Affiliation:
1. Research Institute of Nuclear Materials, Joint-Stock Company
2. High Temperature Electrochemistry of the Ural Branch of the Russian Academy of Sciences
Abstract
At present, technologies are being developed for the regeneration of mixed nitride uranium-plutonium spent nuclear fuel (MNUP SNF) for the BREST-OD-300 reactor plant, including the use of a pyrochemical method of mild chlorination in alkali metal chloride melts to separate fuel from fuel rod claddings made from high radiation resistance of ferritic-martensitic steel EP-823. The paper gives the results of EP-823 static corrosion tests in KCl–LiCl and KCl–LiCl–nPbCl2 molten salts at the temperature of 500 and 650°С during 24 h. Corrosion behaviour of EP-823 steel in non-oxidized and thermal air oxidized state with oxide film thickness up to ~12.5 µm has been investigated using neutron-activation analysis. EP-823 steel samples, irradiated in IVV-2M reactor up to neutron fluence of ~2.9 · 1017 n/cm2, have been examined. It has been shown that corrosion impact of 2KCl–3LiCl and 2KCl–3LiCl–nPbCl2 molten salts on EP-823 element corrosion is selective. It has been established that EP-823 steel in 2KCl–3LiCl molten salts of eutectic composition is highly corrosion-resistant. An increase in the test temperature and the introduction of PbCl2 into the KCl–LiCl salt melt in the amount of one mole percent leads to an increase in the corrosion rate and the removal of steel corrosion products by almost two orders of magnitude. It has been established that oxide films on EP-823 steel surface does not restrain corrosion rate in 2KCl–3LiCl–nPbCl2 molten salts. The values of the constants given in Table 6, make it possible to calculate the values of the average corrosion rates of EP-823 steel and its components (Fe, Cr, Mn) in molten salts 2KCl–3LiCl and 2KCl–LiCl–nPbCl2 at various temperatures.
Publisher
The Russian Academy of Sciences
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