The corrosion of steels in liquid metal lead （Pb） and bismuth （Bi） is a critical challenge in the development of accel-erator driven systems （ADS）. Using a first-principles method with a slab model, we theoretically investigate the interaction between the Pb （Bi） atom and the iron （Fe） （100） surface to assess the fundamental corrosion properties. Our investigation demonstrates that both Pb and Bi atoms favorably adsorb on the （100） surface. Such an adsorption decreases the energy required for the dissociation of an Fe atom from the surface, enhancing the dissolution tendency significantly. The seg- regation of six common alloying elements （Cr, A1, Mn, Ni, Nb, and Si） to the surface and their impacts on the corrosion properties are also considered. The present results reveal that Si seems to have a relatively good performance to stabilize the surface and alleviate the dissolving trend caused by Pb and Bi.
Effects of Si^3＋ and H＋ irradiation on tungsten were investigated by internal friction （IF） technique. Scanning electron microscope （SEM） analysis revealed that sequential dual Si＋H irradiation resulted in more serious damage than single Si irradiation. After irradiation, the IF background was significantly enhanced. Besides, two obvious IF peaks were initially found in tem- perature range of 70～330 K in the sequential Si＋H irradiated tungsten sample. The mechanism of increased IF background for the irradiated samples was suggested to originate from the high density dislocations induced by ion irradiation. On the other hand, the relaxation peak PL and non-relaxation peak PH in the Si＋H irradiated sample were ascribed to the interaction process of hydrogen atoms with mobile dislocations and transient processes of hydrogen redistribution, respectively. The obtained experimental results verified the high sensitivity of IF method on the irradiation damage behaviors in nuclear materials.
HU JingZHANG YanwenWANG XianpingZHAO ZiqiangFANG QianfengLIU Changsong
We investigate the segregation behavior of alloying atoms （Sr, Th, In, Cd, Ag, Sc, Au, Zn, Cu, Mn, Cr, and Ti） near Z3 （ 111 ） [1]-0] tilt symmetric grain boundary （GB） in tungsten and their effects on the intergranular embrittlement by performing first-principles calculations. The calculated segregation energies suggest that Ag, Au, Cd, In, Sc, Sr, Th, and Ti prefer to occupy the site in the mirror plane of the GB, while Cu, Cr, Mn, and Zn intend to locate at the first layer nearby the GB core. The calculated strengthening energies predict Sr, Th, In, Cd, Ag, Sc, Au, Ti, and Zn act as embrittlers while Cu, Cr, and Mn act as cohesion enhancers. The correlation of the alloying atom＇s metal radius with strengthening energy is strong enough to predict the strengthening and embrittling behavior of alloying atoms; that is, the alloying atom with larger metal radius than W acts as an embrittler and the one with smaller metal radius acts as a cohesion enhancer.
首先简要介绍第一代到先进的第四代核能系统的发展、与核能系统发展密切的抗辐照结构材料研发进展、第四代核能系统结构材料辐照性能研究新方法。第四代核能系统发展中,辐照引起材料性能退化是一个需要研究和解决的瓶颈问题。现有中子源都不能满足第四代核能系统结构材料高剂量中子辐照性能研究的要求。为此,发展了用于核能系统结构材料高剂量辐照性能快速检测加速器重离子辐照方法和第四代核能系统实际辐照工况模拟的重离子与氢和氦三束同时辐照新方法,文中进行了详细的介绍。最后介绍了中国原子能科学研究院核能系统结构材料辐照性能研究现状和近期发展计划。该院在HI-13串列加速器器上建立了多种不同用途的重离子辐照装置、三个独立加速器构成的重离子与氢和氦三束同时辐照实验平台,开展了一系列核能结构材料,例如国产改进型奥氏体钢、CLAM钢、1515钢、钽、钨等的辐照性能的系统测试和研究。为了更好地开展核能结构材料性能研究,从国外引进了一台超导直线加速器和一台可变能量重离子回旋加速器。结合现有2×13 Me V,2×1.7 MV串列加速器、30 Me V和100 Me V质子回旋加速器、高压倍加器,中国实验快堆、中国先进研究堆、微堆等,CIAE将建成一个比较完整和先进的核能系统结构材料辐照实验平台系统供国内外用户使用。
