棱柱式高温气冷堆堆芯间隙旁流模拟研究

棱柱式高温气冷堆的堆芯由燃料组件砖块分层、分区垒砌组成,考虑到加工误差以及结构装配,组件之间需要保证一定尺寸的间隙,形成的间隙流道将分流一部分堆芯冷却剂流量,简称间隙旁流。间隙旁流是堆芯结构以及堆内构件设计需要分析的重要因素,为了研究其对于反应堆热工流体性能的影响,采用商用计算流体力学(Computational Fluid Dynamics,CFD)程序ANSYS CFX针对MHTGR-350(Modular High-Temperature Gas Reactor 350 MWt)堆型堆芯活性区内流动、传热的复杂现象开展三维数值模拟,通过建立组件砖块、燃料孔道、冷却剂通道以及间隙流道的详细模型,计算得到区域内的流量分配以及温度分布情况。选取关键参数开展敏感性分析,结果显示:进入狭长间隙流道的冷却剂流量主要由堆芯的结构布置以及间隙的尺寸大小决定,间隙越大、旁流占比越大,冷却效率越差,燃料的局部温度越高。同时,在反应堆运行寿期内,间隙尺寸将随着组件形变而发生变化,引起堆内温度分布以及出口温度发生波动,间隙越大引起的波动幅度也越大,不利于堆芯运行的安全性和稳定性。     

湍流模型在堆芯热工水力性能分析中的应用

针对正三角形布置堆芯棒束燃料通道内冷却剂充分发展湍流流场模拟,对比分析了计算流体动力学软件湍流模型对复杂流道内湍流流场模拟结果的影响。结果表明：湍流模型选取的不同对模拟结果有着显著影响,由于堆芯几何结构复杂,冷却剂流动为复杂三维流动,湍流呈高度各向异性。基于各向同性假设的湍流模型不能准确捕捉堆芯内冷却剂的二次流现象。基于求解雷诺应力输运方程的雷诺应力模型（RSM）能够较好地预测复杂流道内的二次流。本工作的研究结果为复杂流道流动换热模拟及深入研究分析堆芯热工水力性能提供了一定借鉴和指导。     

350MW环型球床高温气冷堆失冷失压事故下的热工特性研究

由于环型球床高温气冷堆特殊的堆芯结构，使其在失冷失压事故下堆内最高温度能够明显低于模块式球床高温气冷堆在相同事故下堆内最高温度。当堆芯热功率有较大幅度提高时，环型堆芯仍然能够凭借自身传热机能将衰变热量及时排出，满足失冷失压事故下燃料最高温度限制。这不仅增大了反应堆的安全性能，同时也能够有效地增加反应堆单堆功率，使环型球床高温气冷堆在经济上更具竞争力。本文研究环型球床高温气冷堆在提高功率水平时，反应堆在失冷失压事故下堆内的热工特性，并综合分析了几个重要的结构尺寸热工参数对失冷失压事故下燃料最高温度的影响。     

基于多孔介质模型的压水堆堆芯温场数值模拟

针对压水堆的复杂结构特点,对堆芯采用多孔介质模型,建立完整的压力容器堆芯模型,使用商用软件CFX对压力容器堆芯的热工水力特性进行数值模拟,得到偏环运行和典型事故工况下冷却剂的热工水力响应特性。计算结果表明:应用多孔介质模型能有效正确直观显示堆芯的冷却剂温度分布情况,在偏环运行工况下堆芯会出现偏心现象,而通过瞬态事故工况计算结果表明堆芯中上部冷却剂温度最高,对压水堆的热工安全具有一定指导作用。     

液体燃料熔盐堆三维稳态分析程序开发

液体燃料熔盐堆的物理热工特性与固体燃料反应堆有很大的不同,在分析计算中必须考虑燃料流动特性的影响,一般分析固体反应堆的程序均不能直接用于分析液体燃料熔盐堆。根据熔盐堆的流动特性,建立了液体燃料熔盐堆的三维中子动力学模型和流动传热模型,开发了针对液体燃料熔盐堆的三维稳态核热耦合程序,并以此分析了稳态情况下MOSART堆的物理热工特性。结果表明,堆芯流速对快中子和热中子影响较小,对堆芯温度和缓发中子分布影响较大。     

