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Study on Neutronics and Thermalhydraulics Characteristics of 1200-MW_(el) Pressure-Channel SuperCritical Water-cooled Reactor (SCWR)

机译:1200MW_(el)压力通道超临界水冷堆(SCWR)的中子学和热工水力特性研究

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Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical-energy generation. The largest group of operating Nuclear Power Plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) have gross thermal efficiencies ranging from 30% and up to 36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5-7.8 MPa / 257-293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fuel - natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation Ⅳ NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., Supercritical Water-cooled Reactors (SCWRs) have to be designed. This path of the thermal-efficiency increasing is considered as a conventional way through which coal-fired power plants gone more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MW_(el) Pressure-Channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulic code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the CFD Fluent code has been used for better understanding of specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical-water loop and developing passive-safety systems.
机译:作为当前和未来发电的基础,核电在全球许多国家变得越来越重要。配备水冷堆的运行中最大的核电厂集团(占所有核电厂的96%)的总热效率在30%到36%之间。如此低的热效率值是由于涡轮机入口处的压力/温度较低(4.5-7.8 MPa / 257-293°C)。但是,现代联合循环发电厂(布雷顿燃气轮机循环和亚临界蒸汽兰金循环,燃料-天然气)和超临界燃煤发电厂的总热效率分别达到62%和55%。因此,具有水冷反应堆的下一代或第四代核电厂的热效率应尽可能接近现代火力发电厂。仅由于在水的临界点以上增加涡轮进口参数,即必须设计超临界水冷反应堆(SCWR),才有可能大大提高水冷NPP的热效率。这种提高热效率的途径被认为是50多年前燃煤电厂停产的常规方法。因此,本论文的目的是研究通用1200 MW_(el)压力通道(PCh)SCWR的中子学和热工水力特性。标准中子学代码DRAGON和DONJON已与基于最新的经验传热相关性开发的新热工代码结合在一起,从而可以更准确地估算PCh SCWR的基本特性。此外,CFD Fluent代码已用于更好地了解超临界水中传热的细节。未来的研究将致力于在堆内超临界水环路中进行材料和燃料测试,并开发无源安全系统。

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