熔盐球床堆径向流堆芯的热工水力分析

核热泉(NHS)堆是一种新型熔盐球床概念设计堆,其冷却剂径向流过堆芯,具有满功率自然循环特性。基于多孔介质局部非热平衡模型,利用计算流体力学(CFD)通用软件Fluent计算核热泉堆径向流堆芯的热工水力特性,并比较了不同的内、外孔板开孔率的影响。结果表明,内孔板开孔率对冷却剂流量分布影响较大;燃料中心温度具有相当的安全裕量,冷却剂横向流过堆芯的阻力远低于浮升力,能够实现全回路的自然循环。     

球床堆流动与传热数值模拟

采用离散元方法(DEM)模拟球床反应堆内燃料球的随机分布,通过计算流体力学(CFD)方法研究球床堆内的流动与传热。结果表明:球与球之间的间隙处压力较低;而流速、温度、涡强度较高。沿径向分布,压力、涡强度、换热系数随孔隙率的增加而降低;流速随孔隙率的增加而增加。     

球床堆内冷却剂流动与传热特性的CFD研究

以计算流体力学(CFD)为基础,利用大型商业软件ProE和CFX,对球床式水冷堆堆芯燃料元件进行三维建模、网格划分和数值计算,对堆芯内冷却剂热工水力特性进行了初步的研究.计算比较了燃料元件球间隙和接触情况下冷却剂的速度场、温度场和压力分布,分析了其对堆芯安全的影响.     

球床氟盐冷却高温堆堆芯热工水力模拟

本文对球床氟盐冷却高温堆堆芯热工流体现象进行了研究。采用计算流体动力学(CFD)方法进行了三维建模和计算,得到了燃料元件球表面温度分布和堆芯冷却剂速度场、温度场和压力的分布,验证了稳态工况下氟盐对堆芯的冷却能力,分析了氟盐的特殊热工流体力学性质对堆芯安全的影响,结果可用于球床氟盐冷却高温堆的初步设计。     

熔盐冷却球床实验堆内流动与传热特性数值研究

采用计算流体动力学（CFD）的方法,对熔盐冷却球床堆（FHR）的堆芯中可能产生热点的位置进行局部数值模拟,从而获得燃料球表面及其内部的温度分布和燃料球附近的流场分布情况,并与所开发的FRAC分析程序结果进行对比,其最大误差为2.9%,可以初步说明FRAC程序的正确性。本文的研究结果对FHR相关实验以及机理的研究具有一定的参考价值。      

球床模块堆停堆后自然对流传热特性数值研究

对冷却流体在球床模块堆内燃料颗粒填充区域中的流动和传热过程进行了研究.数值模拟突然停堆后燃料颗粒区在温差作用下的自然对流过程,分析了瑞利数Ra对燃料填充区域内流场、温度场和局部努塞尔数Nu以及壁面摩擦阻力系数的影响.计算结果表明:当球床模块堆突然停堆时燃料填充区域可形成加热壁面流体上升流动、冷却壁面下降流动的自然循环流动;随着Ra数增大,回流中心向上移动;沿轴向壁面局部Nusselt数和摩擦阻力系数存在极值,并且极值点随Ra数增大而向上移动;与氮气相比,氦气作为冷却介质停堆后具有更均匀的堆芯轴向温度分布.     

熔盐冷却球床堆热通道热工水力特性数值分析

基于计算流体力学(Computational Fluid Dynamics,CFD)通用计算程序Fluent,研究了模块化熔盐冷却球床堆(Pebble Bed Advanced High Temperature Reactor,PB-AHTR)中心热通道稳态热工水力行为。利用已开发的多孔介质流固两相局域非热平衡模型计算了球床堆中的压降、冷却剂的温场分布以及固相球床的温场分布,计算并比较了不同的多孔介质阻力因子(Ergun与KTA)对通道内的冷却剂流动以及温场分布的影响,并对丧失部分冷却剂情况下通道内的冷却剂及燃料温度进行了计算分析。结果表明使用不同的阻力因子对堆芯压降计算结果和流场的分布影响较大;而冷却剂温场及固相球床温场和球心的温度分布在不同的阻力因子下的差别较小,在PB-AHTR的设计参数下堆芯产生的热量能够被有效的输出,设计具有较大的安全裕度。计算结果对于球床堆的优化设计提供了一定的参考价值。     