熔盐球床堆堆芯入口热工水力特性数值分析

堆芯入口流场设计是小型固态燃料熔盐堆系统项目内容之一,它对反应堆结构的稳定性、堆芯温度和流场分布有着非常重要的影响。研究了熔盐流道流通面积变化对堆芯入口温度、流场分布及压降的影响,优化熔盐流道几何结构。以小型熔盐球床堆模型为研究对象,取符合实际边界条件的输入参数,通过改变熔盐流道流通面积,使用计算流体力学(Computational Fluid Dynamics,CFD)通用程序Fluent 16.0对堆芯入口内熔盐的热工水力特性进行数值模拟。在考虑实际下反射层流道的流通面积占比最大为18.14%下,研究了熔盐流道流通面积占比在区间[0,15.00%]变化。结果表明,堆芯活性区熔盐最高局部热点温度随熔盐流道流通面积比的增大而增高;堆芯入口内的压降随下反射层熔盐流道流通面积比的减小而增大;在径向方向上流进孔道的熔盐流速随着孔道远离堆芯位置而增大。本研究可为小型固态燃料球床熔盐堆优化设计提供一定的参考价值。     

TOPAZ-Ⅱ型反应堆堆芯热工水力数值模拟

利用成熟的软件建立了TOPAZ-Ⅱ型反应堆三维流动及换热的计算模型,以结构化为主非结构化为辅相结合的方式进行了网格划分,以符合实际的边界条件等作为输入参数,对堆芯热工水力行为进行了数值模拟,经过数值求解得到了三维流场和温度场结果,并得到了冷却剂流量分配、压力分布、速度分布以及堆内慢化剂、端部及侧反射层等部件的温度分布。经比较,数值模拟结果与俄方设计结果符合较好,本研究为此类型研究堆热工水力设计优化积累了经验。     

窄矩形通道两相流动实验回路设计研究

为了更好地研究某燃料堆内窄矩形冷却剂通道的流动传热特性,调研了窄矩形通道传热的研究现状和发展趋势,结合NP工程反应堆相关参数和实验需求,搭建了一套热工水力两相流实验回路用来研究分析某燃料冷却剂通道传热特性。该实验回路根据实际的反应堆运行操作要求,设计了与实际冷却剂通道一致的实验段,流动方向设定为竖直向下流动,采用双面加热,流道间隙尺寸设定为2.3 mm,通道宽度为67 mm（加热宽度为62 mm）,流道长度为1000 mm（加热长度为750 mm）,并通过实验对其进行了验证。结果表明,本文装置正确可行。      

“华龙一号”反应堆精细化全堆芯大规模CFD数值模拟研究

精细化全堆芯大规模计算流体力学(CFD)数值模拟是"华龙一号"和数字化反应堆研究设计过程中的重要方法。本文通过一系列合理简化,建立了"华龙一号"反应堆全堆芯几何结构模型,并采取分组网格划分的方式对堆芯燃料组件进行离散,得到全堆芯CFD分析模型;通过精细化全堆芯大规模CFD数值模拟,可以获得堆芯完整流场分布特性和热工水力参数,验证"华龙一号"反应堆堆芯参数设计的合理性,为反应堆优化设计和安全运行提供参考。研究结果表明,由于"华龙一号"反应堆堆芯1/4对称结构和"三进三出"的1/3冷却剂进出口对称结构共同作用,堆芯流量分配因子在径向呈现先增加后减小的趋势,流量最大处不在堆芯正中心;在入口管嘴横截面上燃料组件最大温度约为331.2℃,温度分布不均匀,在径向总体呈现先增加后减小的趋势,最大温度区域也不在堆芯正中心,这与堆芯流量分配因子的趋势类似,是堆芯功率分布与冷却剂流量分配共同作用的结果。     

