基于弥散燃料颗粒开裂的金属基体裂纹特征模型

金属基弥散燃料元件在特殊工况下会发生表面起泡失效。燃料颗粒开裂是金属基体开裂的前提条件,只有当金属基体开裂后元件才会发生表面起泡。燃料颗粒开裂后,裂纹宽度和塑性区长度等裂纹特征决定了金属基体开裂行为。基于弹塑性断裂力学和应力平衡条件,建立了基于弥散燃料颗粒开裂的金属基体裂纹特征模型。计算结果表明：裂纹张开位移随退火温度和燃耗深度的升高而增加;裂纹尖端塑性区长度主要与退火温度相关。裂纹张开位移和塑性区长度的计算结果与实验数据均符合较好,验证了金属基体裂纹特征模型的有效性。     

基于弥散燃料颗粒开裂的金属基体裂纹特征模型

金属基弥散燃料元件在特殊工况下会发生表面起泡失效。燃料颗粒开裂是金属基体开裂的前提条件,只有当金属基体开裂后元件才会发生表面起泡。燃料颗粒开裂后,裂纹宽度和塑性区长度等裂纹特征决定了金属基体开裂行为。基于弹塑性断裂力学和应力平衡条件,建立了基于弥散燃料颗粒开裂的金属基体裂纹特征模型。计算结果表明:裂纹张开位移随退火温度和燃耗深度的升高而增加;裂纹尖端塑性区长度主要与退火温度相关。裂纹张开位移和塑性区长度的计算结果与实验数据均符合较好,验证了金属基体裂纹特征模型的有效性。     

裂变气体气泡尺寸对弥散燃料颗粒内部特征的影响规律

在前期均匀裂变气体气泡尺寸弥散燃料颗粒开裂模型基础上,基于不同尺寸气泡压力作用于燃料相的米塞斯(Mises)应力相等这一假设条件,建立了非均匀气泡尺寸的燃料颗粒开裂模型,并通过模型计算了裂变气体气泡尺寸对燃料相等效层厚度、气泡中气体原子数、气泡压力、燃料相最大张应力等内部特征的影响规律。计算结果表明:当气泡半径较大时,燃料相等效层厚度与气泡半径近似呈线性关系,当气泡尺寸较小时,等效层厚度与气泡半径之比随气泡半径减小急剧增加;随着气泡半径减小,气体原子数浓度增加;在升温过程中气泡内壁最大张应力的增大速率明显高于开裂阻力,气泡半径越小,燃料颗粒开裂温度越低。     

基于断裂强度的陶瓷燃料颗粒开裂模型

基于陶瓷燃料断裂强度建立弥散型燃料中陶瓷燃料颗粒开裂行为的数学模型。以铜基弥散型燃料为例,通过计算预测燃料颗粒的开裂温度与燃耗的关系,分析基体金属、环境约束、燃料相的体积、燃料颗粒尺寸对开裂温度的影响,探讨提高燃料颗粒开裂温度的途径。结果表明,燃料颗粒开裂温度与燃耗深度近似呈幂律关系,随燃料相体积的增加近似直线下降;裂变气体气孔率和孔径的增加利于提高颗粒的开裂温度。     

弥散型燃料板的辐照起泡机理分析

弥散型燃料在研究堆和动力堆中有着广泛的应用。起泡是弥散型燃料特有的失效模式,起泡的发生将导致堆芯传热性能恶化,威胁反应堆的运行安全。在分析总结国内外弥散型燃料板的辐照后起泡退火试验结果的基础上,从微观尺度到宏观尺度分析了起泡发生的机理,重点研究了弥散型燃料板的一种重要起泡模式——孔洞连通模式,剖析了孔洞连通发生的3个基本过程。同时应用孔洞连通机理,在估算裂变气体压力的前提下,通过力学计算给出了可引起起泡的孔洞连通的圆形区域尺度约为1.8mm,这与实验观察结果相符。本文分析表明,燃料板的孔洞连通起泡机理涉及到高燃耗效应、燃料相的肿胀和开裂、裂变碎片损伤和应力腐蚀开裂等过程,建立起泡模型需做弹塑性力学和断裂力学的数值计算。     

TRISO燃料颗粒三维多物理场耦合计算模型开发

三向同性燃料(TRISO)颗粒是高温气冷堆弥散型燃料和全陶瓷微密封(FCM)耐事故燃料芯块的裂变区。为研究TRISO燃料颗粒在辐照环境中的复杂行为,基于COMSOL有限元软件开发了TRISO燃料颗粒的三维多物理场耦合性能分析模型。通过采用随辐照条件变化的材料物性参数和行为模型,可模拟燃料颗粒在稳态运行和事故工况下复杂的堆内热-力学行为,以及CO气体产生和裂变气体释放、裂变产物扩散等重要物理过程,还可以计算燃料颗粒的失效概率。基于COMSOL开发三维分析模型的计算结果与美国BISON程序对TRISO燃料颗粒的计算结果相比同样符合较好,说明了所开发模型的合理性。     

