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Journal Articles

Modeling and simulation of redistribution of oxygen-to-metal ratio in MOX

Hirooka, Shun; Kato, Masato; Watanabe, Masashi

Transactions of the American Nuclear Society, 118, p.1624 - 1626, 2018/06

This study suggested the time development of oxygen-to-metal ratio (O/M) redistribution model with oxygen-related properties in MOX. Irradiation simulation including the suggested O/M redistribution and pore migration with vaporization-condensation model which bares density redistribution was demonstrated. The simulation results showed that O/M redistribution proceeded at lower temperature than density redistribution, which indicated that oxygen diffusion got influential at lower temperature than vaporization-condensation of MOX. Another find was that O/M redistribution was very slow at the surface because temperature kept low. However, near the surface (inside from the surface) where the temperature exceeded 1000 K, O/M redistribution was rather recognizable with oxygen flown from inner region to the near-surface. The results will be evaluated by comparison with post-irradiation examination data.

Journal Articles

Sound speeds in and mechanical properties of (U,Pu)O$$_{2-x}$$

Hirooka, Shun; Kato, Masato

Journal of Nuclear Science and Technology, 55(3), p.356 - 362, 2018/03

 Times Cited Count:1 Percentile:65.67(Nuclear Science & Technology)

The sound speeds of longitudinal and transverse waves in the uranium-plutonium mixed oxide (MOX) pellets were measured as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The effect of each parameter was well fitted by a linear function and the equations were obtained to calculate the sound speeds. Mechanical properties were evaluated with the sound speeds and the result of Young's modulus showed that porosity was the most important factor to decrease Young's modulus. Temperature dependence on Young's modulus was also evaluated with previously reported thermal expansion. Decrease of Young's modules with increasing temperature was in good agreement with available literature.

Journal Articles

Mechanical and thermal properties of (U,Pu)O$$_{2-x}$$

Hirooka, Shun; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 6 Pages, 2017/06

Young's modulus of MOX pellets was evaluated by measuring the sound velocities of longitudinal and transverse waves in the pellets as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The results showed that porosity was the most important factor that 20% of the porosity decreased Young's modulus by neatly 100 GPa while O/M and plutonium content could change the Young's modulus by ~20 GPa. From the measured sound velocities, temperature dependence on Young's modulus and specific heat capacity were calculated on the Debye model by leveraging the thermal expansion data. The temperature dependence that Young's modulus decreases with increasing temperature is in good agreement with literature data. The specific heat capacity also agrees with that of calculated value by Kopp's method, taken the Schottky term and the excited term into account.

Journal Articles

Oxygen potentials, oxygen diffusion coefficients and defect equilibria of nonstoichiometric (U,Pu)O$$_{2pm x}$$

Kato, Masato; Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Akashi, Masatoshi

Journal of Nuclear Materials, 487, p.424 - 432, 2017/04

 Times Cited Count:1 Percentile:75.88(Materials Science, Multidisciplinary)

Oxygen potential of (U,Pu)O$$_{2pm x}$$ was evaluated based on defect chemistry using an updated experimental data set. The relationship between oxygen partial pressure and deviation $$x$$ in (U,Pu)O$$_{2pm x}$$ was analyzed, and equilibrium constants of defect formation were determined as functions of Pu content and temperature. Brouwer's diagrams were constructed using the determined equilibrium constants, and a relational equation to determine O/M ratio was derived as functions of O/M ratio, Pu content and temperature. In addition, relationship between oxygen potential and oxygen diffusion coefficients were described.

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

Sintering behavior of (U,Ce)O$$_{2}$$ and (U,Pu)O$$_{2}$$

Nakamichi, Shinya; Hirooka, Shun; Sunaoshi, Takeo*; Kato, Masato; Nelson, A.*; McClellan, K.*

Transactions of the American Nuclear Society, 113(1), p.617 - 618, 2015/10

Cerium dioxide has been used as a surrogate material for plutonium dioxide. Dorr et al reported the use of hyper-stoichiometric conditions causes the start of shrinkage of (U,Ce)O$$_{2}$$ at low temperature compared with the sintering in reducing atmosphere. However, the precise stoichiometry of the samples investigated was not controlled or otherwise monitored, preventing any quantitative conclusions regarding the similarities or differences between (U,Ce)O$$_{2}$$ and (U,Pu)O$$_{2}$$. The motivation for the present work is therefore to compare the sintering behavior of MOX and the (U,Ce)O$$_{2}$$ MOX surrogates under controlled atmospheres to assess the role of oxygen defects on densification in both systems.

