Abstract
鋯合金長期使用於反應器爐心作為燃料護套的材料, 腐蝕問題一直是 鋯合金使用中備受注意的焦點, 最新的研究顯示: 腐蝕動力學的機制和氧 化膜結構, 氧化膜中析出物分佈, 及界面微結構等現象有著密切的關聯. 同時因為鋯合金使用於強烈輻射照射的爐心環境, 輻射傷害的影響是絕不 可忽視的因素. 因此本研究之目的是在探討輻射傷害對鋯二合金均勻腐蝕 的各種影響, 尤其是均勻腐蝕過程中氧化膜微結構的變化與腐蝕機制的關 係. 利用2MeV能量的質子, 於350C下, 對鋯二合金板材進行劑量達到 一個原子位移(1dpa)的輻射照射, 然後於高溫高壓釜中進行400C過熱水蒸 汽的模擬腐蝕實驗, 以及320C純水中的均勻腐蝕測試, 以瞭解輻射傷害對 於均勻腐蝕的影響. 利用上述的腐蝕試片進行X-光薄膜繞射分析, 可以瞭 解氧化膜中結構的變化. 在氧化物微結構的觀察部份, 除利用氧化膜上視 法得到初期氧化膜的型態與結構外, 並將前述經均勻腐蝕測試的試片製成 橫截面電子顯微鏡試片, 以此觀察整個氧化膜自金屬與氧化膜界面到外表 面的所有氧化物微結構的變化, 並且利用高分辨電子顯微鏡進行更細微區 域的影像觀察. 均勻腐蝕結果顯示: 不管是400C過熱水蒸汽, 還是320 C純水的環境下, 輻射傷害都具有加速鋯二均勻腐蝕速率的作用, 且其加 速效果介於1.5到3倍之間. X-光薄膜繞射分析上, 首次發現: 輻射傷害後 的鋯合金經400C過熱水蒸汽環境下, 所形成的均勻腐蝕氧化膜其中長方( tetragonal)堆積結構的量大增. 這種現象除了牽涉到應力的發展外, 也 與溶質在輻射傷害後重新進入基地有關. 微結構的觀察上發現: 均勻腐蝕 的氧化膜都是由兩種型態不同的氧化物所組成, 一是外層, 或是較早期形 成的等軸晶粒區; 二是較內層的柱狀組織區域. 這兩種氧化物都是單斜( monoclinic)結構. 等軸晶粒區域的深度約為200nm左右, 而柱狀組織由小 而大, 氧化膜大部份由這些柱狀組織所組成. 輻射傷害對於等軸晶粒區域 的氧化物大小及此層厚度, 並未產生明顯的差異, 但是輻射傷害對於柱狀 組織中柱狀氧化物晶粒的大小, 卻產生極為明顯的差別- 輻射傷害產生較 小的, 但卻數量較多的柱狀組織晶粒. 在金屬與養化膜界面區域, 並無任 何的中間層氧化物結構, 鋯合金的界面是由柱狀組織的氧化物直接接觸金 屬鋯基地. 此外在界面區域的金屬側, 也可以觀察到由於巨大壓應力所形 成的行變帶狀區域. 本研究除提出機制解釋鋯合金表面氧化膜的過渡 期現象外, 並更進一步, 以溶質元素再分佈進入基地的溶質機制, 來解釋 鋯二合金的均勻腐蝕加速現象. Zirconium alloys have been used as fuel cladding materials since 1960s. The researches on its corrosion problem have been studied extensively.The recent results for uniform corrosion phenomenon show there is a close relationship among the corrosion kinetics mechanism, the microstructure andpreciputate distribution in oxide scale, and the metal/oxide interface region.Meanwhile, zirconium alloys is used in thhe reactor core, the radiation damageeffect should not be negligible. The aim of this thesis is to understand theeffects of radiation damage on the uniform corrosion behavior of Zircaloy-2,and specically to study the relationship between the microstructure evolutionof oxide film and thhe mechanism of uniform corrosion. Zircaloy-2 plate specimens were irradiated with 2 MeV proton at 350C up toa total dose of 1 dpa. Two uniform corrosion tests were carried out to study the effects of radiation damage on the uniform corrosion behavior of Zircaloy-2The one was performed at 400C and 10.3 MPa superheated steam, and another is conducted at 320C pure water environment. The X-ray power diffraction analysiswas carried out to obtain the phase evolution of oxide film. The plane view specimens of oxide scales were conducted to examine the early stage of corrosion phenomenon. And the observation of cross-sectional TEM specimenswere done to examine the evolution of oxide film, from outer surface to metal/oxide interface. High resolution electron microscope was also used to inspectthe details of oxide scale. The results of uniform corrosion tests show that radiation damage enhancesthe uniform corrosion rate of Zircaloy-2 by 1.5 to 3 times. From X-ray powerdiffraction study, the amount of tetragonal phase was found to increase forthe specimens after proton irradiation tested at 400C, 10.3 MPa steam. It isbelieved that the redistribution of solute atoms into Zircaloy-2 metal matrixunder radiation could be involved. Two type of oxide layers were found: the one in the early stage which isequiaxial with about 200nm in length, and thhe other formed later on is columnar. They are both monoclinic structure. And most of the oxide scale iscomposed of columnar oxide. It was found that the radiation damage dose nothave a clear effect on the equiaxial oxide region. But radiation damage hasan influence on the size of columnar oxide grain ie. the radiation effect induces Zircaloy-2 to form smaller oxide grain. The metal/oxide interfaceis the columnar oxide directly contacts metal matrix, no other interlayeris found. It is believed that the stress induces this oxide morphologyevolution. And there is a relationship between the microstructure evolution of oxide and tranistion. A solute- redistributed model is proposed to explain the uniform corrosionbehavior of Zircaloy-2 under irradiation.