快堆严重事故工况下燃料组件盒破损机理的研究

在钠冷快增殖堆安全性分析中,六角形不锈钢燃料组件盒的破损时间和位置是一个重要的问题.对于严重的局部事故而言,钢盒破损的可能性基本上等同于事故向邻近燃料组件蔓延的可能性.本文以SCARABEE-N系列实验和SIMBATH系列实验为基础,对快堆严重事故工况下六角形钢组件盒的破损机理进行了研究.对于冷却状况良好的组件盒,提出了一种新的熔穿机理:局部热侵蚀进而诱发钠侧局部烧干,随后发生熔穿.在此基础上,对中国实验快堆(CEFR)在单个燃料组件瞬间完全堵流事故工况下组件盒破损的时间进行了预测.预测结果为,相邻燃料组件的六角形钢盒应该在堵流后7.2～8 s发生熔穿,随后事故开始向相邻的燃料组件蔓延.     

CANDU堆热工水力设计方法和计算机软件

ＣＡＮＤＵ堆热工水力设计采用目前一些典型的分析方法,并通过一些重要的实验研究计划对这些典型的分析方法进行强化支持。这些研究计划是为了发展和验证数学模型。主要针对ＣＡＤＵ堆特有的热工水力学。本文介绍了ＣＡＮＤＵ堆设计所用的计算机模型,以及为确认其     

Effects of cooling channel blockage on fuel plate temperature in Tehran Research Reactor

In this study, the variation of the temperature distribution of the fuel plate in Tehran Research Reactor core was studied in case of coolant channels blockage. While the experimental method is not possible, both the analytical and simulation methods were used to obtain the more reliable data. The results show that one channel blockage will increase the fuel temperature to about 100%, but it does not lead to clad melt down still. With further calculation and simulation it is understood that if the coolant velocity drops to 90% of its nominal value, it may causes the clad melt-ing down. At least two channels with complete blockage even at the positions far from the core center can also melt down the clad.     

中国先进研究堆标准燃料组件堆外水力稳定性试验

中国先进研究堆（CARR）标准燃料组件由滚压在两块侧板上的21块燃料板组成。堆外水力试验的目的是考验在水力冲刷条件下燃料组件的结构稳定性。试验件是按照正式产品制造工艺制造的贫铀组件,试验平均流速为12m/s,是满功率运行流速的120％。先后试验了2个组件,第1个组件试验60d,是满功率运行时间的120％,试验后观察到固定下定位梳的销钉松动,下定位梳严重磨损了燃料板;工艺改进后制造的第2个组件试验120d,是满功率运行时间的240％,试验表明,第2个组件结构完整。试验中对组件结构稳定性和燃料板腐蚀性能,诸如组件的压差、燃料板振动、包壳表面腐蚀深度等进行了研究。     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

自主化反应堆计算流体力学软件WINGS-CFD的研发与应用

为满足反应堆高精度流动传热数值仿真需求,提出了一种基于面向对象的、分层架构设计的、高可扩展的自主化反应堆计算流体动力学（CFD）软件WINGS-CFD 。从理论模型、数值离散方法、软件架构等方面介绍WINGS-CFD软件的总体设计,并采用WINGS-CFD软件对反应堆典型场景流动传热工况进行了数值计算。结果表明,WINGS-CFD计算结果的精度与商用CFD软件的结果相当;WINGS-CFD软件具备优秀的并行性能,可支持亿级网格大规模数值仿真以及中子输运与流动传热耦合仿真,为反应堆系统的精细化多物理场分析提供了自主化数值技术手段。     

Transient behaviour of a parallel pump in nuclear research reactors

A theoretical analysis of the fast transients of a parallel pump, based on inertia of the rotating parts and inertia of the fluid, is proposed. It leads to total torque, total head, and total system resistance during transient periods. The equations indicate that an increase in coolant inertia increases the acceleration head. While an increase in the moment of inertia of rotating parts decreases the acceleration head. The model is used to analyze the behaviour of the Tehran Research Reactor (TRR) primary coolant loop parallel pump during a fast startup. The results of present model are compared with similar studies and good agreement is observed.     

