肋片对圆环通道内超临界水传热特性影响的数值分析

以计算流体力学(CFD)商业软件CFX为计算平台,对圆环通道内超临界水(SCW)的传热特性进行数值模拟。通过对几种湍流模型的对比,选取对超临界条件适用性相对较好的SST模型进行计算;比较光滑圆环通道与带肋片圆环通道内的壁温分布及传热系数的变化;研究肋片形状、高度、间距及宽度等因素对传热特性的影响。结果表明:圆环通道内肋片的存在会导致局部传热强化;在本文计算的尺寸范围内,肋片形状和宽度对传热的影响很小;一定范围内,增加肋片高度以及减小肋片间距都有利于传热强化。     

Design,Analysis and Optimization of the First Wall Cooling System of the CFETR Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak which is designed by China National Integration design Group for Magnetic Confinement Fusion. CFETR Blanket, as a plasma-facing component withstand very high heat load, is very critical for fusion reactor operation. The first wall (FW) is one of the most significant components of the blanket. The cooling system of the FW has been designed. Meanwhile, thermal–dynamic calculations are performed to obtain the coolant feature and temperature distribution of the FW using ANSYS CFX code. Besides, thermo-mechanical coupling analysis is carried out using the temperature distribution from thermal–dynamic calculation as boundary condition. In addition, cooling channel optimization is proposed according to the analysis results. Analysis results of the optimization cooling channel indicate that the maximum temperature and thermal stress satisfy the design requirements of the FW.     

Thermal–hydraulic and thermo-structural analysis of first wall for Indian DEMO blanket module

The first wall (FW) is one of the most important components of any fusion blanket design. India has developed two concepts of breeding blanket for the DEMO reactor: the first one is Lead–Lithium cooled Ceramic Breeder (LLCB), and the second one is Helium-Cooled Ceramic Breeder (HCCB) concept. Both the concept has the same kind of FW structure. Reduced Activation Ferritic Martensitic steel (RAFMS) used as the structural material and helium (He) gas is used to actively cool the FW structure. Beryllium (Be) layer of 2 mm is coated on the plasma side of the FW as the plasma facing material. Cooling channels running in radial–toroidal–radial direction in the RAFMS structure are designed to withstand the maximum He pressure of 8 MPa. Heat transfer coefficients (HTC) obtained form the correlations revealed that required cooling could be achieved by artificially roughened surface towards the plasma-side wall of He cooling channel which helps to keep the RAFMS temperatures below the allowable limit. A 1D analytical and 2D thermal–hydraulic simulation studies using ANSYS has been performed based on the heat load obtained from neutronics calculations to confirm the heat removal and structural integrity under various conditions including ITER transient events. The required helium flow through the cooling channels are evaluated and used to optimize the suitable header design. The detail design of FW thermal–hydraulics, thermo-structural analyses, and He flow distribution network will be presented in this paper.     

Numerical investigation of heat transfer enhancement in ribbed channel for the first wall of DFLL-TBM in ITER

As an important component of Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM), the first wall (FW) must withstand and remove the heat flux from the plasma (q″ = 0.3 MW/m²) and high nuclear power deposited in the structure at normal plasma operation scenario of ITER. In this paper the transverse ribs arranged along the plasma facing inner wall surface were used to enhance the heat transfer capability. After the validation compared with empirical correlations the Standard k–ω model was employed to do the numerical simulation using FLUENT code to investigate the heat transfer efficiency and flow performance of coolant in the ribbed channel preliminarily. The perforation on the bottom of rib was proposed near the lower heat transfer area (LHTA) to improve the heat transfer performance according to results of analyses.     

