铀锆合金燃料制备技术研究
发布时间：2018-08-29 19:50
【摘要】：人类正面临着能源短缺和化石燃料使用带来的环境压力,核能因其具有能量密度高、不排放温室气体等优点是解决能源问题的一个重要途径。目前,核能领域正发展的聚变-裂变混合堆、快中了堆、ADS能源系统等新型核能系统拟采用金属型核燃料,而金属型核燃料在制备和使用过程中存在成分不均匀、辐照肿胀、与包壳材料在高温发生反应,从而导致金属型核燃料存在长期稳定性问题,因此对燃料制备过程中元素扩散、界而反应及使用过程中辐照稳定性的系统研究非常必要。本论文选择铀锆合金为研究对象,开展的研究内容和结果如下:铀锆合金熔炼铸造技术和组织结构研究。采用二次合金化技术真空感应熔炼制备了 U-2wt.%Zr、U-4wt.%Zr、U-6wt.%Zr、U-8wt.%Zr、U-10wt.%Zr、U-12wt.%Zr和U-15wt.%Zr等铀锆合金,制备的公斤级U-10wt.%Zr合金锆含量偏差小于0.5%,杂质含量低于1000 μg/g。研究表明,铸态铀锆合金即使Zr含量很低,也是由过饱和固溶的α-U相和δ-UZr2相组成,即所有富铀的铸态铀锆合金都为双相组织。铀锆合金在铸造冷却过程中,首先发生γ→γ1+γ2调幅分解,然后γ1相和γ2相分别发生固态反应,其中γ1相发生了 γ1→α-U马氏体相变,γ2相则先通过晶面坍塌形成ω相,然后有序化形成δ-UZr2相。TEM研究表明,α-U和δ-UZr2之间共格,位向关系为(010)[001]_α//(0(?)10)[(?)110]_δ。多孔U-10wt.%Zr合金粉末冶金制备技术研究。U-10wt.%Zr合金中的孔隙在使用过程中可以容纳裂变产物,本论文采用氢化-去氢化法制备U-10wt.%Zr合金粉末,通过冷等静压制备坯料,真空烧结制备了不同孔隙率的多孔U-10wt.%Zr合金。氢化-去氢化法可以得到不被氧化、粒径可控的U-10wt.%Zr合金粉末,U-10wt.%Zr合金粉末冷等静压坯料的密度对烧结后致密度影响不大,通过控制压坯真空烧结的时间和温度可以得到相对密度为70～91%的多孔铀锆合金。孔隙度对U-10wt.%Zr合金热导率的影响符合Loeb关系。Zr-4合金包覆U-10wt.%Zr合金技术研究。采用真空热压扩散法和热等静压法应用Zr-4合金包覆U-10wt.%Zr合金。真空热压扩散实验表明,为了得到U-10wt.%Zr合金燃料与包壳材料Zr-4合金结合良好的界面,应精确控制热压温度和时间,避免在界面处形成过多的UZr2金属间化合物。α-U基体与δ-UZrr共共格,UZr2在靠近Zr-4合金界面铀元素富集,UZr_2与Zr-4合金共格。热等静压包覆结果表明,合金元素体扩散受到各向相等压力的抑制,激活能明显增大,扩散速率降低,而晶界扩散受到的影响较小,铀在Zr-4合金中扩散表现为晶界择优扩散。在U-10wt.%Zr/Zr-4合金界面生成了 UZr2金属问化合物层,在U-10wt.%Zr合金与UZr_2层之间生产了纯锆层。U-10wt.%Zr与Zr-4合金之间相容性研究。U-10wt.%Zr合金和Zr-4合金之间的界面扩散行为受扩散速率控制,U元素向Zr-4合金的扩散速率相对较快,扩散层生长厚度与时间的关系符合抛物线规律。U-10wt.%Zr/Zr-4合金体系在800℃～1100℃C温区的互扩散系数为1× 10~(-15)～1 × 10~(-13) m~2/s,在锆含量小于40 at.%时扩散系数相对较小。在380℃～1100℃温度区间,U-10wt.%Zr/Zr-4合金界面扩散层生长行为分为三个区间,分别受不同的扩散生长机制控制。在380℃～600℃温区,受到铀元素在δ-UZr_2相内扩散的控制,生长速率常数为9.85×10~4m~2/s,激活能为340.93kJ/mol。在650℃～800℃温区,U原子在α-Zr中以置换机制扩散,扩散层生长受此机制控制,生长速率常数为4.10× 10~(-4) m~2/s,激活能为206.33 kJ/mol;在900℃～1100℃温区,元素在Y(U,Zr)固溶体中以空位机制扩散,生长速率常数为3.43×10-6 m~2/s,激活能为158.50 kJ/mol。实验结果表明,在温度小于500℃时U-10wt.%Zr合金燃料与Zr-4合金之间相容性良好。
[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.
【学位授予单位】：中国科学技术大学
【学位级别】：博士
【学位授予年份】：2017
【分类号】：TL21
本文编号：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.
【学位授予单位】：中国科学技术大学
【学位级别】：博士
【学位授予年份】：2017
【分类号】：TL21
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