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Journal Articles

Phase-field model for crystallization in alkali disilicate glasses; Li$$_2$$O-2SiO$$_2$$, Na$$_2$$O-2SiO$$_2$$ and K$$_2$$O-2SiO$$_2$$

Kawaguchi, Munemichi; Uno, Masayoshi*

Journal of the Ceramic Society of Japan, 128(10), p.832 - 838, 2020/10

 Times Cited Count:0 Percentile:100(Materials Science, Ceramics)

This study developed phase-field method (PFM) technique in oxide melt system by using a new mobility coefficient ($$L$$). The crystal growth rates ($$v_0$$) obtained by the PFM calculation with the constant $$L$$ were comparable to the thermodynamic driving force in normal growth model. The temperature dependence of the $$L$$ was determined from the experimental crystal growth rates and the $$v_0$$. Using the determined $$L$$, the crystal growth rates ($$v$$) in alkali disilicate glasses, Li$$_2$$O-2SiO$$_2$$, Na$$_2$$O-2SiO$$_2$$ and K$$_2$$O-2SiO$$_2$$ were simulated. The temperature dependence of the $$v$$ was qualitatively and quantitatively so similar that the PFM calculation results demonstrated the validity of the $$L$$. Especially, the $$v$$ obtained by the PFM calculation appeared the rapid increase just below the thermodynamic melting point ($$T_{rm m}$$) and the steep peak at around $$T_{rm m}$$-100 K. Additionally, as the temperature decreased, the $$v$$ apparently approached zero ms$$^-1$$, which is limited by the $$L$$ representing the interface jump process. Furthermore, we implemented the PFM calculation for the variation of the parameter $$B$$ in the $$L$$. As the $$B$$ increased from zero to two, the peak of the $$v$$ became steeper and the peak temperature of the $$v$$ shifted to the high temperature side. The parameters $$A$$ and $$B$$ in the $$L$$ increased exponentially and decreased linearly as the atomic number of the alkali metal increased due to the ionic potential, respectively. This calculation revealed that the $$A$$ and $$B$$ in the $$L$$ were close and reasonable for each other.

Journal Articles

Thermodynamic and thermophysical properties of the actinide nitrides

Uno, Masayoshi*; Nishi, Tsuyoshi*; Takano, Masahide

Comprehensive Nuclear Materials, 2nd Edition, Vol.7, p.202 - 231, 2020/08

On the thermodynamic and thermophysical properties of the actinide nitrides in Comprehensive Nuclear Materials published by Elsevier as the first edition in 2012, we have revised them by adding some brand-new data. The main topics added are the solid solubility of the actinide nitrides into the zirconium nitride matrix for transmutation fuel, the lattice expansion of actinide nitrides induced by self-irradiation damage, the influence of defects accumulation on thermal conductivity, and the thermal expansion in curium nitride lattice.

Journal Articles

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

Ikusawa, Yoshihisa; Hirooka, 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

A Study on sodium-concrete reaction in presence of internal heating

Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800$$^{circ}$$C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800$$^{circ}$$C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.

Journal Articles

Synergistic enhancement of the extraction and separation efficiencies of lanthanoid(III) ions by the formation of charged adducts in an ionic liquid

Okamura, Hiroyuki; Mizuno, Masayoshi*; Hirayama, Naoki*; Shimojo, Kojiro; Naganawa, Hirochika; Imura, Hisanori*

Industrial & Engineering Chemistry Research, 59(1), p.329 - 340, 2020/01

 Times Cited Count:0 Percentile:100(Engineering, Chemical)

The synergistic ionic-liquid extraction and extraction equilibrium of lanthanoid(III) (Ln(III)) ions have been investigated using 2-thenoyltrifluoroacetone (Htta) and trioctylphosphine oxide (TOPO) in the ionic liquid 1-butyl-3-methylimidazolium bis(trifluoromethanesulfonyl)imide ([C$$_{4}$$mim][Tf$$_{2}$$N]). The selective synergistic effect for heavier Ln(III) ions was found using a combination of Htta and TOPO in [C$$_{4}$$mim][Tf$$_{2}$$N], leading to enhanced separability among Ln(III) ions. The extracted Ln(III) species and the extraction constants in the Htta-TOPO system were determined by three-dimensional extraction equilibrium analysis. The selective synergism for heavier Ln(III) ions in the Htta-TOPO system was ascribed to the formation of hydrophobic, charged adducts, such as Ln(tta)$$_{2}$$(TOPO)$$_{3}$$$$^{+}$$ and Ln(tta)(TOPO)$$_{3}$$$$^{2+}$$, in [C$$_{4}$$mim][Tf$$_{2}$$N].