弥散燃料颗粒裂纹起源的有限元模拟分析

通过建立含多气泡的燃料颗粒模型,采用有限元方法分析了燃料颗粒在裂变气体气泡内压作用下的应力分布,统计了燃料颗粒内部气泡位置对气泡内壁处的最大拉应力的影响,并结合实验结果探寻了弥散燃料颗粒在辐照后退火时的裂纹起源。结果表明:当弥散燃料颗粒内部含有多个裂变气体气泡时,受气泡内压作用,气泡内壁径向应力为压应力,环向应力为拉应力;气泡位置距燃料颗粒心部越远,气泡内壁处的最大环向拉应力越大;表层气泡的最大环向拉应力远大于心部气泡的;燃料颗粒裂纹起源于表层气泡内壁。     

UO2燃料裂变气体渗流模型研究

为分析UO₂燃料晶界气泡连通导致裂变气体间歇性释放的动力学过程,从而解决目前扩散模型预测的沿芯块径向释放份额与实验测量不符的问题,采用二维渗流模型模拟UO₂燃料晶界气泡网络的演化及与燃料棒内自由空间连通的释放过程。研究结果表明,渗流模型预测沿芯块径向的裂变气体释放份额在芯块中间部分出现局部峰值,并随着时间向芯块外侧推进,与辐照试验观察到不同燃耗下径向裂变气体分布现象定性符合。因此,本研究建立的渗流模型能够从机理上解释此前扩散模型未能预测的UO₂燃料裂变气体释放份额沿径向非单调分布现象。      

弥散型燃料裂变产物释放行为分析研究

为了获得弥散型燃料裂变产物向一回路冷却剂的释放特性,开展了弥散型燃料裂变产物释放行为研究,开发了适用于弥散型燃料的裂变产物源项计算程序,并对裂变产物源项进行了影响分析。结果表明:沾污铀和起泡破损后裂变产物的核素谱存在一定差异;裂变产物的释放与起泡当量直径的平方成正比;对于弥散型燃料而言,起泡破损中通过反冲释放的占比较低;相同破口条件下的弥散型和陶瓷型燃料中裂变产物的释放存在量级的差别。本文开发的程序能够用于分析弥散型燃料的裂变产物源项,为后续相关研究工程设计奠定基础。      

高燃耗陶瓷燃料颗粒的参数化建模及强度分析

弥散燃料芯体中的陶瓷燃料颗粒在辐照条件下会形成裂变气孔,燃料颗粒内部气孔间的相互干涉作用及气孔内压的增长致使局部拉应力超过材料强度极限,进而导致燃料颗粒开裂。本文考虑高燃耗燃料颗粒内气孔尺寸和位置分布的非均匀性,实现了颗粒内部的细观结构参数化建模。运用有限元方法计算并分析了气孔尺寸、基体约束压应力、温度和气孔分布方式对颗粒内部最大拉应力的影响,研究了颗粒内开裂危险区的分布规律。结果表明,陶瓷燃料颗粒最大拉应力随气孔尺寸和温度的增加而增大,随基体约束压应力的增加而减小;燃料相的断裂强度减小,开裂危险区面积增大;燃料颗粒从内部多处开裂破坏,而表层处开裂的概率更大。本文为弥散燃料失效研究及优化设计提供了分析方法及数值参考。     

弥散型燃料的裂变气体行为研究

探讨了弥散型燃料中对辐照肿胀有重要影响的裂变气体的行为机理。裂变气体原子聚集成气泡引起燃料相肿胀,气泡的尺寸分布是影响辐照肿胀的重要因素。决定气泡生长的裂变气体的行为机理主要有：裂变气体原子的产生和热扩散迁移,气泡的成核和聚合长大,气泡内气体原子的重溶,燃料相的辐照亚晶化等过程。燃料中各种尺寸的气泡浓度随时间的变化率可用气泡生长的动力学速率方程组来描述。当裂变密度较高时,辐照产生的缺陷引起燃料相的     

Growth of the interaction layer around fuel particles in dispersion fuel

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl_x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.     

