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首页> 外文会议>International conference on nuclear engineering >Thermal-hydraulic and Neutronic Analysis of a Re-entrant Fuel Channel Design for Pressure-Channel Supercritical Water-cooled Reactors
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Thermal-hydraulic and Neutronic Analysis of a Re-entrant Fuel Channel Design for Pressure-Channel Supercritical Water-cooled Reactors

机译：压力通道超临界水冷堆折返式燃料通道设计的热工和中子分析

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To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation Ⅳ. The Generation Ⅳ International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts one of which is the Supercritical Water-cooled Reactor (SCWR). Among the Generation Ⅳ nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation Ⅱ nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30-35% while the evolutionary SCWR will have a thermal efficiency of about 40-45%. In terms of a pressure boundary SCWRs are classified into two categories, namely, Pressure Vessel (PV) SCWRs and Pressure Channel (PCh) SCWRs. A generic pressure channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel channel design (for PCh SCWR). Second, a nuclear fuel and fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and re-entrant channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant fuel channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles. A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the fuel centerline temperature of a newly designed fuel bundle with UO_2 as a reference fuel. The results show that the maximum fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for fuel and sheath, respectively.
机译：为了满足开发具有更高热效率的新型核反应堆的需要，包括加拿大在内的一些国家已经启动了国际合作，以开发下一代核反应堆，即第四代核反应堆。第四代国际论坛（GIF）计划将核反应堆的设计选择范围缩小到六个概念，其中一个就是超临界水冷堆（SCWR）。在第四代核反应堆概念中，只有超临界水冷堆使用水作为冷却剂。 SCWR概念被认为是压水反应堆（PWR）和沸水反应堆（BWR）的发展，它们占当前运行核反应堆舰队的81％，归类于第二代核反应堆。后者的水冷反应堆的热效率在30-35％的范围内，而进化型SCWR的热效率约为40-45％。就压力边界而言，SCWR可分为两类，即压力容器（PV）SCWR和压力通道（PCh）SCWR。本文重点介绍的通用压力通道SCWR在25 MPa的压力下工作，入口和出口冷却液温度分别为350和625°C。冷却剂的高出口温度和压力使得可以提高热效率。另一方面，冷却剂的高工作温度和压力对材料选择和型芯设计提出了挑战。按照这种观点，SCWR的进一步发展需要解决两个主要问题。首先，应根据燃料通道设计（用于PCh SCWR）设计反应堆堆芯。其次，应选择核燃料和燃料循环。第三，应根据材料测试和实验结果选择核心部件和其他关键部件的材料。已经提出了用于短波堆的几种燃料通道设计。这些燃料通道设计可分为两类：直流通道和折返通道概念。本文的目的是研究折返式燃料通道设计的热液压和中子学方面。为此，在MATLAB中开发了一种热工代码，用于计算燃料中心线温度，护套温度，冷却液温度和传热系数曲线。为了确定芯内部的功率分布，使用了晶格码和扩散码。然后，将具有最大热功率的通道中的热通量用作热工代码的输入。本文介绍了以UO_2为参考燃料的新设计燃料束的燃料中心线温度。结果表明，燃料和护套的最高燃料中心线温度和护套温度分别超过了1850°C和850°C的温度极限。

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