Journal Articles

Visualizing an ignition process of hydrogen jets containing sodium mist by high-speed imaging

Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 56(6), p.521 - 532, 2019/06

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

Journal Articles

Melting behavior and thermal conductivity of solid sodium-concrete reaction product

Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 56(6), p.513 - 520, 2019/06

 Times Cited Count:1 Percentile:58.8(Nuclear Science & Technology)

This study revealed melting points and thermal conductivities of four samples generated by sodium-concrete reaction (SCR). We prepared the samples using two methods such as firing mixtures of sodium and grinded concrete powder, and sampling depositions after the SCR experiments. In the former, the mixing ratios were determined from the past experiment. The latter simulated the more realistic conditions such as the temperature history and the distribution of Na and concrete. The thermogravimetry-differential thermal analyzer (TG-DTA) measurement showed the melting points were 865-942$$^{circ}$$C, but those of the samples containing metallic Na couldn't be clarified. In the two more realistic samples, the compression moldings in a furnace were observed. The observation revealed the softening temperature was 800-840$$^{circ}$$C and the melting point was 840-850$$^{circ}$$C, which was 10-20$$^{circ}$$C lower than the TG-DTA results. The thermodynamics calculation of FactSage 7.2 revealed the temperature of the onset of melting was caused by melting of the some components such as Na$$_{2}$$SiO$$_{3}$$ and/or Na$$_{4}$$SiO$$_{4}$$. Moreover, the thermal conductivity was $$lambda$$=1-3W/m-K, which was comparable to xNa$$_{2}$$O-1-xSiO$$_{2}$$ (x=0.5, 0.33, 0.25), and those at 700$$^{circ}$$C were explained by the equation of $$NBO/T$$.

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

Numerical modeling of radiation heat transfer under sodium spray combustion in sodium-cooled fast reactors

Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Uno, Masayoshi*

Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10

Heat radiation is one of dominant heat transfer process during a sodium fire event which is a concern in sodium-cooled fast reactor plants. This study aims to model radiation heat transfer from combusting droplets. Radiation energy transport on the combustion flame surface around a sodium droplet is formulated considering emission, absorption and scattering through a similar approach to the formulation of the wall boundary condition. The improved model is tested trough a simple verification analysis and a benchmark analysis on an upward sodium spray combustion experiment. As the result, overestimation of atmospheric temperature and pressure is mitigated by the improved model due to increase in heat transfer to structure.

Journal Articles

A Study on self-terminating behavior of sodium-concrete reaction, 2

Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 55(8), p.874 - 884, 2018/08

 Times Cited Count:3 Percentile:40.19(Nuclear Science & Technology)

As parts of severe accident studies in sodium-cooled fast reactor, experiments were performed to investigate the termination mechanism of sodium-concrete reaction (SCR). In the experiment, the reaction time was controlled to investigate the distribution change of sodium (Na) and the reaction products in the pool and around the reaction front. In the results, the Na around the reaction front decreased from the enough amount with the reaction time. The concentrations were 18-24 wt.% for Na, and 22-18 wt.% for Si after the termination. From the thermodynamics calculations, the stable materials around the reaction front comprised more than 90 wt.% solid products such as Na$$_{2}$$SiO$$_{3}$$, and no Na. Further, the distribution of Na and reaction products could be explained by a steady-state sedimentation-diffusion model. At the early stage of SCR, the reaction products were suspended as particles in the Na pool because of the high H$$_{2}$$-generation rate. As the concrete ablation proceeds, they start settling down due to the decreased H$$_{2}$$-generation rate, thereby allowing SCR termination. It was concluded that SCR termination was caused by the sediment of the reaction products and the lack of Na around the reaction front.

Journal Articles

Discussion about sodium-concrete reaction in presence of internal heater

Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the presence of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater (800$$^{circ}$$C) were performed to investigate the chemical reaction under the internal heater. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. Because the internal heater hindered the transport of Na, the moisture in the concrete, and reaction products, Na could permeate and react with the surface concrete at the periphery of the internal heater. As the SCR proceeded, the reaction products accumulated under the internal heater and disturbed the Na diffusion. Therefore, the Na concentration under the internal heater decreased relatively lower, and the concrete ablation depth under the internal heater decreased compared to that under the periphery of the internal heater. However, the Na concentration around the reaction front was about 30 wt.% despite the position of the internal heater. The Na concentration was similar to that of Na$$_2$$SiO$$_3$$, which was almost same as that in our past study. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.

