辐照肿胀与蠕变对弥散型核燃料力学行为影响的研究
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摘要
弥散型核燃料与传统的核燃料相比具有高燃耗和高热导的优点,不仅在各种试验研究堆中得到了广泛的应用,而且在商业堆、核动力舰船以及核废料处理等方面有着良好的发展前景。
弥散型核燃料元件是由金属包壳和弥散型核燃料芯体组成。且芯体是由核燃料颗粒弥散的分布在基体材料中所构成。在裂变的过程中,核裂变颗粒不仅会产生裂变热和辐照肿胀,而且会产生快中子和其它裂变碎片。在长时间工作的情况下,燃料元件会由于快中子和裂变碎片的辐照作用而发生辐照蠕变、材料硬化变脆以及辐照生长等效应。由于快中子穿透性很强,可以认为基体中快中子流近似均匀分布,与之相反,裂变碎片只在燃料颗粒附近区域有分布,因此,裂变碎片会导致燃料颗粒附近会形成局部的“蠕变增强”区。不均匀蠕变会导致燃料元件堆内的力学行为发生一系列变化。为了保证弥散型核燃料元件的完整性,有必要对其堆内的辐照力学行为进行研究。本文重点考察辐照肿胀和辐照蠕变对弥散型核燃料堆内力学行为的影响。
本文首先利用解析的方法对稀疏分布的弥散型核燃料进行了分析。针对基体蠕变不均匀以及燃料颗粒附近蠕变比较大的特点,建立了三相模型。并利用Eshelby理论和粘弹性理论的相关知识对模型进行了求解,得到了应力场的解析解。研究发现,基体上的不均匀蠕变导致了第一主应力和Von Mises等效应力最大值出现在离颗粒和基体边界有一段距离的地方,而不是出现在颗粒和基体的边界上。不均匀蠕变区域越大,第一主应力和Von Mises等效应力最大值越小；当不均匀蠕变区域不变时,颗粒的半径越小,第一主应力和Von Mises等效应力最大值越小。基体和颗粒的体积模量之比越小,剪切模量之比越小,第一主应力和Von Mises等效应力最大值就越小。
其次,针对燃料颗粒密集排布的弥散型核燃料元件,本文利用ANSYS有限元软件对其进行了数值计算和研究。不论是板状弥散型核燃料元件,还是棒状弥散型核燃料元件,本文都采用了既能反映其结构特点又能切实考虑边界条件的有限元模型。并考虑了不均匀蠕变区域蠕变率的连续变化。研究表明,相对于均匀蠕变的情况,考虑裂变产物的影响后,基体应力应变的分布会发生相应的变化：基体区域的蠕变率增大,而且离颗粒越近增加的幅度就越大,从而导致基体Von Mises等效应力减小,离颗粒越近减小的幅度越大。而另一方面由于包壳离颗粒较远,受裂变碎片非均匀蠕变效应的影响很小,包壳应力应变的分布和均匀蠕变时的结果相差很小。
本文的研究有助于弥散型核燃料元件的优化设计,能为实际的试验提供一定的理论参考,减少其盲目性。
Compared with the traditional nuclear fuel rods, the dispersion nuclear fuel elements have higher burn-up and higher thermal conductivity, so they have been extensively used in the research reactors, and they have good prospects in the nuclear vessels, nuclear waste disposal and commercial reactors.
The dispersion nuclear fuel element consists of the metal cladding and the dispersion nuclear fuel meat. The fuel meat is distinguished by having the nuclear fuel particles dispersed through the matrix. Inside the demanding environment of the reactors, the fuel particles generate heat by nuclear fissions, the fission product accumulation results in irradiation-induced swelling of the fuel particles. At the same time, fast neutron and fission products release from the fuel particles. Under the long-time-work, the in-pile materials creep under the attack of fast neutron and fission products. Because of the strong penetrability of fast neutron, the distribution of fast neutron flux in the matrix can be considered as uniform. But the fission products are only distributed near the fuel particles. So, the matrix materials near the particle get enhanced creep. The inhomogeneous creep in the matrix can result in different in-pile mechanical behaviors. In order to ensure the safety and reliability of the dispersion fuel rod, it is necessary to study the in-pile irradiation behaviors, especially the effects of irradiation-swelling and irradiation-creep on the in-pile irradiation behaviors.
