Oxidation behaviors of wrought nickel-based superalloys in various high temperature environments

Oxidation characteristics of Alloy 617 and Haynes 230 at 900 oC in simulated helium environment,hot steam environment containing H2 as well as in air and pure helium conditions were investigated.Compared to air condition,the oxidation rate of Alloy 617 was not significantly affected in helium and hot steam environments,while Haynes 230 showed lower oxidation rate in helium environment.On the other hand,the oxide morphology and structure of Alloy 617 were strongly affected by the environments,but those of Haynes 230 were less dependent on the environments.For Haynes 230,a Cr2O3 inner layer and a protective MnCr2O4 outer layer were formed in all environments,which contributed to the better oxidation resistance.As the mechanical properties,such as creep and tensile properties,were significantly affected by the oxidation behaviors,surface treatment methods to enhance oxidation resistance of these alloys should be developed.

Research on Material Irradiation in HANARO

HANARO (High-Flux Advanced Neutron Application Reactor) is a 30MWth multi-purpose research reactor generating a very high neutron flux for scientific and technical applications. Since its initial criticality in Feb. 1995, it has worked as a national neutron source for neutron beam research, RI production and related research, material and fuel irradiation research, and neutron transmutation doping. The installation of a cold neutron source and related beam instrument makes HANARO serve as a regional source for beam research as well. As Korea is developing nuclear power systems and fuels, irradiation research is a main utilization subject in HANARO, and irradiation facilities have been actively utilized for various nuclear fuel and material irradiation tests requested by users from research institutes, universities, and industries. Most of the irradiation tests were related to the R & D relevant to the ageing management and safety evaluation of the present nuclear power reactors. Korea is conducting R & D programs relevant to new nuclear systems including a research reactor, System-integrated Modular Advanced Reactor (SMART), and future nuclear systems such as a very high temperature reactor system (VHTR) and sodium cooled fast reactor system (SFR). To effectively support the R&D for new nuclear systems, advanced irradiation technologies such as high-temperature irradiation, instrumentation, and long-term irradiation are being developed. In addition, research on the irradiation characteristics of super-conductor materials and new chemical materials is being conducted as a part of fundamental research.

第4代核能系统研发介绍

介绍国际核能界正在着手开发的第4代核能系统的背景和要达到的目标。第4代核能系统国际论坛根据第4代核能系统目标选择了6种最有希望的概念进行研发,文章简要介绍了它们的技术特点、主要设计参数以及技术上尚待解决的问题。

确定GH3128高温拉伸性能设计许用值的方法

国产高温镍基合金GH3128在高温条件下具有良好的强度,有望成为超高温气冷堆(VHTR)供热和发电利用的中间换热器(IHX)的主要结构材料。该文针对美国机械工程师协会(ASME)高温强度许用值设计标准应用于GH3128时无法给出清晰一致有效设计余量的不足,采用正态分布直接计算的方法给出了容限范围母体比例为99%的设计许用值曲线,并同ASME设计标准给出的设计许用值曲线进行了对比。结果表明:容限范围母体比例为99%的设计许用值曲线可以在整个温度区间连续给出可靠的设计许用值以及一致有效的设计余量,可以作为高温气冷堆中间换热器结构设计的参考依据。

GH3128高温拉伸强度设计方法的优化

国产镍基高温合金GH3128在高温下具有良好的强度,有望成为超高温气冷堆(very high temperature reactor,VHTR)核心部件中间换热器(intermediate heat exchanger,IHX)的主要结构材料。该文针对美国机械工程师协会(American Society of Mechanical Engineers,ASME)标准应用于GH3128时无法在全温度段得到一致可靠设计余量的不足,采用正态分布直接计算的方法,通过整体多项式、分段多项式以及分段指数这3种拟合方式,对拉伸性能设计许用值进行了计算。分析结果表明:针对现有的GH3128拉伸性能数据,应分别采用分段多项式拟合及分段指数拟合来求抗拉强度和屈服强度的设计许用值,并且这种方法给出的设计许用值曲线可以在全温度区间给出一致可靠的设计余量,可作为超高温气冷堆中间换热器结构设计的参考依据。