全隐二阶迎风差分格式在快堆模拟中的应用

鉴于一般差分格式在模拟快堆系统中间热交换器的事故时会产生温度振荡现象，为了保证数值解法的稳定性和差分精度，采用全隐二阶迎风差分格式对中间热交换器进行了流量失衡模拟，彻底解决了温度振荡问题，所得结果和实验结果吻合良好，而且还可以达到超实时的模拟效果。最后，还将该方法用于对堆芯失流工况的模拟，也取得了良好的模拟效果。     

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。     

非均匀硼稀释瞬态三维流场计算

采用CFX10.0模拟了反应堆发生非均匀硼稀释事故时的瞬态三维流场,得到堆芯冷却剂的硼浓度分布和温度分布。比较三种不同堆芯温度工况计算结果,发现随着堆芯温度的增加,运动阻力下降,清水越快到达堆芯,清水与硼水的搅混时间减少,搅混效果减弱,堆芯中心处冷却剂的硼浓度偏低。     

HTR-PM堆芯出口热气混合实验相似性分析

球床模块式高温气冷堆核电站（HTR-PM）堆底设有热气混合结构，使堆芯流出的氦气混合均匀。堆芯出口热气混合实验用于测量和分析该混流结构的混合性能及其阻力特性。为使设计的热气混合实验系统及实验工况能反映HTR-PM的混流结构的实际混合性能和阻力特性，在确保实验经济成本的前提下，根据相似性准则，分析确定了堆芯出口热气混合实验系统的设计准则和具体参数，并利用Fluent软件对所设计的实验装置内的流场和温度分布进行了数值模拟。该混合实验系统及其工况与HTR-PM实际堆底混流结构具有相似性，在此实验的基础上，可通过理论分析和数值模拟得到HTR-PM实际堆底混流结构的混合性能和阻力特性。     

“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。     

百万千瓦级压水堆堆芯氙稳定性分析

由于压水堆堆芯对于总功率振荡具有固有的稳定性,且控制和保护系统对堆芯总功率不稳定提供了保护,因此关于堆芯稳定性的讨论仅限于氙致功率空间振荡(即氙稳定性)以及对径向功率分布和轴向功率分布的影响。本文主要通过三维及一维程序进行多种氙瞬态构造,对轴向氙振荡及径向氙振荡进行分析,以验证百万千瓦级压水堆堆芯在氙振荡方面是固有稳定的或可被控制的。计算结果表明,该堆芯在径向氙振荡方面固有稳定,首循环在轴向振荡方面稳定性较平衡循环略差,但可通过堆芯控制系统进行控制,使堆芯处于一个安全状态,燃料设计限值不被突破。     

压水堆内钍-铀增殖循环研究——堆芯设计

在全UOX（铀氧化物）堆芯平衡循环的基础上,研究了UOX/PuThOX（钚钍混合氧化物）混合堆芯和UOX/U3ThOX（工业级²³³U-钍混合氧化物）混合堆芯的燃料管理方案设计,实现了钍 铀增殖循环。U3ThOX燃料组件在堆内停留6个燃料循环,平均循环长度较参考的全UOX堆芯增加5 EFPD;U3ThOX燃料组件卸料后冷却1年时易裂变核素存量较装料时增加了7%。为比较分析,设计了UOX/MOX（钚铀混合氧化物）混合堆芯的燃料管理方案。核特性分析结果表明：1）装载PuThOX燃料对堆芯核特性产生的影响与装载MOX燃料类似,硼微分价值和控制棒价值减小、临界硼浓度变大、慢化剂温度系数更负、停堆裕量减小、多普勒亏损更大;2） UOX/U3ThOX混合堆芯和参考的全UOX堆芯具备相似的核特性。     

压力容器环形下降段内含硼溶液流动混合特性的数值模拟分析

发生堆芯应急冷却安注时,外部注入的含硼冷却剂与稀释水团将在环形下降段内发生混合,含硼冷却剂与稀释水团混合不均匀可能导致堆芯重返临界。本文基于Fluent 18.0对环形下降段内的流动混合特性进行分析。横截面的速度分布显示,入口截面的水平方向速度随周向位置的增加而显著衰减,而环形下降段下部区域横截面的速度分布趋于平缓;三维流线图显示,流体进入压力容器后在环腔内壁发生剧烈碰撞,随后绕环形下降段呈放射状流动。通过自定义硼酸溶液,并模拟其与稀释水团之间混合,数值结果与相关的实验研究结果较为一致;三维浓度分布显示,雷诺数较低时入口硼酸溶液将停滞在环形下降段上部空间,增加入口雷诺数有利于搅混均匀。     