TMSR堆芯CFX多孔介质传热建模分析

钍基熔盐堆(TMSR)是一种使用石墨包覆颗粒作为燃料,熔盐作为冷却剂的第4代反应堆。TMSR堆芯区域的球形燃料增加了反应堆热工水力分析的复杂程度,为了分析反应堆在发生丧失强迫循环后堆芯的温度分布情况,需对整个堆芯进行CFD建模模拟。本文对TMSR堆芯进行几何建模和网格划分,并使用ANSYS CFX进行了多孔介质模型的建模模拟。在主要考虑导热换热和浮力影响以及两种不同的保温层厚度情况下,对堆芯稳态运行时的温度分布和发生事故后60s的瞬态温度分布进行了初步分析。研究结果证明了利用CFX及其多孔介质模型对TMSR堆芯进行模拟的可行性,并与REALP5-3D结果进行比较,初步验证了在该简化模型的边界条件下,堆芯熔盐短时间内不会发生沸腾。     

Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core

The thermal hydraulic calculations of the 10 MW high temperature gas-cooled-test module (HTR-10) are among the most important indications to judge the reactor performance under design conditions. The power distribution, the temperature distribution and the flow distribution of the HTR-10 are calculated for initial and equilibrium core in this paper. The temperature distribution includes the temperature parameters of fuel elements, the helium coolant and the main components in the reactor. In the temperature calculation of fuel elements, several uncertain factors are considered carefully, including non-uniform burnup, power distribution deviation, manufacture deviation of fuel elements, graphite balls mixed with fuel balls in the core, calculation deviation of heat transfer and so on. In the flow distribution calculation, the conservative pebble bed core flow value is selected. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.     

Optimization of a pebble bed configuration for quasi-direct numerical simulation

A high temperature reactor (HTR) is envisaged to be one of the renewed reactor designs to play a role in nuclear power generation including process heat applications. The HTR design concept exhibits excellent safety features due to the low power density and the large amount of graphite present in the core which gives a large thermal inertia in the event of an accident such as loss of coolant. However, the possible appearance of hot spots in the pebble bed cores of HTR may affect the integrity of the pebbles. This has drawn the attention of several scientists to understand this highly three-dimensional complex phenomenon. A good prediction of the flow and heat transport in such a pebble bed core is a challenge for CFD based on the available turbulence models and computational power. Such models need to be validated in order to gain trust in the simulation of these types of flow configurations. Direct numerical simulation (DNS), while imposing some restrictions in terms of flow parameters and numerical tools corresponding to the available computational resources, can serve as a reference for model development and validation. In the present article, a wide range of numerical simulations has been performed in order to optimize a pebble bed configuration for quasi-DNS which may serve as reference for validation.     

The impact of design on the decay heat removal capabilities of a modular pebble bed HTR

The decay heat removal capabilities are an important safety feature of the modular pebble bed HTR. It is designed in a way that also during loss of cooling accidents the decay heat can be removed purely by passive means without exceeding predefined temperature limits for fuel and structures. Such a plant design, however, yields limitations on the power output. Thus, from the thermal hydraulic point of view a reactor with maximum power which still obeys the temperature limits of fuel and components, represents an optimal design of a modular pebble bed HTR.In this paper, design options for a modular pebble bed HTR are discussed with respect to their capabilities of decay heat removal.Both pressurized and depressurized loss of coolant accidents are investigated. Optimization of design features is considered with reference not only to the maximum fuel temperature during the accidents, but also to the temperature of structures, mainly that of the reactor pressure vessel. It is pointed out that annular cores can produce higher power without exceeding fuel temperature limits, especially during depressurization accidents. This is mainly due to geometrical effects. Heat storage effects of the inner column also have an influence on the maximum fuel temperature by increasing the time at which this temperature is reached. While a thermal insulation of the core and the reflector increases the fuel temperature, maximum temperature of the pressure vessel and the core barrel is decreased. Thus, carbon blocks represent an important element for optimization of the design.     

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF₂ (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.     

