重水堆密集化乏燃料干式贮存设施热工安全研究

为解决秦山第三核电有限公司（简称:秦三厂）计划延寿导致乏燃料增加、已有乏燃料干式贮存模块容量不足的问题,在原有的1~6号（QM-400）乏贮模块基础上,研发了密集化乏燃料干式贮存设施（M1型乏贮模块）。与QM-400乏贮模块相比,M1型乏贮模块贮存容量更大,能量密度更高。为论证M1型乏贮模块的热工安全性,采用RELAP5/MOD3程序,根据保守的初始假设条件建立其热工分析模型,对极端气候条件下模块正常运行和事故工况下各区域温度进行了计算,同时采用了三维流体计算流体力学（CFD）程序对RELAP5程序计算结果进行了验证,综合RELAP5程序和CFD程序的计算结果,论证了M1型乏贮模块的热工安全。      

乏燃料干式储存模块衰变热导出的三维仿真计算分析

建立了QM-400储存模块内自然对流换热的三维流动及换热的计算模型,计算了QM-400储存模块内部的三维流场和温度场,得到了压力分布和速度分布以及储存筒表面的温度分布.在储存筒内部,建立了三维流场和换热的计算模型,并进行了计算,获得了乏燃料包壳的表面温度.     

乏燃料贮存罐内的自然对流传热特性数值分析

利用三维流场计算程序PHOENICS 3 2研究了水平布置干式乏燃料贮存罐内的自然对流传热特性 ,将计算结果与文献的实验数据及计算结果进行了对比分析 ,符合较好。采用同样的布置方式及 2 0 0MW低温供热堆燃料组件的结构及余热功率参数 ,分别采用氮气及水作为工质进行了计算 ,以研究乏燃料贮存罐应用于低温供热堆的可能性。从计算结果中还发现水作为工质的换热效率比氮气高     

过冷沸腾自然对流两相CFD模拟及应用

采用计算流体力学（CFD）方法,开展过冷沸腾自然对流两相模拟与应用研究。对侧壁加热圆柱水箱过冷沸腾自然对流实验采用两相CFD瞬态模拟,模拟时间为1 500 s,通过模型设置与模拟方法研究,再现了过冷沸腾发生后实验的温度阶跃,得到与实验较一致的温度分布、气泡产生时间与产生位置,确保了数值计算的合理性与准确性。在此基础上,对以欧洲ESBWR（经济简化沸水堆）非能动安全壳冷却系统（PCCS）为原型的ISP-42实验进行了两相CFD模拟,获得与实验一致的温度分布,确定采用两相CFD数值模拟对非能动安全壳冷却系统及非能动余热排出系统进行应用研究可行,为下一步计算传热系数、构建自然对流传热模型建立了良好基础。该项研究对工程应用中探寻非能动安全壳冷却系统及非能动余热排出系统的两相自然循环传热特性具有较大价值。     

过冷沸腾自然对流两相CFD模拟及应用

采用计算流体力学(CFD)方法,开展过冷沸腾自然对流两相模拟与应用研究。对侧壁加热圆柱水箱过冷沸腾自然对流实验采用两相CFD瞬态模拟,模拟时间为1 500 s,通过模型设置与模拟方法研究,再现了过冷沸腾发生后实验的温度阶跃,得到与实验较一致的温度分布、气泡产生时间与产生位置,确保了数值计算的合理性与准确性。在此基础上,对以欧洲ESBWR(经济简化沸水堆)非能动安全壳冷却系统(PCCS)为原型的ISP-42实验进行了两相CFD模拟,获得与实验一致的温度分布,确定采用两相CFD数值模拟对非能动安全壳冷却系统及非能动余热排出系统进行应用研究可行,为下一步计算传热系数、构建自然对流传热模型建立了良好基础。该项研究对工程应用中探寻非能动安全壳冷却系统及非能动余热排出系统的两相自然循环传热特性具有较大价值。     

快中子通量实验堆失流事故三维数值模拟

为验证计算流体动力学（CFD）方法在钠冷快堆失流事故模拟计算中的可靠性和可行性,针对快中子通量实验堆（FFTF）,建立了包含冷池、热池、堆芯在内的全三维模型,其中堆芯组件简化为多孔介质模型,堆芯保留了盒间特征,各类隔板简化为无厚度面。失流事故下主要参数计算结果与实验数据的对比表明,CFD方法能有效捕捉冷池、热池以及盒间复杂的流动换热现象,堆芯最热组件的位置在瞬态过程发生了变化,热管段出口温度与实验值符合良好,装有温度测点的组件出口温度模拟值较实验值低。CFD方法仍需针对组件盒间进行相应的模型开发和验证,此外还需进行大量全堆级别的实验验证,以保证计算结果的合理性。     

