共查询到19条相似文献,搜索用时 156 毫秒
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为完成核电厂操纵员的培训和考试,需开展适用于高温气冷堆的模拟机研究。根据球床堆芯的特点,利用流体网络与传热网络建立了可实时计算的热工水力模型,讨论了球床中氦气沿径向的流动及换热对堆芯热工水力性能的影响。结果表明,在正常运行工况下,径向流动与换热对轴向流动的影响较小,模拟结果差别不大;而在失流不失压工况下,径向流动与换热对堆芯自然对流的形成、余热导出及整个瞬态过程的影响均较为明显,考虑径向流动与换热的仿真结果与设计软件的结果符合更好。考虑到模拟机仿真范围和逼真度的要求,在高温堆模拟机堆芯建模中需加入径向流动及换热模块。 相似文献
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石墨粉尘通过高温气冷堆堆芯球床结构的运动行为研究 总被引:1,自引:1,他引:0
高温气冷堆在运行过程中产生带有放射性的石墨粉尘,对反应堆的运行安全和环境安全造成一定影响。本文选取二维球床流场,采用离散相模型分析了堆芯球床结构对石墨粉尘颗粒的扩散和沉积的影响。计算结果表明:球床结构能有效阻碍石墨粉尘颗粒的扩散;沉积在球床结构上的石墨粉尘颗粒数目随堆芯内氦气流速的增加而增大,而由于受到颗粒惯性及热泳力的作用其增长趋势逐渐放缓;石墨粉尘颗粒在球床结构上的沉积效率随粒径的逐渐增加呈现"几乎不变-快速增长-缓速增长"的态势。 相似文献
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《中国原子能科学研究院年报》2019,(0)
<正>堆芯支承作为示范快堆中的重要设备,承担着支承、堆芯约束、流量分配等七大功能。堆芯支承的三大部件大栅板联箱、小栅板联箱及堆芯围桶中均包含螺栓零件,起到重要连接作用。1螺栓设计准则1)对小栅板联箱,螺栓预紧力需满足在SL1地震工况下小栅板联箱与大栅板联箱的结合面不发生离缝,且螺栓及对应的套筒(螺母)满足强度 相似文献
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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。 相似文献
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在聚变堆氦冷固态包层氚增殖区,球床通道内氦气流动压降特性对泵功率的设计具有重要意义。以氦冷固态包层氚增殖区为背景,研究了氦气流速、球床颗粒直径及球床通道长度对球床通道内氦气流动压降特性的影响。实验段采用20 mm×20 mm×500 mm的矩形通道,实验中氦气流速为0.1~0.6 m/s,球床颗粒直径为0.5、0.8、1.0、1.5、2.0 mm。实验结果表明,压降与氦气流速以及球床通道长度呈正相关,与球床颗粒直径呈负相关。对比Ergun关系式发现,在球床颗粒直径较小时,Ergun关系式预测值低于实验值,这主要是由于氦气可压缩性的影响。通过动量方程,理论推导出经可压缩性修正的Ergun关系式,结果发现修正后的Ergun关系式预测值与实验值符合良好。本研究为氦冷固态包层氚增殖区设计提供了数据支撑,为球床通道内流动特性的数值模拟提供了验证手段。 相似文献
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The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow. 相似文献
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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. 相似文献
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更准确地模拟球床式高温气冷堆堆芯温度分布,是反应堆安全分析尤其是超高温运行研究中的关键问题之一。由于堆芯球流运动具有不确定性,石墨块和碳砖等结构材料采用散体布置,堆内冷却剂流道复杂,对热工水力准确模拟造成困难,可进一步优化。本文结合HTR 10的结构特点和流道特征,简要分析了堆芯传热过程,说明了在热工模拟中准确划分结构和流道对获取更精确的堆芯温度分布的重要意义。详细梳理了冷却剂流动路径,改进了在THERMIX程序下建立的HTR 10原有热工分析模型,更合理地模拟了堆芯冷却剂漏流行为,使得模型对堆芯冷却剂流动和传热过程的描述更准确。与试验数据对比,改进后的模型对堆芯外围系统的温度分布模拟准确性显著提升。计算结果表明,反应堆在额定设计工况下满功率稳态运行时,燃料和反射层最高温度均未超过材料的耐热限值。 相似文献
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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。 相似文献
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Vladimir Sobes Author Vitae Benoit Forget Author Vitae 《Nuclear Engineering and Design》2011,241(1):124-133
Scientists at the German AVR pebble bed nuclear reactor discovered that the surface temperature of some of the pebbles in the AVR core were at least 200 K higher than previously predicted by reactor core analysis calculations. The goal of this research paper is to determine whether a similar unexpected fuel temperature increase of 200 K can be attributed solely or mostly to elevated power production resulting from exceptional configurations of pebbles. If it were caused by excessive pebble-to-pebble local power peaking, there could be implications for the need for core physics monitoring which is not now being considered for pebble bed reactors. The PBMR-400 core design was used as the basis for evaluating pebble bed reactor safety. Through exhaustive Monte Carlo modeling of a PBMR-400 pebble environment, no simple pebble-to-pebble burn-up conditions were found to cause a sufficiently high local power peaking to lead to a 200 K temperature increase. Simple thermal hydraulics analysis was performed which showed that a significant core coolant flow anomalies such as higher than expected core bypass flows, local pebble flow variation or even local flow blockage would be needed to account for such an increase in fuel temperature. The identified worst case scenarios are presented and discussed in detail. The conclusion of this work is that the stochastic nature of the pebble bed cannot lead to highly elevated fuel temperatures but rather local or core-wide coolant flow reductions are the likely cause. 相似文献
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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. 相似文献
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A high temperature gas-cooled reactor (HTGR) is one of the renewed reactor designs to play a role in nuclear power generation. This reactor design concept is currently under consideration and development worldwide. The combination of coated particle fuel, inert helium gas as coolant and graphite moderated reactor makes possible to operate at high temperature yielding a high efficiency. In this study the simulation of turbulent transport for the gas through the gaps of the spherical fuel elements (fuel pebbles) was performed using the large eddy simulation. This would help in understanding the highly three-dimensional, complex flow phenomena caused by flow curvature in the pebble bed. Resolving all the scales of a turbulent flow is too costly, while employing highly empirical turbulence models to complex problems could give inaccurate simulation results. The large eddy simulation (LES) method would overcome these shortcomings. An attempt to obtain experimental velocity flow patterns using particle image velocimetry technique combined with matched refractive index liquid was pursued. 相似文献
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球床包层混合堆与板状元件包层混合堆相比较,前者在核燃料生产和安全方面可能具有更多的优越性。本应用THERMIX程序和辅助程序对我国开发的托卡马克堆芯氮气冷却球床包层聚变-裂变合堆的包层进行了热工计算。计算中考虑了不同的燃料球材料及稳态,卸压和断流事故工况。计算结果表明,只要选用合适的燃料球材料和设置适当的控制保护系统,具有快速卸料罐的托卡马克堆芯氦气包层聚变-裂变混合堆的概念设计在安全上的可行的。 相似文献