鈾鋯合金燃料制備技術(shù)研究
發(fā)布時間：2018-08-29 19:50
【摘要】：人類正面臨著能源短缺和化石燃料使用帶來的環(huán)境壓力,核能因其具有能量密度高、不排放溫室氣體等優(yōu)點是解決能源問題的一個重要途徑。目前,核能領(lǐng)域正發(fā)展的聚變-裂變混合堆、快中了堆、ADS能源系統(tǒng)等新型核能系統(tǒng)擬采用金屬型核燃料,而金屬型核燃料在制備和使用過程中存在成分不均勻、輻照腫脹、與包殼材料在高溫發(fā)生反應(yīng),從而導(dǎo)致金屬型核燃料存在長期穩(wěn)定性問題,因此對燃料制備過程中元素擴散、界而反應(yīng)及使用過程中輻照穩(wěn)定性的系統(tǒng)研究非常必要。本論文選擇鈾鋯合金為研究對象,開展的研究內(nèi)容和結(jié)果如下:鈾鋯合金熔煉鑄造技術(shù)和組織結(jié)構(gòu)研究。采用二次合金化技術(shù)真空感應(yīng)熔煉制備了 U-2wt.%Zr、U-4wt.%Zr、U-6wt.%Zr、U-8wt.%Zr、U-10wt.%Zr、U-12wt.%Zr和U-15wt.%Zr等鈾鋯合金,制備的公斤級U-10wt.%Zr合金鋯含量偏差小于0.5%,雜質(zhì)含量低于1000 μg/g。研究表明,鑄態(tài)鈾鋯合金即使Zr含量很低,也是由過飽和固溶的α-U相和δ-UZr2相組成,即所有富鈾的鑄態(tài)鈾鋯合金都為雙相組織。鈾鋯合金在鑄造冷卻過程中,首先發(fā)生γ→γ1+γ2調(diào)幅分解,然后γ1相和γ2相分別發(fā)生固態(tài)反應(yīng),其中γ1相發(fā)生了 γ1→α-U馬氏體相變,γ2相則先通過晶面坍塌形成ω相,然后有序化形成δ-UZr2相。TEM研究表明,α-U和δ-UZr2之間共格,位向關(guān)系為(010)[001]_α//(0(?)10)[(?)110]_δ。多孔U-10wt.%Zr合金粉末冶金制備技術(shù)研究。U-10wt.%Zr合金中的孔隙在使用過程中可以容納裂變產(chǎn)物,本論文采用氫化-去氫化法制備U-10wt.%Zr合金粉末,通過冷等靜壓制備坯料,真空燒結(jié)制備了不同孔隙率的多孔U-10wt.%Zr合金。氫化-去氫化法可以得到不被氧化、粒徑可控的U-10wt.%Zr合金粉末,U-10wt.%Zr合金粉末冷等靜壓坯料的密度對燒結(jié)后致密度影響不大,通過控制壓坯真空燒結(jié)的時間和溫度可以得到相對密度為70～91%的多孔鈾鋯合金。孔隙度對U-10wt.%Zr合金熱導(dǎo)率的影響符合Loeb關(guān)系。Zr-4合金包覆U-10wt.%Zr合金技術(shù)研究。采用真空熱壓擴散法和熱等靜壓法應(yīng)用Zr-4合金包覆U-10wt.%Zr合金。真空熱壓擴散實驗表明,為了得到U-10wt.%Zr合金燃料與包殼材料Zr-4合金結(jié)合良好的界面,應(yīng)精確控制熱壓溫度和時間,避免在界面處形成過多的UZr2金屬間化合物。α-U基體與δ-UZrr共共格,UZr2在靠近Zr-4合金界面鈾元素富集,UZr_2與Zr-4合金共格。熱等靜壓包覆結(jié)果表明,合金元素體擴散受到各向相等壓力的抑制,激活能明顯增大,擴散速率降低,而晶界擴散受到的影響較小,鈾在Zr-4合金中擴散表現(xiàn)為晶界擇優(yōu)擴散。在U-10wt.%Zr/Zr-4合金界面生成了 UZr2金屬問化合物層,在U-10wt.%Zr合金與UZr_2層之間生產(chǎn)了純鋯層。U-10wt.%Zr與Zr-4合金之間相容性研究。U-10wt.%Zr合金和Zr-4合金之間的界面擴散行為受擴散速率控制,U元素向Zr-4合金的擴散速率相對較快,擴散層生長厚度與時間的關(guān)系符合拋物線規(guī)律。U-10wt.%Zr/Zr-4合金體系在800℃～1100℃C溫區(qū)的互擴散系數(shù)為1× 10~(-15)～1 × 10~(-13) m~2/s,在鋯含量小于40 at.%時擴散系數(shù)相對較小。在380℃～1100℃溫度區(qū)間,U-10wt.%Zr/Zr-4合金界面擴散層生長行為分為三個區(qū)間,分別受不同的擴散生長機制控制。在380℃～600℃溫區(qū),受到鈾元素在δ-UZr_2相內(nèi)擴散的控制,生長速率常數(shù)為9.85×10~4m~2/s,激活能為340.93kJ/mol。在650℃～800℃溫區(qū),U原子在α-Zr中以置換機制擴散,擴散層生長受此機制控制,生長速率常數(shù)為4.10× 10~(-4) m~2/s,激活能為206.33 kJ/mol;在900℃～1100℃溫區(qū),元素在Y(U,Zr)固溶體中以空位機制擴散,生長速率常數(shù)為3.43×10-6 m~2/s,激活能為158.50 kJ/mol。實驗結(jié)果表明,在溫度小于500℃時U-10wt.%Zr合金燃料與Zr-4合金之間相容性良好。
[Abstract]:Nowadays, the fusion-fission hybrid reactor, fast neutralization reactor, ADS energy system and other new nuclear energy systems are planned to adopt metal. Type I nuclear fuel, while metal-type nuclear fuel in the preparation and use of the process of uneven composition, irradiation swelling, and cladding materials at high temperatures, resulting in metal-type nuclear fuel long-term stability problems, so in the process of fuel preparation element diffusion, boundary reaction and irradiation stability in the use of systematic research non- The research contents and results are as follows: Uranium-zirconium alloy melting and casting technology and microstructure research. U-2wt.% Zr, U-4wt.% Zr, U-6wt.% Zr, U-8wt.% Zr, U-10wt.% Zr, U-12wt.% Zr and U-15wt.% Zr were prepared by vacuum induction melting with secondary alloying technology. Zirconium content deviation of U-10wt.