基于系统程序的压水堆核电厂热力系统建模

利用核电站最佳估算热工水力系统程序RELAP5,以大亚湾核电站的核岛和常规岛为模型,对压水堆核电站一、二回路整体的热工水力系统进行建模分析.研究了传统核电站安全分析建立的基本系统模型和常规岛二回路主要的系统模型,主要针对汽轮机回路的建模进行研究分析.稳态数值计算结果与核电站满功率运行数据基本一致.     

反应堆功率控制系统的建模及闭环验证

基于MATLAB/SIMULINK仿真软件,对岭澳一期核电站功率控制系统进行建模.通过与RELAP5完成的核电厂主回路热工水力模型相连接,对该控制系统模型进行了闭环瞬态工况测试.将计算机仿真结果与电厂试验曲线进行了比较,两者非常吻合.定性和定量地验证了本控制模型和热工模型的正确性.     

压水堆核电站堆芯物理/热工水力耦合特性研究

采用RELAP5-HD作为堆芯耦合计算程序,以秦山核电二期工程反应堆堆芯为研究对象,建立堆芯活性区的物理/热工水力耦合模型,在此基础上进行了稳态计算和掉棒事故仿真研究。结果表明,使用RELAP5-HD计算得到的结果与电厂实测值符合较好,获得的掉棒事故参数曲线能准确反映事故工况下的参数变化趋势。稳态和事故工况的计算结果均符合堆芯物理/热工水力反馈效应的理论分析,证实了所建立的堆芯耦合模型的准确性,为下一步进行核电站系统的仿真分析提供基础。     

核电厂汽轮机详细数值建模研究及其瞬态分析

以岭澳一期核电厂汽轮机部件为原型,利用系统程序RELAP5对其进行详细数值建模研究。通过在100%功率稳态工况下的计算证明,详细的汽轮机数值建模弥补了简化建模中焓值计算误差较大的缺陷。将详细的汽轮机数值建模整合到全范围核电厂热力系统模型中进行瞬态分析,并与岭澳一期核电厂原始实验报告中汽轮机负荷从97%功率水平阶跃变化至87%功率水平瞬态运行工况的数据曲线进行对比。结果表明,稳态模型的焓计算值与电厂实际值误差在2%以内,瞬态模型的分析参数趋势符合电厂实际情况。     

用Simulink扩展RELAP5的控制与保护系统仿真功能

RELAP5程序本身具有模拟核电站控制与保护系统的功能，但是，由于该程序采用文本输入方式进行建模，编写复杂，可读性不强，小适合于对大型复杂控制系统进行仿真。而Simulink程序采用图形化建模方式，能够高效、便捷地对核电厂复杂控制与保护系统进行建模。因此，本文将RELAP5程序与Simulink耦合，并利用Simulink扩展RELAP5的控制与保护系统的模拟功能。为了验证两程序耦合方法的准确性，将用Simulink实现的控制与保护系统的仿真结果，与已通过验证的RELAP5实现的具有相同功能的控制与保护系统的仿真结果进行对比，结果表明二者符合较好。     

反应堆热工分析程序中可视化建模技术应用

采用可视化建模技术及可扩展标记语言(XML),建立适用于热工水力安全分析系统程序RELAP5的组件模型库.通过调用库内组件,构造模型系统图,建立可输入热工水力参数的可视化人-机界面;将各组件和对应界面连接后,输入参数,经查错最终生成RELAP5输入卡.通过对简单管道内液体流动分析,证明本技术简化了RELAP5程序的使用.     

RELAP5作为核电站模拟器热工水力系统程序的改造

RELAP5程序由于其非实时计算、无动态输入输出功能以及计算流程难以控制等原因．不适合作为核电站模拟器的热工水力系统程序、RELAPSIM程序在RELAP5基础上经过实时计算功能改造、数据动态交互功能改造、计算流程控制功能改造后,能够完成实时热工水力计算,数据动态交互以及启动、停止、冻结、运行、快照、复位计算流程等功能,满足了作为核电站模拟器的热工水力系统程序的要求。本文主要介绍了RELAP5程序的改造方法和原理以及改造后的RELAPSIM程序测试和结果。     

RELAP5应用于液态燃料熔盐堆的扩展及验证

为将RELAP5程序应用于液态燃料熔盐堆的建模分析,需要对RELAP5的模型进行扩展。基于RELAP5原有的点堆模型和热工水力模型,新增了液态燃料熔盐堆点堆模型和相应的带有内热源的热工水力模型,并通过熔盐实验堆(MSRE)实验数据进行验证。结果表明,扩展后的RELAP5程序能够适用于液态燃料熔盐堆系统建模分析。     

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。     

基于核电站完整二回路的数学模型

采用RELAP5程序对岭澳二期1000 MW核电站进行了核岛和常规岛的热工水力系统以及控制系统建模,分析了汽轮发电机负荷从100%FP瞬时阶跃下降到90%FP后再从90%FP瞬时阶跃上升到100%FP瞬态运行工况。通过对瞬态计算结果进行分析,显示了完整闭合的二回路模型在分析核电站负荷阶跃瞬变过程的优越性。     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Dimensionally adaptive dynamic switching and adjoint sensitivity analysis: new features of the RELAP5/PANBOX/COBRA code system for reactor safety transients

This paper highlights two novel features that have been implemented into the coupled RELAP5/PANBOX2/COBRA3 (R/P/C) code system. On the one hand, the R/P/C code system has been extended to include a dimensionally adaptive algorithm that uses the underlying physical phenomena to switch dynamically between three-dimensional (3D), one-dimensional (1D), and point-kinetics models, thereby reducing computational times up to a factor of five while preserving the accuracy, within user-defined error criteria, of the 3D reference calculation. On the other hand, the R/P/C system has also been extended to include the Adjoint Sensitivity Analysis Procedure (ASAP) for the RELAP5/MOD3.2 two-fluid model with non-condensables, thus enabling the efficient calculation of local sensitivities of RELAP5 results to various parameters in the RELAP5 code.     

