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Kurata, Yuji; Sato, Hidetomo*; Yokota, Hitoshi*; Suzuki, Tetsuya*
Proceedings of 2nd International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-2), p.177 - 188, 2012/12
Liquid lead-bismuth is a candidate material used for innovative nuclear systems such as accelerator driven reactors and fast reactors. Development of corrosion resistant materials in liquid lead-bismuth is one of important research subjects to realize these systems. In this study, improvement of corrosion properties in lead-bismuth by Al alloy coating was investigated. Aluminum alloys were coated on 316SS using sheets made of Al, Ti and Fe powders, and laser heating. Adjustments of chemical composition in sheets and scanning rate of laser beam enabled us to control Al concentration in coating layers. The corrosion tests were conducted at 550
C for 1,000 hours or 3,000 hours in oxygen-controlled lead-bismuth. As a result of the corrosion tests, it was found that the coating with Al concentration of 5 to 8 mass% showed good corrosion resistance.
Ioka, Ikuo; Suzuki, Jun; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 2nd International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-2), p.391 - 400, 2012/12
Optimization of composition for high Cr-W-Si Ni base alloy has been studied to apply to a nitric acid with high oxidation-reduction potential of advanced reprocessing plants. The corrosion resistance of the Ni base alloy is superior to that of conventional stainless steels. In addition, The Ni base alloy has an excellent resistance of weld crack and ability of plastic deformation caused by extra high purity (EHP) refining technology. However, the Ni base alloy has a technical limitation in hot working and welding for practical use. Several Ni base EHP alloys with different content of Si and W were manufactured to choose an optimum composition range without losing corrosion resistance. High strain rate tensile tests at high temperature, corrosion tests and weldability tests were carried out to examine the optimum composition range of Ni base EHP alloy.
Kim, G.; Shiba, Kiyoyuki; Sawai, Tomotsugu; Ioka, Ikuo; Kiuchi, Kiyoshi; Nakayama, Jumpei*
Proceedings of 2nd International Workshop on Structural Materials for Innovative Nuclear Systems (SMINS-2), p.273 - 279, 2012/12
Obara, Satoshi; Wakai, Takashi; Asayama, Tai; Yamada, Yoshiyuki*; Nakazawa, Takanori*; Yamazaki, Masayoshi*; Hongo, Hiromichi*
no journal, ,
This paper describes the effect of heat treatment on long-term creep properties of 9Cr-W-Mo-V-Nb steel which in the specification of ASME Grade 92 as a part of development of high Cr steel for fast breeder reactor (FBR). The effects of normalizing temperature and tempering temperature and time on long-term creep properties were investigated from the viewpoint of microstructures.
Shibata, Taiju; Sumita, Junya; Takagi, Takashi*; Makita, Taiyo*; Fujita, Ichiro; Sawa, Kazuhiro
no journal, ,
VHTR is one of the Generation-IV reactor systems and being developed internationally. Since the core components in the VHTR will be used at severe temperature condition, it is important to develop heat-resistant ceramic composite material. JAEA is carrying out the application study on two-dimensional (2D-) C/C composite to the control rod of VHTR. 2D-C/C composite of CX-270G grade, graphitized at 2800 C, was used to evaluate the change of material properties by neutron irradiation. The samples were irradiated by JMTR at 600 C up to the neutron fluence of 8.2
10
n/m
(E
1.0MeV) corresponding to 1.2 dpa. The irradiation-induced dimensional change, thermal conductivity, coefficient of thermal expansion and Young's modulus were evaluated. The test results were evaluated in comparison with the irradiation database of graphite of JAEA. It was shown that the graphite database is effective to support the evaluation of the irradiation test results of the graphitized C/C composite.
Donomae, Takako; Maeda, Koji
no journal, ,
Boron carbide is used as a material for control rods in fast reactors because of its superior properties. So far, boron carbide pellet swelling was measured only at low irradiation temperatures up to a burnup of 8010
cap/m
. From this investigation, it was revealed that boron carbide pellet swelling behavior is the same up to a high burnup of 10010cap/m regardless of the difference in irradiation temperatures.
