基于FR-Sdaso的法国凤凰堆寿期末自然循环实验分析

系统分析程序是开展反应堆安全分析的重要工具之一,也可用于开展系统瞬态实验过程的分析。法国凤凰堆(Phenix)在停运之前开展的自然循环实验是钠冷快堆领域非常重要的系统瞬态实验,为研究钠冷快堆的瞬态特点提供了很好的参考。为分析此实验过程,利用自主研发的系统分析程序FR-Sdaso对凤凰堆进行建模,对其自然循环实验开展计算分析,并将主要参数的计算值与实验值进行了对比分析。结果表明,FR-Sdaso可较好地模拟此实验的瞬态过程,可用于开展钠冷快堆此类瞬态的安全分析。     

ATHLET程序的钠冷快堆应用扩展及其验证

系统分析程序是对钠冷快堆的冷却剂回路系统进行全局模拟、瞬态及事故安全分析的重要工具。本工作对德国核设施与反应堆安全机构（GRS）开发的轻水堆最佳估算系统程序ATHLET进行修改,增加了钠的物性公式和传热关系式,将其适用范围扩展到钠冷快堆。为验证修改过的ATHLET程序,对法国凤凰（Phenix）反应堆系统建模,并对其自然对流实验进行模拟,将计算结果与实验数据进行比较。结果显示,ATHLET程序的钠冷快堆应用扩展具有良好的适用性。     

池式钠冷快堆系统分析程序开发

针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序（FASYS程序）。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。     

钠冷快堆系统程序FR-Sdaso开发

为解决600 MW示范快堆（CFR600）事故分析和工况设计中的实际问题,自主开发了钠冷快堆系统程序FR-Sdaso,其建模范围包括堆芯、一回路、二回路、三回路、四回路和事故余热排出系统,主要物理模型包括点堆模型、单通道堆芯热工模型、多区钠池模型、四区蒸汽发生器模型等核岛设备或部件分析模型,汽轮机、凝汽器、给水加热器、除氧器等常规岛设备采用集总参数模型,泵、阀门、管道及控制体等采用通用模型。对程序进行了初步验证,结果表明,FR-Sdaso程序可用于分析全厂瞬态工况及超功率、失流、失热阱等典型事故过程。目前,FR-Sdaso程序已用于CFR600的设计和安全分析。     

钠冷快堆瞬态热工水力及安全分析程序开发

钠冷快堆是第4代核反应堆的主力堆型,瞬态热工水力及安全特性是其设计研发和安全评审的重要工作,需要专用的分析工具。本文基于模块化建模思想,建立了钠冷快堆系统关键部件的热工水力模型和辅助模型,采用具有高稳定性和自动变步长能力的Gear算法,开发了钠冷快堆瞬态热工水力及安全分析软件THACS,并通过了国际基准题EBR-Ⅱ的有保护失流事故实验SHRT-17的初步验证。结果表明,THACS程序能较好模拟此实验的瞬态过程,具备钠冷快堆瞬态热工水力及安全分析的能力,可为我国钠冷快堆研发提供支持。     

钠冷快堆瞬态热工水力及安全分析程序开发

钠冷快堆是第4代核反应堆的主力堆型,瞬态热工水力及安全特性是其设计研发和安全评审的重要工作,需要专用的分析工具。本文基于模块化建模思想,建立了钠冷快堆系统关键部件的热工水力模型和辅助模型,采用具有高稳定性和自动变步长能力的Gear算法,开发了钠冷快堆瞬态热工水力及安全分析软件THACS,并通过了国际基准题EBR-Ⅱ的有保护失流事故实验SHRT-17的初步验证。结果表明,THACS程序能较好模拟此实验的瞬态过程,具备钠冷快堆瞬态热工水力及安全分析的能力,可为我国钠冷快堆研发提供支持。     

