共查询到17条相似文献,搜索用时 578 毫秒
1.
压力容器流场特性是反应堆热工水力设计的重要依据之一。论文采用三维数值模拟方法,建立了包括进口及环形下降段、下腔室及堆芯进口段、堆芯段的华龙一号反应堆压力容器下腔室分析模型,并采用多孔介质模拟堆芯段压降及流动,在网格数量级敏感性分析的基础上确定了最终网格模型,对运行工况下压力容器下腔室冷却剂的流动特性进行了研究。结果表明,下腔室出现逆时针漩涡流动,冷却剂在冲刷格架板后在下腔室底部汇集并向上流入堆芯;通过分析格架板的上、下表面压差发现大、小格架板所受水力冲击方向相反,载荷大小相近;对下堆芯板流水孔归一化流量分配进行了分析。通过求解附加标量浓度输运方程以标记并跟踪冷却剂的分布和交混,结果表明冷却剂随着流动发生逆时针横向交混,平均有43.7%的冷却剂份额会偏移至逆时针的相邻堆芯进口位置,表明交混特性较好。 相似文献
2.
秦山二期核电厂反应堆下腔室交混特性CFD分析研究 总被引:1,自引:0,他引:1
运用CFD方法对秦山二期核电厂反应堆下腔室的冷却剂流动及交混特性进行了计算分析,并与反应堆整体水力模拟试验结果进行对比。结果显示:对于堆芯入口流量分配特性,无论采用迎风差分格式还是高精度差分格式,CFD计算结果均与试验结果符合较好;对于下腔室交混特性,两种差分格式的计算结果均与试验结果差异较大,相对而言,迎风格式的计算结果在最大与最小交混因子方面与试验结果更接近。进一步分析发现,是否考虑主泵引起的螺旋流动很可能是造成计算与试验结果偏差的主要原因。 相似文献
3.
4.
5.
为研究小型压水堆下腔室的交混特性,本文基于比例模化方法,开展小型压水堆1∶3比例模型水力学实验,通过测量溶液浓度变化,获得在冷管流量均衡和非均衡工况下堆芯入口的交混因子矩阵。研究结果表明,均衡流量工况下,冷管流量的变化对堆芯入口交混因子矩阵未产生明显影响;非均衡流量工况下,靠近出口管的燃料组件交混因子受流量不均衡的影响较大,而中心区域的交混因子变化幅度较小。由此可见,小型压水堆在均衡流量下具有较稳定的下腔室交混特性,而在非均衡工况下需要重点关注出口附近燃料组件交混特性的变化。 相似文献
6.
压水反应堆各个环路中的冷却剂在下腔室发生剧烈湍流交混,下腔室腔体内产生大量涡流,会导致堆芯燃料组件入口流量随机震荡,引发堆芯瞬态流动不稳定性,可能影响到反应堆热工、结构安全或传热性能。本文对反应堆内燃料组件区域流动特性开展研究,通过水力学试验手段获得反应堆堆芯在多种运行工况下,下腔室安装流量分配裙和不安装流量分配裙时的堆芯燃料组件入口流量脉动数据,试验结果表明,流量分配裙对下腔室涡流的抑制效果明显,在碎涡整流作用下,堆芯流量脉动明显降低;随着运行环路数的减少,下腔室流场对称性降低,涡流增强,堆芯流量脉动明显增大;下腔室涡流还会对堆芯入口流量分配均匀度造成不利影响,流量脉动偏大区域对应的流量分配因子明显较小。 相似文献
7.
8.
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。 相似文献
9.
10.
本研究利用子通道程序,基于已有的实验数据,对棒束通道的单相和两相交混模型进行了评估。单相交混主要考虑横流和湍流交混,横流由守恒方程决定并在流量分布中占主导作用,湍流交混取决于交混系数,对湍流交混研究发现Sadatomi模型预测结果与实验结果吻合较好。两相交混由横流、湍流交混和空泡漂移共同作用,通过已有模型预测结果与实验数据对比分析,推荐两相交混中空泡漂移采用Hotta模型、湍流交混系数采用Sadatomi模型和两相乘子采用Beus模型,这是一个预测结果较为保守的组合模型,有利于反应堆安全的保守性评估。 相似文献
11.
12.
The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. During the design process multiple studies have been performed to develop safety codes for the reactor. Two major accidents of interest are the Pressurized Conduction Cooldown (PCC), and the Depressurized Conduction Cooldown (DCC) scenario. Both involve loss of forced coolant to the core, except the latter involves a pressure loss in the main coolant loop. During normal operation a circulator pumps the coolant into the upper plenum and down through the core, but following a loss of forced coolant the natural convection causes the flow to reverse to go through the core into the upper plenum. Computer codes may be developed to simulate the phenomenon that occurs in a PCC or DCC scenario, but benchmark data is needed to validate the simulations; previously there were no experimental test facilities to provide this. This study will cover the design, construction, and preliminary testing of a 1/16th scaled model of a VHTR that uses Particle Image Velocimetry (PIV) for flow visualization in the upper plenum. Three tests were run for a partially heated core at statistically steady state, and PIV was used to generate the velocity field of three naturally convective adjacent jets; the turbulent mixing of the jets was observed. After performing a sensitivity analysis the flow rate of a single pipe was extracted from the PIV flow field, and compared with an ultrasonic flowmeter and analytic flow rate. All the values lied within the uncertainty ranges, validating the test results. 相似文献
13.
14.
在华龙一号反应堆结构设计中,为提高反应堆结构的安全性,将堆芯测量探测器从反应堆下腔室引出改为从反应堆压力容器顶盖引出,下腔室结构发生改变,影响了堆芯入口流场的均匀性,故需要重新设计下腔室搅混结构以使流场分布均匀。通过对比百万千瓦级国产化二代改进型压水堆(CNP1000)、百万千瓦级先进非能动型压水堆(AP1000)及欧洲先进压水堆(EPR)3种堆型反应堆下腔室结构,结合华龙一号自身下腔室结构特点,借鉴其他堆型以及提出新型结构,共提出了4种结构优化方案,分别对不同方案进行建模并利用计算流体力学(CFD)分析软件进行计算,从结构、制造、安装及流场分析等方面对4种新型下腔室搅混结构和CNP1000下腔室搅混结构进行对比分析,得出采用流量分配板结构的反应堆下腔室搅混结构为最优方案,其既能均匀搅混下腔室流场,又能使堆芯入口流量分配均匀。 相似文献
15.
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。 相似文献
16.
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. 相似文献