UMo/Al弥散燃料板内裂变气体肿胀的静压效应研究
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摘要
将核燃料的裂变气体肿胀与静水压力计算相耦合,并考虑重要的辐照蠕变,编制了定义其复杂力学本构关系的子程序。将定义各部分材料热-力学本构关系的用户子程序引入ABAQUS软件,获得了燃料板细观尺度下辐照-热-力耦合行为的计算模拟方法,并计算分析了核燃料裂变气体肿胀的静压效应。与不考虑裂变气体肿胀静压相关性的计算结果对比发现,在裂变气体肿胀计算中引入静压的影响,将使得核燃料颗粒内的辐照肿胀应变显著减小,引起板内最高温度降低,并减弱燃料颗粒和基体间的力学相互作用,减小燃料颗粒内的等效蠕变应变,致使基体内最大Mises应力和第一主应力减小。
The calculations of fission gas swelling and hydrostatic pressure in the nuclear fuel were coupled with each other.With the important irradiation creep also considered,the user subroutine was programmed to define its complex mechanical constitutive relation.With the user subroutines of defining all the materials introduced into ABAQUS,the computational method was developed for simulating the meso-scale irradiationinduced thermo-mechanical coupling behavior in a UMo/Al dispersion fuel plate,and the hydrostatic pressure effect of the fission gas swelling was analyzed.Comparing the calculated results without considering the hydrostatic pressure effect,the results with the hydrostatic pressure effect show that the irradiation swelling strain in the fuel particles reduces remarkably,the highest temperature decreases,the mechanical interaction between the fuel particles and the matrix is weakened,and the equivalent creep strains in the fuel particles decrease together with the decrease of the maximum Mises stress and the first principal stress in the matrix.
The calculations of fission gas swelling and hydrostatic pressure in the nuclear fuel were coupled with each other.With the important irradiation creep also considered,the user subroutine was programmed to define its complex mechanical constitutive relation.With the user subroutines of defining all the materials introduced into ABAQUS,the computational method was developed for simulating the meso-scale irradiationinduced thermo-mechanical coupling behavior in a UMo/Al dispersion fuel plate,and the hydrostatic pressure effect of the fission gas swelling was analyzed.Comparing the calculated results without considering the hydrostatic pressure effect,the results with the hydrostatic pressure effect show that the irradiation swelling strain in the fuel particles reduces remarkably,the highest temperature decreases,the mechanical interaction between the fuel particles and the matrix is weakened,and the equivalent creep strains in the fuel particles decrease together with the decrease of the maximum Mises stress and the first principal stress in the matrix.
引文
[1]刘利剑,尹昌耕,陈建刚,等.弥散燃料板芯体中U-Mo/Al-Si合金基体反应层性质研究[J].核动力工程,2014,35(2):161-165.LIU Lijian,YIN Changgeng,CHEN Jiangang,et al.Study on characteristics of U-Mo/Al-Si interaction layers of dispersion fuel plates[J].Nuclear Power Engineering,2014,35(2):161-165(in Chinese).
[2]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(1-3):37-46.
[3]CHURCHMAN A T,BARNES R S,COTTRELL A H.Effects of heat and pressure on the swelling of irradiated uranium[J].Journal of Nuclear Energy,1958,7(1-2):88-90.
[4]JR M M H,FLINN J E.Stress state dependence of in-reactor creep and swelling,Part 2:Experimental results[J].Journal of Nuclear Materials,2010,396(1):119-129.
[5]KARAHAN A.Modelling of thermo-mechanical and irradiation behavior of metallic and oxide fuels for sodium fast reactors[R].USA:Massachusetts Institute of Technology,2009.
[6]OGATA T.Development and validation of ALFUS:An irradiation behavior analysis code for metallic fast reactor fuels[J].Nuclear Technology,1999,128(1):113-123.
[7]MATTHEWS J R,WOOD M H.A simple operational gas release and swelling model,Ⅱ:Grain boundary gas[J].Journal of Nuclear Materials,1980,91(2):241-256.
[8]KIM Y S,HOFMAN G L.Fission product induced swelling of U-Mo alloy fuel[J].Journal of Nuclear Materials,2011,419(1-3):291-301.
[9]REST J.A model for the effect of the progression of irradiation-induced recrystallization from initiation to completion on swelling of UO2 and U-10Mo nuclear fuels[J].Journal of Nuclear Materials,2005,346:226-232.
[10]CUI Yi,DING Shurong,CHEN Zengtao,et al.Modifications and applications of the mechanistic gaseous swelling model for UMo fuel[J].Journal of Nuclear Materials,2015,457:157-164.
[11]LIU X,LU T C,XING Z H,et al.Modeling the swelling behavior of UMo alloys for Almatrix dispersion fuel[J].Journal of Alloys and Compounds,2011,509(23):6 589-6 594.