中国实验快堆超功率瞬态分析研究

在核电厂的安全分析中,系统程序能够对多种事故瞬态进行瞬态分析。随着我国快堆技术的发展,在快堆系统程序的开发和应用等方面也做了一些工作。本文运用法国原子能委员开发的快堆系统分析程序DYN4G建立了中国实验快堆(CEFR)的系统模型,包括堆芯、一回路、二回路、汽水回路和事故余热排出系统,并计算了满功率下的稳态,与设计值进行了比对,同时完成了超功率事故的瞬态分析,并与《CEFR最终安全分析报告》中的计算进行了对比验证。计算结果表明程序能较好的模拟CEFR的稳态和超功率情况,为进一步开展CEFR的安全研究及钠冷快堆的安全分析打下了基础。     

SARAX程序系统在钠冷快堆瞬态分析中的应用

无保护事故下的瞬态分析是钠冷快堆安全分析的重要内容。基于OECD/NEA发布的MOX-3600和MET-1000基准题,本文利用SARAX程序系统对不同钠冷快堆进行了瞬态计算,分析了堆内各种反应性反馈效应,并计算了无保护失流(ULOF)事故和无保护超功率运行(UTOP)事故下燃料温度和冷却剂温度的变化。计算结果表明：SARAX程序系统在快堆瞬态分析中可给出合理的参数预测结果;ULOF事故对于钠冷快堆是更为严重的事故瞬态,会导致堆内的钠沸腾进而发生严重事故。     

SARAX程序系统在钠冷快堆瞬态分析中的应用

无保护事故下的瞬态分析是钠冷快堆安全分析的重要内容。基于OECD/NEA发布的MOX-3600和MET-1000基准题,本文利用SARAX程序系统对不同钠冷快堆进行了瞬态计算,分析了堆内各种反应性反馈效应,并计算了无保护失流(ULOF)事故和无保护超功率运行(UTOP)事故下燃料温度和冷却剂温度的变化。计算结果表明:SARAX程序系统在快堆瞬态分析中可给出合理的参数预测结果;ULOF事故对于钠冷快堆是更为严重的事故瞬态,会导致堆内的钠沸腾进而发生严重事故。     

超功率下金属燃料钠冷快堆的动态仿真

编制了计算金属燃料钠冷快堆在超功率事故下的动态过程的仿真程序ＭＦＴＯＰ，并对它对美国池式钠冷快堆ＥＢＲ－Ⅱ在启动和功率运行工况下的反应引入事故瞬态进行了大量的分析计算，所得结果与国外大型程序ＮＡＴＤＥＭＯ的相应预测结果符合良好。本程序可用于其它钠冷快堆的超功率瞬态计算。     

Development of System Code FR-Sdaso for Sodium cooled Fast Reactor

To solve actual problems in the accident analysis and working condition design of the 600 MW demenstration fast reactor (CFR600), the sodium-cooled fast reactor (SFR) system code FR-Sdaso was developed, which could be used to model the reactor core, primary system, secondary system, tertiary system, quadruple system and the decay heat removal system of the SFR. The physical models can be divided into three categories: The models for nuclear island equipment including point reactor model, single-channel core thermal model, multi-zone sodium pool model and four-zone steam generator model, etc., the lump parameter models for conventional island equipment, including turbine, condenser, feed water heater, deaerator, etc., and the general models for pump, valve, pipe and control volume. Preliminary V&V work for FR-Sdaso was conducted, and the results show that FR-Sdaso can be used to analyze the transient conditions of the whole plant and typical SFR accidents such as overpower, loss of flow, and loss of heat sink. FR-Sdaso was used in the design and safety analysis of the CFR600.     

Development of natural circulation analysis methods for a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to insure safety of a sodium cooled fast reactor (SFR) of 1500 MWe adopting a natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can be applied to safety evaluation for SFR licensing taking into account the temperature flattening effect due to buoyancy force in the core, and a three-dimensional fluid flow analysis which can evaluate thermal-hydraulics for local convection and thermal stratification in the primary system and DHRSs. The one-dimensional safety analysis method and the three-dimensional fluid flow analysis method have been validated using the test results of a water test apparatus and a sodium test loop for some typical transient events selected from the design basis events of the SFR. Finally, it has been confirmed that a good agreement between the test results and analysis results has been obtained, and reliability of each method has been demonstrated.     

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

A natural circulation evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analysis adopting the turbulence model. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.     