RELAP5程序与三维时空中子动力学模型的耦合以及改进研究

引入堆芯物理计算的两群三维时空中子动力学模型,对RELAP5程序的点堆中子动力学模型进行了改进,同时设计了可视化界面,可方便地实现人.机交互操作.计算结果与实际应用表明,改进后的RELAP5程序计算功能和精度得到提高,使用更加方便,在核动力装置的仿真方面有很好的应用前景.     

循环工况转换过程中的燃料棒热应变分析

利用RELAP5程序建立分析模型,结合强迫循环工况转换自然循环工况的试验数据,对转换过程中的冷却剂温度变化速率和燃料棒的热应变进行计算分析.计算结果表明:结合RELAP5程序进行燃料棒的热应变分析是合适的;自然循环转换过程中燃料棒热应变的变化剧烈,对燃料棒的机械性能产生影响;结合RELAP5瞬态分析结果,也可以分析其他工况变化过程对燃料棒的热应力冲击问题.     

Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA.     

基于RELAP5的汽轮机仿真模型研究

建立了一个能准确反映级内部非等熵过程及动态运行特性的汽轮机模型,并将其加载到RELAP5程序中,完成RELAP5汽轮机模型的改进。改进的汽轮机模型是基于级内蒸汽的流动和做功特点,充分考虑了汽轮机结构参数以及汽轮机湿蒸汽流的非平衡两相凝结而形成的凝结冲波现象的影响。通过RELAP5程序内部耦合接口的建立和输入处理子程序的修改,实现了汽轮机模型的加载。以秦山一期300 MW核电厂汽轮机部件为对象,分别利用原RELAP5汽轮机模型和改进的汽轮机模型对其进行稳态和动态的仿真计算和比较分析。结果表明,改进的汽轮机模型能更准确地反映汽轮机动态运行特性。     

Development of a new bubbly-slug interfacial friction model for RELAP5

The development of a new bubbly-slug interfacial friction model for the Pressurized Water Reactor (PWR) safety code RELAP5 is described. The model is based on a set of best-estimate void fraction correlations which cover the full range of geometries and flow conditions encountered in PWR safety analysis. By exploiting the relationship between void fraction and interfacial friction that exists for steady, fully developed flow conditions, the correlations are converted into effective interfacial friction coefficients that can be applied in the code for transient as well as steady-state conditions. Assessments against separate effects tests indicate that the new model is more accurate than the existing model in many situations, particularly rod bundle geometries, and should never be significantly less accurate. The model has been implemented in a local version of RELAP5/MOD2 and in a pre-release version of RELAP5/MOD3 at Idaho National Engineering Laboratory (INEL).     

On the use of RELAP5-3D as a subchannel analysis code

The capabilities of the RELAP5-3D code to perform subchannel analyses in sodium-cooled fuel assemblies were evaluated. The motivation was the desire to analyze fuel assemblies with traditional (solid pins) as well as non-traditional (e.g., annular pins with internal cooling, bottle-shape) geometries. Since no current subchannel codes can handle such fuel assembly designs, a new flexible RELAP5-based subchannel model was developed. It was shown that subchannel analysis of sodium-cooled fuel assemblies is indeed possible through the use of control variables in RELAP5. The subchannel model performance was then verified and validated in code-to-code and code-to-experiment analyses, respectively. First, the model was compared to the SUPERENERGY II code for solid fuel pins in a conventional hexagonal lattice. It was shown that the temperature predictions from the two codes agreed within 2% (<3.5 °C). Second, the model was applied to the Oak Ridge 19-pin test, and it was found that the measured outlet temperature distribution could be predicted with a maximum error of 8% (<7 °C). Furthermore, the use of semicircular ribs on the duct wall to flatten the temperature distribution in a traditional hexagonal assembly was explored by means of the newly developed RELAP5-3D subchannel model; the results are reported here as an example of the model capabilities.     

RELAP5大空间自然对流管束传热模型改进及验证

基于经验证的单相和两相大空间自然对流管束传热模型,对RELAP5进行了改进,使得程序具备了模拟单相和两相大空间自然对流管束传热的能力。采用改进后的系统程序RELAP5和改进前的系统程序RELAP5对试验模拟体进行了对比计算,并采用试验数据对改进后的程序进行了验证,结果表明,改进后的系统程序计算结果与试验数据吻合较好。      

CPR1000全厂断电事故瞬态特性分析

用RELAP5/MOD3.4程序对CPR1000压水堆一回路系统进行整体建模,分析全厂断电事故下一回路主要参数的瞬态热工水力特性,并将RELAP5模型计算结果与THEMIS程序的计算结果进行对比,二者符合得较好。计算结果表明：该模型可较准确地模拟CPR1000在事故下的热工水力特性。