Interaction of CsOH with structural steel 316L and simulated products of the steel-cladding reaction under assumed conditions relevant for the accident at the Fukushima Dai-Ichi Unit 2 and 3

A high radiation dose rate under the shield plugs of the Units 2 and 3 reactors at the Fukushima Dai-Ichi Nuclear Power Station (1F), raised an important question on volatile radioactive Cs migration, adsorption and retention on various structural materials during the accident. The high dose rate interferes with decommissioning work. The purpose of this study is to present basic knowledge on contamination mechanisms of a BWR reactor structural materials and post-accident debris by Cs compounds to help decommissioning companies consider the ways of decontamination and debris handling during 1F debris retrieval. Having this goal, a series of tests on CsOH contamination mechanism under assumed accident conditions for Units 2 and 3 was performed. The results may improve the accuracy of estimating the situation inside the reactor from the perspectives of accident analysis and materials’ properties. The test temperature and compounds were chosen based on the probable accident scenario. In particular, Cs adsorption was checked using preoxidized 316L stainless steel, a product of control blade degradation (steel-B ₄C eutectic) and a product of steel-channel box and cladding interaction (Zr-Fe based multiphase material). Penetration ability of Cs into the oxidized and metallic layers of materials was evaluated. CsOH most probable adsorption temperature range and new Cs containing phases’ formation temperature range were followed.