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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

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

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

 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:2 Percentile:16.17(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:1 Percentile:64.68(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

AA2016-0211.pdf:2.83MB

 Times Cited Count:1 Percentile:64.68(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:5 Percentile:48.35(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

JAEA-Review-2011-037.pdf:4.52MB

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:66.23(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.

Journal Articles

Effect of americium and simulated fission products addition on oxygen potential of uranium-plutonium mixed oxide fuels

Tanaka, Kosuke; Osaka, Masahiko; Kurosaki, Ken*; Muta, Hiroaki*; Uno, Masayoshi*; Yamanaka, Shinsuke*

Materials Research Society Symposium Proceedings, Vol.1215, p.95 - 100, 2010/10

In order to investigate the effect of MAs and FP addition on the oxygen potential of MOX fuels, thermogravimetric analysis (TGA) were carried out. MOX fuels with Am and 26 kinds of fission product elements (FPs), simulating low-decontamination MOX fuel and high burn-up to 250 GWd/t, were prepared by a conventional powder metallurgical route in a glove box. The oxygen potentials for simulating low-decontamination MOX fuels were higher than the fuels without FPs and increased positively with increasing simulated burn-up.

Journal Articles

Thermal conductivities of Cs-$$M$$-O ($$M$$ = Mo or U) ternary compounds

Tokushima, Kazuyuki*; Tanaka, Kosuke; Kurosaki, Ken*; Gima, Hiromichi*; Muta, Hiroaki*; Uno, Masayoshi*; Yamanaka, Shinsuke*

Materials Research Society Symposium Proceedings, Vol.1215, p.151 - 156, 2010/10

The thermal diffusivities of Cs$$_{2}$$MoO$$_{4}$$ and Cs$$_{2}$$UO$$_{4}$$ using samples fabricated by hot press and SPS techniques were measured by a laser flash method in the range from room temperature to 823 K for Cs$$_{2}$$MoO$$_{4}$$ and to 900 K for Cs$$_{2}$$UO$$_{4}$$. The thermal conductivities of these cesium ternary oxides were quite low compared with UO$$_{2}$$ and MOX fuel. This is consistent with previous findings. These results would be useful for evaluating the thermal performance of MOX fuels at the high burn-up region in the fast reactors.

Journal Articles

Oxygen potential of (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$

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

Materials Research Society Symposium Proceedings, Vol.1215, p.199 - 203, 2010/10

Oxygen potentials of (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$ were experimentally determined by means of thermogravimetric analysis as a function of non-stoichiometry at 1173 and 1273 K. Oxygen potentials of (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$ at each temperature increased with increase of O/M ratio (= 2-x) and steep increases of the oxygen potentials toward O/M = 2 were observed. These characteristics are typical for non-stoichiometric fluorite-type actinides dioxides. The oxygen potentials of (Th$$_{0.7}$$Ce$$_{0.3}$$)O$$_{2-x}$$ were similar to those of CeOO$$_{2-x}$$ when they were plotted as a function of average Ce valence.

Journal Articles

Chemical states of fission products and actinides in irradiated oxide fuels analyzed by thermodynamic calculation and post-irradiation examination

Kurosaki, Ken*; Tanaka, Kosuke; Osaka, Masahiko; Oishi, Yuji*; Muta, Hiroaki*; Uno, Masayoshi*; Yamanaka, Shinsuke*

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 4 Pages, 2010/10

It is important to understand the behavior of fission products and actinides under irradiation. In the present study, the chemical states of fission products and actinides in irradiated oxide fuels were evaluated by both thermodynamic equilibrium calculation and post-irradiation examination.

JAEA Reports

Studies on planning and conducting for reducing water inflow due to underground construction in crystalline rock

Mikake, Shinichiro; Yamamoto, Masaru; Ikeda, Koki; Sugihara, Kozo; Takeuchi, Shinji; Hayano, Akira; Sato, Toshinori; Takeda, Shinichi; Ishii, Yoji; Ishida, Hideaki; et al.

JAEA-Technology 2010-026, 146 Pages, 2010/08

JAEA-Technology-2010-026.pdf:41.08MB
JAEA-Technology-2010-026-appendix(CD-ROM).zip:83.37MB

The Mizunami Underground Research Laboratory (MIU), one of the main facilities in Japan for research and development of the technology for high-level radioactive waste disposal, is under construction in Mizunami City. In planning the construction, it was necessary to get reliable information on the bedrock conditions, specifically the rock mass stability and hydrogeology. Therefore, borehole investigations were conducted before excavations started. The results indicated that large water inflow could be expected during the excavation around the Ventilation Shaft at GL-200m and GL-300m Access/Research Gallery. In order to reduce water inflow, pre-excavation grouting was conducted before excavation of shafts and research tunnels. Grouting is the injection of material such as cement into a rock mass to stabilize and seal the rock. This report describes the knowledge and lessons learned during the planning and conducting of pre-excavation grouting.

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