光滑方环管内超临界水流动与传热特性数值计算研究

采用CFD数值模拟方法对光滑方环管内超临界水流动与传热特性进行初步研究。计算结果表明,光滑方环管内超临界水的传热特性存在强烈的周向不均匀性,角通道处的传热状况较窄通道处的好,角通道处的加热壁面温度低于同一截面上窄通道处的加热壁面温度。周向传热不均匀因子随主流焓的增加而增加,在拟临界点附近增长的速度最快。造成方环管内超临界水传热周向非均匀的原因主要是流道的形状。超临界水冷堆中,窄通道处最易出现传热恶化,在堆芯设计中需给予更多关注。     

环形通道内再淹没过程先驱冷却传热特性研究

骤冷前沿推进是失水事故后再淹没过程中堆芯冷却速率的重要标志，先驱冷却传热对骤冷前沿的推进起到关键作用，对先驱冷却传热特性的研究十分必要。本文通过数值求解二维非稳态导热方程获得先驱冷却传热系数，并基于环形通道内底部再淹没实验数据，分析先驱冷却传热系数受初始壁温、入口温度和入口质量流速等参数的影响规律。研究结果表明，初始壁温对先驱冷却传热系数的影响不显著，先驱冷却传热系数随入口质量流速的增加而明显增加，随与骤冷前沿距离的增加而减小，基于实验数据得到本参数范围内先驱冷却传热关系式。     

Experimental and numerical study on transient heat transfer for a twisted plate with different length in helium gas at various velocities

This study is aimed to investigate the transient heat transfer process between the solid surface and the coolant (helium gas) in very high temperature reactor or intermediate heat exchanger. Transient heat transfer from a twisted plate with different length in helium gas was experimentally and theoretically studied. The heat generation rate was increased with an exponential function, Q = Q₀exp(t/τ), where t is time and τ is period. Experiment was carried out at various periods ranged from 35 ms to 14 s. Platinum plates were twisted with the same helical pitch of 20 mm, and the effective lengths are 26.8, 67.8 and 106.4 mm (pitch numbers of 1, 3 and 5), respectively. It was clarified that the average heat transfer coefficient approaches quasi-steady-state value when the period τ is larger than about 1 s, and it becomes higher when τ is shorter than about 1 s. The heat transfer coefficient decreases with the increase of plate length. An empirical correlation for forced convection heat transfer for a twisted plate with various lengths was obtained based on the experimental data. Moreover, numerical simulation results were obtained for average surface temperature difference, heat flux and heat transfer coefficient of the twisted plates with different length and showed reasonable agreement with experimental data. Through the numerical simulation, distribution of heat transfer coefficient on heater surface, temperature distribution and velocity distribution were clarified.     

聚变驱动次临界堆双冷嬗变包层第一壁结构热应力计算与分析

使用有限元程序对聚变次临界堆双冷嬗变包层第一壁进行数值模拟 ,给出不同载荷条件下的温度场和应力场分布 ,结果证明典型氦气系统设计满足热工要求。依据数值模拟结果对第一壁氦气载热能力进行分析 ,并考虑了流道形状对结构热应力的影响。     

聚变次临界堆双冷嬗变包层第一壁结构和氦气流道优化分析

在聚变次临界堆双冷嬗变包层第一壁结构初步设计基础上 ,对第一壁结构尺寸和氦气流道形状进行优化分析 ,利用有限元分析软件对第一壁结构进行应力数值模拟 ,在满足结构应力及部件可靠性的前提下 ,给出最佳优化方案。     

Experimental and analytical studies on turbulent heat transfer performance of a fuel rod with spacer ribs for high temperature gas-cooled reactors

Turbulent heat transfer performance of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was studied for various Reynolds numbers using an annular channel at the same coolant condition as the reactor operation, maximum outlet temperature of 1000 °C and pressure of 4 MPa, and analytically by a numerical simulation using the k- turbulence model. The turbulent heat transfer coefficients of the fuel rod were 18–80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the empirical correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer ribs and the axial velocity increase due to a reduction in the annular channel cross-section.     

A new cooling scheme for the HCLL TBM

The Helium Cooled Lithium Lead (HCLL) blanket is one of the two blanket concepts selected by the European Union to be tested in ITER. It is based on the use of Eurofer as structural material, helium as coolant and eutectic lithium–lead as breeder/neutron multiplier material. The design of the corresponding Test Blanket Module (TBM) for ITER has undergone several revisions in the last years. This paper presents an alternative cooling scheme for the HCLL-TBM, where the First Wall (FW) is cooled by vertical (poloidal) instead of horizontal (toroidal) channels. New Finite Element models have been developed and thermal and thermo-hydraulical analyses of the new design have been performed. Results show that the new cooling scheme presents several advantages with respect to the previous one: (i) the total number of cooling channels in the FW can be reduced; (ii) the overall pressure drops in one cooling channel are lower; (iii) the temperature profile in the breeding zone is more uniform.     

