| NUCLEAR MATERIALS |
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| The Compatibility of Zirconium Hydride in a Sodium Environment |
| FU Xiaogang, ZHANG Jinquan, QIN Bo, MA Haoran, LONG Bin
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| China Institute of Atomic Energy, Beijing 102413 |
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Abstract In the process of fast reactor design and application, the zirconium hydride is a useful neutron moderator for fast reactor. However, there are few reports on the compatibility of the zirconium hydride in sodium at high temperature. And no data are available to support the application at high temperature sodium. As a result, the compatibility of the zirconium hydride in sodium was studied in this work. The ZrH1.8 was tested at 500 ℃, 600 ℃ and 700 ℃ in sodium. Another, some samples prepared with a ZrO2 layer were also tested at 650 ℃ in sodium to evaluate the capabi-lity of preventing hydrogen loss. The test results showed that a ZrO2 layer grew on the surface of zirconium hydride when it was tested at 500 ℃, 600 ℃ and 700 ℃ in sodium. There was no hydrogen emission of ZrH1.8 at 500 ℃. However, the hydrogen content decreased obviously after 120 h at 600 ℃ and 700 ℃. Moreover, a ZrO2 layer prepared for zirconium hydride could not prevent hydrogen loss after exposure at 650 ℃ in sodium. In the process of hydrogen emission at high temperature, due to the opposite movement of hydrogen and oxygen, ZrO2 transformed into Zr3O gradually at the ZrO2/ZrH1.8 interface.
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Published: 21 May 2019
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| About author:: Xiaogang Fu received master’s degree from Department of Materials Science and Engineering in Tsinghua University in July 2009. He is working at China Institute of Atomic Energy as a senior engineer and focusing on research and development of moderator and heat resis-tant steel for fast reactor.Bin Long, Ph.D. of the Ecole Polytechnique Fédérale de Lausanne (EPFL), a professor of China Institute of Atomic Energy(CIAE), focus on the research of nuc-lear reactor materials and liquid metal technology. |
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2019, Vol. 33 
