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  • 學位論文

304L不銹鋼和316L不銹鋼於模擬沸水式反應器起動狀態之水化學環境中的應力腐蝕龜裂行為研究

Stress Corrosion Cracking of Type 304L Stainless Steels and Type 316L Stainless Steels in Simulated Boiling Water Reactor Water Chemistry Environments under Start-up Conditions

指導教授 : 葉宗洸

摘要


04L不銹鋼和316L不銹鋼對於應力腐蝕龜裂(Stress Corrosion Cracking, SCC)有較高的敏感性,為了確保沸水式反應器(Boiling Water Reactor, BWR)結構組件的完整性,目前已有相當多的電廠採用了加氫水化學(Hydrogen Water Chemistry, HWC)技術。利用降低腐蝕電位(Electrochemical Corrosion Potential, ECP)來達到腐蝕抑制的效果。然而由於技術運轉規範的限制,台灣現行的核電廠設計在功率達到90%後才啟動加氫系統,考量到反應爐起動時特殊的水化學狀態,在此運轉條件下材料的腐蝕狀況值得加以重視和研究。 本研究採用慢應變速率拉伸實驗(Slow Strain Rate Test, SSRT)探討兩種低碳不鏽鋼在不同熱處理後,於一般水環境 (Normal Water Chemistry, NWC) 以及加氫水化學等兩種水化學環境下的應力腐蝕龜裂行為。試棒為棒狀圓柱,主要實驗溫度為288℃、250oC和200oC,拉伸速率為3.0 × 10-7 /s。接著以掃描式電子顯微鏡(Scanning Electron Microscopy, SEM)分析斷裂面型態以及表面樣貌。結果顯示316L不銹鋼相較於304L不銹鋼有較好的抗沿晶腐蝕能力。同時,在一般水化學環境下,對兩種不銹鋼而言較低的溫度都有著較低的應力腐蝕龜裂敏感性,其中316L在150oC下和304L在200oC下都無應力腐蝕龜裂出現,然而在250oC和288oC的結果都有著傾向多重裂縫起始的破裂形貌,呈現了較嚴重的應力腐蝕龜裂特徵。而加氫水化學的採用對於敏化316L抵抗應力腐蝕龜裂劣化有良好的效果,然而對於304L其優化效果卻不如預期。推論由於加氫水化學環境上和材料本身以及熱處理的差異造成不同的結果。其中加氫水化學中的氫/氧莫耳比、不銹鋼中的鉬和鎳含量造成了304L不銹鋼對於穿晶應力腐蝕龜裂可能有著較高的敏感性。

並列摘要


Hydrogen water chemistry (HWC) was proposed and widely adopted in BWRs worldwide to mitigate the problem of SCC. However, the hydrogen injection can only be applied during normal plant operation due to some limits inherent in the original design. The pressure and temperature gradient during plant start-up could cause the tensile stress conditions. Regarding the water environment, the above factors may contribute to the initiation of SCC. In this study, we compared the SCC behavior between two type of stainless steel with different heat treatments at various temperature in the NWC and HWC during start-up. The SCC behavior and high temperature tensile properties were investigated by performing slow strain rate tensile (SSRT) tests at a constant strain rate of 3.0 × 10-7/s mainly at 200°C, 250°C and 288°C. After SSRT tests, scanning electron microscopy (SEM) was conducted to characterise the fractured surface. Basically, these two materials in the NWC test conditions showed lower SCC susceptibility with the lower testing temperature. Specimens tested under 250oC, however, showed higher tendency toward multiple cracks initiations. On the other hand, the performance of sensitized 316L in the aspects of tensile properties and fractography was substantially improved after introducing the HWC. Nevertheless, the unexpected performances were found on sensitized 304L after injecting hydrogen into the system. The slight differences between materials, heat treatment and water chemistry in the testing conditions were considered as the mian causes of these results.

並列關鍵字

BWRs SCC SSRT HWC

參考文獻


[1] 郭榮卿, 核能電廠材料劣化與對策研究:現況與規劃. 台灣: 楊義卿, 2006.
[2] G. R. M. Horn. et al, "Experience and assessment of stress corrosion cracking in L-grade stainless steel BWR internals," Nuclear Engineering and Design, pp. 313-325, 1997.
[3] P. L. Andresen, "Stress Corrosion Cracking of Current Structural Materials in Commercial Nuclear Power Plants," Corrosion, vol. 69, pp. 1024-1038, 2013.
[4] International Atomic Energy Agency, “Stress Corrosion Cracking In Light Water Reactor: Good Practices And Lessons Learned,” IAEA, Vienna, NUCLEAR ENERGY SERIES No. NP-T-3.13, 2011.
[5] U. Ehrnstén et al, “Intergranular Cracking of AISI 316NG Stainless Steel in BWR Environment,” ENVDG 2001, Nevada, USA, August 2001.

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