高温气冷堆非能动舱室冷却系统排热功率计算分析

非能动舱室冷却系统(RCCS)是模块式球床高温气冷堆(HTR-PM)的重要安全设施，准确预测事故工况下其与反应堆压力容器间的传热量对于RCCS设计具有重要意义。本文依托HTR-PM热态调试阶段反应堆压力容器壁面温度分布，采用计算流体动力学(CFD)方法，开展了RCCS全比例三维辐射传热及对流换热模拟。结果显示，Realizable k-ε湍流模型与Discrete Ordinates辐射传热模型可准确预测RCCS的排热功率，数值结果与测量结果相对误差在10%左右。基于THERMIX程序计算得到的事故工况后反应堆压力容器壁面温度分布，计算分析了投入不同列数RCCS及不同冷却水温度下的排热功率，并给出了不同工况时水冷壁与混凝土温度分布计算结果。     

压水堆装载MOX燃料以及引入干式贮存的燃料循环成本计算分析

针对压水堆装载MOX燃料以及引入干式贮存燃料循环情景,建立了燃料循环成本经济性计算模型,分别对基准情景、装载MOX燃料情景和引入干式贮存燃料循环情景进行具体计算,并就天然铀价格、贴现率对3种燃料循环模式的平准化发电成本的影响进行了敏感性分析,计算结果能为分析压水堆装载MOX燃料以及引入干式贮存经济可行性提供参考。     

核电厂堆芯出口温度测量表征性分析

堆芯出口温度测量对于掌握反应堆运行状态有着重要的意义，本文通过计算流体动力学（CFD）方法对堆芯出口温度测量的表征性进行分析。通过对燃料组件及仪表管结构进行模拟计算，获得了仪表管内冷却剂流场和温度分布；通过对9种典型功率分布下堆芯出口温度测量结果的定量分析，获得了堆芯出口温度表征性与燃料组件功率的关系。结果显示，测点平均温度与燃料棒功率基本呈线性关系，其测点温度随燃料棒功率的增加而增加，测温表征性随燃料棒功率的升高而变差。研究结果为堆芯出口温度测量的校正提供了一定的依据。     

恰希玛核电厂压力容器各关键部位在PTS瞬态下的温度场研究

承压热冲击现象在核电厂延寿评估中应被重点关注。本文针对恰希玛核电厂1号机组的压力容器及堆内构件建立了完整的CFD模型，计算了正常工况下压力容器内冷却剂的速度场和温度场分布，计算结果与试验结果符合良好。本文详细研究了蒸汽发生器传热管破裂事故工况下压力容器接管及下降段中冷却剂的热工水力特性，并将计算结果与RELAP5计算结果进行对比，结果表明二者符合良好。本文研究可为反应堆压力容器老化管理评估的计算分析工作提供重要参考。     

Estimation of fuel and clad temperature of a research reactor during dry period of de-fuelling operation

Heat transfer and fluid flow studies related to spent fuel bundle of a research reactor in fuelling machine has been carried out. When the fuel is in reactor core, the heat generated in the fuel bundle is removed by heavy water under normal reactor operation. However, during the de-fuelling operation, the fuel bundle is exposed to air for some period called dry period. During this period, the decay heat from fuel bundle has to be removed by air flow. This flow of air is induced by natural convection only. In this period, the temperatures of fuel and clad rise. If clad temperature rises beyond a certain limit, structural failure may occur. This failure can result into release of fission products from fuel rod. Hence the temperature of clad has to be within specified limit under all conditions. The objective of this study is to estimate the clad temperature rise during the dry period.In the CFD simulation, the turbulent natural convection flow over fuel and radiation heat transfer are accounted. Standard k-? model for turbulence, Boussinesq approximation for computing the natural convection flow and IMMERSOL model for radiation are used.The steady state and transient CFD simulation of flow and heat is performed, using the CFD code PHOENICS. The steady state analysis provides the maximum temperature the clad will attain if fuel bundle is left exposed to air for sufficiently long time. For safe operation, the clad temperature should be limited to a specified value. From steady state CFD analysis, it is found that steady state clad temperature for various decay powers is higher than the limiting value. Hence transient analysis is also performed. In the transient analysis, the variation of clad temperature with time is predicted for various decay powers. Safe dry time, i.e. the time required for clad to reach the limiting value, is predicted for various decay powers. Determination of safe dry time helps in deciding the time available to the operator to drop the bundle in light water pool for storage. The analysis is found useful in optimizing the de-fuelling process.     