方家山核电站延伸运行期间氙振荡控制

对于大型商用反应堆,运行过程中出现一定扰动后,容易产生氙振荡。小幅度的氙振荡不会影响堆芯安全。大幅度的氙振荡会使堆芯局部区域的温度持续升高,可能会突破燃料设计限值,使燃料元件熔化。方家山核电站2号机组延伸运行期间出现了轴向氙振荡加剧的情况,通过适时移动控制棒组、降低单次降功率变化量、选择氙振荡负周期前3/4时段降功率的方式,有效地抑制了氙振荡,在保证堆芯安全的前提下顺利实施了延伸运行。     

热管熔盐堆堆芯倾角对堆芯熔盐自然对流影响研究

热管熔盐堆堆芯倾角对堆芯温度分布和局部热点具有重要影响。为获得堆芯在不同倾角下内部熔盐的自然对流换热特性、优化堆芯设计和提高系统安全性，对堆芯进行三维建模，通过Fluent软件进行数值模拟，对横置和竖置放置2种情况下堆芯内熔盐自然对流的温度场和流场进行了分析，同时讨论了堆芯倾角变化对堆芯温度场及局部热点的影响。研究结果表明：局部热点始终出现在堆芯上部，相对于竖置，横置时堆芯温度场及流场更加不稳定。当倾角在5°～10°范围内，局部热点温度最高，竖置时热点温度最低。模拟结果揭示了堆芯内熔盐的自然对流特性，并为热管熔盐堆热工方面的概念设计提供了一定参考。     

Establishment of the design requirements for a flow blockage detection system through a LES analysis of the temperature fluctuation in the upper plenum

When an assembly in a core is partially blocked, the temperature in the upper plenum fluctuates at an early stage. Therefore, the temperature fluctuation in the upper plenum can provide the information about a local blockage of an assembly. For developing the detection algorithm for the partial blockage, we analyzed the temperature fluctuation in the upper plenum due to the partial blockage in an assembly. The LES turbulence model in the CFX code was used for analyzing the temperature fluctuation in the upper plenum because the LES is suitable for analyzing the time dependent turbulence variables. After analyzing the temperature fluctuations in the upper plenum, we established basic design requirements for the flow blockage detection system through a FFT analysis and some statistical analysis. We concluded that response time of a measuring device was less than 13 m s and that it should cover a high temperature range of 1000 K. In addition, the resolution of the thermocouple was less than 2 K and its location should be within 25 cm from the exit of each assembly.     

压水堆上腔室冷却剂温度振荡现象研究

为了解压水堆上腔室冷却剂的温度振荡现象,利用专业计算软件CFX,采用大涡模拟（LES）方法,对简化上腔室内的瞬态流场进行数值模拟,并与实际数据进行对比分析。结果表明,LES模型可较好地模拟上腔室的温度振荡现象;上腔室出口区域的温度波动多集中于低频部分,未呈现出明显的周期性;出口位置对流场温度分布有明显影响,自入口至上腔室出口中心线所在平面,外围及中心部分温度波动幅度较大,其余区域温度变化幅度较小;而自上腔室出口中心线所在平面至顶部区域,温度波动逐渐趋缓。     

Large eddy simulation of thermal striping in the upper plenum of the PGSFR

A computational study of thermal striping in the upper plenum of the prototype generation-IV sodium-cooled fast reactor (PGSFR) being developed at Korea Atomic Energy Research Institute is presented. First, previous experimental and numerical studies on the thermal striping are briefly discussed. Both Reynolds-averaged Navier–Stokes (RANS) and large eddy simulation (LES) approaches are employed for the simulation of thermal striping in the upper plenum of the PGSFR. For the RANS approach, the conventional k ? ? turbulence model is employed and the LES is performed using the wall-adapting local eddy viscosity model. From the RANS results, the time-averaged velocity components and temperature field in the complicated upper plenum of the PGSFR are calculated. In the LES results, the time history of temperature fluctuation at several locations of upper internal structure (UIS) and intermediate heat exchanger (IHX) are additionally stored. Comparisons of the predicted time-averaged velocity and temperature between the two methods show that the prediction by the LES shows faster thermal mixing than that by the k ? ? turbulence model. From the computed results of the temporal variation of temperature, it was possible to find the amplitude and frequency of the temperature fluctuation at the several locations of the UIS and IHXs. It was found that the location where the thermal stress is largest in the upper plenum of the PGSFR is the ?-shape region of the first grid plate.     

