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。     

RELAP5与CFX程序耦合研究

以RELAP5与CFX程序为基础,利用并行虚拟机技术和CFX用户函数进行编程,开发了RELAP5/CFX耦合程序。在单相范围内,首先利用水平圆管喷放问题验证了程序间耦合的正确性。然后,针对双T型接管混合实验进行了模拟,相对于单独的RELAP5程序,耦合程序能更好地揭示真实的物理现象。通过后续的开发完善,耦合程序可用于反应堆安全分析中存在着显著三维混合现象的问题。     

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。     

可视化交互式核动力系统模拟平台VITARS的开发与验证

针对核反应堆系统热工分析软件RELAP5存在的人机交互不友好的问题,本文开发了可视化交互式核动力系统模拟平台VITARS。VITARS在可视化用户操作界面的基础上,满足RELAP5模块化建模、计算运行控制、计算结果实时显示和多RELAP5耦合计算的功能需求,还具备一个反应堆控制逻辑计算系统。通过VITARS用户可在计算过程中对阀门开度、主泵转速等参数进行手动调节。另外,得益于普适性的数据交互接口,VITARS还可作为一个交互平台,具有与其他热工水力程序耦合的可行性。最后通过对岭澳核电站机组的稳态模拟验证了VITARS的计算准确性。     

RELAP5/FLUENT耦合程序的开发

热工水力数值模拟是反应堆系统设计和安全分析的重要内容,以RELAP5为代表的系统程序可对瞬态或事故工况进行快速分析,同时以FLUENT为代表的计算流体动力学(CFD)程序对堆芯局部三维现象的分析也越来越重要。为综合利用两者的优点,以RELAP5/FLUENT为基础,利用对RELAP5程序源代码的二次开发和FLUENT的用户自定义函数(UDF)进行编程,开发了RELAP5/FLUENT耦合程序。利用flibe熔盐在水平圆管流动问题验证了程序耦合的正确性;针对2 MW熔盐堆进行了稳态模拟,耦合程序能详细分析熔盐堆的热工水力行为;模拟了2 MW熔盐堆功率突变的瞬态热工水力行为,相对于单独的RELAP5,耦合程序能更好地揭示熔盐堆系统和堆芯的三维物理现象。该耦合程序可用于解决熔盐堆热工水力分析中存在的显著三维混合现象的问题。     

RELAP5堆芯功率计算模型的扩展

为了更好地将反应堆热工水力最佳估算程序RELAP5应用于分析控制棒控制的反应堆堆芯的功率瞬变过程,堆芯功率计算模块除保留原程序中使用的点堆中子动力学模型外,还必须向轴向一维中子动力学模型进行扩展。本文通过在现有轴向一维物理程序基础上进行改造和开发,实现了RELAP5程序与一维物理程序的耦合,并且通过例题验证了耦合的正确性。     

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。     

用OECD/NEA堆芯瞬态基准题验证RELAP5-TDNK程序系统

利用经济发展与合作组织核能机构(OECD/NEA)压水堆堆芯弹棒瞬态基准题对RELAP5-TDNK进行了验证.使用RELAP5-TDNK建立了弹棒基准题模型,分析了两种弹棒问题,对程序的数据交换能力、耦合方法和瞬态事故分析能力进行了检验.与国际上多种程序进行比较,结果表明:RELAP5-TDNK程序模拟结果较好,能够分析事故或瞬态过程中堆芯内局部功率和热工参数的相互作用,具有分析强反馈现象的能力.     

核电站控制保护及辅助系统仿真方法研究

分析了大亚湾核电站控制保护及辅助系统仿真模型,采用面向对象编程方法以自编软件形式对大亚湾核电站模型进行模拟,并在开发过程中采用UML统一建模语言进行分析设计,实现了对该大亚湾核电站控制与保护系统的动态仿真.瞬态工况测试结果显示该仿真能较好地模拟反应堆一、二回路的控制与保护功能.本仿真程序具有通用、便携以及廉价等优点.     

AP1000冷却剂强迫流动全部丧失事故瞬态特性分析

AP1000反应堆是目前国际上典型的"三代"核电厂,利用RELAP5/MOD 3.3程序对AP1000核电厂一回路系统进行整体建模,分析冷却剂强迫流动全部丧失事故下一回路主要热工水力参数的瞬态特性,并与COAST和LOFTRAN程序的计算结果进行了对比,发现两者具有相同的分布规律,表明利用RELAP5程序建立的计算模型可以准确模拟AP1000冷却剂强迫流动全部丧失事故下的热工水力特性。     

Development and application of an extensible engineering simulator for NPP DCS closed-loop test

