RELAP5—3D程序对RBMK-1500反应堆的模型验证

鉴于堆芯中子响应对RBMK．1500反应堆的重要作用，通过对伊格那林核电站（Ignalina nuclear power plant，INPP）中RBMK-1500反应堆特定瞬态的模拟，验证了RELAP5-3D程序对RBMK．1500的有效性。本文开发了一个适用于RBMK-1500的RELAP5-3D最佳估算模型，其计算结果与INPP的测量数据基本匹配。此外，基于RELAP5-3D最佳估算模型对单独主循环回路预测的热工水力参数和物理过程与RBMK-1500主回路发生的实际过程吻合良好，且计算得出的反应性和堆芯瞬态总功率与电厂的测量值也非常一致，这表明了该程序准确地模拟了堆芯中子响应过程。通过对RELAP5-3D模型的验证表明，该程序可以成功地运用于未来的RBMK-1500安全性计算。     

反应性动态加入对脉冲堆中子脉冲波形的影响

在点堆模型基础上开发了脉冲模拟程序,该程序考虑了缓发中子效应、裂变热反馈效应和脉冲棒的行进过程,可以描述脉冲堆爆发脉冲全过程的中子强度、裂变能以及反应性的变化,并经过实验数据验证。用该程序计算CFBR-Ⅱ堆在反应性动态加载过程中全波形、峰功率、反应性等脉冲特征参数,结果表明:脉冲引发的越晚,其峰功率和裂变产额越大,而且其最大裂变产额与静态爆发脉冲情况下的裂变产额相同。     

Simulink仿真软件在船用堆参数快速计算中的应用

以6组缓发中子点堆模型和堆芯双区集总参数模型为基础,建立Simulink仿真模型,对船用反应堆动态过程进行仿真分析,并与相关文献作比较.结果表明:Simulink仿真软件能以较长的平均步长对反应堆动态过程作高精度仿真,能够在普通微机上实现物理热工参数超时计算,对船用反应堆安全运行有重要意义.     

脉冲堆物理设计分析

本文介绍了我国自行设计建造的首座脉冲型实验反应堆的堆芯核设计、计算模型和程序以及计算结果与零功率实验值的分析比较。     

5MW核供热堆的控制模型

从分析５ＭＷ核供热堆的物理过程入手，以集总参数的形式，建立了适用于研究该堆控制方式的简化模型，导出了各环节的传递函数，并以实验和分析相结合的方法为各参数赋值。实验曲线表明该模型的动态特性与实际系统相近。     

脉冲堆堆芯结构抗震分析

脉冲堆堆芯结构类似于圆柱形刚性块体,由一铝质支撑简体支承。根据结构特点,我们分别采用刚体模型和有限元模型进行了响应分析,并将其结果作了比较。动态特性实验研究表明,分析结果是可靠的,采用刚体模型处理脉冲堆堆芯结构这种类型的结构是一种行之有效的方法。最后对关键部位作出了安全评价。     

先进空间快堆安全特性分析

以200kW空间快堆RAPID－L为对象，建立瞬态分析模型，分析了在无保护超功率事故UTOP和无保护失流事故ULOF下的瞬态特性。计算结果表明:快速型锂膨胀模块（LEM）可以随着冷却剂温度变化自动快速的响应，能够在不停堆的情况下保证反应堆的安全，因此，RAPID－L具有固有安全特性。     

堆芯中子动力学实时仿真模型

介绍了堆芯中子动力学实时仿真的三种不同模型：点堆模型，三维绝热模型和改进的准稳态模型。并采用了美国Ａｇｏｎｎｅ实验室一维基准问题及秦山６００ＭＷ反应堆弹棒事故对三种模型进行了运算比较。综合计算精度及运算时间表明，随着计算机运算能力的提高，用于堆芯动态仿真，改进的准稳态是国为理想的求解模型。     

