排放载荷分析方法合理性论证及关键影响因素研究

在自主化三代核电厂的设计中,为了满足超压保护功能,一般会使用较为先进的先导式安全阀。先导式安全阀较以前M310电厂的阀门,具有开启时间更短等特点。三代核电厂的稳压器安全阀如果在某些情况下开启,其上游水封将会对下游排放管道造成非常大的冲击力。因此,排放载荷分析是先进的三代电厂的稳压器排放管线设计中的一个关键性技术,只有掌握了排放载荷分析方法,才能为应力计算提供可靠参考,进而对管线设计、管道支撑布置的合理性进行论证。使用RELAP5程序进行热工水力计算,对输出信息进行合理的处理可获得载荷计算结果。本文将对该分析方法的合理性进行验证,并且在此基础上,选取对载荷影响非常关键的因素,包括:阀门开启时间、水封温度、水封体积等进行研究,为载荷优化提供合理性建议。     

三门核电厂稳压器安全阀误开启事故研究

稳压器安全阀用于核电站一回路系统和设备的超压保护,如果发生故障卡开,将造成冷却剂丧失事故(LOCA)。本文使用机理性分析程序对三门核电厂1号机组进行建模,并对稳压器安全阀误开启导致的LOCA事故进行模拟分析,研究在稳压器水位较高的情况下,非能动安全设施对LOCA事故的响应情况。之后,为验证三门核电站对类似三哩岛事故的应对能力,假设丧失给水叠加稳压器安全阀卡开事故并进行相应事故分析。通过以上两个事故的分析表明,三门核电厂的非能动安全设计能够应对稳压器安全阀故障造成的LOCA事故,保证对一回路补水,不会造成非常严重的事故后果。     

两相排放载荷分析方法研究

建立了一套两相排放载荷计算方法,并以典型压水堆核电厂排放管线为例,计算当有水封存在时稳压器安全阀下游管道所受的载荷。计算结果与法玛通公司提供的参考数据基本一致,表明本文建立的两相排放载荷计算方法是正确的。     

AP1000自动降压系统误开启事故仿真分析

针对AP1000的具体结构和运行特点,采用FORTRAN程序设计语言,开发了AP1000瞬态热工水力计算程序RETAC。利用RETAC对AP1000自动降压系统(ADS)误开启事故进行仿真分析,得到稳压器压力、堆芯归一化热功率、堆芯归一化流量、堆芯平均温度、燃料中心最高温度和最小偏离核态沸腾比（MDNBR）等主要系统参数的响应特性。分析结果表明,在稳压器低压停堆保护的作用下,燃料中心最高温度和MDNBR未超出规定限值,满足安全准则要求。并将计算结果与美国西屋公司AP1000分析软件LOFTRAN的计算结果进行对比,对比趋势符合良好,证明了RETAC建模和自动降压系统临界流模型计算的合理性。     

AP1000稳压器安全阀动态排放特性的CFD仿真分析

AP1000核电厂的稳压器顶部安装有两台弹簧式安全阀,用于提供超压保护。由于安全阀的口径较大,其排放能力及排放特性无法通过工业试验台架来验证。本文建立了安全阀的CFX动网格,并在此基础上进行动网格二次开发,编写外部命令,根据开启时间轴,实时替换计算网格,以模拟安全阀开启过程中截面的不断变化,进而分析安全阀的动态特性,为安全阀的排放特性评估提供了一定的依据。     

主泵两相降级对大破口失水事故的影响研究

大破口失水事故过程中,主泵的工作范围覆盖了单相液、气液两相和单相气工况。在两相工况下,主泵的扬程和转矩发生降级。对于AP1000核电厂,WCOBRA/TRAC被用于大破口失水事故分析,其现有的主泵两相降级数据来源于西屋W93A主泵。为正确模拟AP1000主泵在大破口失水事故过程中的热工水力特性,需对其两相降级特性进行研究。本研究分别采用国际上广泛使用的SEMISCALE和EPRI/CE主泵的两相降级数据进行AP1000冷段双端断裂事故的计算分析,并与原有W93A的计算结果进行对比。结果表明,AP1000主泵两相降级特性对反应堆冷却剂系统压力、破口流量和安注箱流量影响不大。相比于SEMISCALE和EPRI/CE,现有的W93A的两相降级数据将导致更低的堆芯冷却流量和更高的包壳峰值温度最大值,计算结果相对偏于保守。     

