棒束通道中燃料棒壁面温度的子通道分析

燃料棒的壁面温度是反应堆设计和运行过程中需要关注的重要参数之一。本文利用子通道程序对燃料棒壁面温度进行模拟,通过与实验数据对比,分别分析了子通道程序中的单相和两相换热模型。单相换热模型中,采用适用于棒束的Weisman公式与常用的D-B公式对比计算并最终选用Weisman公式。两相换热模型中,选用RELAP4公式包进行计算并对其判断准则做了简要分析。最终通过对空泡份额模型的对比,选择了Modified Armand模型,获得了较为准确的计算结果。     

行波堆TP-1堆芯热工水力单通道与子通道分析方法研究

以泰拉能源公司提出的钠冷行波堆TP-1为研究对象,通过钠冷行波堆瞬态安全分析程序TAST得到堆芯各组件内冷却剂、包壳和燃料棒的平均温度分布。用子通道分析程序SACOS-Na对TAST计算得到的最热组件进行详细分析计算,得到该组件内冷却剂的温度、压力和流速分布,并得到燃料棒和包壳的温度场。结果表明：单通道与子通道的结合使用能有效提高计算效率,提高反应堆设计的安全性。     

子通道分析程序LINDEN的开发与初步验证

中国广东核电集团有限公司自主开发的子通道分析程序LINDEN采用基于同位网格有限差分技术的四方程漂移流模型以及面向对象的模块化编程技术。该程序具备分析计算的可靠性、稳定性。通过LINDEN和COBRA-Ⅳ程序分别对大亚湾1#、2#机组稳态工况进行了计算分析。结果表明,LINDEN程序和COBRA-Ⅳ程序的计算结果总体吻合较好,LINDEN程序可适用于大型压水堆的热工水力分析。     

基于双区域模型的钠冷快堆组件子通道分析程序的开发与验证

子通道分析程序是钠冷快堆堆芯热工水力设计和安全分析的重要工具。本文为计算和分析钠冷快堆组件在径向均匀与倾斜功率分布工况下的热工水力特性,利用双区域绕丝交混模型开发了一款适用于钠冷快堆组件分析的子通道程序SPLICA,并与FFM2A 19棒束实验数据与WARD 61棒束实验数据进行了对比验证。由于本文开发的子通道分析程序SPLICA使用了详细的绕丝交混模型,与经过二次开发后的COBRA程序的计算结果相比,对于FFM2A实验SPLICA程序计算得到的结果与实验结果符合得更好。这两个实验数据的验证结果证明了本文开发的子通道分析程序的准确性以及对高流量工况和低流量工况均具有良好的适用性。本程序能为钠冷快堆组件热工水力分析提供有效的设计和研究手段。     

5MW 低温供热堆棒束的稳态子通道分析

本文介绍了 CHAN-2T 程序的热工水力模型及求解方法的特点。给出了 5MW 低温供热堆棒束一维稳态子通道分析结果,并与三维分析的结果进行了比较。     

快堆燃料组件的子通道分析

开发了ＴＨＡＳ－ＰＣ２子通道分析微机程序，用于计算稳态和瞬态工况下快堆燃料组件的流量、温度和压力等参量分布。对ＥＦＲ燃料组件的稳态和瞬态计算结果如下：堆芯出口平均温度和温长分别为５５７℃和１５７℃，最高包壳表面温度为６０１℃，它发生在中心燃料棒上，最大冷却剂温度为５９３℃；主泵断电二次停堆事故作为瞬态计算，算得的最高冷却剂温度和最高包壳表面温度分别为６３０℃和６３７℃（当ｔ＝２ｓ时），它们都远低于     

子通道分析软件CORTH的研发

子通道分析软件CORTH基于具有滑速比的四方程模型,适用于反应堆堆芯或加热棒束实验热工水力分析。CORTH软件的研发采用模块化设计和面向对象的编程语言,针对输入和输出特别设计了图形化的用户界面。软件通过了独立的第三方测试,检验了编码的可靠性和规范性。利用核电厂实测数据、国际基准题和AP1000额定工况对软件进行验证。结果表明,CORTH软件的计算精度较高,与国际同类软件相当,能够满足工程设计与分析需求。     

子通道分析中的湍流交混研究综述

在子通道分析中,湍流交混是冷却剂通道间横向交混的重要组成部分,是由于流体脉动时自然涡团扩散引起的非定向交混。湍流交混的强弱程度将影响通道的局部热工参数,从而影响临界热流密度的预测,是反应堆热工水力设计与分析重点关注的对象。本文针对湍流交混的相关研究进行了综述,包括机理与模型、湍流交混系数、实验方法、计算流体动力学(CFD)方法和子通道软件中的模型等,可作为自主化燃料组件设计和自主化子通道分析软件开发的参考。     

