定位格架对超临界水冷堆堆芯换热的影响

在压力为25 MPa、质量流速800kg/(m~2·s)、热流密度600kW/(m2·K)的条件下,对超临界水冷堆堆芯(燃料棒直径D=8.0mm,节径比P/D=1.2)类三角形子通道内超临界水的换热特性进行研究。用结构化的六面体网格以及计算流体动力学软件ANSYS CFX对堆芯的换热情况进行了模拟。以无定位格架子通道为基准,对比阻流片型定位格架对通道内流体换热的作用。结果表明,在超临界压力下阻流片型定位格架能够明显增强换热,降低最高包壳温度。另外,不同焓值区定位格架对换热的影响存在差异。     

The influence of thermal stratification on hydrogen fuel flow and heat transfer in cooling channel with combining fin and dimple

The rockets engines combustion chamber wall suffers from high single-side heating and small cooling channel size, and this can cause serious cross-section thermal stratification and local heat transfer deterioration problem, leading to the decreasing of the flow and heat transfer efficiency and thermal protection problem. In this paper, to effectively reduce this non-uniform distribution phenomenon to enhance the flow heat transfer capacity, a new vortex generator with combing fin and dimple is proposed. The CFD software is used to numerically solve this problem and analyzed the influence of thermal stratification on hydrogen fuel flow and heat transfer. The investigation results indicate that for Re_in = 42000, Nu of new dimple and fin structure improves about 94%, and the maximum heat temperature decreases 15.4% than smooth channel, while the friction factor is about 2.6 times smooth channel. The temperature non-uniform index decreases the 12.24% than smooth channel. Comparing with smooth and dimple structure, the new structure effectively reduced the thermal stratification phenomenon to improve the hydrogen fuel flow and heat transfer performance, thus the overheated structure is better protected.     

Temperature distributions in a system of two cylindrical bodies

Consideration is made of transient heat transfer in a cylindrical fuel element of the nuclear reactor. The fuel rod, whose power dissipation is assumed to take place in the fuel element by a sinusoidal pulse, and cladding are separated by a thin layer. The conjugate boundary-value problem involving the equation for heat conduction in the fuel rod and cladding, and boundary conditions with regard for non-ideal thermal contact is solved by the Laplace transformation. The results on a temperature distribution for uranium dioxide fuel rod and zirconium cladding when cooled by liquid sodium are presented graphically.     

Preliminary safety analysis for the SCWR-FQT facility

In recent years, a substantial number of theoretical, numerical and experimental R&D activities are carried out on the supercritical water-cooled reactor (SCWR), which proposed as one of the Generation IV nuclear power plants by the Generation IV International Forum (GIF). A research plan has been proposed by GIF on the designing and licensing of a SCWR prototype, which is planned to be constructed and operated in the near future. In the preliminary stage of this research plan a fuel assembly, with its experimental loop, will be constructed and tested in SCWR operating conditions. This article reviews the research activities carried out in the Supercritical Water Reactor Fuel Qualification Test (SCWR-FQT) project in Europe and Super Critical Reactor In-Pipe Test Preparation (SCRIPT) project in China. These research activities studied both neutronic and thermal–hydraulic behavior of the test fuel assembly and auxiliary systems of SCWR fuel test facility. CFD simulations, subchannel analysis and system simulations coupled with neutronics code are performed to study the performance of the tested fuel assembly, especially safety related aspects.     

Comparative study of fuel gas production for SOFC from steam and supercritical-water reforming of bioethanol

Theoretical study of fuel gas (H₂ + CO) production for SOFC from bioethanol was carried out to compare performances between two reforming technologies, including steam reforming (SR) and supercritical-water reforming (SCWR). It demonstrates that the fuel gas productions are comparable among the two reforming systems; however, SCWR requires the operation at much higher temperature and pressure than SR. The maximum hydrogen yield can be obtained at 850 K, atmospheric pressure, ethanol to water molar feed ratio of 1:20 for SR system and at 1300 K, 22.1 MPa, and ethanol to water feed ratio of 1:20 for SCWR. The use of a distillation column to purify the bioethanol feed was proven to improve the fuel conversion efficiency of both systems. The analysis reveals that SCWR is a promising system for fuel production for SOFC when a gas turbine is incorporated to the system for energy recovery. Further, it is not necessary to distil bioethanol to obtain too high ethanol recovery (i.e. >90%) as higher energy consumption at the distillation column could lead to lower overall thermal efficiency.     