% Zr alloy is less than 0.5% and impurity content is less than 1000 ug/g. The results show that as-cast Uranium-Zirconium alloy is composed of supersaturated solid solution of a-U phase and delta-U Zr 2 phase even though the content of Zr is very low, that is to say, all as-cast Uranium-Zirconium alloys rich in uranium are biphase structure. Amplitude modulation decomposition followed by solid state reaction of gamma-1 phase and gamma-2 phase respectively, in which gamma-1 phase undergoes gamma-1_a-U martensitic transformation, and gamma-2 phase first forms_phase through the collapse of crystal plane, and then forms delta-U Zr-2 phase orderly. TEM studies show that there is a coherent relationship between a-U and delta-U Zr-2, and the phase orientation is (010) [001]_a/(0(?) 10) [(?) 110]_delta.porous U-10wt.% Zr alloy powder. In this paper, U-10wt.% Zr alloy powders were prepared by hydrogenation-dehydrogenation method. The billets were prepared by cold isostatic pressing, and the porous U-10wt.% Zr alloys with different porosity were prepared by vacuum sintering. Porous U-10wt.% Zr alloy powders with controllable particle size and cold isostatic pressing billets of U-10wt.% Zr alloy powders have little effect on the density after sintering. Porous U-Zr alloy with relative density of 70-91% can be obtained by controlling the time and temperature of vacuum sintering. The effect of porosity on the thermal conductivity of U-10wt.% Zr alloy is in accordance with Loeb relationship. U-10wt.% Zr alloy was coated with Zr-4 alloy by vacuum hot-pressing diffusion method and hot isostatic pressing method. The vacuum hot-pressing diffusion experiment showed that in order to obtain a good interface between U-10wt.% Zr alloy fuel and cladding material Zr-4 alloy, the hot-pressing temperature and time should be accurately controlled to avoid excessive formation at the interface. UZr2 intermetallic compound. Alpha-U matrix and delta-UZrr coincide, UZr2 enriches near the Zr-4 alloy interface, and UZr2 and Zr-4 alloy coincide. The results of HIP coating show that the diffusion of alloy elements is inhibited by isobaric pressure, the activation energy increases obviously, the diffusion rate decreases, but the diffusion of uranium at the Zr-4 alloy boundary is less affected. UZr2 intermetallic compound layer was formed at the interface of U-10wt.% Zr / Zr-4 alloy, and pure zirconium layer was produced between U-10wt.% Zr alloy and UZr_2 layer. The interdiffusion coefficient of U-10wt.% Zr / Zr-4 alloy system is 1 (-15) (-13) m~2/s at 800 ~1100 6550 The growth behavior of the interfacial diffusion layer of% Zr/Zr-4 alloy can be divided into three regions, which are controlled by different diffusion growth mechanisms. The growth rate constant is 9.85 *10~4 m~2/s and the activation energy is 340.93 kJ/mol. In the temperature range of 650 ~800 C, the U atom is expanded by substitution mechanism in a-Zr at the temperature range of 380 600 C. The growth rate constant is 4.10 (-4) m 2/s and the activation energy is 206.33 kJ/mol. In the temperature range of 900 ~1100, the elements diffuse by vacancy mechanism in Y (U, Zr) solid solution. The growth rate constant is 3.43 (-6) m 2/s and the activation energy is 158.50 kJ/mol. The compatibility between alloy fuel and Zr-4 alloy is good.
