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Verification of the TIGRSP Computer Code as Applied to a 19-Rod Fuel Assembly Using CFD Computations

机译：使用CFD计算验证应用于19杆燃料组件的TIGRSP计算机代码

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Abstract— For assessing the core critical heat flux ratio in VVER-type nuclear reactors, the lattice computation code TIGRSP (steady-state thermal and hydraulic analysis of parameters in fuel rod bundles) is used. Despite the fact that quite a large amount of verification studies has been accomplished and that the code has passed the certification procedure, work on improving the code capabilities is still being continued. One of the modern tendencies is the wide-scale involvement of CFD codes for studying local parameters in fuel assemblies (FAs). Along with comparison with experimental data, it is advisable to carry out cross verification of lattice codes with the results of 3D thermal-hydraulic computations for conditions for which experimental data are either lacking or limited for some or other reasons. This article considers thermal-hydraulic processes during coolant flow in a 19-rod assembly with parameters close to the nominal conditions in the FAs of a VVER-1000 reactor. The 3D model is verified with respect to empirical formulas and experimental data. Based on the results of test and verification computations, we determined the mesh and turbulence model parameters using the ANSYS CFX code by means of which it is possible to obtain reliable results in estimating the hydraulic characteristics of FAs fitted with spacer grids, heat transfer in the single-phase region, and coolant flow distribution over the FA cross section. The CFD computation of coolant flow for the 19-rod assembly is carried out, and the coolant mass velocity and temperature in different sections over the length and corresponding cells of the TIGRSP code are calculated. The TIGRSP code has been verified against experimental data and taking into account the CFD simulation of the 19-rod FA, and a software module for graphically visualizing the lattice code output results has been developed. The presented data can be used for verification of thermal-hydraulic codes and in elaborating the design of FAs for VVER-type reactors.
机译：摘要—为了评估VVER型核反应堆的堆芯临界热通量比，使用了晶格计算代码TIGRSP（燃料棒束参数的稳态热力和水力分析）。尽管已经完成了大量的验证研究并且代码已通过认证程序，但仍在继续进行改进代码功能的工作。现代趋势之一是广泛使用CFD代码研究燃料组件（FA）中的局部参数。在与实验数据进行比较的同时，建议针对3D热工液压计算的结果对晶格代码进行交叉验证，以针对由于某些或其他原因而缺乏或受限制的实验数据的条件。本文考虑了在19杆组件中冷却剂流动期间的热工过程，其参数接近VVER-1000反应堆FA中的标称条件。相对于经验公式和实验数据验证了3D模型。根据测试和验证计算的结果，我们使用ANSYS CFX代码确定了网格和湍流模型参数，通过这些参数，可以获得可靠的结果，可估算出装配有间隔栅的FA的水力特性，热传递。单相区域和冷却剂在FA横截面上的流量分布。对19杆组件的冷却液流量进行CFD计算，并计算长度和TIGRSP代码的相应单元中不同部分的冷却液质量速度和温度。 TIGRSP代码已针对实验数据进行了验证，并考虑了19杆FA的CFD模拟，并且开发了用于以图形方式可视化晶格代码输出结果的软件模块。所提供的数据可用于验证热工水力代码，以及用于VVER型反应堆的FA设计。

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