Firstly, mechanical behaviors of the sparsely dispersed nuclear fuel are investigated through analytic solutions, considering the thermal effect, irradiation swelling of the fuel particle and the inhomogeneous irradiation-creep of the surrounding matrix. Analytic solutions of the stress fields are obtained with the method of solving the Eshelby's inclusion problem in a three-phase, spherically concentric solid. The result indicates that the maximum of the first principal stress in the matrix appears at some distances away from the interface between the matrix and the inclusion instead of on the interface, and that is induced by a locally enhanced creep of the matrix. When the ratio of the volume modulus or shear modulus between the matrix and the inclusion is smaller, the matrix is more difficult to perform fracture considering the same inclusion or matrix.
Subsequently, mechanical behaviors of the densely dispersed nuclear fuel are investigated through the finite element method. Two classical computing models, considering the design feature of the elements and the practical boundary conditions, are adopted to simulate the dispersion fuel plate and the dispersion fuel rod separately. The results indicate that:compared with the situation of a homogeneous distribution of the creep rate in the matrix, different distributions of stress and strain perform in the matrix, when considering the effect of fission product:the creep rate increases in the matrix, and the creep rate increases more intensely when the point has a closer distance from the particle; Accordingly, the Von Mises stress decreases in the matrix, and the Von Mises stress decreases more intensely when the point has a closer distance from the particle. Because of the distance between the cladding and particle, the similar distributions of stress and strain in the cladding perform both on the conditions of homogeneous creep rate and inhomogeneous creep rate in the cladding.
This study is useful to the optimal design of the dispersion nuclear fuel elements, and it could provide numerical reference basis for the actual operation of the fuel element.
弥散型核燃料元件是由金属包壳和弥散型核燃料芯体组成。且芯体是由核燃料颗粒弥散的分布在基体材料中所构成。在裂变的过程中,核裂变颗粒不仅会产生裂变热和辐照肿胀,而且会产生快中子和其它裂变碎片。在长时间工作的情况下,燃料元件会由于快中子和裂变碎片的辐照作用而发生辐照蠕变、材料硬化变脆以及辐照生长等效应。由于快中子穿透性很强,可以认为基体中快中子流近似均匀分布,与之相反,裂变碎片只在燃料颗粒附近区域有分布,因此,裂变碎片会导致燃料颗粒附近会形成局部的“蠕变增强”区。不均匀蠕变会导致燃料元件堆内的力学行为发生一系列变化。为了保证弥散型核燃料元件的完整性,有必要对其堆内的辐照力学行为进行研究。本文重点考察辐照肿胀和辐照蠕变对弥散型核燃料堆内力学行为的影响。
本文首先利用解析的方法对稀疏分布的弥散型核燃料进行了分析。针对基体蠕变不均匀以及燃料颗粒附近蠕变比较大的特点,建立了三相模型。并利用Eshelby理论和粘弹性理论的相关知识对模型进行了求解,得到了应力场的解析解。研究发现,基体上的不均匀蠕变导致了第一主应力和Von Mises等效应力最大值出现在离颗粒和基体边界有一段距离的地方,而不是出现在颗粒和基体的边界上。不均匀蠕变区域越大,第一主应力和Von Mises等效应力最大值越小；当不均匀蠕变区域不变时,颗粒的半径越小,第一主应力和Von Mises等效应力最大值越小。基体和颗粒的体积模量之比越小,剪切模量之比越小,第一主应力和Von Mises等效应力最大值就越小。
其次,针对燃料颗粒密集排布的弥散型核燃料元件,本文利用ANSYS有限元软件对其进行了数值计算和研究。不论是板状弥散型核燃料元件,还是棒状弥散型核燃料元件,本文都采用了既能反映其结构特点又能切实考虑边界条件的有限元模型。并考虑了不均匀蠕变区域蠕变率的连续变化。研究表明,相对于均匀蠕变的情况,考虑裂变产物的影响后,基体应力应变的分布会发生相应的变化：基体区域的蠕变率增大,而且离颗粒越近增加的幅度就越大,从而导致基体Von Mises等效应力减小,离颗粒越近减小的幅度越大。而另一方面由于包壳离颗粒较远,受裂变碎片非均匀蠕变效应的影响很小,包壳应力应变的分布和均匀蠕变时的结果相差很小。
本文的研究有助于弥散型核燃料元件的优化设计,能为实际的试验提供一定的理论参考,减少其盲目性。
Compared with the traditional nuclear fuel rods, the dispersion nuclear fuel elements have higher burn-up and higher thermal conductivity, so they have been extensively used in the research reactors, and they have good prospects in the nuclear vessels, nuclear waste disposal and commercial reactors.