【學(xué)位授予單位】：中國科學(xué)技術(shù)大學(xué)
【學(xué)位級別】：博士
【學(xué)位授予年份】：2017
【分類號】：TL21
本文編號：2212239
[Abstract]:Nowadays, the fusion-fission hybrid reactor, fast neutralization reactor, ADS energy system and other new nuclear energy systems are planned to adopt metal. Type I nuclear fuel, while metal-type nuclear fuel in the preparation and use of the process of uneven composition, irradiation swelling, and cladding materials at high temperatures, resulting in metal-type nuclear fuel long-term stability problems, so in the process of fuel preparation element diffusion, boundary reaction and irradiation stability in the use of systematic research non- The research contents and results are as follows: Uranium-zirconium alloy melting and casting technology and microstructure research. U-2wt.% Zr, U-4wt.% Zr, U-6wt.% Zr, U-8wt.% Zr, U-10wt.% Zr, U-12wt.% Zr and U-15wt.% Zr were prepared by vacuum induction melting with secondary alloying technology. Zirconium content deviation of U-10wt.% Zr alloy is less than 0.5% and impurity content is less than 1000 ug/g. The results show that as-cast Uranium-Zirconium alloy is composed of supersaturated solid solution of a-U phase and delta-U Zr 2 phase even though the content of Zr is very low, that is to say, all as-cast Uranium-Zirconium alloys rich in uranium are biphase structure. Amplitude modulation decomposition followed by solid state reaction of gamma-1 phase and gamma-2 phase respectively, in which gamma-1 phase undergoes gamma-1_a-U martensitic transformation, and gamma-2 phase first forms_phase through the collapse of crystal plane, and then forms delta-U Zr-2 phase orderly. TEM studies show that there is a coherent relationship between a-U and delta-U Zr-2, and the phase orientation is (010) [001]_a/(0(?) 10) [(?) 110]_delta.porous U-10wt.% Zr alloy powder. In this paper, U-10wt.% Zr alloy powders were prepared by hydrogenation-dehydrogenation method. The billets were prepared by cold isostatic pressing, and the porous U-10wt.% Zr alloys with different porosity were prepared by vacuum sintering. Porous U-10wt.% Zr alloy powders with controllable particle size and cold isostatic pressing billets of U-10wt.% Zr alloy powders have little effect on the density after sintering. Porous U-Zr alloy with relative density of 70-91% can be obtained by controlling the time and temperature of vacuum sintering. The effect of porosity on the thermal conductivity of U-10wt.% Zr alloy is in accordance with Loeb relationship. U-10wt.% Zr alloy was coated with Zr-4 alloy by vacuum hot-pressing diffusion method and hot isostatic pressing method. The vacuum hot-pressing diffusion experiment showed that in order to obtain a good interface between U-10wt.% Zr alloy fuel and cladding material Zr-4 alloy, the hot-pressing temperature and time should be accurately controlled to avoid excessive formation at the interface. UZr2 intermetallic compound. Alpha-U matrix and delta-UZrr coincide, UZr2 enriches near the Zr-4 alloy interface, and UZr2 and Zr-4 alloy coincide. The results of HIP coating show that the diffusion of alloy elements is inhibited by isobaric pressure, the activation energy increases obviously, the diffusion rate decreases, but the diffusion of uranium at the Zr-4 alloy boundary is less affected. UZr2 intermetallic compound layer was formed at the interface of U-10wt.% Zr / Zr-4 alloy, and pure zirconium layer was produced between U-10wt.% Zr alloy and UZr_2 layer. The interdiffusion coefficient of U-10wt.% Zr / Zr-4 alloy system is 1 (-15) (-13) m~2/s at 800 ~1100 6550 The growth behavior of the interfacial diffusion layer of% Zr/Zr-4 alloy can be divided into three regions, which are controlled by different diffusion growth mechanisms. The growth rate constant is 9.85 *10~4 m~2/s and the activation energy is 340.93 kJ/mol. In the temperature range of 650 ~800 C, the U atom is expanded by substitution mechanism in a-Zr at the temperature range of 380 600 C. The growth rate constant is 4.10 (-4) m 2/s and the activation energy is 206.33 kJ/mol. In the temperature range of 900 ~1100, the elements diffuse by vacancy mechanism in Y (U, Zr) solid solution. The growth rate constant is 3.43 (-6) m 2/s and the activation energy is 158.50 kJ/mol. The compatibility between alloy fuel and Zr-4 alloy is good.
【學(xué)位授予單位】：中國科學(xué)技術(shù)大學(xué)
【學(xué)位級別】：博士
【學(xué)位授予年份】：2017
【分類號】：TL21
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