UMo/Zr单片式燃料板堆内热力耦合行为研究
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摘要
UMo/Zr单片式燃料板在堆内辐照环境下会经历复杂的多场耦合及多尺度关联的行为。针对均匀辐照的堆内工况条件,建立了对UMo/Zr单片式燃料板的堆内行为进行多尺度模拟的方法,并计算分析了元件的温度场、变形和主要应力场随燃耗演化的规律,获得了芯体与包壳界面层间应力的分布与演化规律。研究结果表明,芯体的最高温度会随着辐照时间持续增长;芯体厚度随着辐照时间而增大,在靠近芯体的边界附近厚度增长较多,与辐照后相关检测结果相符;芯体的Mises应力要远小于包壳中的Mises应力;芯体和包壳界面正应力最大值位于靠近芯体角部的位置,界面角部区域较大的界面拉应力易导致此处首先产生界面破坏。
UMo/Zr monolithic fuel plates experience complex multi-field-coupling and multi-scale behavior during the in-pile radiation.In this study,the method of multi-scale simulation of the in-pile behavior in the UMo/Zr monolithic fuel plate under a homogeneous irradiation condition is built.The distributions and evolutions of the temperature field,the strain field and stress field,and the interfacial stress between the cladding and the fuel meat are calculated and analyzed.The research results indicate that the peak temperature increases with the irradiation time,the thickness of the plate increases with the irradiation time,and the peak value occurs near the edge of the fuel meat; the Mises stress in the fuel meat is far less than that in the cladding; and the maximum interfacial normal stress is located near the corner of the fuel meat,where the tensile stress may cause interfacial failure.
UMo/Zr monolithic fuel plates experience complex multi-field-coupling and multi-scale behavior during the in-pile radiation.In this study,the method of multi-scale simulation of the in-pile behavior in the UMo/Zr monolithic fuel plate under a homogeneous irradiation condition is built.The distributions and evolutions of the temperature field,the strain field and stress field,and the interfacial stress between the cladding and the fuel meat are calculated and analyzed.The research results indicate that the peak temperature increases with the irradiation time,the thickness of the plate increases with the irradiation time,and the peak value occurs near the edge of the fuel meat; the Mises stress in the fuel meat is far less than that in the cladding; and the maximum interfacial normal stress is located near the corner of the fuel meat,where the tensile stress may cause interfacial failure.
引文
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[4]Lopes D A,Zimmermann A J O,Silva S L,et al.Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel[J].Nuclear Technology,2016,473:136-142.
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[9]Zhao Yunmei,Ding Shurong,Zhang Xunchao,et al.Effects of fuel particle size and fission-fragmentenhanced irradiation creep on the in-pile behavior in CERCER composite pellets[J].Journal of Nuclear Materials,2016,482:278-293.
[10]Gong X,Zhao Y,Ding S.A new method to simulate the micro-thermo-mechanical behaviors evolution in dispersion nuclear fuel elements[J].Mechanics of Materials,2014,77:14-27.
[2]Kim Y S,Hofman G L.Fission product induced swelling of U-Mo alloy fuel[J].Journal of Nuclear Materials,2011,419:291-301.
[3]Kim Y S,Hofman G L,Cheon J S,et al.Fission induced swelling and creep of U-Mo alloy fuel[J].Journal of Nuclear Materials,2013,437:37-46.
[4]Lopes D A,Zimmermann A J O,Silva S L,et al.Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel[J].Nuclear Technology,2016,473:136-142.
[5]Meyer M K,Gan J,Jue J F,et al.Irradiation performance of U-Mo monolithic fuel[J].Nuclear Engineering and Technology,2014,46:169-182.
[6]Rest J,Hofman G L.An alternative explanation for evidence that xenon depletion,pore formation,and grain subdivision begin at different local burnups[J].Journal of Nuclear Material,2000,277:231-238.
[7]Cui Y,Ding S,Chen Z,et al.,Modifications and applications of the mechanistic gaseous swelling model for UMo fuel[J].Journal of Nuclear Materials,2015,457:157-164.
[8]Zhao Y,Gong X,Ding S.Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic UMo/Zr fuel plates under a heterogeneous irradiation condition[J].Nuclear Engineering and Design,2015,285:86-97.
[9]Zhao Yunmei,Ding Shurong,Zhang Xunchao,et al.Effects of fuel particle size and fission-fragmentenhanced irradiation creep on the in-pile behavior in CERCER composite pellets[J].Journal of Nuclear Materials,2016,482:278-293.
[10]Gong X,Zhao Y,Ding S.A new method to simulate the micro-thermo-mechanical behaviors evolution in dispersion nuclear fuel elements[J].Mechanics of Materials,2014,77:14-27.