高温气冷堆侧反射层纵向窄缝中的旁流研究

高温气冷堆的堆内构件由大量石墨块与碳砖构成,石墨块之间的窄缝会造成堆芯旁流,影响堆芯的流量与温度分布,需细致研究。石墨侧反射层有垂直方向的窄缝,是主要的旁流通道之一,氦气可能从冷氦联箱通过这些窄缝直接流入热氦联箱,也会与球床中的氦气发生横向交混。通过对球床流动及垂直窄缝中的旁流建立流体网络模型,分析了横向交混对窄缝旁流的影响,并讨论在不同窄缝大小及窄缝分布情况下旁流的变化规律。研究结果表明,球床边缘的氦气横向交混对旁流量影响较为明显,需在旁流分析中考虑,尺寸较大的窄缝对整个旁流的影响较为明显,窄缝尺寸较大时,堆芯的旁流量也更大。     

Investigating leakage and bypass flows in an HTR using a CFD methodology

An area that has been identified as significantly important in the development of a high temperature reactor (HTR) is the prediction of leakage and bypass flows. It is therefore essential to understand the influence of leakage and bypass flows on the thermal performance of an HTR.A methodology was developed to conduct an integral thermal analysis of a reactor using a CFD approach. One of the main objectives was to include leakage and bypass flow paths in order to provide a capability for simulating these very detailed flows.This paper investigates leakage and bypass flows through the PBMR reactor unit. It was found that, although these flows are dependent on the pressure drop through the pebble bed, a change in pebble bed pressure drop does not result in a similar change in the predicted leakages flows. It is also shown that the ability to account for leakage and bypass flows in an integral manner can help designers to focus their efforts on the specific regions that need to be targeted for the improvement of the life expectancy of the graphite blocks. Furthermore, leakage and bypass flows were found to reduce the pressure drop across the reactor unit while increasing the peak fuel temperatures.     

Neutronic and thermal hydraulic design of the graphite moderated helium-cooled high flux reactor

A new design concept for a high flux reactor was investigated, where a graphite moderated helium-cooled reactor with pebble fuel elements containing (²³⁵U, ²³⁸U)O₂ TRISO coated particles was taken as its design base. The reactor consists of an annular pebble bed core, a central heavy water region, and inner, outer, top, and bottom graphite reflectors. The maximum thermal neutron flux in its central heavy water region as high as 10¹⁵ cm⁻² s⁻¹ with thermal neutron spectral purity of more than two orders of magnitude and a large usable volume of more than 1,000 liters can be achieved by (1) diluting the fissile material in the core and (2) optimizing the core-reflector configuration. The in-core thermal-hydraulic analysis was done to obtain adequate information about the coolant flow pattern and pressure distribution, core temperature profile, as well as other safety aspects of the design. To protect the reactor during off-normal or accident events, a large margin of temperature difference between the operating fuel temperature and the fuel limit temperature is confirmed by lowering the coolant inlet and core rise temperatures.     

Development of a thermal–hydraulic analysis code for the Pebble Bed Water-cooled Reactor

The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly.Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper.     

Turbulence-induced Heat Transfer in PBMR Core Using LES and RANS

This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core. In the core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution.     

高温球床壁面区域有效导热系数模型优化

有效导热系数用来表征高温气冷球床堆堆芯综合传热能力,提高球床有效导热系数的预测精度对于高温气冷球床堆的热工设计和安全分析十分重要。为了优化球床壁面区域有效导热系数模型,本文针对无序石墨球床有效导热系数开展数值研究,分析了无序堆积球床主体区域、近壁面区域以及壁面区域有效导热系数的分布特性。结果表明：壁面区域有效导热系数相对于主体区域和近壁面区域显著降低,其平均降幅约为22%。因此引入了修正系数C_w对ZBS模型在壁面区域进行优化,对于球床主体区域及近壁面区域修正系数C_w=1,对于壁面区域,修正系数C_w=0.78。通过与前期无序球床实验数据和南非HTTU实验数据的对比,验证了优化后的ZBS模型能较好地预测球床壁面区域有效导热系数。