Assessment of High Chromium Oxide Dispersion Strengthened Steels for Nuclear Applications

I.S. Kim, J.S. Lee (Sp), C.H. Jang, Center for Advanced Reactor Research / KAIST, Daejeon (Korea, Republic); A. Kimura, Kyoto Universtiy (Japan) 
The use of oxide dispersion strengthened (ODS) steels in fusion and advanced fission reactors requires a high resistance of thermal aging and hydrogen embrittlement. In this work, the effects of thermal aging and hydrogen on the mechanical properties of ODS steels were investigated. The materials used were five kinds of ODS steels, produced by changing Cr contents from 13 to 22 weight percent with keeping yittria contents within 0.36-0.38 weight percent. Small punch (SP) and tensile specimens were sampled from the extruded rod in such way that the axis direction is parallel to longitudinal(L) or transverse(T)-direction with respective to the extruded direction. Specimens were thermally aged at 693 K for 322 hours. Hydrogen was cathodically charged into the specimens at room temperature for 30 min and during tensile testing. Tensile and SP tests were conducted at room temperature and from 293 to 77 K, respectively. Ductile to brittle transition behavior of SP tests was strongly affected by not only the Cr and Al contents, but the specimen sampling direction. Furthermore, after thermal aging treatment, SP-ductile to brittle transition temperature (SP-DBTT) and microhardness was significantly increased as a function of Cr contents. The shift of SP-DBTT was approximately 10 and 73K in 13 and 19 Cr ODS steel, respectively. Based on the TEM observation, the hardening is thought to be associated with the precipitation of high density fine chromium-rich ferrite. In addition, hydrogen charging resulted in the considerable loss of ductility accompanied by a change in fracture mode from ductile to intergranular and quasi-cleavage fracture.

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