重水堆停堆工况下单相自然循环热阱有效性分析

对重水堆核电厂停堆冷却剂丧失强迫循环后,单相自然循环热阱的有效性进行了计算分析。通过分析发现,每环路内一台或两台蒸汽发生器可用时,主热传输系统都可以建立稳定的自然循环,排出堆芯热量。一台蒸汽发生器可用时,两燃料通道内包壳由于冷却条件的不同有温差存在。在同一堆芯衰变功率水平下,主系统内自然循环流量受环路内可用蒸汽发生器数量影响较小。     

先进PWR堆腔淹没情况下的自然循环研究

在先进压水堆核电厂中,堆腔注水是重要的严重事故管理措施之一。采用RELAP5程序对堆腔淹没情况下反应堆压力容器(RPV)外壁与保温层间隙内的自然循环进行研究。通过对韩国APR1400堆腔注水系统实验的模拟,证明RELAP5程序模拟此类自然循环的适用性。对美国AP1000堆腔自然循环的分析表明,保温层间隙内建立的自然循环有足够的热量排出能力,可以有效防止RPV下封头的热失效。敏感性分析结果表明,冷却水入口面积越大,自然循环的流量越大;为了使产生的蒸汽顺利排出以建立稳定的自然循环,汽-水出口面积应足够大,且汽-水出口应布置在不会被堆腔水淹没的位置;在堆腔注水措施的实施中,应保证堆腔被淹没到足够高的位置以提供足够的循环静压驱动力。     

摇摆条件下海上浮动堆全厂断电事故分析

为研究海洋条件对海上浮动堆全厂断电事故后的事故进程及非能动安全系统运行特性的影响,通过建立海洋条件加速度场模型,基于RELAP5程序开发获得了适用于海上浮动堆的系统分析程序,并对程序进行了实验验证。利用所开发的程序通过建立双环路海上浮动堆及二次侧非能动余热排出系统的计算模型,开展了不同摇摆运动参数下海上浮动堆全厂断电事故的计算分析。计算结果表明,船体的横摇运动可加快全厂断电事故后浮动堆系统压力和温度的下降速度,堆芯余热能够被二次侧非能动余热排出系统有效导出;但横摇运动会造成事故后堆芯自然循环流量的显著降低,引起一回路系统和非能动余热排出系统中自然循环流量的大幅度振荡及周期性倒流。本文计算结果可为海上浮动堆非能动安全系统的设计提供参考。     

某天然铀重水研究堆改造设计方案研究

某重水研究堆设计采用天然铀作燃料,具有生产武器级钚的能力。为去除武器级钚的产出能力,本文对该堆的改造设计方案进行了研究。采用CITATION+WIMSD-4程序、MCNP程序和RELAP5/MOD3.2程序进行物理、热工计算,验证了将该堆的燃料由天然铀改为稍加浓缩铀的可行性。利用ORIGEN2程序计算分析表明,改造后的重水堆已不具备武器级钚生产堆的特性,同时民用放射性同位素辐照生产能力和科学研究技术指标亦有所提高。     

超临界水堆系统分析程序的改进

针对超临界水堆特殊的水物性参数和独立的慢化剂通道设计,对堆芯计算程序PARCS和热工水力程序RELAP5进行了适应性改造。使用改造后的耦合程序PARCS/RELAP5分析了美国超临界水冷参考堆,发现了慢化剂逆向流动和最高功率组件不同于最高外表面包层温度组件的现象,根据这些经验,对中国的超临界水堆分析程序的改进和研发提出了相关意见。     

核动力装置自然循环及其过渡过程计算模型的建立

为准确分析含反应性反馈的核动力装置自然循环及其过渡过程中重要参数的响应特性,以核动力装置瞬态最佳估算程序RELAP5/MOD3为基础,采用两群三维时空中子动力学模型替代RELAP5/MOD3的点堆模型,并建立三维空间内中子物理与热工水力的耦合模型,编制相应的计算程序。利用所研制的程序对实际核动力装置的自然循环及其过渡过程进行分析计算,并与试验结果进行比较。结果表明:本文建立的时空中子动力学计算模型克服了点堆方程不能准确计算反应性反馈的缺点,计算精度高,研制的程序可作为核动力装置强迫循环与自然循环及其过渡过程的计算分析工具。     

基于RELAP5和CTF的西安脉冲堆多尺度耦合瞬态特性研究

基于系统程序RELAP5和子通道程序CTF,建立了适合于自然循环冷却池式脉冲堆的区域分解多尺度耦合方法和满足不同尺度参数间精确转换的网格映射方法。通过脉冲堆稳态和瞬态工况分析,验证耦合方法正确性和可行性,计算结果表明：所建立的多尺度耦合方法能够应用于自然循环冷却池式脉冲研究堆的耦合分析,其耦合计算结果在准确性方面优于单一系统程序计算结果;2 MW稳态工况下的燃料芯块温度准确性提高约62.50%,脉冲瞬态工况下燃料芯块温度准确性提高约76.71%;同时,耦合计算能够获得高精度、高分辨率的堆芯热工水力分布和瞬态特性。     

