等速入口三喷口流动的温度振荡数值模拟

由于来自核反应堆堆芯不同通道冷却剂的温度存在差异,它们在上腔室混合后将产生复杂的温度振荡现象,可能干扰堆芯出口的温度测量,也可能造成周围固体结构的热疲劳破坏。本文基于FLUENT软件,采用大涡模拟湍流模型(LES)对三维平行三喷口流动进行数值模拟,并与实验结果进行了比较。结果表明：随着入口速度的增加,流体混合越来越剧烈,同时温度振荡的最大振幅逐渐减小,而相应的频率则逐渐增加。     

用于同轴喷射流引起的温度振荡数值模拟的湍流模型有效性研究

本文采用Fluent软件对同轴喷射流冷热流体搅混引起的三维温度振荡现象进行数值模拟。模拟中，分别选取几种代表性的湍流模型，如大涡模拟(LES)、雷诺应力模型(RSM)及标准k-ε模型。并将模拟结果与实验结果进行了对比，发现：LES是最适合预测冷热流体搅混特性的模型；LES可准确预测温度振荡在半径、高度和方位角方向的平均温度，而雷诺平均法(RANS)的结果则延长了流体的混合过程；LES可预测流体的瞬态温度振荡，RANS则无法预测。     

上游带弯头的T型管内混合流动数值模拟的湍流模型评价

T型管是研究冷热流体混合流动及其引起的温度振荡现象的典型几何模型,而上游带弯头的T型管又是一不可忽视的特殊情形。本文运用计算流体力学软件,采用3种湍流模型（RNG k-ε 模型、SSG雷诺应力模型、LES模型）对上游带弯头T型管内冷热流体的交混现象进行模拟,并与实验数据进行了对比。结果表明：非混合区域如上游弯头内,RNG k-ε 模型、SSG雷诺应力模型的模拟结果与实验结果较吻合,而在混合区内LES模型的模拟结果更能表征实际流动。     

稳压器波动管内热纹振荡现象数值模拟研究

对波动管内热纹振荡现象展开研究,获得波动管内热纹振荡现象的产生原因以及启堆升功率工况下热纹振荡现象造成波动管内壁面温度振荡的幅值和频率。采用计算流体力学(CFD)方法中的动态Smagorinsky涡粘模型对波动管内热纹振荡现象进行了数值模拟。结果表明:波动管内产生热振荡现象的原因是冷热流体相对流动所产生的涡状流使得冷热交界面产生上下振荡;波动管内的热纹振荡现象使得波动管内壁面温度的振荡幅值在4~20℃范围内,频率低于5 Hz。     

基于大涡模拟(LES)的格架外条带区域压力和速度瞬态特性研究

分别运用雷诺平均模拟(RANS)方法和大涡模拟(LES)方法对燃料组件格架外条带区域的流场和压力场进行瞬态求解。分析发现:采用2种方法所计算出的速度分布大体一致,但是RANS方法不能捕捉到速度和压力的瞬态变化特性,而LES方法能够有效观察到流场的瞬态特性。通过2种方法的对比,可发现部分局部区域计算结果有较大差异。LES计算结果呈现出计算域内明显的压力波动,频谱分析发现格架不同区域具有不同的压力波动特性。     

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。      

立式倒U型管蒸汽发生器倒流现象及初步分析

文章涉及中国核动力研究设计院自然循环实验装置单相稳态自然循环实验过程中立式倒U型管蒸汽发生器(UTSG)模拟体一次侧流体的流动特性。实验观察到：1)UTSG模拟体进口腔室压力低于出口腔室压力；2)UTSG模拟体入口腔室温度较热段温度有一陡降。通过对该实验现象的分析可以判定，在单相自然循环工况下，UTSG模拟体中某些传热管内出现了倒流。实验结果表明，倒流的出现使UTSG模拟体自然循环工况下的流动阻力系数较强迫循环工况下的明显增大。      

日本文殊原型快堆堆芯出口腔室热分层现象数值模拟

本文利用商业CFD程序STAR-CCM+,采用合理的网格生成技术及物理模型,对日本文殊原型快堆堆芯出口腔室建立近似1∶1的模型,模拟分析40%额定功率停堆过程中堆芯出口腔室的瞬态工况,获得腔室内较为完整的热分层进程。结果表明：停堆2 min后腔室内出现稳定热分层现象;10～21 min时热分层通过上升桶桶顶位置;10～140 min热分层处于上升筒顶端位置附近期间,腔室内流型不稳定;140 min后热分层完全处于上升桶顶,桶内流型稳定且接近于停堆前。模拟结果与实验数据对比表明,停堆初期4 min内两者符合较好,表明本文模拟方法适用于停堆工况堆芯出口腔室热分层进程模拟;之后模拟进程明显快于实验,分析其偏差主要来自模拟边界及结构与实际的差异。     

