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Stress Corrosion Cracking Behavior of Cast Stainless Steels

机译:铸造不锈钢的应力腐蚀开裂行为

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Casting of austenitic stainless steels offers the possibility of directly producing large and/or relatively complex structures, such as the first wall shield modules or the diverter cassette for the ITER fusion reactor. Casting offers major cost savings when compared to fabrication via welding of quarter modules machined from large forgings. However, the strength properties of such cast components are typically considered inferior to those of conventionally forged and annealed components. To improve and validate cast stainless steel as a substitute for wrought stainless steel, a development and testing program was initiated, utilizing nitrogen and manganese additions to promote improved performance. This paper focuses on the response of the first set of developmental alloys to neutron-irradiation and susceptibility to stress corrosion cracking. These cast materials may also have applications for different components in light water reactors. Results showed that all steels exhibited irradiation-induced hardening and a corresponding drop in ductility, as expected, although there is still considerable ductility in the irradiated samples. The cast steels all exhibited reduced hardening in comparison to a wrought reference steels, which may be related to a larger grain size. Higher nitrogen contents did not negatively influence irradiation performance. Regarding stress corrosion cracking susceptibility, the large difference in grain size limits the comparison between wrought and cast materials, and inclusions in a reference and archive cast alloy tests complicate analysis of these samples. Results suggest that the irradiated archive heat was more susceptible to cracking than the modified alloys, which may be related to the more complex microstrueture. Further, the results suggest that the modified cast steel is at least as SCC resistant as wrought 316LN. The beneficial effect of nitrogen on the mechanical properties of the alloys remains after irradiation and is not detrimental to SCC resistance.
机译:奥氏体不锈钢的铸造提供了直接生产大型和/或相对复杂的结构的可能性,例如用于ITER聚变反应堆的第一壁屏蔽模块或分流器盒。与通过焊接大型锻件加工的四分之一模块进行制造相比,铸造可节省大量成本。但是,通常认为这种铸造部件的强度性能不如常规锻造和退火的部件。为了改进和验证铸造不锈钢替代锻造不锈钢,已启动开发和测试计划,利用氮和锰的添加来提高性能。本文着重研究第一类发育合金对中子辐照的响应以及对应力腐蚀开裂的敏感性。这些铸造材料还可用于轻水反应堆中的不同组件。结果表明,尽管在被辐照的样品中仍然具有相当大的延展性,但是所有钢都表现出了辐照引起的硬化,并且延展性相应降低。与可锻参比钢相比,铸钢均显示出降低的硬化,这可能与更大的晶粒尺寸有关。较高的氮含量不会对辐照性能产生负面影响。关于应力腐蚀开裂敏感性,晶粒尺寸的巨大差异限制了锻造材料与铸造材料之间的比较,参考和归档铸造合金测试中的夹杂物使这些样品的分析变得复杂。结果表明,辐照的存档热比改性合金更易于破裂,这可能与更复杂的微观结构有关。此外,结果表明,改性铸钢至少具有与锻造316LN一样的耐SCC性。氮对合金力学性能的有益影响在辐照后仍然存在,并且对耐SCC性无害。

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