Journal Articles

Application of multi-dimensional sodium fire analysis code AQUA-SF to severe accident; Benchmark analysis of upward spray combustion experiment

Aoyagi, Mitsuhiro; Takata, Takashi; Ohno, Shuji; Uno, Masayoshi*

Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00374_1 - 17-00374_13, 2018/03

no abstracts in English

Journal Articles

Thermophysical properties of sodium-concrete reaction products

Kawaguchi, Munemichi; Miyahara, Shinya; Uno, Masayoshi*

Netsu Sokutei, 45(1), p.2 - 8, 2018/01

Liquid sodium (Na) has been used as the coolant of fast reactors for the various merits, such as the high thermal conductivity. On the other hand, it is postulated that a steel liner may fail and lead to a sodium-concrete reaction (SCR) during the Na-leak accident. Because of concrete ablation and release of hydrogen gas due to the chemical reactions between Na and concrete components, the SCR is one of the important phenomena in the Na-leak accident. In the study, fundamental experiments related to the SCR were performed using Na and concrete powder. Here, the used concrete powder is milled siliceous concrete which is usually used as the structural concrete in Japanese nuclear power plants. The obvious temperature changes at 3 temperature regions were observed for the reaction process such as Na-melt, NaOH-SiO$$_{2}$$ and Na-H$$_{2}$$O-SiO$$_{2}$$ reaction, which occurred around 100, 300 and 500$$^{circ}$$C, respectively. Especially, the violent reaction around 500$$^{circ}$$C caused the temperature peak to $$836 sim 853^{circ}$$C, and the reaction heat of $$0.15 sim 0.23$$ kW/g was estimated under the Na-concrete mixing ratio such as $$gammaapprox 0.32$$. The main components of the reaction products was identified as Na$$_{2}$$SiO$$_{3}$$ with X-ray diffraction technique. Moreover, the measured thermophysical properties such as melting point, density, specific heat, thermal conductivity and viscosity were similar to those of $$x$$Na$$_{2}$$O-$$(1-x)$$SiO$$_{2}$$ ($$xleq 0.5$$).

Journal Articles

Oxide-metal ratio dependence of central void formation of mixed oxide fuel irradiated in fast reactors

Ikusawa, Yoshihisa; Maeda, Koji; Kato, Masato; Uno, Masayoshi*

Nuclear Technology, 199(1), p.83 - 95, 2017/07

 Times Cited Count:2 Percentile:80.48(Nuclear Science & Technology)

Based on thermal computation results obtained using an irradiation behavior analysis code, we have evaluated the effect of O/M ratio on fuel restructuring from the results of PIEs for the B14 irradiation test fuel, which was a mixed oxide fuel and was irradiated in the experimental reactor Joyo. The thermal computation results showed that fuel restructuring in the stoichiometric oxide fuel was accelerated, though the fuel temperature in the stoichiometric oxide fuel was evaluated as lower than that of the hypo-stoichiometric one. We explained this behavior as follows: first, the fuel temperature decreased due to the high thermal conductivity at stoichiometry; second, the pore migration velocity increased due to the increase in vapor pressure caused by the high vapor pressure of UO$$_{3}$$, which was derived from the high oxygen potential at stoichiometry. In addition, our results indicated that the central void diameter strongly depended on not only fuel temperature, but also vapor pressure.

Journal Articles

Cs-Te corrosion depth dependence on distribution of chromium carbide precipitation in high chromium steel

Sasaki, Koei; Fujimura, Ryota*; Tanigaki, Takanori; Matsubara, Masanori*; Fukumoto, Kenichi*; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 54(2), p.139 - 146, 2017/02


 Times Cited Count:3 Percentile:50.46(Nuclear Science & Technology)

In an attempt to investigate Cs-Te corrosion depth dependence on distribution of chromium carbide precipitation in high chromium steel, Cs-Te corrosion out-pile tests of two 9Cr steels with different distributions of chromium carbide were carried out at 975K for 100h and their corrosion depths were compared. The corrosion is obviously more advanced in a specimen which has grain boundary carbide than in the one that does not. A considerable reason of the result is that the carbide distributed at grain boundaries promoted the corrosion reaction and the corrosion extended along the grain boundary. This is the first case in which the Cs-Te corrosion depth dependence on distribution of chromium carbide precipitation in Fe-Cr steel is clarified experimentally.