基于节点模型的空间堆系统动态特性分析

建立集总参数法的碱金属冷却空间堆系统动态特性分析的节点模型,利用Simulink软件开发了空间堆系统动态特性分析程序,并利用设计参数对程序进行验证。分析了控制鼓转角和外部负载电阻阶跃变化时的系统动态响应特性。结果显示:在控制鼓角度阶跃变化引入正反应性时,堆芯功率迅速上升尔后由于负反馈而达到新的稳定状态,但热电偶(TE)电功率的输出有一定的延迟。在外部负载电阻阶跃变化时,TE热电转换电功率输出快速升高,使得TE热端温度升高,堆芯温度升高,由于负反应反馈导致堆芯温度下降。比较两者瞬态响应,外部负载电阻的变化较控制鼓角度的变化引起TE电功率输出的响应要快速。     

基于双群点堆模型的小型压水堆反应性扰动的动态模拟

本文采用双群点堆动力学模型耦合传热集总参数模型,分别对小型压水堆高、低功率条件下反应性扰动进行模拟,并与三维仿真模型进行比较.结果表明:本模型可较好地模拟小型压水堆反应性扰动情况下的功率、温度变化趋势及峰值,且分析时间短,能满足工程精度要求,可用于小型反应堆正常运行以及事故状态下反应性扰动的现场超时预测.     

基于CFD的池式快堆多物理耦合分析方法研究

基于临界/次临界点堆中子动力学模型、燃料棒传热模型、热交换器和多孔介质等辅助热工水力模型,采用显式迭代和动态链接库技术（DLL）,利用商用计算流体力学（CFD）程序FLUENT的用户自定义函数（UDF）实现中子动力学、燃料棒热传导等和快堆堆池冷却剂流动换热的耦合计算,开发池式快堆多物理耦合计算程序CFD/PF。采用CFD/PF开展小型自然循环铅铋快堆SNCLFR-10无保护超功率事故（UTOP）模拟,并与国际知名快堆多物理耦合分析程序SIMMR-III的计算结果开展Code-to-Code对比分析。研究结果表明:CFD/PF与SIMMER-III的分析结果吻合良好,耦合程序的开发取得了初步成功,可用于分析池式快堆堆池内的复杂三维流动和换热现象。      

多物理耦合方法在严重事故仿真软件SimSA开发中的应用

为满足核电厂全范围模拟机对严重事故过程仿真的需求,自主开发了严重事故仿真软件SimSA,能模拟从设计基准事故到严重事故的主要事故过程,并能准确给出相关进程的计算结果。SimSA包含3大主要模块：热工水力模块（Therm）、堆芯行为模块（Core）以及安全壳行为模块（Cont）。其中,Therm与Core两个模块的耦合过程中采用了SCDAP/RELAP5相似的基于过程机理的耦合方法。本文结合SimSA软件的具体情况介绍了这种耦合方法的实现过程,并采用耦合后的程序对大破口叠加安注失效及全厂断电叠加辅助给水丧失两个典型初因事故导致的严重事故序列进行了计算,将计算结果与相同初始条件下MAAP4的计算结果进行对比分析。结果表明,SimSA中采用的这种耦合方式是成功的。     

Numerical simulations of gas–liquid–particle three-phase flows using a hybrid method

For the analysis of debris behavior in core disruptive accidents of liquid metal fast reactors, a hybrid computational tool was developed using the discrete element method (DEM) for calculation of solid particle dynamics and a multi-fluid model of a reactor safety analysis code, SIMMER-III, to reasonably simulate transient behavior of three-phase flows of gas–liquid–particle mixtures. A coupling numerical algorithm was developed to combine the DEM and fluid-dynamic calculations, which are based on an explicit and a semi-implicit method, respectively. The developed method was validated based on experiments of water–particle dam break and fluidized bed in systems of gas–liquid–particle flows. Reasonable agreements between the simulation results and experimental data demonstrate the validity of the present method for complicated three-phase flows with large amounts of solid particles.     

Dynamic behavior of a solid particle bed in a liquid pool: SIMMER-III code verification

Dynamic behavior of solid particle beds in a liquid pool against pressure transients was investigated to model the mobility of core materials in a postulated disrupted core of a liquid metal fast reactor. A series of experiments was performed with a particle bed of different bed heights, comprising different monotype solid particles, where variable initial pressures of the originally pressurized nitrogen gas were adopted as the pressure sources. Computational simulations of the experiments were performed using SIMMER-III, a fast reactor safety analysis code. Comparisons between simulated and experimental results show that the physical model for multiphase flows used in the SIMMER-III code can reasonably represent the transient behaviors of pool multiphase flows with rich solid phases, as observed in the current experiments. This demonstrates the basic validity of the SIMMER-III code on simulating the dynamic behaviors induced by pressure transients in a low-energy disrupted core of a liquid metal fast reactor with rich solid phases.     