强内热源球床通道单相对流换热特性实验研究

球床水冷反应堆的堆芯为球形燃料元件堆积成的多孔通道,具有显著的强化换热作用。球床通道内的孔隙因具有多变性、随机性的特点,换热情况非常复杂,相关研究较少。为了研究含内热源球床通道内的换热特性,本文用直径为8 mm碳钢球堆积形成球床,以蒸馏水为工质,采用电磁感应加热方式对球床进行整体加热,研究球床通道内部的换热特性。通过对实验数据进行分析,得到了球床通道内部的功率分布和换热系数随热流密度、工质Re的变化规律,根据实验数据拟合得到了球床通道内平均换热系数的无量纲准则关联式,拟合结果与实验结果的相对偏差在12%以内,符合良好。     

高温气冷堆球床模拟研究

本文描述了漏斗形高温气冷球床堆的模拟计算方法及数据转换情况,克服了原来VSOP程序系统只能将漏斗形堆芯等效成一个圆柱体的局限性,新的程序系统CHTRP可依照实验测得的球流速度曲线剖分几何网格层,对于不同尺寸的反应堆锥体都可进行模拟计算,并作了实例计算,取得了令人满意的结果,为高温堆物理设计和分析提供了有力的工具。     

球床堆燃料元件球放热分布研究

本文主要研究球床堆燃料元件球放热分布规律及其主要影响因素。在雷诺数9600到55000范围内,研究了单球、不完整的菱形球床及完整的菱形球床中球元件的放热分布。认为:单球局部放热的强弱主要受边界层发展过程的支配。球床中,边界层的发展受到死点的限制,元件球的放热主要受气流交混扰动的影响;随着雷诺数的增加,元件球平均放热系数升高,放热不均匀系数下降。不同的球阵布置对元件球放热分布有明显的影响。     

Analysis of heat source distribution in internal circulating surface heat transfer molten salt reactor

The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods.     

Burnup optimization of once-through molten salt reactors using enriched uranium and thorium

The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thori...     

Improvements and validation of the transient analysis code MOREL for molten salt reactors

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective.     

An experimental study on pressure drop and dryout heat flux of two-phase flow in packed beds of multi-sized and irregular particles

This paper is concerned with debris bed coolability in a postulated severe accident of light water reactors, where the debris particles are irregular and multi-sized. To obtain and verify the friction laws predicting the hydrodynamics of the debris beds, the drag characteristics of air/water single- and two-phase flow in a particulate bed packed with multi-sized spheres or irregular sand particles were investigated on the POMECO-FL test facility. The same types of particles were then loaded in the test section of the POMECO-HT facility to obtain the dryout heat fluxes of the particulate beds heated volumetrically. The effective (mean) particle diameter is 2.25 mm for the multi-sized spheres and 1.75 mm for the sand particles, determined from the Ergun equation and the measured pressure drop of single-phase flow through the packed bed. Given the effective particle diameter, both the pressure drop and the dryout heat flux of two-phase flow through the bed can be predicted by the Reed model. The experiment also shows that the bottom injection of coolant improves the dryout heat flux significantly and the first dryout position is moving upward with increasing bottom injection flowrate. Compared with top-flooding case, the dryout heat flux of the bed can be doubled if the superficial velocity of coolant injection is 0.21–0.27 mm/s. The experimental data provides insights for interpretation of debris bed coolability (how to deal with the multi-sized irregular particles), as well as high-quality data for validation of the coolability analysis models and codes.     

