共查询到19条相似文献,搜索用时 250 毫秒
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给出了无量纲分析法与频域法相结合的稳定性分析方法的详细描述,并定义了影响稳定性的关键无量纲数。针对垂直加热通道内超临界水进行了密度波稳定性分析,并建立了稳定性边界。对系统入口阻力因数、出口阻力因数、摩擦因数、进出口压降和流动方向等进行了参数敏感性分析,结果表明高的入口阻力因数有利于系统的稳定,但高的出口阻力因数和高的摩擦因数不利于系统的稳定,系统进出口压差对系统的稳定性影响较小,向上流动比向下流动更有利于系统的稳定。计算结果对超临界水堆的堆芯和系统设计具有指导性作用。 相似文献
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混合能谱超临界水堆失流事故缓解措施研究 总被引:1,自引:1,他引:0
使用改进的系统程序RELAP5建立了一个混合能谱超临界水堆(SCWR-M)模型。为研究混合能谱超临界水堆失流事故特性,以获取缓解混合能谱超临界水堆失流事故的措施,选取反应堆冷却剂泵惰转时间、压力容器上部储水空间容积和安注流量作为主要参数进行分析。研究表明,混合能谱超临界水堆系统的设计是可行的。反应堆冷却剂泵惰转15 s,压力容器上部水空间容积大于27 m3,以及安注流量高于系统满功率稳态流量的5%是缓解混合能谱超临界水堆失流事故的主要措施。 相似文献
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本文研究了混合能谱超临界水冷堆(SCWR-M)在发生控制棒失控提升事故和弹棒事故这两类反应性引入事故后的反应堆系统响应。首先利用修改的可用于超临界条件下的系统程序RELAP5对混合能谱超临界水冷堆进行系统建模,并计算分析在功率运行工况下事故过程中功率、流量及包壳温度等重要参数的变化趋势,最后对反应性参数如控制棒价值、控制棒抽出速率和负反馈系数进行了参数效应分析。结果表明,在设计工况下混合能谱超临界水冷堆系统可有效地将衰变热导出堆芯,保证了燃料棒的完整性。另外,反应性参数对控制棒失控提升事故的安全性影响不大,但对弹棒事故的包壳峰值温度影响很大,过于保守的反应性参数估计会使安全裕量大为减小。 相似文献
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超临界水的流动不稳定性特征研究是超临界水冷堆热工水力设计的重点,为进一步获得超临界水流动不稳定性发生的内部机理,采用系统分析程序RELAP5对已有实验本体进行建模,并基于已有超临界水不稳定性实验数据开展了计算方法的验证;系统研究了并联通道内超临界水的流动不稳定性规律,并对比研究了超临界水与亚临界水的不稳定边界。结果表明,超临界水的流动不稳定界限功率随入口温度的增加存在变化拐点;相同入口温度下,随压力上升,不稳定界限功率增加,超临界水相比亚临界气-液两相流具有更好的稳定性;无量纲准则数在超临界条件下具有适用性,超临界水不稳定性变化规律与亚临界水具有相似性。 相似文献
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超临界水冷堆堆芯子通道稳态热工分析 总被引:1,自引:1,他引:1
超临界水冷堆(SCWR)作为6种第四代未来堆型中唯一的水冷堆,冷却剂出口温度可达500℃,具有良好的经济性.本文采用改进的COBRA-IV程序对超临界水冷堆方形组件子通道进行稳态热工分析.对计算结果进行分析可知:减小慢化剂通道中给水质量流量份额和加大慢化剂通道与相邻子通道之间的热阻,可以降低热管焓升,后者还可以得到较好的慢化效果.通过热通道的传热恶化分析发现,超临界水冷堆的设计不能避免传热恶化,必须精确计算传热恶化条件下的包壳温度才能确定包壳能否保证其完整性. 相似文献
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The paper describes a novel core concept for a nuclear reactor cooled with supercritical water, in which the coolant is heated up from 280 °C at the reactor inlet to 500 °C at the outlet in four steps: a first heat-up step is provided by heat transfer from fuel assemblies to the moderator water in gaps and moderator boxes, a second step is foreseen in a central “evaporator” and two further steps in a first and a second superheater surrounding it. The coolant flow scheme includes upward and downward flow through the core with intermediate mixing in chambers above and below the core to eliminate hot streaks. A preliminary single channel analysis, concentrating on an average flow channel and on the hottest one only, indicates that such core design can match the limits of cladding materials available today. Even though the resultant pressure drop of the coolant will be higher than usual, it is expected that the assembly boxes can be designed with acceptable deformations. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(6):537-548
The feasibility of the sliding pressure startup of a high-temperature supercritical-pressure light water reactor (super LWR, SCLWR-H) is assessed from both thermal and stability considerations. In the sliding pressure startup, nuclear heating starts at subcritical pressure and the reactor is pressurized to supercritical pressure at a low power and high enough flow rate. The reactor power and flow rate are then raised gradually to the rated normal values at constant supercritical operating pressure. During startup, the maximum cladding surface temperature must not exceed 620°C. For two-phase flow at subcritical pressures, the homogeneous equilibrium model is used. The thermal-hydraulic and coupled neutronic thermal-hydraulic stabilities during pressurization and power-raising are investigated by a frequency-domain linear analysis for both supercritical-pressure and subcritical-pressure operating conditions. The same stability criteria as those of BWRs are used. From the analysis results, a sliding pressure startup procedure is proposed for super LWR. The thermal criteria are satisfied by keeping the core power between the maximum allowable limit and minimum limit required for turbine startup and operation. The thermal-hydraulic stability and coupled neutronic thermal-hydraulic stability can be maintained by applying an orifice pressure drop coefficient at the inlet of fuel assembly and by controlling the power and flow rate during startup. 相似文献
