1)SCWR超临界水冷堆
1.Steady-State Thermal-Hydraulic Analysis of SCWR Assembly;超临界水冷堆堆芯子通道稳态热工分析
2.Investigations on the thermal-hydraulic behavior in the SCWR fuel assembly have obtained a significant attention in the international SCWR community.目前国际上对超临界水冷堆进行了大量的研究,但对其堆芯内超临界流体流动传热的认识还十分欠缺。
英文短句/例句
1.Research on Designing the Outlet Nozzle in Main Vessel of Supercritical Water Reactor;超临界水冷堆主容器出口管的设计研究
2.Optimization Design of Reduced-Activation Ferritic/Martensitic Steels for SCWR Fuel Cladding Materials超临界水冷堆燃料包壳管用低活性F/M钢的优化设计
3.Studies and Design of the Fast Zone Assembly in SCWR Mixed Core超临界水冷混合堆快谱组件研究与设计
4.A neutron-kinetics/thermal-hydraulics analysis of the fast zone assembly in the SCWR mixed core超临界水冷混合堆快谱区组件物理——热工分析
5.Analysis of Candidate Core Materials in SCWR Based on Boiler Materials of SCU基于超临界火电站锅炉用材的超临界水冷却反应堆堆芯候选材料分析
6.Coupled Thermal-Hydraulics and Neutron-Physics Analysis of Supercritical Water Cooled Reactor With Mixed Spectrum Core混合能谱超临界水堆堆芯热工-物理性能分析
7.Corrosion Behaviors of Candidate Materials for Supercritical-Cooled Water Reactor超临界水堆候选材料的腐蚀特性研究
8.The Investigation of Heat-transfer Property of Super-critical Waterwall Tubes;超临界锅炉水冷壁管传热特性的研究
9.Application of Degree of Similarity in Choice of Correlations for Supercritical Water Reactor贴近度在超临界水堆实验关联式选择中的应用
10.Neutron-Kinetics/Thermal-Hydraulics Performance Analysis of Fast Zone Assembly in Supercritical Water Reactor Mixed Core超临界水混合堆快谱区组件物理-热工性能分析
11.Neutronics/thermal-hydraulics coupled axial 1D model for SCWR core static analysis超临界水堆堆芯轴向一维物理热工耦合稳态分析
12.High Temperature Corrosion of Water Cooling Wall of Extrasupercritical Units超超临界机组水冷壁高温腐蚀问题的探讨
13.Numerical Calculation on Temperature Fields of Vertical Waterwall in Ultra-supercritical Boilers超超临界锅炉垂直水冷壁温度场的数值计算
14.Supercritical Concurrent Boiler Water Cooling Wall Hydrodynamic Force Characteristic Research;超临界直流锅炉水冷壁水动力特性研究
15.Numerical Simulation of Heat Transfer in Pipe of Supercritical Boiler Water-cooling Wall;超临界锅炉水冷壁管内传热特性的数值模拟
16.CAUSE ANALYSIS OF HIGH TEMPERATURE CORROSION ON WATER WALL OF SUPERCRITICAL ONCE-THROUGH BOILERS AND COUNTERMEASURES THEREOF超临界锅炉水冷壁高温腐蚀分析及对策
17.Thermal Behavior of Membrane Waterwall of 600 MW Supercritical Pressure Boiler Furnace600MW超临界锅炉炉膛膜式水冷壁的热行为研究
18.Research on Wall Temperature of Spiral Tube Water Wall for 600MW Supercritical Boiler600MW超临界锅炉机组螺旋管圈水冷壁管壁温度
相关短句/例句
supercritical water cooled reactor超临界水冷堆
1.The supercritical water cooled reactor(SCWR) is essentially light water reactor(LWR) operating at higher pressure and temperature beyond the thermodynamic critical point of water(374 ℃,22.超临界水冷堆(SCWR)是在高于水的临界点(374℃,22。
