MATERIAL DEVELOPMENT FOR SUPERCRITICAL WATER-COOLED REACTOR
The 5th International Symposium on SuperCritical Water-cooled Reactors - 2011 March 13-16


Presented at:
The 5th International Symposium on SuperCritical Water-cooled Reactors
2011 March 13-16
Location:
Vancouver,Canada
Session Title:
Materials I (General)

Authors:
Junya Kaneda (Hitachi Limited)
Shigeki Kasahara (Hitachi Limited)
Fumihisa Kano (Toshiba Corporation)
Norihisa Saito (Toshiba Corporation)
Tatsuo Shikama (Tohoku University)
Hideki Matsui (Kyoto University)
  

Abstract

Material properties of candidate alloys, including ferritic/martensitic steels, austenitic stainless steels, Ni based alloys, Ti alloys have been investigated to evaluate their applicability for fuel claddings of the supercritical water-cooled reactor (SCWR) in Japanese national projects since 2000. Modified austenitic stainless steels such as fine grain materials and Zr-modified materials have also been investigated. In the projects, high temperature tensile tests, creep tests, general corrosion tests, stress corrosion cracking (SCC) tests, and manufacturability tests were conducted under the predicted conditions in the SCWR. The following results were obtained from the tests. Several of the test materials satisfied the tensile properties required for the fuel cladding design. Requirement of the creep rupture lifetime is 50,000 h under 30 – 50 MPa at 700˚C. Several materials such as alloy 625, type 310S and Zr-modified type 316L and type 310S were expected to meet the design requirements. Ni based alloys, type 310S, Zr-modified type 310S and the fine grain materials had good general corrosion resistance. SCC did not appear in any test materials except type 304, 310S, Ti alloy and Alloy 625. Austenitic stainless steels could be manufactured by the process proven for the seamless tube with about 4.5 mm in outside diameter. Alloy 690 and the fine grain material could be manufactured with a minor process modification as well. According to the calculation of He yield, He embrittlement is a serious concern for Ni-based alloys because of their the high Ni content. As a result of the comprehensive evaluation, Zr-modified type 310S is the first candidate alloy, and Ti-modified type 310S is the second candidate alloy for the JSCWR fuel rod. Considering the standard of mechanical design, the existing material is favorable to be considered for the power plants. Thus, type 310S with reduced carbon content within the alloy specification is recommended as the second candidate material.

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