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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$$_{2}$$ ($$xleq 0.5$$).

JAEA Reports

A Study on density, melting point, thermal expansion, creep, thermal diffusivity and thermal conductivity of the simulated rock-like oxide (ROX) fuels

Yanagisawa, Kazuaki; Omichi, Toshihiko*; Shirasu, Noriko; Muromura, Tadasumi; *

JAERI-Tech 99-032, 65 Pages, 1999/03

JAERI-Tech-99-032.pdf:3.23MB

no abstracts in English

Journal Articles

Out-of-pile tests of simulated rock-like oxide (ROX) fuels

Yanagisawa, Kazuaki; Omichi, Toshihiko*; Muromura, Tadasumi; *; Shirasu, Noriko

Journal of Nuclear Science and Technology, 36(2), p.160 - 168, 1999/02

 Times Cited Count:4 Percentile:35.7(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Application of neutron radiography to visualization of direct contact heat exchange between water and low melting point alloy

Nishi, Yoshihisa*; *; *; *; Matsubayashi, Masahito; Tsuruno, Akira

Fifth World Conf. on Neutron Radiography, 0, p.548 - 555, 1996/00

no abstracts in English

Journal Articles

Structure molten salts near the melting point

Furukawa, Kazuo

Discuss.Faraday Soc., 32, 53 Pages, 1962/00

no abstracts in English

Oral presentation

Characterization of melt-solidified (U, Gd, Zr)O$$_{2-x}$$ as simulated corium debris

Morimoto, Kyoichi; Hiroka, Shun; Akashi, Masatoshi; Watanabe, Masashi

no journal, , 

The influence of Gd on characteristics of debris is important for removing the debris from the reactors of Fukushima Daiichi Nuclear Power Plant because subassemblies of nuclear fuels containing Gd$$_{2}$$O$$_{3}$$ were loaded in the some reactor cores. Additionally, it is important to assess the distribution state of Gd from the anxiety of re-criticality caused by the relocation of debris while removing them. In this study, sintered pellets of (U$$_{0.95-y}$$Gd$$_{0.05}$$Zr$$_{y}$$)O$$_{2-x}$$ (y=0,0.5, 2-x=1.989-2.000) were melted and solidified to prepare specimens of simulated corium debris. Phase states and fundamental properties of them were evaluated.

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