3)Supercritical water-cooled reactor超临界水冷却堆
4)Review of Supercritical Water Cooled Reactor超临界水冷堆述评
5)Supercritical water reactor超临界水堆
1.This paper introduces the software extension in the reactor core analysis code PARCS and thermodynamics code RELAP5 to accommodate the special water thermal physical properties and separated moderator channel design in supercritical water reactor(SCWR).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。
2.Research activities of supercritical water reactor(SCWR) have been carried out worldwide,aiming at cost reduction by system simplification and higher thermal efficiency.由于超临界水堆(SCWR)在系统简化、降低成本和提高热效率上的优势,SCWR的研究在全球范围内得到广泛关注。
6)supercritical water cooled reactor超临界水堆
1.Feasibility of the steel was analysed for in-core component and fuel cladding application in supercritical water cooled reactor.分析了9Cr3W钢用作超临界水堆堆芯内部件及包壳材料的可行性,其高温力学性能远优于Zr合金包壳材料;拉伸性能与T91钢相当,且韧脆转变温度低于T91钢,冲击吸收功上限高于T91钢,具有优良的冲击韧性;9Cr3W钢的高温瞬时强度低于奥氏体316不锈钢,成为制约其用于超临界水堆堆芯内部件及包壳的因素之一。
2.Recently,a new supercritical water cooled reactor(SCWR) conceptual design was proposed on the basis of a mixed spectrum core concept consisting of a thermal zone and a fast zone.针对一种新型的超临界水堆设计方案——混合能谱超临界水堆(SCWR-M)进行分析。
3.As the only water cooled reactor among the six generation-Ⅳ reactor,supercritical water cooled reactor(SCWR) has its special characteristics,and takes up attentions extensively.超临界水堆作为6种第4代未来堆型中唯一的水冷堆,具有一些独特的特点,受到了广泛重视。
延伸阅读
石墨水冷堆石墨水冷堆water cooled graphite moderated reactor 石墨水冷堆(water eooled graphitemoder-ated reactor)以石墨为慢化剂、水为冷却剂的热中子反应堆。核工业发展初期，石墨水冷堆主要用以生产核武器装料—怀、氖等。这种堆一般以天然铀金属元件做燃料。在堆内天然铀中的铀235吸收中子发生核裂变反应，放出中子和能量，这些中子一部分用于维持链式核裂变反应，一部分则为天然铀中的铀238所吸收，转化为怀239及其他钵同位素。 结构石墨水冷堆用核纯石墨砌体作慢化剂与反射层。在石墨砌体内有二三千个水平孔道(卧式堆)或垂直孔道(立式堆)，在这些孔道中插有可更换的石墨套管，套管中插铝合金工艺管，将冷却水同石墨慢化剂隔开。在工艺管内壁有凸肋以保持 工艺管与燃料元件之间的间隙。石 墨砌体各部分的温度是不均匀的， 通过改变石墨套管与工艺管之间的 间隙和工艺管内的水流量，可部分 地调整砌体温度，使其温度分布较 为平坦。燃料元件一般均做成棒状， 直径约35一38mm，长度约100~ ZOomm，为了提高比功率和使元件 径向燃耗均匀，也有用管状燃料元 件的。 生产堆发展初期曾采用过开式 冷却方式。即使河水直流堆芯，带出 热量的水再排到河里。由于这种方 式耗水量大，排水中放射性水平高、 环保问题突出等原因，业已停止使用，而普遍采用闭式冷却方式，即冷却水从堆芯流过，将热量导出堆外，通过热交换器将热传导给另一回路侧的水，再经主泵返回堆芯，形成闭合循环的主冷却回路，或称一回路。对导出一回路水的热量的处理方式有两种:一种是将一回路热量通过热交换器导给二回路水，经过冷却水塔或河水冷却，将热量排到环境中去。另一种方式是通过热交换器将热传给余热利用系统，作为热源向外界供热或发电。 特点天然铀石墨水冷堆重要特点之一是后备反应性很小。 早期石墨水堆的反应性随其温度升高而升高，堆功率也随之升高(即所谓的正温度效应)，从而又导致了反应性上升，直到反应堆置于外部引人中子吸收体(控制棒等)的控制下，或造成堆芯熔化等恶性事故。1986年，切尔诺贝利核事故后，正温度效应问题更加引起各方面的重视，在堆物理设计方面必须获得负温度效应，以确保反应堆具有至关重要的自稳性。 美国自1943年起建造了8座石墨水冷军用怀生产堆，1座生产发电两用堆(NPR)，后者热功率为4 00OMW。