Studies on advanced water-cooled reactors beyond generation III for power generation

China’s ambitious nuclear power program motivates the country’s nuclear community to develop advanced reactor concepts beyond generation III to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics, sustainability and technology availability. It is a logical extension of the generation III PWR technology in China. The status of international R&D work is reviewed. A new supercritical water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydraulics method is carried out. It shows good feasibility for the new design proposal.     

Forced Convective Heat Transfer in a Plate Channel Filled with Solid Particles

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基于Hermite插值法的板型燃料不对称冷却计算

通过Hermite插值积分理论列出积分平均值与边界条件的关系,进而建立板型燃料芯体及其左右包壳的导热方程,利用Fortran科学计算语言对所建立的数学模型编译求解程序。将求解程序加入到反应堆热工水力实时仿真程序THEATRe中进行不对称冷却问题的计算,并对THEATRe程序的输入卡进行修改。通过计算中国先进研究堆(CARR)的标准燃料组件和跟随体燃料组件的稳态温度分布,与现有参考结果进行对比验证求解程序的正确性。最后模拟分析板型燃料组件流道堵塞事故。     

Heat recirculating reactors: Fundamental research and applications

Worldwide emphasis on fuel efficiency, low emissions, and use of low-quality fuels such as biogas continues to drive the development of combustors that operate over a wider range of fuel/air ratios and with higher burning velocities than their conventional counterparts. Enhancement of reaction rates is required to increase burning velocities and widen fuel/air operating ranges over values achievable in conventional combustors, and extensive research over the last few decades has shown that transferring heat in a reactor from hot combustion products to incoming reactants can accomplish this enhancement without external energy addition. These reactors, called heat recirculating reactors, use various geometries and flow strategies to optimize the heat transfer. In this paper, research on heat recirculating reactors is reviewed with an emphasis on the most important designs and applications. The basic characteristics of a heat recirculating reactor are encompassed in a simple configuration: a flame stabilized in a tube with high thermal conductivity. More complex designs that have evolved to further optimize heat transfer and recirculation are then described, including porous reactors with or without flame stabilization and channel reactors consisting of parallel tubes or slots. Advanced designs introduce additional means of heat transfer, such as transverse heat transfer from hot products through channel walls to incoming reactants, thereby leading to the counter-flow channel reactor. The flexibility of heat recirculating reactors to operate on a variety of fuels and over wide operating ranges has led to many applications including fuel reformers, radiant heaters and thermal oxidizers, and important work on these applications is reviewed. Finally, future research directions are discussed.     

Transient characteristics of small molten salt reactor during blockage accident

This paper reports the results from a transient core analysis of a small molten salt reactor (MSR) when a duct blockage accident occurred. The focus of this study is a numerical model employed in order to consider the interaction among fuel salt flow, heat transfer, and nuclear reactions. The numerical model comprises continuity and momentum conservation equations for fuel salt flow, two‐group neutron diffusion equations for fast and thermal neutron fluxes, transport equations for six‐group delayed neutron precursors, and energy conservation equations for fuel salt and graphite moderators. The analysis results show the following: (1) the effect of the self‐control performance of the MSR on the effective multiplication factor and thermal power output of the reactor after the blockage accident is insignificant, (2) fuel salt and graphite moderator temperatures increase drastically but locally at the blockage area and its surroundings, (3) the highest fuel salt temperature after the blockage accident is 1,363 K; this value is lower than the boiling point of fuel salt and the melting temperature of the reactor vessel, (4) the change in the distributions of fast and thermal neutron fluxes after the blockage accident when compared with the distributions at the rated condition is very slight, and (5) delayed neutron precursors, especially the first delayed neutron precursor, accumulate at the blockage area due to its large decay constant. These results imply that the safety of the MSR is assured in the case of a blockage accident. © 2006 Wiley Periodicals, Inc. Heat Trans Asian Res, 35(6): 434–450, 2006; Published online in Wiley InterScience ( www.interscience.wiley.com ). DOI 10.1002/htj.20123