The dispersion nuclear fuel element consists of the metal cladding and the dispersion nuclear fuel meat. The fuel meat is distinguished by having the nuclear fuel particles dispersed through the matrix. Inside the demanding environment of the reactors, the fuel particles generate heat by nuclear fissions, the fission product accumulation results in irradiation-induced swelling of the fuel particles. At the same time, fast neutron and fission products release from the fuel particles. Under the long-time-work, the in-pile materials creep under the attack of fast neutron and fission products. Because of the strong penetrability of fast neutron, the distribution of fast neutron flux in the matrix can be considered as uniform. But the fission products are only distributed near the fuel particles. So, the matrix materials near the particle get enhanced creep. The inhomogeneous creep in the matrix can result in different in-pile mechanical behaviors. In order to ensure the safety and reliability of the dispersion fuel rod, it is necessary to study the in-pile irradiation behaviors, especially the effects of irradiation-swelling and irradiation-creep on the in-pile irradiation behaviors.
Firstly, mechanical behaviors of the sparsely dispersed nuclear fuel are investigated through analytic solutions, considering the thermal effect, irradiation swelling of the fuel particle and the inhomogeneous irradiation-creep of the surrounding matrix. Analytic solutions of the stress fields are obtained with the method of solving the Eshelby's inclusion problem in a three-phase, spherically concentric solid. The result indicates that the maximum of the first principal stress in the matrix appears at some distances away from the interface between the matrix and the inclusion instead of on the interface, and that is induced by a locally enhanced creep of the matrix. When the ratio of the volume modulus or shear modulus between the matrix and the inclusion is smaller, the matrix is more difficult to perform fracture considering the same inclusion or matrix.
Subsequently, mechanical behaviors of the densely dispersed nuclear fuel are investigated through the finite element method. Two classical computing models, considering the design feature of the elements and the practical boundary conditions, are adopted to simulate the dispersion fuel plate and the dispersion fuel rod separately. The results indicate that:compared with the situation of a homogeneous distribution of the creep rate in the matrix, different distributions of stress and strain perform in the matrix, when considering the effect of fission product:the creep rate increases in the matrix, and the creep rate increases more intensely when the point has a closer distance from the particle; Accordingly, the Von Mises stress decreases in the matrix, and the Von Mises stress decreases more intensely when the point has a closer distance from the particle. Because of the distance between the cladding and particle, the similar distributions of stress and strain in the cladding perform both on the conditions of homogeneous creep rate and inhomogeneous creep rate in the cladding.
This study is useful to the optimal design of the dispersion nuclear fuel elements, and it could provide numerical reference basis for the actual operation of the fuel element.
引文
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