基于CFD的钠冷快堆子通道绕丝搅浑系数选取

钠冷快堆能够提高铀资源的利用率,减少核废料的产生,是非常有前景的第四代核能系统堆型之一。同时,钠冷快堆也因其使用金属绕丝对燃料棒进行固定,具有更复杂的堆内构造,探究钠冷快堆堆芯内因绕丝而引起的搅浑效应对钠冷快堆的堆芯设计及安全分析具有重要意义。本文针对钠冷快堆的堆芯设计,采用CFD软件建立带绕丝的7根燃料棒束模型,针对大流量工况（工况1）、中流量工况（工况2）进行工况计算,根据流场的雷诺应力获得绕丝的湍流搅浑系数。并基于自主研发的子通道计算程序SAC-SUB建立相同的几何模型,将湍流搅浑系数输入子通道计算程序中,获得内通道、边通道、角通道温度分布,并将两种软件的计算结果进行了对比。对比结果表明,对于不同的通道而言,两种计算软件内通道的温度偏差最小（2.5℃）,角通道的温度偏差最大（13.2℃）。对于不同的流量而言,中流量工况（工况2）温度偏差更小,最小温差只有0.8℃。该工作为后续快堆子通道分析搅浑系数的选取提供了技术基础。     

液态金属冷却快堆子通道分析软件SACOS-LMR研发与工程应用

子通道分析方法是反应堆堆芯设计和热工水力分析的重要手段之一,对于我国提出的压水堆-快堆-聚变堆三步走核能发展战略,开发适用于液态金属冷却快堆热工安全分析的子通道分析程序具有重要意义。本文基于西安交通大学热工水力研究室自主开发的压水堆子通道程序SACOS,通过添加液态金属快堆特有的模型,如绕丝模型、盒间流模型、液态金属对流换热模型等,扩展至适用于液态金属快堆的子通道分析程序SACOS-LMR,该程序具备对液态金属快堆组件开展稳态和瞬态热工水力分析的功能。结合卡尔斯鲁厄开展的37棒钠冷瞬态实验,完成了SACOS-LMR程序的瞬态功能验证。基于验证后的SACOS-LMR程序,对欧洲铅冷快堆(ALFRED)堆芯开展了稳态工况和瞬态事故工况下的热工安全特性分析,计算结果合理,且与同类程序保持一致,表明SACOS-LMR程序可用于液态金属快堆的堆芯设计和热工水力分析研究。     

Spacer Effect Model for Subchannel Analysis

Mean velocity and velocity fluctuation in a test channel that consisted of five subchannels with and without ferrule-type spacer were measured using air as a working fluid, to clear turbulence intensity enhancement due to spacer. Measurements were performed at Reynolds number of 0.5–1.2×10⁵, which simulated vapor flow velocity of annular-dispersed flow in BWR condition. It was confirmed that magnitudes of velocity fluctuations in radial direction were proportional to Reynolds number and square root of friction factor downstream from a spacer. New spacer effect model to describe turbulence intensity enhancement due to the spacers was developed. In the model, dependence of the velocity fluctuation on ferrule thickness was correlated by blockage ratio. It was found that the present spacer model is applicable to prediction of turbulence intensity enhancement due to spacer.     

Method of Critical Power Prediction Based on Film Flow Model Coupled with Subchannel Analysis

A new method was developed to predict critical powers for a wide variety of BWR fuel bundle designs. This method couples subchannel analysis with a liquid film flow model, instead of taking the conventional way which couples subchannel analysis with critical heat flux correlations. Flow and quality distributions in a bundle are estimated by the subchannel analysis. Using these distributions, film flow rates along fuel rods are then calculated with the film flow model. Dryout is assumed to occur where one of the film flows disappears. This method is expected to give much better adaptability to variations in geometry, heat flux, flow rate and quality distributions than the conventional methods.

In order to verify the method, critical power data under BWR conditions were analyzed. Measured and calculated critical powers agreed to within ±7%. Furthermore critical power data for a tight-latticed bundle obtained by LeTourneau et al. were compared with critical powers calculated by the present method and two conventional methods, CISE correlation and subchannel analysis coupled with the CISE correlation. It was confirmed that the present method can predict critical powers more accurately than the conventional methods.     

Application of Galerkin Finite Element Method to Subchannel Analysis

Most of the computer codes based upon subchannel analysis, which are important for thermal hydraulic analysis of the reactor core, use the finite difference method to solve the set of equations. In the present study, however, the Galerkin finite element method was tried, with the result that more accurate solutions and more efficient calculations were obtained than those by the finite difference method. The results of error evaluation obtained herein are useful for application of this method to actual subchannel analysis codes and to other general thermal hydraulic analysis codes. As an example, steady-state single-phase subchannel analysis was performed.     