核电厂稳压器水封结构数值模拟研究

建立了含不凝性气体的气-液两相流三维计算流体力学(CFD)模型,运用Fluent软件对核电厂稳压器新型水封结构流场进行模拟,通过水密封建立过程中稳压器压力和不凝性气体含量的影响分析,研究了水密封建立过程的热工特性。结果表明,对于新型水封结构,水密封建立时间随稳压器压力的增大而缩短,随不凝性气体含量的增大而增长。     

AP1000非能动余热排出系统建模与瞬态数值分析

针对AP1000非能动余热排出系统(PRHRS)的具体结构,采用FORTRAN程序设计语言自主开发了瞬态分析程序RETAC-PRHRS(REactorTransientAnalysisCode-PassiveResidualHeatRemovalSystem)。利用编制的程序对PRHRS误开启事故进行分析,得到了堆芯归一化热功率、流量、最小偏离核态沸腾比(MDNBR)、系统压力、PRHRS流量等主要系统参数的响应特性。分析结果表明,在PRHRS误开启事故发生时,主要系统参数未超出规定限值,不会触发反应堆停堆。并将计算结果与热工水力分析软件,包括西屋公司开发的LOFTRAN及GSE公司开发的Topmeret/THEATRe的计算结果进行对比。对比趋势符合良好,从而证明了AP1000PRHRS建模的合理性。     

TRANTH软件基于稳压器安全阀组流量试验的确认与评估

热工水力瞬态分析软件TRANTH用于分析核电厂安全性,其中,考虑了两区质量守恒和能量守恒的关键模型之一稳压器模型可对稳压器安全阀、释放阀、电加热器、喷淋和相关系统进行模拟。在软件开发完成后需进行相关软件验证,故结合方家山核电厂1号机组稳压器安全阀流量试验数据和软件模拟结果,验证稳压器模型。结果表明,模拟计算结果与现场试验数据符合度高,模型精度满足工程设计要求。      

主泵参数变化对失水事故后果影响分析

基于100D主泵和ANDRITZ主泵的差异,分析主泵相似特性曲线和自由容积的变化对失水事故(LOCA)后果的影响。针对岭澳核电站二期反应堆冷却剂系统,应用CATHARE GB程序和CONPATE4程序分析大破口LOCA事故堆芯热工水力后果;应用ATHIS和FORCET程序分析失水事故喷放阶段的反应堆冷却剂主管道水力载荷。结果表明,主泵相似特性曲线的变化对大LOCA事故再淹没阶段的堆芯热工特性影响很大,采用不同主泵时的最高峰值包壳温度(PCT)相差很大;而主泵自由容积对失水事故喷放阶段的卸压波传递影响较大,导致采用不同主泵时的反应堆冷却剂主管道水力载荷有所不同。     

非凝性气体对汽-气稳压器稳压的影响研究

通过分析相间的传热传质过程以及非凝性气体存在时壁面蒸汽冷凝过程,建立了汽 气稳压器模型,研究了非凝性气体对稳压过程的影响,描述了稳压器的稳压特性,并将模型计算结果与MIT稳压器实验数据进行了对比。结果表明：当不含非凝性气体时,计算精度高,相对偏差在0.8%内,压力峰值为0.647 MPa;当非凝性气体含量从0增至20%时,计算精度相对减小,最高相对偏差为15.4%;压力峰值从0.647 MPa增至1.02 MPa。研究表明非凝性气体对稳压器稳压过程具有重要影响作用,随着非凝性气体的种类和含量的变化,稳压器内稳压过程发生显著变化。     

Unsteady piping forces caused by hot water discharge from suddenly opened safety/relief valves

Various fluid transients in nuclear pressure vessels create the possibility of liquid discharge from safety/relief valves designed for steam release. Piping systems which carry the flow from safety/relief valves to an appropriate discharge environment are designed to withstand the unsteady reaction forces associated with steam flows. If liquid discharge occurs, the piping system may be subject to different unsteady reaction forces. This study provides a means of estimating discharge rates for saturated and subcooled water and resulting pipe forces, for the limiting case of a suddenly opened valve. It was found that pipe forces caused by the moving shock are about the same for either steam or water discharge. Pipe forces caused by a moving steam-water/air interface during water discharge may exceed those forces caused by the steam-air interface resulting from steam discharge. It was determined that water discharge would result in a longer opening time for a piston-type valve. It is therefore expected that with realistic estimates of valve opening time, the magnitude of water discharge forces will be approximately the same as steam discharge piping forces.     

