On the corrosion behavior of zircaloy-4 in spent fuel pools under accidental conditions

O. Lavigne, Tetsuo Shoji, K. Sakaguchi

Research output: Contribution to journalArticle

1 Citation (Scopus)

Abstract

After zircaloy cladding tubes have been subjected to irradiation in the reactor core, they are stored temporarily in spent fuel pools. In case of an accident, the integrity of the pool may be affected and the composition of the coolant may change drastically. This was the case in Fukushima Daiichi in March 2011. Successive incidents have led to an increase in the pH of the coolant and to chloride contamination. Moreover, water radiolysis may occur owing to the remnant radioactivity of the spent fuel. In this study, we propose to evaluate the corrosion behavior of oxidized Zr-4 (in autoclave at 288°C for 32 days) in function of the pH and the presence of chloride and radical forms. The generation of radicals is achieved by the sonolysis of the solution. It appears that the increase in pH and the presence of radicals lead to an increase in current densities. However, the current densities remain quite low (depending on the conditions, between 1 and 10 μA cm -2). The critical parameter is the presence of chloride ions. The chloride ions widely decrease the passive range of the oxidized samples (the pitting potential is measured around +0.6 V (vs. SCE)). Moreover, if the oxide layer is scratched or damaged (which is likely under accidental conditions), the pitting potential of the oxidized sample reaches the pitting potential of the non-oxidized sample (around +0.16 V (vs. SCE)), leaving a shorter stable passive range for the Zr-4 cladding tubes.

Original languageEnglish
Pages (from-to)120-125
Number of pages6
JournalJournal of Nuclear Materials
Volume426
Issue number1-3
DOIs
Publication statusPublished - 2012 Jul 1

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Materials Science(all)
  • Nuclear Energy and Engineering

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