ACP100反应堆整体水力模拟试验研究

模块式先进小型压水堆（ACP100）是一种新型一体化小型反应堆,直流蒸汽发生器和主泵均直接集成在压力容器上,紧凑的结构导致其内部流场复杂。本研究应用1:3缩比模型模拟ACP100反应堆内部流场,开展反应堆整体水力模拟冷态试验。试验得到反应堆模型的总压降和分段压降,获得了反应堆模型总阻力系数以及主要流道分段阻力系数;并得到堆芯入口各燃料组件的流量分配因子。模型试验结果显示,主流道内的流动已进入第二自模区,流体的流型、流速分布以及阻力系数与原型反应堆相同;流动进入自模区后,反应堆模型的阻力系数为常数,阻力系数值为8.02,可直接用于原型反应堆压降计算;额定运行工况下,堆芯入口的流量分配因子值在0.91~1.08,满足设计需求;流量分配罩具有良好的整流作用,模拟失流事故工况下的流量分配仍较均匀。     

堵塞条件下紧密栅湍流交混特性研究

开展堵塞工况下紧密栅内流体子通道间隙湍流交混研究,对事故工况下燃料组件热工水力行为的预测具有重要意义。本文采用CFD方法对紧密栅内堵塞工况的流体流动现象进行了模拟,模拟结果与相关文献结果吻合较好。进一步对比分析了不同堵塞工况下,堵塞段及堵塞下游的速度场、涡结构以及湍流交混系数分布。所得不同堵塞工况下的横向与轴向湍流交混系数变化规律,可为子通道分析程序的参数设置提供参考。     

中国先进研究堆堵流事故分析

堵流事故是板状燃料研究堆发生概率相对较高的一种事故。本文进行CARR堵流事故分析时,假定额定功率运行下,1盒燃料组件的入口全部堵塞。用Relap5\Scdapsim\Mod3．2程序分析堵流事故时反应性、堆功率、压力、流量和温度等的变化,并分析发生堵流的标准燃料组件对周边燃料组件的影响。结果表明,发生堵流的燃料组件将烧毁,但不向相邻的组件扩展。CARR设计采取的安全措施可以把堵流事故后果控制在可接受范围之内。     

低流量下蒸汽发生器一次侧流量分配研究

采用CFD方法对低流量下倒U型管式蒸汽发生器一次侧内的流动特性进行了分析研究.结果表明,在自然循环模拟试验装置的U型管结构参数条件下,倒流只发生在内层的短管内.低流量时,随单管平均流量的增大,倒流管内的流量相对减小,倒流管数量相对减少,最终倒流消失;随二次侧温度升高,倒流管内的流量也相对减小,倒流管数量相对减少.     

快堆单个燃料组件完全堵流事故的建模及其验证

为了预测正常功率下快堆单个燃料组件入口完全堵流所导致的事故序列,根据SCARABEE-N系列实验建立了相关的计算模型.冷却剂的沸腾及其两相流动的描述采用两流体模型;包壳的流动、燃料的熔化及其塌陷采用类似SURFASS程序的简单方法处理.对于事故后期形成的UO2-钢混合沸腾池,采用一维半经验模型描述,即:用漂移速度模型来预测空泡份额分布;用修正后的Greene关系式计算沸腾池和壁面之间的传热系数;用焓方法(enthalpy method)求解包裹沸腾池的固化壳的温度场及厚度.为了验证本文建立的模型,对SCARABEE BE 1实验结果进行了校核计算,其结果与实验结果基本吻合.     

板状燃料组件流量分配CFD研究与优化

板状燃料组件被广泛应用于研究堆中,组件内的流量分配是设计时需要考虑的一项重要内容。计算流体动力学（Computational Fluid Dynamic, CFD）方法是研究流量分配的重要手段,但有限的计算资源限制了其在板状燃料组件流量分配研究中的推广。针对板状燃料组件冷却剂流道狭长、封闭的特点,提出了部分建模迭代求解的计算方式,将无流量分配组件与有流量分配组件两种工况下各流道流量的计算值与直接完整建模的结果进行了对比,最大误差分别为0.56%与0.81%。鉴于前者对计算资源的需求远小于后者,部分建模迭代求解可以作为板状燃料组件流量分配CFD研究的合理可信的优化方案。     

Development of CONTHAC-3D and hydrogen distribution analysis of HPR1000

An in-house code,CONTHAC-3D,was developed to calculate and analyze thermal-hydraulic phenomena in containments during severe accidents.CONTHAC-3D is a three-dimensional computational fluid dynamics code that can be applied to predict gas flow,diffusion,and steam condensation in a containment during a severe hypothetical accident,as well as to obtain an estimate of the local hydrogen concentration in various zones of the containment.CONTHAC-3D was developed using multiple models to simulate the features of the proprietary systems and equipment ofHPR1000 and ACP100,such as the passive cooling system,passive autocatalytic recombiners and the passive air cooling system.To validate CONTHAC-3D,a GX6 test was performed at the Battelle Model Containment facility.The hydrogen concentration and temperature monitored by the GX6 test are accurately predicted by CONTHAC-3D.Subsequently,the hydrogen distribution in the HPR1000 containment during a severe accident was studied.The results show that the hydrogen removal rates calculated using CONTHAC-3D for different types of PARs agree well with the theoretical values,with an error of less than 1％.As the accident progresses,the hydrogen concentration in the lower compartment becomes higher than that in the large space,which implies that the lower compartment has a higher hydrogen risk than the dome and large space at a later stage of the accident.The amount of hydrogen removed by the PARs placed on the floor of the compartment is small;therefore,rais-ing the installation height of these recombiners appropriately is recommended.However,we do not recommend installing all autocatalytic recombiners at high positions.The study findings in regard to the hydrogen distribution in the HPR1000 containment indicate that CONTHAC-3D can be applied to the study of hydrogen risk containment.     