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The HPLWR (high performance light water reactor) is the European concept design for a SCWR (supercritical water reactor). This unique reactor design consists of a three pass core with intermediate mixing plena. As the supercritical water passes through the core, it experiences a significant density reduction. This large change in density could be used as the driving force for natural circulation of the coolant, adding an inherent safety feature to this concept design. The idea of natural circulation has been explored in the past for boiling water reactors (BWR). From those studies, it is known that the different feedback mechanisms can trigger flow instabilities. These can be purely thermo-hydraulic (driven by the friction – mass flow rate or gravity – mass flow rate feedback of the system), or they can be coupled thermo-hydraulic–neutronic (driven by the coupling between friction, mass flow rate and power production). The goal of this study is to explore the stability of a natural circulation HPLWR considering the thermo-hydraulic–neutronic feedback. This was done through a unique experimental facility, DeLight, which is a scaled model of the HPLWR using Freon R23 as a scaling fluid. An artificial neutronic feedback was incorporated into the system based on the average measured density. To model the heat transfer dynamics in the rods, a simple first order model was used with a fixed time constant of 6 s. The results include the measurements of the varying decay ratio (DR) and frequency over a wide range of operating conditions. A clear instability zone was found within the stability plane, which seems to be similar to that of a BWR. Experimental data on the stability of a supercritical loop is rare in open literature, and these data could serve as an important benchmark tool for existing codes and models. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(12):1166-1175
This paper summarizes the analysis results of the thermal-hydraulic stability of a high-temperature reactor cooled and moderated by supercritical-pressure light water (SCLWR-H). A linear stability analysis code in the frequency domain was developed to study the thermal-hydraulic stability of SCLWR-H at constant supercritical pressure. The analysis method is based on linearization by perturbation of numerically-discretized one-dimensional single-channel single-phase conservation equations. The effect of water rods on stability is considered. The thermal-hydraulic stability of SCLWR-H for full-power and partial-power normal operations was investigated by frequency domain method. Our analysis reveals that though SCLWR-H has low coolant flow rate and large density change in the core, the thermal-hydraulic stability can be maintained both at normal operation and during power raising phase of constant pressure startup by applying an orifice pressure drop coefficient at the inlet of the fuel assemblies. A parametric study was also carried out to determine the parameters affecting the stability. 相似文献
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针对带定位格架的超临界水冷堆堆芯垂直上升类三角形子通道,开展超临界水的流动传热试验研究。反应堆堆芯类三角形子通道棒束直径为8 mm、栅距比为1.4,试验参数范围为:热流密度q=200~600 kW/m2、压力P=23~28 MPa、质量流速G=700~1300 kg/(m2·s)。分析了热流密度、压力和质量流速等热工参数对超临界水传热特性的影响。试验结果表明:定位格架处质量流速升高,流体扰动性增强,换热系数提升显著;在超临界压力下,提高压力会导致内壁温度上升,换热系数峰值降低;过高的热流密度会导致换热系数峰值降低,适当减小热流密度可提高换热性能;提高质量流速会导致内壁温度降低,换热系数峰值上升,能够显著提高换热性能。压力变化对定位格架区域传热特性影响较小,适当提升压力可提高系统安全性。 相似文献
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综合考虑辐照试验指标与燃料试验安全、高通量工程试验堆(HFETR)运行要求、试验段压差波动等因素,基于HFETR开展了快堆燃料短棒辐照试验方案设计与分析,确定了铅铋合金层厚度、冷却水流道结构、阻力塞结构、冷却水流量等关键参数,获得了热棒包壳最高温度为(490±60)℃的高线功率密度辐照试验方案。试验结果表明,热棒最大线功率密度为68~85 kW/m时,包壳与燃料芯体温度满足辐照试验要求且留有余量;在200~300 kPa堆芯压差范围内,相同压差下试验段流量的计算流体力学(CFX)计算值比试验值偏小9% ~11%;试验段外侧窄缝流道的流量份额为7.3%,显著低于该流道的流通面积份额,满足线功率密度为85 kW/m时燃料短棒的冷却要求。本文提出的辐照试验方案可为快堆燃料棒的高线功率密度辐照试验提供参考。 相似文献
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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。 相似文献