基于RELAP5和FLUENT的多尺度耦合分析方法研究

基于二次开发得到的铅冷快堆一维系统程序RELAP5_LEAD和三维计算流体力学程序FLUENT,利用动态链接库技术和FLUENT用户自定义函数,开发了多尺度耦合分析程序RELAP5/FLUENT。在单相范围内,分别利用耦合程序RELAP5/FLUENT开展简单铅冷串联管道的瞬态流动和传热模拟、简单铅冷闭式回路的瞬态流动模拟,并与RELAP5_LEAD计算结果开展Code-to-Code对比分析。研究结果表明,RELAP5/FLUENT计算结果与RELAP5_LEAD模拟结果吻合良好,耦合程序的开发取得了初步成功,可用于分析铅冷快堆堆内的复杂三维热工水力现象。     

Development of multicomponent vaporization/condensation model for a reactor safety analysis code SIMMER-III: Theoretical modeling and basic verification

It is believed that the numerical simulation of thermal-hydraulic phenomena of multiphase, multicomponent flows in a reactor core is essential to investigate core disruptive accidents (CDAs) of liquid-metal fast reactors. A new multicomponent vaporization/condensation (V/C) model was developed to provide a generalized model for a fast reactor safety analysis code SIMMER-III, which analyzes relatively short-time-scale phenomena relevant to accident sequences of CDAs. The model characterizes the V/C process associated with phase transition through heat-transfer and mass-diffusion limited models to follow the time evolution of the reactor core under CDA conditions. The heat-transfer limited model describes the nonequilibrium phase-transition processes occurring at interfaces, while the mass-diffusion limited model is employed to represent effects of noncondensable gases and multicomponent mixture on V/C processes. Verification of the model and method employed in the multicomponent V/C model of SIMMER-III was performed successfully by analyzing a series of multicomponent phase-transition experiments.     

基于CFD程序的物理热工耦合计算不确定性分析

基于计算流体力学（CFD）程序FLUENT的用户自定义函数（UDF）,耦合中子动力学计算模型、燃料棒热传导计算模型、不确定性分析程序SIMLAB,开发了物理热工耦合计算不确定性分析平台CFD/PFS,并开展了小型自然循环铅基快堆SNCLFR-10的无保护超功率（UTOP）事故的不确定性量化,最后对计算结果进行不确定性分析和敏感性分析。研究表明,CFD/PFS平台的物理热工耦合计算具有良好的可靠性、精确性;总反应性峰值、功率峰值等瞬态安全参数的名义值均处于95/95双侧容忍限值内,且名义值与限值相对偏差小于3.95%;燃料多普勒系数是主要不确定性来源,对反应堆安全影响最大。     

Simulation of molten metal freezing behavior on to a structure

In the severe accident analysis of liquid metal reactors (LMRs), understanding the freezing behavior of molten metal onto the core structure during the core disruptive accidents (CDAs) is of importance for the design of next-generation reactor. CDA can occur only under hypothetical conditions where a serious power-to-cooling mismatch is postulated. Material distribution and relocation of molten metal are the key study areas during CDA. In order to model the freezing behavior of molten metal of the postulated disrupted core in a CDA of an LMR and provide data for the verification of the safety analysis code, SIMMER-III, a series of fundamental experiments was performed to simulate the freezing behavior of molten metal during penetrating onto a metal structure. The numerical simulation was performed by SIMMER-III with a mixed freezing model, which represents both bulk freezing and crust formation. The comparison between SIMMER-III simulation and its corresponding experiment indicates that SIMMER-III can reproduce the freezing behavior observed on different structure materials and under various cooling conditions. SIMMER-III also shows encouraging agreement with experimental results of melt penetration on structures and particle formation.     

Addressing the heavy liquid metal – Water interaction issue in LBE system

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.