鋯合金因為擁有好的抗腐蝕能力，較低的熱中子捕獲截面積還有良好的機械性質，被廣泛地使用在核能電廠的護套材料上。但是當意外發生時，特別是在冷卻水喪失的意外中，爐心的溫度會不斷上升，此時鋯合金燃料護套將會喪失其抗氧化能力並且與環境中的水蒸氣劇烈反應生成二氧化鋯並且放出氫氣，除了護套本身毀壞會導致合然料外洩，釋放出的氫氣也可能產生氫爆，因此改善鋯合金護套材料在高溫下的抗腐蝕能力是相當重要的議題。在本研究中，我們透過非平衡磁控濺鍍系統在鋯四合金上鍍著氮化鉻薄膜，並研究鍍膜的鋯四合金在400℃到600℃的高溫水氣環境中2，4及6天的抗腐蝕能力。實驗結果顯示在400℃，500℃的高溫水氣中進行長達2到6天的實驗後，從重量變化觀察到鍍膜可以降低重量增加將近九成左右，鍍膜試片在實驗後依然維持薄膜完整結構，未鍍膜試片則是在表面形成數微米的二氧化鋯層。在600℃高溫水氣中經過6天後，沒鍍膜試片已經全部氧化成二氧化鋯並且變成粉末狀，而鍍膜試片表層氮化鉻薄膜仍然附著於基材，並且薄膜表面開始形成三氧化二鉻薄膜，整體試片依舊保持完整性。由實驗結果證明鍍上氮化鉻薄膜確實能夠提升基材抗腐蝕能力。此外，我們發現鋯合金在600℃後抗腐蝕能力會驟然下降，主要是因為氧在氧化鋯中的擴散係數在600℃會突然增加，所以沒鍍膜的試片在600℃水氣環境長達六天之後，會因為氧化成二氧化鋯而整塊粉碎，而在400℃與500℃的水氣環境6天後，雖然也有二氧化鋯生成，但因為生成速率較慢，試片依然能保持其完整性。

Zr-based alloys have been widely used as the cladding materials in the nuclear reactors, due to their desirable properties of high corrosion resistance, low thermal-neutron capture cross-section, and good mechanical properties. However, in the accident such as the loss-of-coolant accident, the temperature of reactor cores will increase rapidly. In this case, Zr-based alloys will lose their corrosion resistance and react with the steam to produce zirconia and hydrogen, which may leads to the leakage of fission products and hydrogen explosion caused by the degradation of fuel cladding and the formation of hydrogen. Therefore, the improvement of corrosion resistance of cladding materials at high temperature is critical. In this study, we deposited chromium nitride (CrN) thin films on investigated Zircaloy-4(Zr-4) substrates by using unbalanced magnetron sputtering (UBMS) system and investigated the corrosion resistance of CrN deposited Zr-4 alloy at high temperatures, ranging from 400℃ to 600℃ in steam environment for 2, 4, and 6 days. The results show that the weight gain of CrN-coated samples is about 90% less than that of uncoated samples for 2 to 6 days at 400℃ and 500℃ in steam environment. In addition, there is no formation of zirconia and the CrN film remain integral in the CrN-coated specimens, while there is a continuous oxide layer which is a few microns thick on the surface of the uncoated specimen. The uncoated specimens completely fractured caused by phase transformation from Zr to porous monoclinic ZrO2, whereas the CrN-coated specimen is still intact and the CrN film is well attached to the substrate with the formation of Cr2O3 layer on the surface after steam tests for 6 days at 600℃. The results show that introducing a CrN thin film can effectively improve the corrosion resistance of Zircaloy-4 at high-temperature steam environment. Furthermore, we find that the corrosion resistance of specimens significantly degrades when temperature increases to 600℃, which is caused by the sudden increase of the diffusion coefficient of oxygen in zirconia. That is why the uncoated specimens still have their integrities despite the formation of ZrO2 at 400℃ and 500℃, whereas they fractured into pieces due to the phase transformation of whole substrate from Zr to ZrO2 at 600℃.

摘要 i
Abstract ii
致謝 iv
目錄 vi
表目錄 viii
圖目錄 ix
第一章 緒論 1
1.1 研究背景 1
1.2 研究目的 2
第二章 文獻回顧 3
2.1 核能反應爐[7][8] 3
2.2 核能電廠意外事故[9][10] 6
2.3 燃料護套材料 6
2.3.1 鋯合金之性質[11] 6
2.3.2 鋯合金的氧化行為 8
2.4 氮化鉻薄膜的抗氧化性質 10
2.5 非平衡磁控濺鍍系統(Unbalanced Magnetron Sputtering System, UBM) 12
第三章 實驗流程與儀器 14
3.1 試片準備與鍍膜程序 14
3.1.1 鋯四合金基材備製 14
3.1.2 氮化鉻薄膜備製 14
3.2 高溫氧化實驗 17
3.3 分析儀器 18
3.3.1 聚焦離子束電子束掃描式顯微鏡(Dual-beam Focused Ion Beam System, FIB) 18
3.3.2 X光繞射分析儀(X-ray Diffraction, XRD) 19
3.3.3 奈米壓痕測試儀 20
3.3.4 原子力顯微鏡(Atomic Force Microscope, AFM) 20
3.3.5 輝光放電分析儀(Glow Discharge Spectrometer, GDS) 20
3.3.6 能量散佈X光能譜儀(Energy-dispersive X-ray Spectroscopy, EDX) 21
第四章 結果 22
4.1 裸材組與鍍膜組性質比較 22
4.2 在400℃下氧化結果觀察 27
4.3 在500℃下氧化結果觀察 32
4.4 在600℃下氧化結果觀察 36
第五章 討論 43
5.1 溫度對於氧化行為之影響 43
5.2 時間對於氧化行為之影響 48
5.3 試片結構對於氧化行為之影響 53
第六章 結論 57
第七章 未來工作 58
參考文獻 59

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