钠冷快堆二回路系统热工水力瞬态分析程序SELTAC的开发与应用

针对钠冷快堆二回路系统的具体结构和运行特点,对中间热交换器、直流蒸汽发生器、钠缓冲罐以及泵、管道等设备和部件建立模型,采用FORTRAN语言自主编制了二回路系统热工水力瞬态分析程序SELTAC。利用中国实验快堆的停堆试验数据对所编制程序进行了初步验证。结果表明,程序计算值与试验值趋势一致,最大相对偏差不超过4.34%,吻合程度较好。将验证后的程序与一回路系统程序耦合,分析了某600 MW钠冷快堆在主热传输系统保持排热能力时的紧急停堆工况,得到了二回路系统的瞬态特性,为大型商用快堆电站的设计提供了参考。     

铅铋自然循环热工水力分析程序开发

铅铋堆内冷却剂的自然循环对于反应堆的正常运行以及事故工况下的堆芯热量导出均至关重要,相关热工水力分析工作对于支持设计及安审均有重要意义。通过对铅铋堆内一回路系统内主要部件,包括堆芯、热交换器、管道等建立热工水力物理模型,开发了适用于铅铋自然循环瞬态过程模拟的热工水力分析程序,并利用铅铋自然循环回路内开展的自然循环启动实验、功率台阶影响实验等的结果进行了程序的初步验证。结果表明,程序计算得到的结果与实验结果符合较好,能够较好模拟铅铋自然循环的瞬态过程。该程序可以为铅铋堆研发过程中自然循环热工水力分析工作提供支持。     

Experimental validation of the TASS/SMR code for an integral type pressurized water reactor

The transient and setpoint simulation small and medium reactor (TASS/SMR) code has been applied to perform the safety analysis and performance evaluation of an integral type pressurized water reactor. Till now, the code has only been verified by using simplified and analytical problems as well as a reliable system code due to the lack of available experimental data. Recently, several kinds of experiments have been performed by focusing on an identification of the heat transfer characteristics at a heat sink and source, and the thermal hydraulic characteristics and the natural circulation performance in an integral effect test facility. In this paper, the TASS/SMR code has been validated by using the experimental data obtained from a separate effect test facility by focusing on the heat transfer characteristics and an integral effect test facility by focusing on the thermal hydraulic characteristics and the natural circulation performance. According to the validation results of the TASS/SMR code against the separate effect test and the integral effect test, the code predicts the overall variation of the thermal hydraulic parameters well, including the system pressure, fluid temperature, mass flow rate, etc., and it is applicable for the safety analysis and performance evaluation of an integral type pressurized water reactor.     

基于RELAP5 MOD3.2的钠冷快堆热工水力系统分析程序开发及验证

对大型核反应堆热工水力分析程序RELAP5 MOD3.2进行了改造,使之适用于钠冷快堆系统安全分析。在不影响原程序功能的基础上添加了气液两相钠物性和液态金属对流换热模型,并改造了相应的初始化模块和计算模块。改造后的程序可正确模拟钠的流体力学特性和热物性,搭建钠冷快堆热工水力流体网络进行分析计算。对EBR-Ⅱ试验堆基准题进行了稳态模拟和失流事故分析,其中稳态计算主要参数与实验值相对偏差小于1%,瞬态计算相对偏差小于10%,各参数变化趋势与实验值相符良好,初步验证了改造程序的可靠性。     

DHR-200池式低温供热堆SBO-ATWS事故及自然循环能力分析

DHR-200池式低温供热堆(简称DHR-200池式堆)设计有自然循环瓣阀,为检验其安全性,选取典型的全厂断电叠加紧急停堆系统失效(SBO-ATWS)事故,使用RELAP5程序对其热工水力参数瞬态特性及其自然循环能力进行分析。结果表明,DHR-200池式堆具有很好的负温度反应性反馈效应,即SBO-ATWS事故后,由于燃料和冷却剂温度升高,引入负反应性,可使反应堆实现热停堆;事故后,通过非能动方式开启自然循环瓣阀,可建立稳定的自然循环,将堆芯衰变热导出至堆水池内,验证了DHR-200池式堆的固有安全性。