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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,例如第一墙壁屏蔽模块或迭代融合反应器的分流器盒。通过从大锻件加工的四分之一模块的焊接相比,铸件可以节省重大成本。然而,这种浇铸部件的强度性质通常被认为是常规锻造和退火组分的那些。为了改善和验证铸造不锈钢作为锻造不锈钢的替代品,开始开发和测试程序,利用氮和锰添加来促进改进的性能。本文侧重于第一组发育合金对中子辐照和易感性对应力腐蚀开裂的响应。这些铸造材料还可在轻水反应器中具有不同组分的应用。结果表明,所有钢都表现出照射诱导的硬化和延展性的相应下降,尽管辐照样品中仍存在相当大的延展性。与锻造的参考钢相比,铸钢均表现出降低的硬化,这可能与较大的晶粒尺寸有关。更高的氮含量不会影响辐照性能。关于应力腐蚀裂缝敏感性,晶粒尺寸的较大差异限制了锻造材料之间的比较,以及参考和归档铸造合金测试中的夹杂物进行复杂分析这些样品。结果表明,辐照的归档热比改性合金更容易破裂,这可能与更复杂的微电池有关。此外,结果表明,改性铸钢至少作为锻造316Ln的SCC抗性。氮对合金的机械性能的有益效果在照射后残留,并且对SCC抗性并不有害。

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