System Thermal-hydraulics Model for Fluoride Salt-Cooled Reactor Based On Small Advanced High Temperature Reactor (smAHTR) Design Concept

IF 0.5 Q4 NUCLEAR SCIENCE & TECHNOLOGY Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-05-05 DOI:10.1115/1.4062500
Shu Jun Wang, Xianmin Huang, B. Bromley
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Abstract

A system thermal-hydraulics model for a fluoride-salt-cooled high-temperature reactor (FHR) based on the small modular advanced high-temperature reactor (SmAHTR) design concept is developed, using RELAP5-3D. The SmAHTR components modelled in the simulations include: the reactor core, lower plenum, upper plenum, top plenum, three Primary Heat Exchangers (PHX's) equipped with three primary pumps, and three Director Reactor Auxiliary Cooling System (DRACS) equipped with three fluid diodes. Flows through the reactor core are represented by 19 individual fuel channels, one reflector-hole channel, and one downcomer channel. In each of the 19 SmAHTR fuel block channels, the fuel elements are modeled in 5 groups using 5 heat structures, each with their corresponding power level. The total reactor power is 125 MWth. Using representative core power distributions for the SmAHTR at beginning-of-cycle (BOC) and at end-of-cycle (EOC), two steady-state system thermal-hydraulic model simulations with RELAP5-3D were performed using a default pressure drop loss factor of 1.5 for all 19 fuel channels. Exit coolant temperatures ranged from 688°C to 739°C (BOC) and from 696°C to 721°C (EOC), while peak fuel centerline temperatures in the highest power block were 1,249°C (BOC) and 1,029°C (EOC). By adjusting the loss factors to modify coolant flow rates in each channel, a more uniform exit coolant temperature was possible.
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基于小型先进高温堆(smAHTR)设计理念的氟盐冷堆系统热工模型
基于小型模块化先进高温堆(SmAHTR)设计理念,利用RELAP5-3D软件建立了氟化物盐冷高温堆(FHR)系统热工模型。模拟的SmAHTR组件包括:堆芯、下静压室、上静压室、上静压室、装有三个主泵的三个主热交换器(PHX)和装有三个流体二极管的三个主任反应堆辅助冷却系统(DRACS)。通过反应堆堆芯的流动由19个单独的燃料通道、一个反射孔通道和一个下降管通道表示。在19个SmAHTR燃料块通道中,燃料元件被分为5组,使用5种热结构,每种热结构都有相应的功率水平。反应堆的总功率为125兆瓦。利用SmAHTR在循环开始(BOC)和循环结束(EOC)时具有代表性的堆芯功率分布,使用RELAP5-3D进行了两个稳态系统热工模型模拟,所有19个燃料通道的默认压降损失系数为1.5。出口冷却液温度范围为688°C至739°C (BOC)和696°C至721°C (EOC),而最高功率块的燃料中心线峰值温度为1,249°C (BOC)和1,029°C (EOC)。通过调整损失系数来调整每个通道的冷却剂流速,可以获得更均匀的出口冷却剂温度。
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来源期刊
CiteScore
1.30
自引率
0.00%
发文量
56
期刊介绍: The Journal of Nuclear Engineering and Radiation Science is ASME’s latest title within the energy sector. The publication is for specialists in the nuclear/power engineering areas of industry, academia, and government.
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