共查询到17条相似文献,搜索用时 218 毫秒
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超临界水冷堆开发现状与前景展望 总被引:1,自引:0,他引:1
超临界水冷堆是被国际上选定为第四代核能系统长远开发的6种堆型之一,是在现有LWR和超临界火电技术基础上发展起来的革新型设计.在技术上,超临界水冷堆可以借鉴现有PWR和超临界火电的设计、建造和运行经验,不存在不可逾越的技术障碍.我国近期和中期目标都是采用压水堆技术,考虑到技术的继承性和可持续发展的要求,开发和研制超临界水冷堆核能系统是必然的选择. 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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超临界水冷堆(SCWR)是第四代国际论坛(GIF)选定的需要进行研究开发的六种反应堆之一。由于SCWR的热效率比较高(大约为45%,比目前的轻水堆33%的效率要高得多),并且可以使电厂显著地简化,所以,SCWR被认为是一种比较有前途的先进核能系 相似文献
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近年来,世界各国提出了许多新核反应堆设计和核燃料循环方案。2001年,美国、法国、日本等10个国家组成了“第四代国际核能论坛”(简称GIF),约定共同研发第四代核能系统,使其在安全性、经济性、可持续发展、防核扩散等方面都有显著的先进性和竞争力。2002年,GIF在法国巴黎举行研讨会,选定了6种堆型作为第四代核能系统优先研发对象,这其中就包括超临界水冷堆。近日,本刊记者就相关问题专访了彭士禄院士。 相似文献
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超临界水堆(SCWR)是第4代核反应堆的优先发展对象之一,它在经济性上的明显优势使其受到广泛关注。本文以混合谱超临界水堆(SCWR-M)为研究对象,建立合理的数学模型,开发了针对超临界水堆系统的瞬态分析程序TACOS。运用TACOS程序对SCWR-M进行了稳态计算和部分失流事故的瞬态分析。稳态计算的结果与设计值符合良好。部分失流事故的分析结果表明,事故中包壳表面最高温度为702.6 ℃,与安全限值相比有很大裕度。部分失流事故过程中不需采取特殊的安全措施,堆芯可自行回到安全状态。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(6):929-935
A supercritical water-cooled reactor (SCWR) was proposed as a kind of generation IV reactor in order to improve the efficiency of nuclear reactors. Although investigations on the thermal-hydraulic behavior in SCWR have attracted much attention, there is still a lack of CFD study on the heat transfer of supercritical water in fuel channels. In order to understand the thermal-hydraulic behavior of supercritical fluids in nuclear reactors, the local fluid flow and heat transfer of supercritical water in a 37-element fuel bundle has been studied numerically in this work. Results show that secondary flow appears and the cladding surface temperature (CST) is very nonuniform in the fuel bundle. The maximum cladding surface temperature (MaxCST), which is an important design parameter for SCWR, can be predicted and analyzed using the CFD method. Due to a very large circumferential temperature gradient in cladding surfaces of the fuel bundle, the precise cladding temperature distributions using the CFD method is highly recommended. 相似文献
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Supercritical water-cooled reactor (SCWR) is the only water-cooled reactor among six Generation IV reactor concepts. Safety analysis is one of the most important tasks for SCWR design. A typical thermal spectrum SCWR with passive safety system during design-basis accident (DBA) and beyond design-basis accident (BDBA) is performed. For DBA, reactor system is modeled based on a revised code ATHLET-SC. Loss of coolant accident is chosen to perform safety analysis and sensitive analysis. The results achieved demonstrate the feasibility of proposed passive cooling system to provide sufficient cooling. However, it should be noted that if one of safety systems fails to actuate during loss of coolant accident, although the likelihood is fairly low, there is potential risk of cladding failure. Consequently, the DBA will develop into the BDBA. For BDBA, a postulated severe accident is analyzed after melt pool is formed in the lower plenum. Heat transfer behavior in the melt pool as well as two-dimensional heat transfer effect in the lower head wall is discussed. Then, key parameters are chosen to perform parametric analysis. Results show that the safety margin to critical heat flux is significant. After considering two-dimensional heat conduction effect in the lower head, the safety margin could be further increased. 相似文献
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The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90° the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR. 相似文献
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The study of thermal characteristics during startup is one of the most important aspects for safety analysis of supercritical water-cooled reactor(SCWR).According to the given sliding pressure mode of SCWR,thermal analysis on temperature-raising phase and power-raising phase of startup are carried out.Considering the radial heterogeneity of power distribution,thermal characteristics for different assemblies during startup are also put forward.The results show that,during temperature-raising phase with core power increased only,the temperature of moderator,coolant and fuel cladding in inner assemblies are increased with little amplitude.During power-raising phase with core power and feed-water flow rate increased,the coolant temperature keeps unchanged,but the moderator temperature is decreased.With a greater variation of power,fuel cladding temperature shows a greater increase.Furthermore,considering the uneven distribution of radial power,thermo-hydraulic characteristics with uneven cladding temperature distribution shows a certain horizontal heterogeneity for different fuel assemblies,which becomes serious as flow rate and power increase.By adjusting flow rate distribution in different fuel assemblies or changing power setting during startup,the cladding temperature difference could be effectively reduced,which provides a certain reference for startup optimization of SCWR. 相似文献
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超临界水堆系统分析程序的改进 总被引:1,自引:1,他引:0
针对超临界水堆特殊的水物性参数和独立的慢化剂通道设计,对堆芯计算程序PARCS和热工水力程序RELAP5进行了适应性改造。使用改造后的耦合程序PARCS/RELAP5分析了美国超临界水冷参考堆,发现了慢化剂逆向流动和最高功率组件不同于最高外表面包层温度组件的现象,根据这些经验,对中国的超临界水堆分析程序的改进和研发提出了相关意见。 相似文献
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超临界水堆(SCWR)的LOCA研究是安全分析的重点和难点,其中压力容器的喷放泄压过程的研究至关重要。本文通过对反应堆压力容器进行简化,建立了简单容器喷放的数学物理模型,开发了超临界流体的喷放瞬态计算程序。将该程序的计算结果与超临界二氧化碳的泄压喷放过程的实验数据进行了比较,计算值与实验结果吻合良好,验证了模型的正确性。运用该验证后的程序对超临界水的容器喷放过程进行了深入研究和分析,分析了不同初始条件、破口面积及加热功率等对泄压过程瞬态特性的影响。结果表明,本文建立的简单容器模型能模拟从超临界到亚临界压力的喷放泄压过程。计算结果可为超临界水堆的LOCA分析提供理论基础。 相似文献