Preliminary study on FROBA-SCO2 with transient models and its application for Reactivity Initiated Accident and loss of flow accident

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2022-09-01 DOI:10.1016/j.jandt.2022.08.001

T.A. Yang, K. Zhang, H.K. Yang, W.X. Tian, S.Z. Qiu

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引用次数: 0

Abstract

As a coolant enabling the reactor with a high efficiency and low volume, the supercritical CO₂ has becoming a hot topic nowadays. However, the fuel rod behaviors under accidents are not studied thoroughly, especially the interaction between the coolant and the cladding. With the previous study of fuel rod behaviors analysis of supercritical CO₂ cooled nuclear reactor under steady state operation, transient state models are implemented and the upgraded code can be used to predict the fuel rod behaviors under RIA and LOFA. After the comparation of the calculation results with SACOS-SCO₂, the tendency and numerical value are close, so the accuracy is warranted. The upgraded FROBA-SCO₂ is used to predict the fuel rod behaviors under RIA and LOFA conditions and the results indicate that the maximum temperature of cladding and fuel is within safety limits.

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基于暂态模型的FROBA-SCO2及其在反应性引发事故和失流事故中的应用初探

超临界CO2作为一种高效、小体积的反应堆冷却剂，已成为当前研究的热点。然而，对事故条件下燃料棒的性能，特别是冷却剂与包壳的相互作用的研究还不够深入。在前人对超临界CO2冷却核反应堆稳态运行下燃料棒性能分析的基础上，实现了暂态模型，升级后的代码可用于RIA和LOFA下燃料棒性能的预测。将计算结果与SACOS-SCO2进行比较，趋势与数值接近，保证了计算的准确性。采用升级后的FROBA-SCO2对RIA和LOFA工况下的燃料棒行为进行了预测，结果表明包壳和燃料的最高温度在安全范围内。

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

International Journal of Advanced Nuclear Reactor Design and Technology

CiteScore

2.50

自引率

0.00%

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