中国先进研究堆冷中子源单相冷包方案可行性分

在现有的冷源设计中,两相氢循环因其换热能力强而被广泛采用,但它最大的缺点是存在含气率影响慢化的稳定性。能否采用单相循环代替两相循环实现高热流密度的热量输出,是待研究的重点。为兼顾循环流量等宏观特性和流场、温度场分布等细节参数的分析,提出了一种基于迭代的耦合算法,将一维理论公式与三维数值仿真模型相结合,用于分析中国先进研究堆单相冷包方案的可行性。研究发现,单相循环只能带走约30%的核发热,但由于冷包增加了氦冷却套,其余热量全部通过氦气对冷包壁面的直接冷却带走。温度场的分析显示液氢和壁面的最高温度分别为21.7和23.7 K。这说明冷包得到了充分冷却,单相循环及单相冷包结构可满足工程需要。     

Heat conduction analyses on rewetting front propagation during transients beyond anticipated operational occurrences for BWRs

Our previous study investigated the rewetting behavior of dryout fuel surface during transients beyond anticipated operational occurrences for BWRs, which indicated the rewetting velocity was significantly affected by the precursory cooling defined as cooling immediately before rewetting. This study further investigated the previous experiments by conducting additional experimental and numerical heat conduction analyses to characterize the precursory cooling. For the characterization, the precursory cooling was first defined quantitatively based on evaluated heat transfer rates; the rewetting velocity was investigated as a function of the cladding temperature immediately before the onset of the precursory cooling. The results indicated that the propagation velocity appeared to be limited by the maximum heat transfer rate near the rewetting front. This limitation was consistent with results of the heat conduction analysis using heat transfer models for the precursory cooling expressed as a function of distance from the rewetting front, the maximum wetting temperature, and the heat transfer coefficients in the wetted region. This paper also discusses uncertainties in the evaluation of transient heat flux from the measured surface temperature, and technical issues requiring further investigation.     

基于RELAP5的双功能液态锂铅实验包层模块安全分析

利用嵌入了液态锂铅（LiPb）的热工水力子模块的系统程序RELAP5/MOD3,对双功能液态锂铅（DFLL）实验包层模块（TBM）的安全特性进行评价。对DFLL-TBM及其辅助冷却系统的稳态运行工况、预期运行事件和相关事故工况进行了建模、计算和分析。计算结果表明,稳态运行时第一壁（FW）结构材料表面最高温度低于允许值550 ℃。事故工况下氦气泄漏引起的ITER真空室（VV）、窗口设备室（port cell）以及托卡马克冷却水系统大厅拱顶（TCWS vault）的增压均低于ITER要求的限值0.2 MPa。实验包层钢结构不会熔化且可通过辐射换热有效地导出衰变余热。DFLL-TBM的设计可满足ITER对其热工水力安全方面的要求。     

Numerical simulation of a blanket cooling system for fusion reactor based on PWR conditions

The simulations of a blanket cooling system were presented to address the choice of cooling channel geometry and coolant input data which are related to blanket engineering implementation. This work was performed using computer aided design (CAD) and computational fluid dynamics (CFD) technology. Simulations were carried out for the blanket module with a size of 0.6 m × 0.45 m in toroidal plane, and the nuclear heat was applied on the cooling system at P_n (neutron wall load) of 5 MW/m². The structure factors and input data of hydraulics were investigated to explore the optimal parameters to match the PWR condition. It was found that the inlet velocity of first wall (FW) channel should be within the range of 2.48–3.34 m/s. As a result, the temperature rise (TR) of the coolant in the FW channel would be 24–25 K. This leads to the remaining space for TR within the range of 15 K in the piping circuits. It also indicated that the FW plays an important role in TR (reaches 60% of the whole cooling system) due to its high level of P_n and heat flux in the zones. It was predicted that the nuclear heat inside blanket module could be removed completely by the piping circuits with an acceptable pipe bore and the related input data. Finally, a possible design range of cooling parameters was proposed in view of engineering feasibility and blanket neutronics design.     