LinAo Nuclear Power Plant (NPP) Phase II is a newly-built CPR1000 reactor in China, and many new technologies including the incorporation of digital control system (DCS) substituting traditional analog control systems have been applied. This is the first time for Chinese engineers to setup and adjust the DCS configurations. Both the lack of the operating experiences and the plant safety requirements from the government make a necessity of the closed-loop DCS test before commercial plant operation. The most practical way is to build a digital plant as the controlled target and this digital plant is used to provide the plant thermal–hydraulic parameters and feedbacks for the DCS. Though the RELAP5 code has been developed for the best-estimate transient simulation of light water reactor coolant systems and is used worldwide, its functionality is too limited to implement a digital plant, such as the simulation of the complicated plant control and protection systems, the 3-dimensional neutron kinetics and the fluid network for the plant auxiliary systems. To overcome these drawbacks, a RELAP5-based extensible simulator has been built to satisfy the new requirements for the implementation of a digital plant. Any simulation code of desired functionality can be integrated into this simulator as a simulation module once it applies a set of well-defined data exchange interfaces. At the present stage, a RELAP5 module, a control system modeling module, a software–hardware data bridge module and some other auxiliary modules have been integrated into the simulator. There are more than 60 systems that need to be tested with the DCS in LinAo Phase II, and the whole testing work is separated into several phases. In this paper, we take the testing of the pressure control system and water level control system of pressurizer as example. A typical transient of 10% load step change from 100%FP (full power) to 90%FP was performed for the closed-loop DCS test. The necessity and capability of this RELAP5-based engineering simulator has been demonstrated.     

核电厂控制与保护系统动态仿真

分析了CHASHIMA核电站的测量系统、控制与保护模型、系统设备及设备失效模型、辅助系统管网模型。然后,基于C语言编制了控制与保护系统动态仿真程序模块PROSYS．并将其用于在工程模拟器,在模拟器上实现了CHASHIMA核电站控制与保护系统的动态仿真该工程模拟器已应用于核电站安全分析,以及为核电站先进主控室设计提供软件支持和验证服务：实际应用结果显示,该仿真软件能较好地模拟反应堆一、二回路的控制与保护功能。     

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).     

Development of nuclear power plant real-time engineering simulator

A nuclear power plant real-time engineering simulator was developed based on general-purpose thermal-hydraulic system simulation code RELAPS. It mainly consists of three parts: improved thermal-hydraulic system simulation code RELAP5, control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power, was simulated by the engineering simulator as an application example. This paper presents structure and main features of the engineering simulator, and application results are shown and discussed.     

AP1000核电厂二回路主管道双端断裂流体喷射力计算分析

AP1000是先进的第三代压水堆核电厂,为确保核电厂在事故工况下的安全性,需对二回路主管道发生双端断裂的工况进行研究。本文采用RELAP5/MOD3.4软件对核电厂二回路突发主管道双端断裂的事故工况进行了数值模拟,计算得到断裂后管道破口处的喷放流量、压强、空泡份额及喷射力等物理参数的变化特性,并将计算结果与ANSI 58.2简化计算方法的结果进行了比较分析。结果表明,RELAP5/MOD3.4计算所得的喷射力小于简化计算方法所得结果。本文分析结果为进行AP1000核电厂的破裂管道甩击防护提供了基础。     

PSB-VVER simulation of Kozloduy NPP “loss of feed water transient”

This paper provides a comparison between the PSB test facility experimental results obtained during the simulation of loss of feed water transient (LOFW) and the calculation results received by INRNE computer model of the same test facility. Integral thermal-hydraulic PSB-VVER test facility located at Electrogorsk Research and Engineering Center on NPPs Safety (EREC) was put in operation in 1998. The structure of the test facility allows experimental studies under steady state, transient and accident conditions.RELAP5/MOD3.2 computer code has been used to simulate the loss of feed water transient in a PSB-VVER model. This model was developed at the Institute for Nuclear Research and Nuclear Energy for simulation of loss of feed water transient.The objective of the experiment “loss of feed water”, which has been performed at PSB-VVER test facility is simulation of Kozloduy NPP LOFW transient. One of the main requirements to the experiment scenario has been to reproduce all main events and phenomena that occurred in Kozloduy NPP during the LOFW transient. Analyzing the PSB-VVER test with a RELAP5/MOD3.2 computer code as a standard problem allows investigating the phenomena included in the VVER code validation matrix as “integral system effects” and ”natural circulation“. For assessment of the RELAP5 capability to predict the “Integral system effect” phenomenon the following RELAP5 quantities are compared with external trends: the primary pressure and the hot and cold leg temperatures. In order to assess the RELAP5 capability to predict the “Natural circulation” phenomenon the hot and cold leg temperatures behavior have been investigated.This report was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory (ANL), under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

核电厂DCS系统功能验证工程模拟机研究

建立了核电厂分布式控制系统(DCS)功能验证工程模拟机系统.该系统采用RELAP5建立热工水力模型,利用MATLAB/Simulink建立电厂主要控制系统数学模型,利用MYSQL建立数据库,利用VisualStudio.NET开发了系统控制台;采用数据采集系统实现工程模拟机与现场DCS系统间的实时信号通讯,从而实现对DCS系统的功能验证.验证结果表明,系统能实现实时运行,并满足DCS系统硬件和逻辑功能测试的要求.     

Benchmark analysis of main circulation pump trip events at the Ignalina NPP using RELAP5 code

The state-of-the-art code RELAP5/MOD3 was originally designed for PWRs. Because of unique RBMK designs the application of this code to RBMK-1500 encountered several problems. A successful best estimate RELAP5 model of the Ignalina NPP has been developed. This model includes the reactor main circulation circuit (MCC) and reactor control and protection system required for this kind of transient analysis. Benchmark analysis of all operating main circulation pump (MCP) trip events was performed. During the analysis the characteristics of isolation control valves and MCP throttling regulating valves were established. Comparison of calculated and measured parameters was also used to establish realistic resistances of different MCC components and realistic behaviour of the controllers of the reactor systems. Calculations performed with the RELAP5 model, which includes these modifications, compare favourably with plant data.