Optimization of the upper plenum of a PBMR to enhance thermal performance in the reactor core

An upper plenum of a PBMR type gas cooled nuclear reactor has been optimized using three-dimensional Reynolds-averaged Navier–Stokes (RANS) analysis and surrogate modeling technique. Shear stress transport turbulence model is used as a turbulence closure. Two geometric design variables viz., ratio of height of upper plenum to diameter of rising channels, and ratio of height of the slot at inlet to that at outlet, are used as design variables for the optimization. Design points are selected by Latin-hypercube sampling. The objective function is defined as a linear combination of uniformity of temperature distribution in the core and pressure drop through the upper plenum. The optimal point is determined through surrogate-based optimization method which uses RANS derived calculations at design points. The results show that the optimization improves considerably both the temperature uniformity and the friction performance.     

Upper Plenum Dump during Reflood in PWR Loss-of-Coolant Accident

Upper plenum dump during reflood in a large break loss-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood.

The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnitude of water subcooling.     

Experimental Study of Two-Dimensional Thermal-Hydraulic Behavior in Core during Reflood Phase of PWR LOCA

Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles due to the nonuniform water accumulation in the upper plenum.     

NHR—200堆芯旁通区三维流动传热数值分析

应用三维CFD软件PHOENICS－3．2,计算了200MW低温供热堆（NHR－200）堆芯旁通区及上腔室的流场和温场。分析了在堆芯与围板间的乏燃料存放区上端不同档板布置方案下的流场和温场,并考虑了旁通流量的影响。自然对流对流场和温场的影响不大,不会改变主流方向。在计算区域内,除主流外,还有由堆芯旁通区的下部流通面积突扩造成的一回流区及上腔室堆芯出口流通面积突扩和自然对流而形成的一大回流区。加挡板可阻挡上部大回流区对堆芯旁通区的影响,降低堆芯旁通区流体温度的变化。     

三环路压水堆压力容器上腔室交混矩阵数值研究

使用STAR-CCM+软件对三环路压水堆压力容器上腔室流场进行了大规模、精细化三维数值模拟,并采用组分跟踪方法分别对157个燃料组件出口冷却剂流动进行计算,构造了一个具有3×157个元素的“上腔室交混矩阵”,用该矩阵即可定量、精确地描述冷却剂从堆芯流出后,经上腔室内交混并再分配到各热管道的复杂流动过程。研究发现堆芯流出的冷却剂在压力容器上腔室内的交混是并不充分的,径向上不同位置燃料组件流出的冷却剂会在上腔室同热管道的接口区域存在明显的对应关系,而燃料组件径向功率分布的差异必然导致热管道中冷却剂热分层现象的产生。      

Initial Thermal-Hydraulic Behaviors under Simultaneous ECC Water Injection into Cold Leg and Upper Plenum in a PWR-LOCA

Core thermo-hydrodynamic characteristics under the combined injection mode before and just after the beginning of bottom reflood of a PWR-LOCA were experimentally studied by performing three tests in Slab Core Test Facility simulating a full radius slab section of a PWR. Emergency core cooling water was simultaneously injected into the upper plenum and the intact cold leg. The subcooling and the radial distribution of the upper plenum injection water were the test parameters.

The core was cooled by falling water before the beginning of bottom reflood. However, the core was finally quenched by bottom reflood. Before the beginning of bottom reflood, the transients of water level in the lower plenum were different among three cases, that is, the water level was rapidly or gradually increased in the first and second cases, respectively, or remained below the bottom of core barrel in the third case. The bottom reflood was much delayed in the last two cases. Even under the conditions with large upper plenum injection rate of subcooled water and with steam escape through the lower plenum, continuous fall back was not observed but the subcooled water was intermittently supported by the upward steam flow generated in the core.     

A thermal-fluid assessment of a cooled-vessel concept for a VHTR

Flow distribution and thermal analyses of a conceptual design of a cooled vessel for a very high temperature reactor (VHTR), which has a forced vessel cooling with an internal coolant path through a permanent side reflector, have been performed. A computational fluid dynamics (CFD) code was employed to investigate flow distributions at inlet and upper plenums of the proposed cooled-vessel concept. Thermal-fluid analyses of the cooled vessel during a normal operation were carried out by using the CFD code with the boundary conditions provided by the GAMMA system analysis code. The transient analyses during postulated accidents were conducted by the GAMMA code itself. According to the results, the flow deviation at the riser holes due to a change of the inlet flow path to the core inlet is about ±20% which results in about a 3-7% core flow deviation from the average value depending on the upper plenum height. The pressure drops in the inlet and upper plenums are estimated to be from 13 to 25 kPa with a change of the upper plenum height. A cooling flow of more than 4 kg/s is sufficient to maintain the RPV temperature within the required limit during a normal operation. Transient analysis reveals that the reactor vessel is exposed to a temperature above its limit of 371 °C but this duration is shorter than the allowable time for a creep region with a sufficient safety margin. The results suggest that the cooled-vessel concept considered in this paper has the potential to be used for a VHTR but further and more detailed studies are required to realize the proposed concept.