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In-Pile Thermal Conductivity Measurement of Uranium-Zirconium Hydride Fuel

机译:铀-锆氢化物燃料的堆中热导率测量

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摘要

The focus of this manuscript is to discuss in-pilernexperimental measurements made to deduce thermalrnconductivity of uranium-zirconium hydride fuels. Thernmethod utilized here relies on in-situ measurements ofrncenterline fuel and outside cladding temperature alongrnwith a calculated value for the power generation raterninside the fuel to determine the thermal conductivity.
机译:该手稿的重点是讨论为推断铀-锆氢化物燃料的导热性而进行的实验性内部测量。这里使用的方法依赖于对中心线燃料和外包层温度的原位测量,以及在燃料内部的发电率的计算值以确定导热率。

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  • 来源
    《Transactions of the American nuclear society 》 |2012年第6期| p.1303-1304| 共2页
  • 作者单位

    Department of Nuclear Engineering, University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720 Oak Ridge National Laboratory, Oak Ridge, TN, 37831;

    Department of Nuclear Engineering, University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720;

    Massachusetts Institute of Technology Nuclear Reactor Laboratory, 138 Albany St., Cambridge, MA 02139;

    Massachusetts Institute of Technology Nuclear Reactor Laboratory, 138 Albany St., Cambridge, MA 02139;

    Massachusetts Institute of Technology Nuclear Reactor Laboratory, 138 Albany St., Cambridge, MA 02139;

    Idaho National Laboratory, Idaho Falls, ID, 83415;

    epartment of Nuclear Engineering, University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720;

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