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Fast Reactor: an Experimental Study of Thermohydraulic Processes in Different Operating Regimes

机译：快速反应器：不同工况下热工过程的实验研究

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摘要

Results of integrated water model studies of temperature fields and a flow pattern of a nonisother-mal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.
机译：在快速中子反应堆（以下称为快速反应堆）一次回路中，以不同方式（例如强制循环（FC））的一次钠对温度场和非等温一次冷却剂在主要回路中的流动模式进行综合水模型研究的结果介绍了自然冷却剂对流过渡到反应堆冷却和紧急冷却的过程。结果表明，在提升力的作用下，上部腔室底部区域周围的非等温冷却液在侧罩上方流动，形成了一个稳定的冷冷却剂等温区，其尺寸随着总水流量的增加而增大。 。在侧罩上方的（热）腔室的外围区域，压力侧腔室和冷侧腔室，升降机挡板，反应堆容器的冷却系统以及反应堆容器的冷却系统中，检测到基本稳定的冷却剂温度分层。中间和自主换热器在不同运行方式下的出口。在分层和再循环地层的界面处记录到大的梯度和温度波动。在所有研究的冷却方式中，与FC相比，核心燃料组件处的冷却剂出口温度降低，而上腔室外围区域中的冷却剂温度升高。带有潜水自动换热器（AHE）的快速反应堆（BN-1200）的被动应急冷却系统的高性能已得到确认。因此，在正常操作状态下，即使在三个潜水式AHE发生故障的情况下，反应器内部设备的温度仍保持在可接受的范围内，并且从反应器中排出的衰减热量不会超过安全操作范围。所获得的结果既可以用于计算机代码验证，又可以用于基于相似性标准的反应堆设备参数的近似估计。

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