BEAF - a computer program for analysis of light water reactor fuel rod behavior

BEAF - a computer program for analysis of light water reactor fuel rod behavior was developed. The BEAF code, which is appropriate for on-line prediction of fuel rod behavior, can analyze fuel rod thermal and mechanical behaviors using the axisymmetric, plane strain approximation and finite difference method to realize a fast running time.In the mechanical analysis, a new cracked pellet compliance model is introduced, in which pellet cracking and crack healing, pellet initial relocation, modified elastic moduli of a cracked fuel pellet, and stress dependent hot pressing are considered. Adding to those capabilities, fission gas flow and diffusion in the fuel-clad gap are analyzed to take into account the slow fission gas dilution effect on the gap conductance during power ramp.     

金属基弥散燃料元件失稳肿胀的静态弹塑性模型

针对金属基弥散燃料元件金属基体开裂导致的失稳肿胀,在不考虑粘塑性变形情况下建立了裂纹面的静态弹塑性模型,采用有限元模拟对静态弹塑性模型进行了验证。当金属基体发生全屈服后,其主要变形方式从弹性变形转变为塑性变形;根据金属基体的主要变形方式,分别建立金属基弥散燃料裂纹面的弹性变形模型和塑性变形模型;结合内应力与弯矩的平衡条件,获得了裂纹面弹塑性变形的临界转变条件。弹性变形模型和塑性变形模型的计算结果与有限元模拟结果符合较好,验证了金属基弥散燃料失稳肿胀的静态弹塑性模型的有效性。      

Crack formation,crack healing and porosity redistribution during irradiation of UO2 and (U,Pu) O2

The irradiation-induced void volume redistribution in the fuel was analysed. The radial crack volume and porosity distributions, the central radii and the radial gap width were measured after irradiation and compared with the calculated values. Short-time (He-loop experiments in the FR2 reactor), medium-time (bundle irradiation in the BR2 reactor) and long-time (trefoil-irradiation in the DFR reactor) irradiated fuel pins were examined. The model of pore migration, used in the computer code SATURN-la, is based on the evaporation-condensation mechanism. Measured swelling rates were extrapolated to higher temperatures and used. The crack volume distribution was calculated on the basis of a multifractured fuel model. One can conclude from the comparison between calculated and measured void volume distributions that several mechanisms redistribute void volume. These are crack formation, crack healing, migration of sinter pores and fission gas bubbles, gas swelling, evaporation-condensation phenomena in the region of the central void, irradiation-induced sintering and increase in diameter of the cladding.     

Computer modeling of steady state fission gas behavior in carbide fuels

A model for the simulation of long-term, steady-state fission gas behavior in carbide fuels is formulated. It is assumed that fission gas release occurs entirely through gas atom diffusion to grain boundaries and cracks. Fission gas bubbles are assumed to remain stationary and to grow as the net result of gas atom precipitation into the bubbles from the matrix solid and gas atom re-solution from the bubbles into the matrix. Furthermore, assuming that local gas atom redistribution process in the immediate neighborhood of a bubble is very rapid, the bubble size is assumed to correspond to the equilibrium size that maintains exact balance between the rate of gas atom re-solution and that of gas atom precipitation.The model also treats the effect of attachment between bubbles and second-phase precipitates; the experimentally observed faster growth rate of precipitate bubbles is simulated using a reduced re-solution parameter for precipitate bubbles. With the grain matrix assumed to be spherical, the model allows the computation of the radial distribution of the intragranular bubbles and the gas atom concentration in the matrix.The flux of gas atoms arriving at the grain boundary is computed. The continual growth of grain boundary bubbles, resulting from the accumulation of gas atoms on the grain boundary, leads to grain boundary interlinkage and all gas atoms that subsequently reach the grain boundary are assumed to be released. Similarly, all gas atoms generated following the interlinkage of intragranular bubbles are also assumed to be immediately released.Application of the model indicates that fission gas swelling is largely due to intragranular bubbles. Grain boundary bubbles, although very large in size, contribute little to fission gas swelling and the contribution from gas atoms in solid solution in the matrix is even less significant.Physical parameters entering the model were assigned numerical values that closely represent the physical characteristics of the irradiation samples. Careful comparisons between the results of sensitivity studies and the experimental data readily identify the re-solution parameter to have the strongest influence on the results predicted by the code and that the grain size, and not the temperature, is the dominant factor affecting gas release.When allowance is made for the uncertainties of the experimental data, the predicted fission gas swelling also correlates well with experiment. The spread in the fuel swelling data, however, indicates that fuel cracking, and not fission gas swelling alone, very often contributes significantly to the fuel external dimensional changes. The linear fission gas swelling rate prediceted by the model exhibits almost a linear variation with temperature. This result correlates well with the linear swelling rate obtained from experimental swelling data if immersion density data alone are used, in order to eliminate the sources of uncertainties associated with fuel cracking.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.