49-2泳池堆低温供热全厂断电ATWS事故分析

针对49-2泳池式反应堆（简称49-2泳池堆）用于城市低温供热的工况,选取典型的全厂断电叠加紧急停堆系统失效(全厂断电ATWS)的超设计基准事故,使用RELAP5/MOD3.2程序对其热工水力参数瞬态特性进行分析。结果显示,49-2泳池堆具有很好的负温度反馈效应,事故后,由于燃料和冷却剂温度升高,从而引入一定的负反应性,使反应堆处于次临界状态;同时堆芯通过与堆水池建立自然循环,将衰变热带出,最终依靠自然循环方式将堆芯余热排出至上部大气环境热阱,验证了49-2泳池堆用于城市低温供热的固有安全性。     

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。     

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。      

Capability of the RELAP5 code to simulate natural circulation behavior in test facilities

Many advanced reactor designs incorporate passive systems mainly to enhance the operational safety and possible elimination of severe accident condition. Some reactors are even designed to remove the nominal fission heat passively by natural circulation without using mechanical pumps e.g. ESBWR, AHWR, CHTR, CAREM, etc. while in most other new reactor concepts, the decay heat is removed passively by natural circulation following the pump trip conditions. The design and safety analysis of these reactors are carried out using the best estimate codes such as RELAP5, TRAC and CATHARE, etc. These best estimate codes have been developed for pumped circulation systems and it is not proven about their adequacy or applicability for natural circulation systems wherein the driving mechanism is completely different. Some of the key phenomena which are difficult to model but are significantly important to assess the natural circulation system performances are – low flow natural circulation mainly because the flow is not fully developed and can be multi-dimensional in nature; flow instabilities; critical heat flux under oscillatory condition; flow stratification particularly in large diameter vessel; thermal stratification in large pools; effect of non-condensable gases on condensation, etc. Though, these best estimate codes use a six equation two-fluid model formulation for the thermal-hydraulic calculation which is considered to be the best representative of two-phase flows, but their accuracies depend on the accuracies of the models for interfacial relationships for mass, energy and momentum transfer which are semi-empirical in nature. The other problem with two-fluid models is the effect of ill-posedness which may cause numerical instability. Besides, the numerical diffusion associated due to truncation of higher order terms can affect the prediction of flow instabilities. All these effects may lead to inability to capture the important physical instability in natural circulation systems and instability characteristics i.e. amplitude and frequency of flow oscillation. In view of this, it is essential to test the capability of these codes to simulate natural circulation behavior under single and two-phase flow conditions before applying them to the future reactor concepts.In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized.     

非能动余热排出系统数学模型研究与运行特性分析

利用某型核动力装置非能动余热排出系统1:10原理性试验的8个稳态工况、6个启动工况的试验数据,验证RELAP5/MOD3.2程序对本类型非能动余热排出系统的适用性。结果表明:垂直管内蒸汽凝结换热系数对两相流自然循环的流动与传热影响大;RELAP5/MOD3.2程序过低估算了垂直管内蒸汽流速对蒸汽凝结换热系数的影响,计算结果与试验结果偏差大。对RELAP5/MOD3.2程序垂直管内的蒸汽凝结换热模型进行修正,修正后的计算结果与试验值基本吻合;采用RELAP5程序对垂直管内两相流自然循环传热进行计算,须选择热前沿跟踪模型。对非能动余热排出系统的稳态与瞬态运行特性进行分析,理论计算与试验结果均表明:稳态工况下,系统可以实现稳定的两相流自然循环,系统排热能力受蒸汽发生器水位的影响大,冷却水入口温度与系统压力的影响相对较小;系统的启动特性良好,可快速地建立环路的自然循环,带走反应堆的衰变热。     

The development and validation of a thermal hydraulic code in rolling motion

An advanced thermal hydraulic code is established on the basis of RELAP5/MOD3.3 code for the investigation of the thermal hydraulic behavior of nuclear power systems. The RELAP5 code is modified by adding a module calculating the effect of rolling motion and introducing new flow and heat transfer models. The experimental data are used to validate the theoretical models and calculation results. It is shown that the advanced flow and heat transfer models could correctly predict the frictional resistance and heat transfer coefficients in rolling motion. The thermal hydraulic code is used to simulate the operation of a natural circulation system in rolling motion. The calculation results are in good agreement with experimental data. The relative discrepancies between calculation results and experimental data are less than 5%.     

