Abstract The study of two-phase flow behavior is a very important requirement in nuclear reactor systems, due to it can appear during normal operating conditions in a BWR or fast and slow transients for both PWR and BWR reactors. This work presents the evaluation at steady-state of two-phase flow prediction performed by THUNDER code (Thermal-Hydraulic Unit of Numerical Development for Elements with Rod-bundles), a new computational code that has been developed to perform a thermal-hydraulic analysis of LWR reactors with rod-type fuel assemblies. The physic and mathematical models used to develop the steady-state analysis of a nuclear reactor core are presented. The conservation equations of mass, momentum, and energy were established based on operating and geometric conditions, for single and two-phase flow using the subchannel and control volume methods, it applied to square arrays to the four sub-channel types (center-typical, center-thimble, edge, and corner) found in a PWR fuel assembly. THUNDER code performs a tridimensional analysis for the whole core solving simultaneously the conservation equations for each control volume. For two-phase flow treatment, the code uses the drift flux model. It provides detailed information of operational conditions e.g. coolant, cladding and center fuel temperatures, core pressure drop, coolant velocity, void fraction, coolant quality, onset of nuclear boiling (ONB) and departure from nucleate boiling ratio (DNBR). To validate the code, we consider experimental data from OECD/NRC benchmark based NUPEC PWR subchannel and bundle tests (PSBT) and CFD calculations from ANSYS-CFX®. It was evaluated the void fraction, average density and pressure drop for individual subchannel as well as for a rod bundle 5 × 5 fuel assembly. THUNDER code reproduced well the benchmark results presenting a mean absolute discrepancy of void fraction equal to 0.039 and density relative discrepancy equal to 0.11. This indicates that THUNDER code performs a two-phase flow evaluation within the acceptance criterion for reactors with rod-type fuel assemblies under steady-state conditions.

中文翻译：

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根据基准结果和 CFD 代码对 THUNDER 代码中的两相模型进行热工水力验证

摘要 两相流动行为的研究是核反应堆系统中一个非常重要的要求，因为它可能出现在 BWR 的正常运行条件下或 PWR 和 BWR 反应堆的快速和慢速瞬变。这项工作介绍了由 THUNDER 代码（具有杆束的元件的数值开发的热-液压单元）执行的两相流预测的稳态评估，这是一种新的计算代码，已开发用于执行热-水力分析装有棒式燃料组件的轻水堆反应堆。介绍了用于开发核反应堆堆芯稳态分析的物理和数学模型。根据操作和几何条件建立质量、动量和能量守恒方程，对于使用子通道和控制体积方法的单相和两相流，它应用于方形阵列到 PWR 燃料组件中的四种子通道类型（中心典型、中心顶针、边缘和角部）。THUNDER 代码对整个核心进行三维分析，同时求解每个控制体积的守恒方程。对于两相流处理，代码使用漂移通量模型。它提供了运行条件的详细信息，例如冷却剂、包壳和中心燃料温度、堆芯压降、冷却剂速度、空隙率、冷却剂质量、核沸腾开始 (ONB) 和偏离核沸腾比 (DNBR)。为了验证代码，我们考虑了来自基于 OECD/NRC 基准的 NUPEC PWR 子信道和束测试 (PSBT) 的实验数据以及来自 ANSYS-CFX® 的 CFD 计算。评估了单个子通道以及棒束 5 × 5 燃料组件的空隙率、平均密度和压降。THUNDER 代码很好地再现了基准结果，显示空隙率的平均绝对差异等于 0.039，密度相对差异等于 0.11。这表明 THUNDER 代码在稳态条件下对带有棒式燃料组件的反应堆的验收标准内进行了两相流评估。