STUDIES ON ZIRCALOY-4 CLADDING CORROSION IN THE HALDEN REACTOR

摘要

This paper presents the results of Zircaloy-4 cladding corrosion tests conducted in the Halden Reactor between 1987 and 2006. The tests studied the effects of operating parameters on the corrosion behaviour of both standard and low-tin Zircaloy-4 under PWR conditions. One test was performed to investigate the effect of power cycling between two cooling regimes, i.e. nucleate boiling and one-phase cooling, on the corrosion process. A further experiment had the objective to assess the effect of an increased coolant lithium concentration on cladding corrosion. The next test in the series was conducted to study different possible causes of accelerated corrosion in Zircaloy-4 cladding under typical PWR operating conditions, while in the final experiment in-pile corrosion and hydriding of both conventional and newly developed Zr-based PWR cladding materials were investigated at burnup levels in excess of 50 MWd/kg UO_2. All experiments included fuelled test rods to achieve representative cladding temperatures. PWR thermal-hydraulic and water chemistry conditions were simulated by means of an external loop system. The corrosion behaviour was investigated by means of interim inspections (comprising oxide thickness measurements) and post irradiation examination. A semi-empirical corrosion model based on EPRI/C-R/KWU correlations has been applied to predict the corrosion behavior. The results indicated that the corrosion rate of the cladding materials may have a relatively lower sensitivity to metal-oxide interface temperature than assumed by model predictions. A clear corrosion enhancement due to high lithium concentration could not be concluded, neither was an aggravating effect due to nucleate boiling detected in the tests.