Hydrogen flux through the heat transfer tube wall in the steam generator of Monju

高速増殖原型炉「もんじゅ」における蒸気発生器伝熱管の水側腐食による拡散水素量の評価

土井 大輔  ; 中桐 俊男  

Doi, Daisuke; Nakagiri, Toshio

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.