針對壓水式反應器(pressurized water reactor, PWR)二次側冷卻水系統，水化學的控制顯得非常重要，避免因為蒸氣產生器的腐蝕造成系統損壞，像是沸騰區域的不純物或是管路內的污泥(Sludge)沈積造成管件內部腐蝕問題的發生。
蒸氣產生器(Steam generator, SG)二次側環境總是存在著一定程度的鉛，鉛是蒸氣產生器污泥沈積物常見的成分之一，極低濃度的可溶性鉛會導致鎳基合金產生應力腐蝕龜裂(SCC)，鉛在劣化過程產生影響並出現在裂縫尖端附近，讓裂縫區域的基材發生局部變化，鉛也可能是破壞保護性氧化層形成的能力，鉛污染可能導致鈍化膜發生變化，導致其發生劣化。
透過PWR二次側水循環模擬迴路中針對600合金進行notched C-ring 應力腐蝕試驗，透過不同濃度的PbO注入，水環境控制在低溶氧狀態下(~數個ppb)的水化學條件，利用氫氧化銨控制其pH值後，在280 oC溫度下600合金的應力腐蝕龜裂行為分析。探討在280 oC、8.3MPa條件下，不同PbO濃度與固定pH值的水環境對於600合金SCC裂縫起始與成長之影響。

For the secondary cooling system of the pressurized water reactor, the control of water chemistry condition is very important. Adjusting the water chemistry condition can prevent the corrosion of the steam generator caused by the impurities in the boiling area or sludge in the pipeline (Sludge) deposits.
In the secondary side of steam generators, the presence of lead is one of the common compositions of sludge deposits. The low concentration of soluble lead can cause stress corrosion cracking (SCC) in nickel-based alloys. Lead may contribute to the severe impacts near the tip of crack in the degradation process. In addition, it may break the formation of protective oxide layer and lead to the degradation of the passive films.
The aim of this study is to investigate the impact of soluble lead concentration on SCC behavior of Alloy 600 in the simulated PWR secondary side water chemistry circuit. The stress corrosion cracking behavior was conducted by notched C-Ring method at the temperature of 280 oC and the pressure of 8.3 MPa. The water chemistry condition was controlled under specific conductivity and pH value by injecting ammonium hydroxide and low dissolved oxygen value(~several ppb) with different PbO concentrations. In addition, the effect of heat treatment process on SCC behavior of Alloy 600 was also discussed in the same environments. The SCC tests were followed by a detailed characterization on the microstructure and the distribution and growth of the cracks generated on the surface of Alloy 600 samples.

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