用過核子燃料乾貯筒材料在含氯環境中的腐蝕行為研究

Corrosion Behavior of Candidate Materials Used in Dry Cask Storage Systems in Chloride-rich Environments

使用過的核燃料先經過濕式燃料池冷卻貯存之後，再移至乾式貯存筒存放。本文之目的為探討乾式貯存筒之應力腐蝕龜裂。以U-bend試片進行實驗，使用固溶及敏化熱處理之304、304L、316L不銹鋼，經過氯化鈉沉積後，於不同溫度下、進行長時間的測試，以電子顯微鏡及光學顯微鏡進行表面分析及裂縫觀察。研究顯示，在高溫的實驗環境與敏化熱處理的條件下，產生較多的裂縫，但，測試時間對不銹鋼裂縫成長的影響並不顯著，而溫度及敏化則有讓裂縫數目增加且氧化情形加劇的效果。

The dry cask storage of spent nuclear fuel is suitable as an interim safer solution than spent fuel pool storage. Most canisters used in dry storage system for spent fuel are fabricated from austenitic stainless steel. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC). The purpose of this study is to evaluate the susceptibility to chloride induced stress corrosion cracking (CISCC) of candidate canister materials (304, 304L,316L stainless steels) by U-bend tests in a chloride-rich environment. A detailed characterization on the microstructure of the samples was analyzed by using the scanning electron microscope (SEM). The number and length of cracks in selected area were also measured. More cracks were found in higher environment temperature and also on the sensitized specimens. The influence of testing time did not show in these test conditions.