Development of an innovative spacer grid model utilizing computational fluid dynamics within a subchannel analysis tool
Development of an innovative spacer grid model utilizing computational fluid dynamics within a subchannel analysis tool
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发表时间：
2007-09
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通讯作者：
M. Avramova;Kostadin Ivanov;L. Hochreiter;J. Mahaffy;Cengiz Camci;Markus Glueck
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作者：
M. Avramova;Kostadin Ivanov;L. Hochreiter;J. Mahaffy;Cengiz Camci;Markus Glueck
在过去的几十年里，改进核反应堆安全分析的需要导致了多维热水力分析的先进方法的快速发展。为了解释在稳态和瞬态轻水反应堆（LWR）条件下预期的许多物理现象，这些方法变得越来越复杂。先进的热液子通道代码COBRA-TF （Thurgood, m.j.等人，1983）在世界范围内用于核反应堆安全余量的最佳估计评估。在宾夕法尼亚州立大学（PSU）和AREVA NP GmbH的联合研究项目框架内，对COBRA-TF的理论模型和数值进行了改进。在F-COBRA-TF的名义下，该代码已经过广泛的验证和验证程序，并已应用于各种低水堆稳态和瞬态模拟。为了使F-COBRA-TF适用于工业应用，包括安全边际评估和设计分析，对代码间隔网格模型进行了修订并进行了实质性改进。隔离栅对抽油杆管束流动热工性能影响的最先进建模方法是通过计算流体动力学（CFD）计算进行数值实验。由于涉及计算成本，CFD代码还不能用于完整的束预测，但它们的功能可以用于开发更高级和更复杂的模型，用于子通道级分析。配备改进物理模型的子信道代码可以成为轻水堆安全性和设计评估的有力工具。这个博士研究的独特贡献被视为发展，实施，
In the past few decades the need for improved nuclear reactor safety analyses has led to a rapid development of advanced methods for multidimensional thermal-hydraulic analyses. These methods have become progressively more complex in order to account for the many physical phenomena anticipated during steady state and transient Light Water Reactor (LWR) conditions. The advanced thermal-hydraulic subchannel code COBRA-TF (Thurgood, M. J. et al., 1983) is used worldwide for best-estimate evaluations of the nuclear reactor safety margins. In the framework of a joint research project between the Pennsylvania State University (PSU) and AREVA NP GmbH, the theoretical models and numerics of COBRA-TF have been improved. Under the name F-COBRA-TF, the code has been subjected to an extensive verification and validation program and has been applied to variety of LWR steady state and transient simulations. To enable F-COBRA-TF for industrial applications, including safety margins evaluations and design analyses, the code spacer grid models were revised and substantially improved. The state-of-the-art in the modeling of the spacer grid effects on the flow thermalhydraulic performance in rod bundles employs numerical experiments performed by computational fluid dynamics (CFD) calculations. Because of the involved computational cost, the CFD codes cannot be yet used for full bundle predictions, but their capabilities can be utilized for development of more advanced and sophisticated models for subchannel-level analyses. A subchannel code, equipped with improved physical models, can be then a powerful tool for LWR safety and design evaluations. The unique contributions of this PhD research are seen as development, implementation,