Development of new cladding types for nuclear fuel

被引：2
作者：
Hozer, Z. [1 ]
Novotny, T. [1 ]
Perez-Fero, E. [1 ]
Horvath, M. [1 ]
Csordas, A. Pinter [1 ]
Szabo, P. [1 ]
Illes, L. [1 ]
Schyns, M. [2 ]
Delville, R. [2 ]
Kim, D. [3 ]
Kim, W. J. [3 ]
Sevecek, M. [4 ]
机构：
[1] Ctr Energy Res, Budapest, Hungary
[2] SCK CEN Belgian Nucl Res Ctr, Mol, Belgium
[3] Korea Atom Energy Res Inst, Daejeon, South Korea
[4] Czech Tech Univ, Prague, Czech Republic
来源：
12TH HUNGARIAN CONFERENCE ON MATERIALS SCIENCE (HMSC12) | 2020年 / 903卷
关键词：
HIGH-TEMPERATURE OXIDATION; E110; ZIRCALOY-4;
D O I：
10.1088/1757-899X/903/1/012004
中图分类号：
O646 [电化学、电解、磁化学];
学科分类号：
081704 ;
摘要：
Three different cladding types were tested for nuclear fuel in traditional light water reactors and generation IV gas-cooled fast reactors. Cr coated Zr cladding was tested in steam atmosphere up to 1200 degrees C to demonstrate moderate oxidation and hydrogen production in accident conditions. 15-15Ti stainless steel alloy and SiCf/SiC cladding tube samples were treated in helium atmosphere with different impurities for several hours at 1000 degrees C. Additional mechanical testing and microstructure examinations were carried out with as-received samples and with specimens after high temperature treatments. The experiments results indicated the applicability of the tested materials for reactor conditions in the investigated range of parameters.
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页数：8
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