[12]邢忠虎,应诗浩.U3Si2-Al弥散型燃料的辐照肿胀研究[J].原子能科学技术,2001,35(1):15-19.XING Zhonghu,YING Shihao.Study on the irradiation swelling of U3Si2-Al dispersion fuel[J].Atomic Energy Science and Technology,2001,35(1):15-19(in Chinese).
[13]JEONG G Y,KIM Y S,SOHN D S.Mechanical analysis of UMo/Al dispersion fuel[J].Journal of Nuclear Materials,2015,466:509-521.
[14]JEONG G Y,KIM Y S,JAMISON L M,et al.Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel[J].Journal of Nuclear Materials,2017,487:265-279.
[15]ZHAO Yunmei,GONG Xin,DING Shurong.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.
[16]OZALTUN H,SHEN M H H,MEDVEDEV P.Assessment of residual stresses on U10Mo alloy based monolithic mini-plates during hot isostatic pressing[J].Journal of Nuclear Materials,2011,419(1-3):76-84.
[17]SPINO J,REST J,GOLL W,et al.Matrix swelling rate and cavity volume balance of UO2fuels at high burn-up[J].Journal of Nuclear Materials,2005,346(2-3):131-144.
[18]REST J.A model for the influence of microstructure precipitate pinning and fission gas behavior on irradiation-induced recrystallization of nuclear fuels[J].Journal of Nuclear Materials,2004,326:175-184.
[19]MASSIH A R,FORSBERG K.Calculation of grain boundary gaseous swelling in UO2[J].Journal of Nuclear Materials,2008,377:406-408.
[2]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(1-3):37-46.
[3]CHURCHMAN A T,BARNES R S,COTTRELL A H.Effects of heat and pressure on the swelling of irradiated uranium[J].Journal of Nuclear Energy,1958,7(1-2):88-90.
[4]JR M M H,FLINN J E.Stress state dependence of in-reactor creep and swelling,Part 2:Experimental results[J].Journal of Nuclear Materials,2010,396(1):119-129.
[5]KARAHAN A.Modelling of thermo-mechanical and irradiation behavior of metallic and oxide fuels for sodium fast reactors[R].USA:Massachusetts Institute of Technology,2009.
[6]OGATA T.Development and validation of ALFUS:An irradiation behavior analysis code for metallic fast reactor fuels[J].Nuclear Technology,1999,128(1):113-123.
[7]MATTHEWS J R,WOOD M H.A simple operational gas release and swelling model,Ⅱ:Grain boundary gas[J].Journal of Nuclear Materials,1980,91(2):241-256.
[8]KIM Y S,HOFMAN G L.Fission product induced swelling of U-Mo alloy fuel[J].Journal of Nuclear Materials,2011,419(1-3):291-301.
[9]REST J.A model for the effect of the progression of irradiation-induced recrystallization from initiation to completion on swelling of UO2 and U-10Mo nuclear fuels[J].Journal of Nuclear Materials,2005,346:226-232.
[10]CUI Yi,DING Shurong,CHEN Zengtao,et al.Modifications and applications of the mechanistic gaseous swelling model for UMo fuel[J].Journal of Nuclear Materials,2015,457:157-164.
[11]LIU X,LU T C,XING Z H,et al.Modeling the swelling behavior of UMo alloys for Almatrix dispersion fuel[J].Journal of Alloys and Compounds,2011,509(23):6 589-6 594.
[12]邢忠虎,应诗浩.U3Si2-Al弥散型燃料的辐照肿胀研究[J].原子能科学技术,2001,35(1):15-19.XING Zhonghu,YING Shihao.Study on the irradiation swelling of U3Si2-Al dispersion fuel[J].Atomic Energy Science and Technology,2001,35(1):15-19(in Chinese).
[13]JEONG G Y,KIM Y S,SOHN D S.Mechanical analysis of UMo/Al dispersion fuel[J].Journal of Nuclear Materials,2015,466:509-521.
[14]JEONG G Y,KIM Y S,JAMISON L M,et al.Effect of stress evolution on microstructural behavior in U-Mo/Al dispersion fuel[J].Journal of Nuclear Materials,2017,487:265-279.
[15]ZHAO Yunmei,GONG Xin,DING Shurong.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.
[16]OZALTUN H,SHEN M H H,MEDVEDEV P.Assessment of residual stresses on U10Mo alloy based monolithic mini-plates during hot isostatic pressing[J].Journal of Nuclear Materials,2011,419(1-3):76-84.
[17]SPINO J,REST J,GOLL W,et al.Matrix swelling rate and cavity volume balance of UO2fuels at high burn-up[J].Journal of Nuclear Materials,2005,346(2-3):131-144.
[18]REST J.A model for the influence of microstructure precipitate pinning and fission gas behavior on irradiation-induced recrystallization of nuclear fuels[J].Journal of Nuclear Materials,2004,326:175-184.
[19]MASSIH A R,FORSBERG K.Calculation of grain boundary gaseous swelling in UO2[J].Journal of Nuclear Materials,2008,377:406-408.