the pellets in the post irradiation examination. Si impurities came to be mixed into the raw materials in the ball milling process, because Si rubber was used as the lining of the milling pot. Content of Si in the pellets was within the specification of the fuel. It is important to investigate the Si state in MOX at high temperatures like the reactor operating temperature of the fuel to evaluate irradiation behavior. In the present work, MOX specimens with mixed SiO$$_{2}$$ impurity were prepared. The ratio of MOX to SiO$$_{2}$$ was controlled at a mol fraction of 3 to 1. The specimens were first heated at 1973K in atmospheres of three different oxygen partial pressures to adjust the O/M ratio. Then these specimens were sealed in a tungsten capsule, and heated at 2273K or 2673K. Compounds consisting of Pu and Si were observed at grain boundaries of the MOX matrix in specimens after heat treatment. These compounds were not observed in grain interior and MOX matrix was not affected significantly by Si impurity. These compounds tended to form in specimens with low O/M ratio and in specimens heated at higher temperatures.

Journal Articles

Effect of oxygen-to-metal ratio on melting temperature of uranium and plutonium mixed oxide fuel for fast reactor

Kato, Masato; Morimoto, Kyoichi; Nakamichi, Shinya; Sugata, Hiromasa*; Konashi, Kenji*; Kashimura, Motoaki; Abe, Tomoyuki

Nippon Genshiryoku Gakkai Wabun Rombunshi, 7(4), p.420 - 428, 2008/12

The melting temperatures of MOX for fast reactor fuel were investigated as functions of Pu content, Am content and oxygen-to-metal (O/M) ratio using thermal arrest technique. Rhenium inner was used for the measurement to prevent the reaction between the sample and capsule materials. The solidus temperatures decreased with increasing Pu and Am content and increased with decreasing O/M ratio. It is considered that the maximum temperature in U-Pu-O system varies in hypostoichiometric composition region. The melting temperatures were evaluated by ideal solid solution model in UO$$_{2}$$-PuO$$_{2}$$-AmO$$_{2}$$-PuO$$_{1.7}$$ system, and the model was derived for calculating solidus and liquidus temperature. The derived model reproduced the experimental data with $$pm$$25 K.

Journal Articles

Thermal conductivities of (U,Pu,Am)O$$_{2}$$ solid solutions

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki; Abe, Tomoyuki

Journal of Alloys and Compounds, 452(1), p.54 - 60, 2008/03

 Times Cited Count:23 Percentile:21.96(Chemistry, Physical)

Plutonium and uranium mixed oxide (MOX) fuel with high Pu content have been developed as a fuel of fast reactor (FR). As the storage time of Pu raw material between reprocessing and fabrication increases, americium content of the fabricated MOX fuel increases up to a few percent. In this work, the thermal conductivity of MOX fuel containing Am was investigated as a part of clarifying the effect of Am content on thermal physical properties. The pellets of (Am$$_{0.007}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$, (Am$$_{0.02}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ and (Am$$_{0.03}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ were prepared. The oxygen to metal ratio (O/M ratio) of sintered pellet was adjusted to 2.00. The thermal diffusivity measurement was carried out in the range of temperature from 900 K to 1700 K by the laser flash method, and thermal conductivity of these pellets was evaluated. The heat capacity for evaluating thermal conductivity was derived from heat capacity of UO$$_{2}$$, PuO$$_{2}$$ and AmO$$_{2}$$ by using the Kopp-Neumann rule.

Journal Articles

Solidus and liquidus of plutonium and uranium mixed oxide

Kato, Masato; Morimoto, Kyoichi; Sugata, Hiromasa*; Konashi, Kenji*; Kashimura, Motoaki; Abe, Tomoyuki

Journal of Alloys and Compounds, 452(1), p.48 - 53, 2008/03

 Times Cited Count:20 Percentile:25.24(Chemistry, Physical)

Plutonium and uranium mixed oxide has been developed as a fuel of a fast reactor. The maximum temperature of the fuel pellet is limited within a design criterion to prevent fuel melting. So, the melting points of the mixed oxide have been investigated since the development of fast reactor started. However the measured data are limited. In this work, the melting points of (U1-yPuy)O$$_{2-x}$$ (y: 0, 0.12, 0.2, 0.3, 0.4) were measured by the thermal arrest method. The evaluated melting point of this study underestimates in case of MOX with high Pu contents of 30% and 40%. The solidus of UO$$_{2}$$, (Pu$$_{0.12}$$U$$_{0.88}$$)O$$_{2.00}$$ and (Pu$$_{0.2}$$U$$_{0.8}$$)O$$_{2.00}$$ were determined to be 3128K, 3077K and 3052K, respectively. The solidus temperature of hypostoichiometric MOX slightly increased with decreasing O/M.

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