Evaluation of system codes for analyzing naturally circulating gas loop

Steady-state natural circulation data obtained in a 7 m-tall experimental loop with carbon dioxide and nitrogen are presented in this paper. The loop was originally designed to encompass operating range of a prototype gas-cooled fast reactor passive decay heat removal system, but the results and conclusions are applicable to any natural circulation loop operating in regimes having buoyancy and acceleration parameters within the ranges validated in this loop. Natural circulation steady-state data are compared to numerical predictions by two system analysis codes: GAMMA and RELAP5-3D. GAMMA is a computational tool for predicting various transients which can potentially occur in a gas-cooled reactor. The code has a capability of analyzing multi-dimensional multi-component mixtures and includes models for friction, heat transfer, chemical reaction, and multi-component molecular diffusion. Natural circulation data with two gases show that the loop operates in the deteriorated turbulent heat transfer (DTHT) regime which exhibits substantially reduced heat transfer coefficients compared to the forced turbulent flow. The GAMMA code with an original heat transfer package predicted conservative results in terms of peak wall temperature. However, the estimated peak location did not successfully match the data. Even though GAMMA's original heat transfer package included mixed-convection regime, which is a part of the DTHT regime, the results showed that the original heat transfer package could not reproduce the data with sufficient accuracy. After implementing a recently developed correlation and corresponding heat transfer regime map into GAMMA to cover the whole range of the DTHT regime, we obtained better agreement with the data. RELAP5-3D results are discussed in parallel.     

直流蒸汽发生器传热管破裂事故分析

通过对直流蒸汽发生器传热管破裂（SGTR）事故的分析,可看出RELAP5瞬态分析程序能较好地模拟一体化反应堆在SGTR事故后的事件响应序列及主要热工水力现象,例如环路的不对称效应、主回路的自然循环等。一体化反应堆在发生SGTR事故后,可通过一系列安全与保护系统的动作得到有效缓解,并最终能应用非能动余热排出系统（PRHRS）的自然循环导出堆芯余热,使反应堆处于安全状态。同时,受事故影响蒸汽发生器压力在PRHRS投入运行后会快速升高,最终与一回路压力相平衡,此后,破口处的泄漏也会终止。此外,本文还研究了破口处临界流量及其积分流量结果不确定性的影响因素,其中主要考虑了采用不同的临界流模型和破口建模方式等两个方面。     

RELAP程序用于PRS分析的适用性研究

以三代核电技术“华龙一号”二次侧非能动余热排出系统（PRS）为研究对象,对RELAP程序用于PRS分析的适用性进行论证。首先通过关键现象识别和程序功能梳理,定性说明RELAP程序相关数理模型能够满足PRS分析的需求;同时通过修改RELAP程序水平/竖直管内蒸汽冷凝相关换热模型并与PRS实验台架结果进行对比,定量评价相关模型的适用性。在上述工作基础上,使用RELAP程序分析全厂断电（SBO）工况下PRS投入运行的瞬态过程,结果表明PRS能够建立自然循环并有效排出热量;最后分析了PRS误投入瞬态工况,结果表明PRS误投入工况下反应堆是安全的,PRS的设计满足要求。      

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.     

Study of startup transients and power ramping of natural circulation boiling systems

Numerical models of a natural circulation test facility and its prototype have been developed with RELAP5/MOD3.4 code and verified for their grid independence by nodal sensitivity studies. The model of the test facility has been validated for its steady state as well as transient predictions with the help of experimental observations. The transient predictions and parametric trends obtained by the numerical model of the prototype have been compared with those of the numerical model of the test facility. Thus, the ability of RELAP5 code to predict the transients during startup of a natural circulation boiling water reactor is verified. A powering procedure for the test facility has been conceptualized with the help of its RELAP5 model and demonstrated experimentally. Based on this, a similar powering procedure for the prototype has been proposed and simulated numerically with its RELAP5 model.     

RELAP5/MOD3.2竖直管束外大容积沸腾换热模型适用性分析

以浸没在大容积水箱内的非能动余热排出换热器为研究对象,采用实验数据对RELAP5/MOD3.2程序计算竖直管束外大容积沸腾换热适用性进行校核。在充分考虑大容积水箱内流体沿管束轴向的自然对流以及径向的气液间热量交换基础上,合理建立了沸腾换热回路节点划分模型。将计算结果与实验数据进行对比,发现沸腾换热系数的计算值与实验值最大相对偏差在50%以上,且沸腾换热系数随热流密度变化的趋势明显不同。由此判断,Chen关系式并不适合计算竖直管束外大容积沸腾的情况。通过与已有的大容积沸腾换热计算关系式对比,发现Kutateladze “new”公式或Rohsenow公式计算值与实验值符合较好。     

针对REPEC加热实验的RELAP5程序模拟与分析

为研究压力容器外部流道的冷却能力及流动传热过程,在反应堆压力容器外部冷却(REPEC, Reactor Pressure vessel External Cooling)实验台架前期加热实验的基础上,采用RELAP5程序对实验工况进行模拟和对比。模拟结果与实验数据一致性较好。随加热热流、进出口面积的增加,系统内自然循环流量也增加;入口欠热度对自然循环流量的影响不是很明显;近饱和沸腾条件下,系统出现明显的两相不稳定流动。