基于辐射-对流-传导热流固耦合模型的乏燃料贮运容器传热特性与优化研究

基于辐射-对流-传导的热流固耦合计算流体动力学（CFD）模型,对比不同放置状态、环境温度、容器内部填充介质和有无环形翅片情况下,各部件最高温度、自然对流流速、外表面辐射及对流功率的变化。结果表明,水平放置有利于强化环形翅片的对流传热;环境温度每升高10℃,燃料包壳温度增加6.5℃,外壁面温度增加8.3℃;燃料篮内部填充物由中子吸收板或铝块更换为氦气后,容器整体导热性能明显降低,容器内部温度升高,但容器壁面温度基本不变;如不考虑太阳暴晒,环形翅片可增加对流传热,使得容器整体温度降低;如考虑太阳暴晒,环形翅片会增加太阳暴晒能量的吸收,使容器整体温度反而比光滑壁面容器高;基于代数分析法和漫灰表面模型对CFD辐射模型进行验证,CFD辐射功率和公式法计算结果基本一致。      

CFX simulation of high temperature thermal–chemical experiment: CS28-2

A computational fluid dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal–chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal–chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (large break loss of coolant accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal–chemical behavior of the 28-element fuel channel, the measured temperatures of the fuel element simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions.     

A numerical investigation of natural convection heat transfer in horizontal spent-fuel storage cask

Natural convection heat transfer in a horizontally placed dry spent-fuel storage cask is numerical investigated. The commercial computational fluid dynamics (CFD) code, -3.2 is used and the laminar and turbulent model are employed. The numerical predictions obtained are compared with the experimental data reported by Nishimura et al. [J. Nucl. Sci. Technol. 33 (1996) 821]. The computational results corresponding to laminar model agree well with the experimental data, but the calculated results of turbulent model are higher. The velocity pattern and the isotherms are drawn. With the increasing of Rayleigh number, the heat transfer in the cask changes from conduction dominant mode to convection dominant mode. In the condition of Ra_m=1.3×10⁹, turbulent model prevails. The convective heat transfer is so strong that almost all temperature changes take place in the region near the wall of the cask. The Rayleigh number Ra_m and the Nusselt number Nu_m characterized by maximum temperature difference are defined to depict the heat transfer characteristics. It is found laminar and turbulent models predict the same trend but different value. The flow patterns in the cask can be divided to three regimes. In these three regimes, modified Nusselt numbers are proportional to the 0.7, 0.25 and 0 power of the modified Rayleigh number, respectively.     

基于OpenFOAM的熔融池自然对流传热与凝固数值研究

熔融物堆内滞留是第3代核电技术重要的严重事故缓解措施之一,堆芯熔融池在压力容器下封头壁面的热流密度分布直接影响该策略的有效性。本文基于开源的数值计算流体力学软件平台OpenFOAM,应用相变模型和浮升力模型二次开发了用于模拟堆芯熔融物由内热源或温差驱动的自然对流传热与相变求解器。应用该求解器模拟了瑞典皇家理工学院开展的二维氧化池与金属层耦合传热试验,获得了氧化池和金属层硬壳的相场,以及熔融池内的温度分布及沿容器壁面的热流密度分布。计算结果表明,该模型可用于熔融物凝固与自然对流的模拟,为深入分析核电厂采用熔融物堆内滞留措施后熔融池的行为奠定了基础。     

Study of heat transfer by using DEM–CFD method in a randomly packed pebble-bed reactor

The pebble-bed reactor is one of the most promising designs for the nuclear energy industry. In this paper,a discrete element method-computational fluid dynamics(DEM-CFD) approach that includes thermal conduction, radiation, and natural convection mechanisms was proposed to simulate the thermal-fluid phenomena after the failure of forced circulation cooling system in a pebble-bed core. The whole large-scale packed bed was created using the DEM technique, and the calculated radial porosity of the bed was validated with empirical correlations reported by researchers. To reduce computational costs, a segment of the bed was extracted, which served as a good representative of the large-scale packed bed for CFD calculation. The temperature distributions simulated with two different fluids in this DEM-CFD approach were in good agreement with SANA experimental data. The influence of the natural convection mechanism on heat transfer must be taken into account for coolants with strong convective capacity. The proposed DEM-CFD methodology offers a computationally efficient and widely applied method for understanding the heat transfer process in a pebble-bed core. The method can also be easily extended to assess the passive safety features of newly designed fluoride-salt-cooled pebble-bed reactors.     

乏燃料后处理玻璃固化产品干法贮存通风方式优化研究及仿真模拟分析

本文介绍了高放发热体干法贮存的一般通风形式,采用隔热材料优化自然通风方式,并运用CFD技术对其进行模拟计算分析。研究发现采用隔热措施能有效降低发热体对混凝土结构的热影响,强化自然通风,达到节能和降低运行维护费用的效果。