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

Development and verification of the thermal behavior analysis code for MA containing MOX fuels

Ikusawa, Yoshihisa; Ozawa, Takayuki; Hirooka, Shun; Maeda, Koji; Kato, Masato; Maeda, Seiichiro

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 6 Pages, 2014/07

In order to develop MA contained MOX (MA-MOX) fuel design method, the analysis models to predict irradiation behavior of MA-MOX fuel have to be developed and the accuracy of irradiation behavior analysis code should be evaluated with the result of post-irradiation examinations (PIEs) for MA-MOX fuels. In this study, we developed the computer module "TRANSIT" to compute thermal properties of MA-MOX fuel. TRANSIT can give thermal conductivity, melting temperature and vapor pressures of MA-MOX. By using this module, we improved the thermal behavior analysis code "DIRAD" and developed DIRAD-TRANSIT code system to compute the irradiation behavior of MA-MOX fuel. This system was verified with the results of PIEs for the conventional MOX fuels and the MA-MOX fuels irradiated in the experimental fast reactor "JOYO". As the result of the verification, it can be mentioned that the DIRAD-TRANSIT system would precisely predict the fuel thermal behavior, i.e. fuel temperature and fuel restructuring, for oxide fuels containing several percent minor actinides.

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

Oxidation and reduction behaviors of plutonium and uranium mixed oxide powders

Hirooka, Shun; Kato, Masato; Tamura, Tetsuya*; Nelson, A. T.*; McClellan, K. J.*; Suzuki, Kiichi

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 8 Pages, 2013/03

As research and development activities for MOX fuel pellet production, oxidation and reduction behaviors of MOX powders were investigated by thermogravimetry and X-ray diffraction measurements. It was observed that the oxidation limit decreased with oxidizing temperature and Pu content. The MOX powders showed a two-step oxidation and kinetic stability under non-stoichiometry. The oxidation rates were evaluated from the isothermal oxidation tests. It was found that the reduction temperature of M$$_{4}$$O$$_{9}$$ + M$$_{3}$$O$$_{8}$$ was higher than that of M$$_{4}$$O$$_{9}$$. This indicated that the reduction of M$$_{4}$$O$$_{9}$$ was prevented by the existence of M$$_{3}$$O$$_{8}$$. Activation energy of the reduction was derived from the non-isothermal reduction tests. The data are expected to contribute to establishing a control technique for O/M ratio during MOX powder storage and pellet production.

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

Oxygen potentials of PuO$$_{2-x}$$

Komeno, Akira; Kato, Masato; Hirooka, Shun; Sunaoshi, Takeo*

Materials Research Society Symposium Proceedings, Vol.1444, p.85 - 89, 2012/09

 Times Cited Count:4 Percentile:6.54

Journal Articles

Oxide fuel fabrication technology development of the FaCT project, 3; Analysis of sintering behavior for MOX pellet production

Hirooka, Shun; Kato, Masato; Takeuchi, Kentaro; Takano, Tatsuo

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12

In this work, the shrinkage behavior and O/M change of MOX pellet during sintering process were investigated with dilatometer and thermo-gravimeter, and equations to analyze the sintering behavior were derived. The derived equations represented the change of density and O/M ratio of MOX pellet during heat treatment as functions of heat treatment conditions such as heating rate, holding temperature and $$P$$$$_{rm H2}$$/$$P$$$$_{rm H2O}$$ ratio in an atmosphere. They contribute the development of advanced pellet production process and would accurately control density and O/M ratio of MOX pellets.

Oral presentation

Evaluation of reduction rate on (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{2+x}$$ raw powder

Hirooka, Shun; Kato, Masato; Morimoto, Kyoichi; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Development of evaluation technique of O/M in sintering process of MOX pellets

Hirooka, Shun; Kato, Masato; Takeuchi, Kentaro; Sunaoshi, Takeo*

no journal, , 

no abstracts in English

Oral presentation

Oral presentation

Effect of oxygen to metal ratio on properties of UO$$_2$$-cladding simulated debris

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

no journal, , 

no abstracts in English

45 (Records 1-20 displayed on this page)