Journal Articles

Thermodynamic and thermophysical properties of the actinide nitrides

Uno, Masayoshi*; Nishi, Tsuyoshi; Takano, Masahide

Comprehensive Nuclear Materials, 2, p.61 - 85, 2012/03

In a review "Comprehensive Nuclear Materials", we wrote the thermophysical and mechanical properties of actinide nitrides. In Chapter 1, the introductive remark was shown. In Chapter 2, the phase diagram and crystal structure of actinide nitrides were summarized. In Chapter 3, the melting or decomposition temperature, equilibrium vapor pressure, heat capacity, Gibbs free energy of formation and thermal conductivity of actinide nitrides were described, including recent experimental data obtained in JAEA. In Chapter 4, the mechanical properties of actinide nitrides and thermal expansion data obtained in JAEA were shown. On the other hand, the characteristic of nitride fuel and its historical development were not described because there appeared in another chapter of this review.

Journal Articles

Preparation and characterization of the simulated burnup americium; Containing uranium-plutonium mixed oxide fuel

Tanaka, Kosuke; Osaka, Masahiko; Miwa, Shuhei; Hirosawa, Takashi; Kurosaki, Ken*; Muta, Hiroaki*; Uno, Masayoshi*; Yamanaka, Shinsuke*

Journal of Nuclear Materials, 420(1-3), p.207 - 212, 2012/01

 Times Cited Count:4 Percentile:61.15(Materials Science, Multidisciplinary)

In order to investigate the effect on fuel thermophysical properties when adding americium and selected fission products to uranium-plutonium mixed fuel, simulated low decontamination MOX fuel with high burn-ups to 250 GWd/t, has been prepared and subjected to characterization tests, elastic moduli measurements, melting temperature measurement.

JAEA Reports

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2010 - March 31, 2011)

Katagiri, Hiromi; Okuno, Hiroshi; Sawahata, Masayoshi; Ikeda, Takeshi; Sato, Sohei; Terakado, Naoya; Nagakura, Tomohiro; Nakanishi, Chika; Fukumoto, Masahiro; Yamamoto, Kazuya; et al.

JAEA-Review 2011-037, 66 Pages, 2011/12


When a Nuclear emergency occurs, Nuclear Emergency Assistance & Training Center (NEAT) of JAEA gives technical advice and information, dispatch specialists as required, supplies emergency equipment and materials to the National Government and local governments. NEAT provides various lectures and training courses concerning nuclear disaster prevention for those personnel taking an active part in emergency response organizations at normal time. NEAT also researches on nuclear disaster prevention and cooperates with international organizations. Concerning about the assistance to the Accident of Fukushima No.1 Nuclear Power Station caused by the Great East Japan Earthquake at 11 March, 2011, JAEA assisted activities including environmental radiation monitoring, environmental radioactivity analyses, resident public consulting etc., with its full scale effort. NEAT served as the center of these supporting activities of JAEA.

Journal Articles

Actinide-handling experience for training and education of future expert under J-ACTINET

Osaka, Masahiko; Konashi, Kenji*; Hayashi, Hirokazu; Li, D.*; Homma, Yoshiya*; Yamamura, Tomoo*; Sato, Isamu; Miwa, Shuhei; Sekimoto, Shun*; Kubota, Takumi*; et al.

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 5 Pages, 2011/12

Summer schools for future experts have successfully been completed under Japan Actinide Network (J-ACTINET) for the purpose of development of human resources who are expected to be engaged in every areas of actinide-research/engineering. The first summer school was held in Ibaraki-area in August 2009, followed by the second one in Kansai-area in August 2010. Two summer schools have focused on actual experiences of actinides in actinide-research fields for university students and young researchers/engineers as an introductory course of actinide-researches. Several quasi actinide-handling experiences at the actinide-research fields have attracted attentions of participants at the first school in Ibaraki-area. The actual experiments using actinides-containing solutions have been carried out at the second school in Kansai-area. Future summer schools will be held every year for the sustainable human resource development in various actinide-research fields.

Journal Articles

Oxygen non-stoichiometries in (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$

Osaka, Masahiko; Miwa, Shuhei; Kurosaki, Ken*; Uno, Masayoshi*; Yamanaka, Shinsuke*

Journal of Nuclear Materials, 408(3), p.285 - 288, 2011/01

 Times Cited Count:3 Percentile:69.54(Materials Science, Multidisciplinary)

Oxygen non-stoichiometry in (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$ were investigated from a viewpoint of features of Ce reduction in several oxide solid solutions. Oxygen non-stoichiometry was experimentally determined by means of thermogravimetric analysis as a function of oxygen potential at 1173, 1273 and 1373 K. The similar features of isotherms of oxygen non-stoichiometry in (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$ with those in actinide and/or lanthanide-oxygen dioxides were observed. Oxygen non-stoichiometry in (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$ were compared with those of CeO$$_{2-x}$$ and (U$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$. It was concluded that features of Ce reduction have some relationships with defect forms and their transformations in the solid solutions.

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