為了解決日益關鍵的用過核燃料貯存問題,將已在濕式貯存燃料池冷卻多年的燃料棒移至乾式貯存將會是良好且安全的解決辦法。乾式貯存分為室內及室外兩種,本實驗之目的為探討此設施在使用上的安全性,是否會因應力腐蝕龜裂之裂痕貫穿密封筒導致輻射外洩。由美國核能管制委員會的報告[1]中得知,不鏽鋼乾貯桶上應力腐蝕龜裂的發生條件,包含其表面必須有氯鹽沉積、環境相對濕度的高低循環變化及張應力。 本實驗以模擬乾貯筒放置之沿海環境的U-bend試片進行實驗,材料為固溶熱處理及敏化熱處理之304、304L、316L不鏽鋼與GGG-40 球墨鑄鐵(Ductile Cast Iron, DCI),以氯化鈉水溶液噴灑,使試片表面形成鹽類沈積之後,於不同溫度、固定相對濕度環境中進行長時間的測試。實驗完成後,以電子顯微鏡及光學顯微鏡觀測其試片表面是否有產生裂紋或孔蝕之現象,並加以分析,接著針對定面積進行裂縫數目及長度量測。根據不鏽鋼裂縫形貌及數量計算的結果,可得知時間並未明顯影響裂縫的成長,敏化熱處理則有讓不鏽鋼材料表面小裂縫數目增加的趨勢,實驗溫度的提高使的氧化情形加劇的效果,同時也觀察到孔蝕與裂縫成長間的關係。
The dry cask storage of spent nuclear fuel is more 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), and the corrosion initiated by the deliquescence of sea salts coupled with susceptible materials and residual stress can lead to stress corrosion cracking. 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 and GGG-40 ductile cask iron) by U-bend tests in a simulated marine atmospheric environment. A detailed characterization on the microstructure of the samples was analyzed by using the scanning electron microscope (SEM) and optical microscope (OM). The number and length of cracks in selected area were also measured. The influence of sensitization heat treatment and temperature was obvious. More cracks were found in sensitized specimens. The oxidation behavior is more serious with higher environment temperature. Besides, the phenomena that the cracks tended to grow between pits was also found.