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Journal Articles

Relative oxygen potential measurements of (U,Pu)O$$_{2}$$ with Pu = 0.45 and 0.68 and related defect formation energy

Hiroka, Shun; Matsumoto, Taku; Sunaoshi, Takeo*; Hino, Tetsushi*

Journal of Nuclear Materials, 558, p.153375_1 - 153375_8, 2022/01

no abstracts in English

Journal Articles

Development of fuel performance analysis code, BISON for MOX, named Okami; Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

Ozawa, Takayuki; Hiroka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*

Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09

 Times Cited Count:0 Percentile:0.01(Materials Science, Multidisciplinary)

To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.

Journal Articles

Oxygen potential measurement of (U,Pu,Am)O$$_{2 pm x}$$ and (U,Pu,Am,Np)O$$_{2 pm x}$$

Hiroka, Shun; Matsumoto, Taku; Kato, Masato; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*

Journal of Nuclear Materials, 542, p.152424_1 - 152424_9, 2020/12

 Times Cited Count:1 Percentile:39.17(Materials Science, Multidisciplinary)

The measurement of oxygen potential was conducted at 1,673, 1,773, and 1,873 K for (U$$_{0.623}$$Pu$$_{0.350}$$Am$$_{0.027}$$)O$$_{2}$$ and at 1,873 and 1,923 K for (U$$_{0.553}$$Pu$$_{0.285}$$Am$$_{0.015}$$Np$$_{0.147}$$)O$$_{2}$$ by using a thermo-gravimeter and an oxygen sensor. Am inclusion in terms of substituting the U significantly increased the oxygen potential. Similarly, the inclusion of Np as a substitute for U increased the oxygen potential; however, the effect was not as large as that with the Pu or Am addition at the same rate. The results were analyzed via defect chemistry and certain defect formations were suggested in the reducing region and the near-stoichiometric region by plotting the relationship between PO$$_{2}$$ and the deviation from the stoichiometry. The equilibrium constants of the defect reactions were arranged to reproduce the experiment such that Am/Np contents were included in the entropy with coefficients fitting the experimental data.

Journal Articles

Effect of O/M ratio on sintering behavior of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2-x}$$

Nakamichi, Shinya; Hiroka, Shun; Kato, Masato; Sunaoshi, Takeo*; Nelson, A. T.*; McClellan, K. J.*

Journal of Nuclear Materials, 535, p.152188_1 - 152188_8, 2020/07

 Times Cited Count:3 Percentile:76.41(Materials Science, Multidisciplinary)

Oxygen-to-metal ratio (O/M) of uranium and plutonium mixed oxide depends on its oxygen partial pressure. To attain the desirable microstructure and O/M ratio of sintered pellets, it is important to investigate the relation between the sintering behavior and the atmosphere of sintering process. In this study, sintering behavior of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2}$$ and (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{1.99}$$ in controlled po$$_{2}$$ atmosphere were investigated. It was found activation energy of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{1.99}$$ was higher than that of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2}$$. On the other hand, it was observed grain growth during sintering was suppressed in hypo-stoichiometric composition.

Journal Articles

Physical properties of non-stoichiometric (U,Pu)O$$_{2}$$

Watanabe, Masashi; Matsumoto, Taku; Hiroka, 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$$_{2}$$. 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$$_{2}$$ are presented. The data set would be available in development of a fuel performance code.

Journal Articles

Oxygen potential and self-irradiation effects on fuel temperature in Am-MOX

Ikusawa, Yoshihisa; Hiroka, Shun; Uno, Masayoshi*

2018 GIF Symposium Proceedings (Internet), p.321 - 327, 2020/05

Research and development of Minor actinides (MAs) bearing MOX fuel for fast reactor has been proceeding from the viewpoint of reducing radioactive waste. In order to develop, MA bearing MOX, it is indispensable to clarify the influence of MA addition on irradiation behavior. The addition of Americium (Am) to MOX affects vapor pressure and thermal conductivity, which are important properties from the perspective of evaluating fuel temperature. This is because vapor pressure affects fuel restructuring, and thermal conductivity affects fuel temperature distribution. Focusing on these physical properties, this study evaluates the influence of Am on fuel temperature using irradiation behavior analysis code to contribute to the development of MA-bearing MOX fuel. An increase in Am content decreases the thermal conductivity and increases the oxygen potential of oxide fuel. Because vapor pressure increases with increasing Am content, pore migration is accelerated, and the central void diameter increases with increasing Am content. As a result, after formation of the central void, the influence of Am content on the fuel center temperature is mild. Alpha particles generated by radioactive decay of transuranium elements cause lattice defects in the oxide fuel pellets. It is well known that this phenomenon, which is called self-irradiation, affects thermal conductivity. Since americium is the typical alpha radioactive nucleus, to evaluate fuel temperature of Am-MOX is necessary to take account of the influence of self-irradiation damage on thermal conductivity. Self-irradiation decreases thermal conductivity, and as the Am content increases, the rate of decrease in thermal conductivity is accelerated. Because it recovers with temperature rise, the decrease in thermal conductivity due to self-irradiation damage has very little effect on fuel center temperature. These results suggest that Am-MOX fuel could be irradiated under the same conditions as conventional MOX fuel.

Journal Articles

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

Hiroka, 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$$_{2-x}$$

Hiroka, Shun; Kato, Masato

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

 Times Cited Count:2 Percentile:33.17(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$$_{2-x}$$

Hiroka, 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; Hiroka, Shun; Akashi, Masatoshi

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

 Times Cited Count:5 Percentile:58.67(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; Hiroka, 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:22.25(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; Hiroka, 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

Oxygen thermochemistry of urania-rare earth system; UO$$_2$$-CeO$$_2$$

Hiroka, Shun; Murakami, Tatsutoshi; Nelson, A. T.*; McClellan, K. J.*

INL/EXT-14-33515, p.34 - 36, 2014/10

Collaborative study has been done on the properties of nuclear materials between the DOE and Japan and the oxygen potential of (U,Ce)O$$_2$$ was measured in this year. Experimental measurements of the oxygen potential were conducted on Ce=20% and 30% composition rates simulating Pu content in advanced MOX fuel in JAEA by gas equilibrium method where oxygen partial pressure in the atmosphere was controlled with mixing dry/wet Ar/H$$_2$$ gas. More than 100 data points were obtained in the O/M range of 1.945 $$sim$$ 2.000 at 1200$$^{circ}$$C, 1400$$^{circ}$$C and 1600$$^{circ}$$C. The experimental results were analyzed by the defect analysis and analytical equations were obtained to calculate O/M as functions of temperature and oxygen potential. From the comparison with that of (U,Pu)O$$_2$$, applicability of the same defect chemistry and S-style curve are common. Also, it is revealed that (U,Ce)O$$_2$$ requires evidently higher oxygen potential for the O/M.

Journal Articles

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

Kato, Masato; Hiroka, 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

Hiroka, 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; Hiroka, 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; Hiroka, 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

Hiroka, 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:37.15(Materials Science, Multidisciplinary)

Journal Articles

Oxidation and reduction behaviors of plutonium and uranium mixed oxide powders

Hiroka, 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; Hiroka, Shun; Akashi, Masatoshi; Komeno, Akira; Morimoto, Kyoichi

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

 Times Cited Count:1 Percentile:60.76

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