A sodium-concrete reaction (SCR) is one of the important phenomena to cause the structural concrete ablation and the release of hydrogen (H 2) gas in the sodium (Na) leak accident. In this study, the long-time SCR test had been carried out to investigate the self-termination mechanism under the condition to keep the temperature of Na on the concrete more than the reaction threshold temperature during 24 hours. The test results showed the SCR terminated by itself even if enough amount of Na remained on the concrete. In addition, quantitative data were collected on the SCR terminating behavior such as the temperature, the concrete ablation depth, the H 2 generation behavior and the concentration profiles of Na, silicon (Si), aluminum (Al) and calcium (Ca) in the reaction products after the test. In the concentration profiles, the calculation by the sedimentation diffusion model of the steady state was comparable with the experimental results. Though the reaction products were suspended by H 2 bubbling and Na ablated the concrete surface with the high H 2 generation rate, the reaction products gradually settled down with decreasing of the H 2 generation rate. Therefore, the Na concentration decreases at the reaction front with time and the SCR terminates of itself.

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In fast breeder reactors (FBRs), hydrogen, one of the major impurities in sodium coolant, is used for water leak detection and tritium control in the interest of operating nuclear plant safety. It is essential to evaluate the hydrogen flux into the sodium coolant through the heat transfer tubes in a steam generator to understand the hydrogen behavior in an FBR plant. This study shows the time and temperature dependence of hydrogen flux using data obtained in the power rising test of the prototype FBR, Monju, and using the analysis code that can simulate the hydrogen distribution in an FBR plant. The hydrogen flux evaluated by the code gradually decreased with operating time, following the parabolic oxidation law and the Arrhenius relation over a wide range of steam temperatures. The results agreed with the hydrogen fluxes of other FBR plants evaluated in the past. It was also found that the hydrogen flux was mainly controlled by permeation through the heat transfer tubes, rather than the corrosion at the water side of the heat transfer tubes.