重力注水下再淹没棒束通道数值模拟研究

基于棒束通道注水再淹没系统,提出重力注水方案,运用RELAP5/MOD3.2建立其再淹没模型,模拟从棒束底部依靠重力注水再淹没高温棒束通道时的骤冷现象。模拟结果显示:再淹没的过程中出现持续的流动振荡,冷却水周期性地注入、逐出,振荡过程可分为初始阶段的剧烈振荡和后续阶段的平稳振荡,在此期间的注入流量,棒束通道汽空间压力,壁面温度和传热系数都出现相应的周期性变化;同时进一步模拟分析了注入水入口尺寸、初始包壳温度、蒸汽出口尺寸三个因素对振荡的影响机制,发现其振荡周期和振荡幅度随注入口的增大,初始包壳温度的升高、蒸汽出口尺寸的减小而增大。     

平行喷口流动引起温度振荡流固热耦合数值模拟研究

本文基于FLUENT软件,采用大涡模拟(LES)方法对平行喷口冷热流体搅混产生的流固热耦合现象进行了数值模拟。首先将模拟值与实验值进行了对比,验证了模拟方法的正确性。其次,针对多种入口工况进行计算和分析,得出:某一时刻,流体温度处于波谷时,热量从固体传递给流体,下一时刻,当流体温度处于波峰时,热量则从流体传递给固体,故而形成了流固间周期性的热传递现象;随着固体厚度的增加,温度振荡的振幅呈非线性减小,但频率不变;随着入口速度的增加,流体和固体温度振荡的主频逐渐增大,当速度从0.5 m/s变化到1.5 m/s时,主频率可从2 Hz升高到11 Hz。     

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.     

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.     

Pressurized water reactor (PWR) hot-leg streaming: Part 1: Computational fluid dynamics (CFD) simulations

A thermo-hydraulic analysis model was developed to analyze thermal stratification phenomena observed in the hot-legs of pressurized water reactors (PWR). The model uses VIPREW code to determine the flow field and temperature distribution in the reactor fuel region. The temperature readings from the thermal couples located at the exit of the reactor core were used to compare with the VIPREW computed results. The predicted values agree well with the measurements. The VIPREW results are then used as the boundary conditions for the CFD analysis. The CFD computational domain includes the upper plenum and hot-legs and the fifty two (52) control rod guiding tubes to properly include the additional obstructions imposed to the fluid. Different fuel loading patterns were studied to investigate the effects of different power distribution and fuel channel exit water temperature on hot-leg thermal stratification magnitude. The analysis results show that the 52 control rod guide tubes have major contribution to the mixing effect in the upper plenum. The sudden expansion of the cross sectional area in the upper plenum leads to the formation of recirculation vortex that prolongs the duration of coolant in the reactor vessel. The hotter coolant from the center portion tends to flow upwards to the top before exiting at the upper portion of the hot-leg pipes. It leads to higher temperature in the upper portion of the hot-legs. Water from the cooler outer fuel channels tends to trap in the recirculation region before exiting from the lower portion of the hot-legs.     

高温气冷实验堆堆芯出口热气联箱混合性能对比试验

１０ＭＷ高温气冷实验堆堆芯出口冷却剂温度径向分布很不均匀，若不使之均匀化，将造成蒸汽发生器件部上过大的热应力，设置在堆底反射层中的堆芯出口热气联箱的作用之一是使冷却剂氦气在其中得到充分的热混合。     

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.     

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.     

用大涡模拟计算流致振动的流体激励力

本文对湍流大涡模拟理论的脉动过滤和Smagorinsky亚格子模型进行了介绍,并用大涡模拟对秦山II期反应堆1∶5模型压力容器和吊篮的环腔内流场进行了数值模拟,计算了吊篮表面压力的时间分布和空间分布,其压力脉动的功率谱密度计算值和实验值在同一个量级内.因此,可以用计算所得的压力作为吊篮振动的激励载荷.     

Effects of Particle Confinement and Recycling on Thermally Stable Regions in D-T Tokamak Plasma

The water accumulation phenomena in the upper plenum of a PWR was investigated by using the data of the Cylindrical Core Test Facility (CCTF) test.

When the liquid inventory in the upper plenum is small, the liquid carryover rate from the upper plenum was governed by the liquid inlet flow rate. After some amount of the water was accumulated in the upper plenum, the liquid carryover rate was governed by the steam up-flow rate and the liquid inventory in the upper plenum. The liquid carryover rate was higher with the higher steam up-flow rate and the larger liquid inventory. This trend on the liquid carryover rate was also observed in small scale model test.

Based on the CCTF data, the empirical correlation on the liquid carryover rate was obtained.