Application of Quick Subchannel Analysis Method for Three-Dimensional Pin-by-Pin BWR Core Calculations

Three-dimensional pin-by-pin core analysis is considered to be a candidate for the next-generation BWR core calculation method. In our previous study, the applicability of the transport and burnup calculations for a three-dimensional pin-by-pin BWR core analysis was investigated. However, the thermal-hydraulics calculation has not yet been studied in this framework. In the conventional core analysis code, the bundlewise thermal-hydraulics calculation is adopted. In the actual core analysis, the power distribution inside a fuel assembly is tilted at the region adjacent to a control blade or the core peripheral region. In these regions, the consideration of the subchannel-wise void distribution has an impact on the fission rate distribution. Therefore, an evaluation of the detailed void distribution inside an assembly, i.e., the incorporation of the subchannel wise void distribution, is desirable for the pin-by-pin BWR core analysis. Although several subchannel analysis codes have been developed, these subchannel analysis codes generally require a large computational effort to estimate the subchannel-wise void distribution in a whole BWR core. Therefore, to analyze a whole BWR core within a reasonable computation time, it was necessary to apply a fast subchannel analysis code. In this paper, a quick subchannel analysis code dedicated to pin-by-pin BWR core analysis is newly developed, and the void distribution of the present subchannel analysis code is compared with the prevailing subchannel analysis code NASCA using three-dimensional single-assembly geometries. Since the present subchannel analysis code is used for a coupled neutronics/thermal-hydraulics analysis, the results of the coupling calculation are also compared with those of NASCA. The calculation result indicates that the void distribution difference between NASCA and the present subchannel analysis code is slightly less than 10%. This result indicates that the prediction accuracy of the present subchannel analysis code will be reasonably appropriate for a pin-by-pin BWR core analysis. Furthermore, the results show that the calculation time of the present subchannel analysis code is only 10 min for a hypothetical three-dimensional ABWR quarter-core geometry using a single CPU. This calculation time is sufficient for a pin-by-pin BWR core analysis.     

子通道程序与CFD程序的耦合方法研究

随着多种新型堆型的发展,堆芯设计将更加紧凑,为了保证堆芯安全,要求在设计阶段尽可能地精确计算出堆芯内热工参数分布,这就需要针对特定瞬态工况开展堆芯多尺度耦合研究。本文在已有的子通道程序COBRA-EN的基础上,采用动态链接库技术将其耦合到流体动力学程序FLUENT中,开发了适用于堆芯多尺度计算的COBRA-EN/FLUENT耦合程序。进一步通过带腔室的棒束通道算例,分别测试了稳态和瞬态情况下耦合程序的计算精度,结果显示COBRA-EN与FLUENT两者的耦合是有效且可靠的。本研究成果将为新型堆芯的设计和安全分析提供可靠的工具。     

钍基先进坎杜堆子通道分析

利用子通道计算程序ASSERT-PV V3R1计算TACR1000在不同钍装填模式、不同功率、不同寿期下的子通道热工水力学特性.根据对子通道质量流密度、空泡份额和干涸起始功率方面的计算,从功率展平及安全性的角度考虑,钍铀粉末交混装填模式明显好于内8根棒为钍的装填模式.     

超临界水堆子通道分析

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。本工作以上海核工程研究设计院的常规压水堆子通道程序为基础,开发编制了适用于超临界水堆的子通道程序,并对典型带有慢化剂水棒的超临界水堆燃料组件进行了模拟计算,得到了堆芯子通道内的温度、燃料棒包壳温度、表面传热系数等参数的分布规律。此外,研究了不同超临界流体换热关系式对计算结果的影响,结果显示,各传热关系式的计算结果存在一定差异。     

超临界水冷堆CSR1000燃料组件子通道分析

利用实验数据和计算流体力学(CFD)商用软件CFX对现有子通道分析模型进行研究,分析其在超临界水冷堆(SCWR)分析中的适用性,并根据分析结果对ATHAS程序进行改进。采用改进的ATHAS程序对超临界水冷堆CSR1000燃料组件进行稳态子通道分析,获得燃料组件冷却剂和包壳温度分布、流动压降等参数。结果表明:减小螺旋肋螺距(Hw)可展平燃料组件冷却剂出口温度分布、降低包壳表面最高温度(MCST),但同时燃料组件流动阻力将增大。     

子通道分析程序在整理临界热流密度关系式上的应用

从反应堆热工水力实验中能获得和临界热流密度有关的各平均参数，子通道分析程序提供了一种手段，把平均参数转化成为ＣＨＦ产生处的局部参数。从而可以整理出带局部参数条件的ＣＨＦ关系式。本文介绍了用ＦＬＩＣＡⅢ－Ｍ整理局部参数ＣＨＦ关系式的详细步骤。