Safety Relief Valve Performance for Two-Phase Flow

The performance of main steam safety relief valve has been evaluated with respect only to the steam. In the present study, two-phase flow and subcooled water blow-out tests with model valves were performed in order to evaluate the valve's characteristics and performance. From the test results, it was made clear that not only for the steam but also for the two-phase flow the measurement data were hardly affected by scaling and also that the reaction force of the fluid to the valve stem was hardly dependent upon the void fraction. Analytical study was performed using the two-phase flow model in the valve. The results of the analysis showed good agreement with the test data. It was shown from the test and analysis results that the reaction force of the two-phase flow and subcooled water to the valve stem was almost as much as that of the steam flow, and the integrity of the safety relief valve could be maintained.     

小型堆稳压器波动管双端破口试验研究

开展了模块化小堆稳压器波动管双端破口试验研究,获得了非能动安全系统的事故响应特性和一回路系统参数变化。试验研究结果表明,在稳压器波动管双端破口极端工况条件下,中压安注箱能在短时间内提供较大的稳定安注流量,及时补充系统水装量;高压安注系统运行过程比较复杂,安注流量与堆芯补水箱压力平衡管线内介质状态和中压安注系统运行状态密切相关,在1.7 h内呈间歇注入运行状态。在整个事故过程中,堆芯一直处于淹没状态,模块化小堆非能动安全系统能够确保稳压器波动管在双端破口极端工况条件下的堆芯安全。      

Target Curve with Maintenance Time in Optimal Xenon Shutdown Problems

A series of sparger tests have been conducted to investigate the performance of a steam sparger, which will be used in a Korean Advanced Power Reactor, APR1400. The tests have been conducted at the Blowdown and Condensation Loop in the Korea Atomic Energy Research Institute using a prototypic steam sparger. The major test parameters include amount of air mass and air temperature in the discharge line, valve opening time, steam mass flow rate, and water temperature and level in the In-Containment Refueling Water Storage Tank. The hydrodynamic loads induced by the air clearing phenomenon during typical operating conditions seem to be dependent on the valve opening time, steam mass flow rate, submergence of a sparger, and distance between the sparger and the structure. The effect of the amount of air mass in the discharge line and the water temperature to the peak load seems to be negligible for a given range of parameters. The hydrodynamic load induced by the prototypic steam sparger was less than those expected in a BWR and the steam sparger tested in this program can be used satisfactorily for the APR1400.     

全厂断电事故下AP1000非能动余热排出系统分析

利用RELAP5/MOD3.3程序对AP1000反应堆一回路及非能动系统进行建模计算,给出了AP1000非能动余热排出系统（PRHRS）在全厂断电事故下的瞬态响应特性。计算结果表明：情况1,PHRH系统由蒸汽发生器低水位与低启动给水流量符合信号启动,稳压器安全阀的开启导致PRHRS发生倒流现象,并会引起堆芯冷却剂过热沸腾、压力容器进出口温差过大等后果;情况2,由断电信号直接触发PRHRS,触发前安全阀不开启,此时PRHRS正常运行。     

二次侧非能动余热排出系统启动特性试验研究

二次侧非能动余热排出（ASP）系统是国内二代加型百万千瓦级压水堆核电厂应对全厂断电事故的重要改进项。为获取ASP系统的启动特性,基于比例模化方法设计建造了ASP系统试验装置。试验结果获取了不同因素对ASP系统启动特性的影响。结果表明：蒸汽发生器二次侧水装量与ASP系统隔离阀动作时间对ASP系统的启动特性影响较小;ASP系统的流量随蒸汽管线与回水管线阻力系数的增大而降低;蒸汽释放阀（VDA）的往复开启引起自然循环流量的波动,当VDA关闭后自然循环可恢复至稳定状态;换热管内初装水的水量影响ASP系统初始流量峰值;所有试验工况中均建立了稳定的自然循环。     

Experimental investigation in the TOSQAN facility of heat and mass transfers in a spray for containment application

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear-pressurized water reactor (PWR) containment. The TOSQAN facility which is highly instrumented with non-intrusive optical diagnostics is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we perform detailed characterization of the two-phase flow.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

瞬态条件下稳压器水位测量特性试验研究

稳压器是压水堆核动力装置压力安全系统的主要设备，其水位波动反映了一回路系统的水容积变化情况，是稳压器运行控制的关键参数之一。本文基于双区非平衡模型模拟蒸汽泄露条件下的稳压器水位变化，并针对稳压器蒸汽泄漏工况开展了水位测量特性试验研究，研究了2.6~7.8 kPa/s压降速率工况下，稳压器内水位测量压差的变化情况。研究发现:采用压差修正液相区密度计算的水位值在压力瞬变情况下有较好的跟随性，能够更好的反应水位特性；表征稳压器内液相区密度变化的压差在压力减小的过程中，过渡时间小于40 s，且过渡时间与压变速率单因素无强相关性。这为稳压器的安全运行控制提供了基础研究数据。