Coolant Hydrodynamics at the Inlet to the FA of the RITM-Type Reactor of a Small Nuclear Power Plant

IF 0.9 Q4 ENERGY & FUELS Thermal Engineering Pub Date : 2024-05-20 DOI:10.1134/S0040601524050057
S. M. Dmitriev, T. D. Demkina, A. A. Dobrov, D. V. Doronkov, D. S. Doronkova, A. N. Pronin, A. V. Ryazanov
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Abstract

The results of an experimental study into the features of the process of coolant flow formation in the inlet section of the fuel assembly (FA) of the core of a RITM-type reactor of a small nuclear power plant (SNPP) are presented. The purpose of the work is to evaluate the influence of different design elements of the inlet section on the formation of coolant flow. To achieve this goal, a series of experiments was completed on a research aerodynamic stand with an air working environment using a large-scale experimental model, including structural elements of the FA inlet section from the throttle orifice to the fuel rod fastening unit to the diffuser, as well as a fragment of the fuel rod bundle between the absorber and spacer grids. The studies were carried out using the pneumometric method and the method of injection of a contrast admixture in several sections along the length of the model. Measurements were made over the entire cross section of the model. Features of the coolant flow are visualized by cartograms of the axial flow velocity of the working medium and the distribution of admixture in the cross section of the model. The research results were used by specialists from the design and calculation departments of OKBM Afrikantov to substantiate engineering solutions when designing new cores of RITM reactors. The results of the experiments were collected into a database and used in the validation of the LOGOS CFD computer program created by employees of the All-Russian Research Institute of Experimental Physics and the Institute for Theoretical and Mathematical Physics of Moscow State University as analogues of foreign computer programs of the same class, which include ANSYS, Star CCM, etc. Experimental data were also used when validating one-dimensional thermal-hydraulic codes used at OKBM Afrikantov in substantiating the thermal reliability of reactor cores. The thermohydraulic code CANAL is also included in this class of computer programs.

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小型核电站 RITM 型反应堆 FA 入口处的冷却剂流体力学
本文介绍了对小型核电厂(SNPP)RITM 型反应堆堆芯燃料组件(FA)入口段冷却剂流形成过程特征的实验研究结果。这项工作的目的是评估入口部分不同设计元素对冷却剂流形成的影响。为实现这一目标,在一个具有空气工作环境的研究空气动力学台架上使用大型实验模型完成了一系列实验,其中包括从节流孔到燃料棒紧固单元再到扩散器的 FA 入口部分的结构元素,以及吸收器和隔栅之间的燃料棒束片段。研究采用了气动测量法和在模型长度方向上的几个部分注入对比度外加剂的方法。对模型的整个横截面进行了测量。工作介质的轴向流速和混合剂在模型横截面上的分布情况通过制图直观地显示出来。在设计 RITM 反应堆的新堆芯时,OKBM 阿夫里坎托夫公司设计和计算部门的专家利用这些研究成果来证实工程解决方案。全俄实验物理研究所和莫斯科国立大学理论和数学物理研究所的员工创建了 LOGOS CFD 计算机程序,作为 ANSYS、Star CCM 等国外同类计算机程序的类似程序。在验证阿夫里康托夫国家核反应堆厂(OKBM Afrikantov)用于证明反应堆堆芯热可靠性的一维热流体力学代码时,也使用了实验数据。热液压代码 CANAL 也属于这一类计算机程序。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
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来源期刊
CiteScore
1.30
自引率
20.00%
发文量
94
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