大庆200MW低温堆换热器水力学模拟实验研究

通过１∶２的实验模型，对大庆２００ＭＷ低温核供热堆主换热器进行了水力学模拟研究。研究结果表明：当Ｒｅ＞５０００时，换热器的阻力系数已进入自模区。给出了换热器达到自模时的阻力系数及各流程间的流动阻力分布。提出了减少出口段流动阻力的优化设计方案。阻力系数的设计值与模拟研究结果相吻合。描述了发生２次流量漂移现象时，各系统参数的变化过程。为大庆２００ＭＷ低温核供热堆主换热器的最终设计提供了实验基础。     

Qualitative assessment of RANS models for Hypervapotron flow and heat transfer

The power densities present within a fusion device and its heating systems require the use of high heat flux (HHF) devices to sustain power densities >10 MW/m² in steady state. One such device, known as the Hypervapotron, uses internal water cooling along with a series of fins and cavities perpendicular to the flow to maximise the heat transfer capability. UKAEA, in collaboration with Cranfield University, have initiated a study whereby computational fluid dynamics (CFD) software will be used to predict the variation of heat transfer coefficients (HTC) throughout the Hypervapotron, allowing accurate calculations of both thermal and thermo-mechanical performance. In this paper, the first steps in this process are presented. In particular, different turbulence models are assessed and best practices identified for predicting single phase flow and heat transfer within Hypervapotron-sized cavities.     

200MW低温供热堆盒间流道流动传热数值分析

采用三维CFD软件Phoenics - 3.2 ,计算了 2 0 0MW低温供热堆燃料组件盒间的流场及温场。研究了旁通流量、控制棒提升等因素的影响。在考虑这些因素之后 ,得出了最佳旁通入流方案。     

堵流条件下矩形并联通道非均匀流动传热模拟方法研究

为建立矩形并联通道非均匀流动传热模拟方法,针对板型燃料元件的安全分析提供新的模拟方法和工具,本研究采用一维两流体模型和燃料元件二维导热模型开发热工水力瞬态分析程序,对堵流条件下非均匀流动传热进行模拟。通过数值模拟得到不同堵流工况下流量分配和燃料温度分布,此外对4种不同功率分布下燃料元件二维导热效应进行研究。研究结果表明,堵流后并联通道流量和传热量将重新分配,二维导热模型使燃料元件截面温度场分布更均匀。本文开发的热工水力瞬态分析程序能够用于板型燃料元件非均匀流动传热现象的模拟。     

半椭球堵流结构对窄通道流动传热特性的大涡模拟研究

为研究半椭球堵流结构对窄通道内冷却剂流动传热特性的影响,针对高度为窄边宽1/2的半椭球堵流结构开展了大涡模拟,分析了堵流结构下游的时均速度场、湍动能和温度分布,研究了堵流结构附近的流动传热特征。半椭球堵流结构下游流场有显著的三维特性,存在绕流及边界层分离形成的回流区、剪切层、主流区和再发展区等特征区域。结合时均温度场及局部努塞尔数（Nu）的变化规律,研究了堵流结构对窄通道传热的影响机制。分析发现,半椭球底面侧壁面附近回流区及其顶部壁面再发展区内,由于热流体聚集,缺乏主流低温流体冲扫,出现局部Nu极小值和局部高温。总体上看,半椭球堵流结构对窄通道速度场和温度场的展向影响范围为5倍半椭球长半轴范围。     

Numerical analysis of thermal-hydraulic behavior of supercritical water in vertical upward/downward flow channels

Investigations on the thermal-hydraulic behavior in the SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding of the heat transfer behavior of supercritical fluids. In this paper, the numerical analysis is carried out to study the thermal-hydraulic behaviour in vertical sub-channels cooled by supercritical water. Remarkable differences in characteristics of secondary flow are found, especially in square lattice, between the upward flow and downward flow. The turbulence mixing across sub-channel gap for downward flow is much stronger than that for upward flow in wide lattice when the bulk temperature is lower than pseudo-critical point temperature. For downward flow, heat transfer deterioration phenomenon is suppressed with respect to the case of upward flow at the same conditions.