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5 MW/m²) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V = 1–3 m/s, P = 2–3 MPa, T = 110–170 °C). The heat flux was gradually varied in the range of 1–10 MW/m² until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2 m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3–8 MW/m²) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.     

Analysis and optimization of the manifolds of the HCCB blanket for CFETR

Thermal-hydraulic performance is a challenging issue in helium-cooled ceramic breeder (HCCB) blanket design due to the extremely complicated working environment and the strict limits of materials temperature. The heat loads deposited on the HCCB blanket comprises of severe surface heat flux from plasma and the volumetric nuclear heat from neutron irradiation, which can be exhausted by the built-in cooling channels of the components. High pressure helium with 8 MPa, distributed from the coolant manifolds is employed as coolant in the blanket. The design and optimization of the manifolds configuration was performed to guarantee the accurate flow control of helium coolant. The flow distribution in the coolant manifolds was investigated based on the structural improvement of manifolds aiming at overall uniform mass flow rates and better flow streamline distribution without obvious vortexes. The peak temperature of different functional materials in the blanket under normal operating condition is below allowable material limits. It is found that the components in the current blanket module could be cooled effectively under the intense thermal loads due to the updated design and optimization analysis of manifolds.     

3×3棒束通道内蒸汽对流换热特性数值分析

采用计算流体力学（CFD）方法，建立3×3棒束模拟体的数值模型，进行蒸汽冷却条件下的对流传热特性分析。结果表明:棒束通道内周向的壁面热流密度不均匀性明显，体现出流固耦合方法相比于均匀热流方法对传热细节模拟的优越性。蒸汽速度场、温度场、热流密度、换热系数等热工参数分布规律受入口效应、壁面效应、热源分布、物性参数等因素影响。压力的升高及氢气的加入均能提升通道内的换热性能。加热段换热系数沿程变化趋势与文献[13]中Deissier的趋势一致，CFD的换热系数结果与WCOBRA/TRAC程序中的关系式吻合较好。本文模拟方法可行，其结果可为后续的实验模拟体设计提供技术支持。      

冷却气体对黑腔系统温度场影响数值模拟

为了研究氦氢冷却气体对黑腔系统温度场的影响，采用CFD数值模拟方法，计算了氘氚靶丸外表面最大温差与填充区域的气体流场随气压、氦气含量变化的规律。通过对冷却壁面施加壁温扰动函数，监测了靶丸外表面平均温度、最大温差随时间的波动。研究结果表明：提高氦氢混合气体的填充压力或减小氦气含量，使得黑腔上下部分冷却气体自然对流强度差异增大，导致靶丸外表面温度场均匀性恶化；但降低冷却气体中氦气含量使气体导热系数减小，比热容增大，使得冷却壁温扰动对靶丸外表面温度场均匀性的影响减弱。     

Optimization of Hot Gas Chamber in High Temperature Reactor

The flow field in the hot gas chamber of the High Temperature Gas Cooled Reactor (HTGCR) was studied with the Computational Fluid Dynamics (CFD) program CFX5. On the basis of the experimental studies, the velocity field, pressure field and temperature field in the hot gas chamber and hot gas duct were obtained, and the simulation's accuracy and reliability were validated by comparison with the results of previous experiments. Two other design schemes of the hot gas chamber were calculated in order to determine which hot gas chamber would be optimal for minimizing temperature differences at the inlet of the heat exchanging components. The results indicated that there was much highly turbulent twisting flow in the hot gas chamber, which was responsible for the excellent temperature mixing effect of the hot gas chamber. But the flow in the rib region was calm, and this fact hindered the heat transfer between the hot and cold gas. The temperature mixing coefficient increases with the increase of the hot gas duct's distance from the hot gas chamber. The hot gas chamber without ribs was more beneficial to the heat transfer between air flows with different temperatures, so the hot gas chamber without ribs was indicated as the optimal design.