基于多孔介质模型的行波堆TP-1堆芯稳态温度场与流场数值模拟

根据TerraPower公司最新设计的钠冷行波堆TP-1的具体结构和运行特点,采用多孔介质模型,使用商用软件CFX对行波堆堆芯的热工水力特性进行数值模拟,得到了TP-1稳态运行条件下堆芯温度场、速度场和压力场分布。结果表明：应用多孔介质模型对行波堆堆芯进行三维热工水力数值模拟的方法直观、快速、有效,将它应用于行波堆堆芯稳态条件下三维流场和温度场分析具有一定的意义。     

The influence of Xe-135m on steady-state xenon worth in thermal molten salt reactors

There has been a resurgence of interest in fuel-in-salt Molten Salt Reactors (MSR); a number of governments and private companies are currently undertaking efforts to develop and commercialize MSR technology. Recent nuclear models used in the TENDL nuclear data library have estimated the cross section of the metastable state of ¹³⁵Xe, ^135mXe, to have a much larger cross section than the ground state of ¹³⁵Xe. Thermal MSRs with continual online noble gas stripping of the fuel salt can operate in a regime where ^135mXe makes up a notable fraction of the xenon worth, necessitating the implementation of these new cross-sections in the neutronic analysis of these advanced reactor types. To estimate the effect of ^135mXe on reactor operation, a simplified mathematical model was produced with one neutron energy group and ^135mXe cross section data from the TENDL-2015 nuclear data library. ²³⁵U and ²³³U systems were investigated. It was found that the steady-state xenon reactivity worth was considerably higher for some modes of operation when ^135mXe was included in the xenon worth calculations. Based on available literature, it was found that proposed MSR concepts may operate in the modes of operation where ^135mXe has a notable impact on steady-state xenon worth. This work highlights the need to include ^135mXe in MSR models and the importance of acquiring evaluated cross-sections for ^135mXe.     

基于球床密实实验装置的熔盐堆分流板设计

在熔盐球床堆设计中,为实现堆芯内部燃料球堆积结构的稳定性,需保证堆芯内部的流场均匀分布。研究基于模拟熔盐球床堆堆芯水力特性的球床密实实验装置(Pebble bed dense experiment facility,PBDE),通过设计不同形状和不同孔道分布的分流板,运用计算流体力学(Computational fluid dynamics,CFD)方法使用FLUENT软件对其堆芯内部的流场分布进行数值模拟,目的是保证实验中堆芯的流场分布均匀稳定。模拟结果表明,平板形分流板较锥形分流板能更好地使堆芯内流场均匀分布;且增加分流板的孔道数目或减小孔径能使堆芯内部的流场更加均匀稳定;比较设计的6种分流板模拟结果,最终给出满足PBDE堆芯流场均匀分布的分流板,为PBDE实验提供了基础,也为熔盐球床堆的堆芯流量分配设计提供技术方案与选型参考。     

200MW低温供热堆盒间流道流动传热数值分析

采用三维CFD软件Phoenics-3.2,计算了200MW低温供热堆燃料组件盒间的流场及温场。研究了旁通流量、控制棒提升等因素的影响。在考虑这些因素之后,得出了最佳旁通入流方案。     

Experimental study on heat transfer and pressure drop in mercury flow system for spallation neutron source

In the design of MW-class spallation target system, using mercury to produce practical neutron applications, keeping the highest level of safety is vitally important. To establish the safety of spallation target system, it is essential to understand the thermal hydraulic properties of mercury. Through thermal hydraulic experiments using a mercury experimental loop, which flows at the rate of 1.2 m³/hr maximum, the following facts were experimentally confirmed. The wall friction factor was relatively larger than the Blasius correlation due to the effects of wall roughness. The heat transfer coefficients agreed well with the Subbotin correlation. Furthermore, for validation of the design analysis code, thermal hydraulic analyses were conducted by using the STAR-CD code under the same conditions as the experiments. Analytical results showed good agreement with the experimental results, using optimized turbulent Prandtl number and mesh size.