Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

ACME台架PRHR管线破口位置敏感性试验研究

为研究先进非能动（AP）型核电厂在非能动系统失效条件下的安全性能,利用我国先进堆芯冷却机理整体试验台架（ACME）开展了非能动余热排出（PRHR）管线破口失水试验研究,分析了主要的试验进程和破口位置对事故过程各阶段关键参数的影响。结果表明,ACME PRHR管线破口试验进程与冷管段小破口失水事故（SBLOCA）进程基本一致,再现了非能动核电厂自然循环阶段、自动卸压系统（ADS）喷放阶段和安全壳内置换料水箱（IRWST）安注阶段的安全特性;在不同破口位置的试验中,非能动堆芯冷却系统（PXS）均可保证堆芯得到补水,堆芯活性区始终处于混合液位以下;破口位置对ACME LOCA事故进程、反应堆冷却剂系统（RCS）初期降压速率、PRHR热交换器（HX）流量、喷放流量、堆芯液位、IRWST安注流量等参数具有显著影响,对堆芯补水箱（CMT）和蓄压安注箱（ACC）安注流量的影响较小。      

Air-water multidimensional impingement flow on vertical flat wall

The emergency core cooling (ECC) water is supplied from the direct vessel injection (DVI) system in the Advanced Power Reactor 1400 MWe (APR1400) during a postulated large-break loss-of-coolant accident (LBLOCA). The velocity of ECC water exceeds 10 m/s in the early high pressure phase of LBLOCA and then is decreased to 2-3 m/s in the late phase of reflood. During the injection the flow behavior exhibits a complex mode involving impingement, bypass, entrainment, sweepout and condensation in the reactor downcomer. There is currently no model to accurately simulate the local and complicated flow behavior in the APR1400 downcomer during a LBLOCA. This study is aimed at developing models for the water film flow and deformation, both of which are expected to sizably affect the other multidimensional flow characteristics in the downcomer. Experimental studies are conducted to benchmark the predictive model by furnishing the boundary conditions for the analysis resorting to the Accelerated Liquid Phase Hydrodynamics Apparatus (ALPHA) and the Kinetic Aerodynamic Physics Parallelepiped Apparatus (KAPPA). The Poisson equation and potential theory are applied to formulate the behavior of the water film and air flow. In both the experimental and numerical studies, the temperature-dependent thermodynamic properties and the reactor vessel curvature are neglected to render the problem at hand tractable. The model is found to reasonably describe the downward film flow behavior. The water film is developed in proportion to the initial injection velocity of the ECC water. The downward velocity of water film is increased with the heights of injection. Regarding the film deformation the calculated results tend to deviate from the experimental data as the injected air velocity is increased. The disagreement is attributed to limitations inherent in the two-dimensional treatment and point source approach.     

基于系统分析软件的IRWST中PRHR HX建模方法研究

本文针对AP1000内置换料水箱(IRWST)热工水力特性缩比实验4种典型的沸腾工况,应用两种不同的系统分析软件（RELAP5/SCDAPSIM mod3.4和COSINE）,将三维模型简化为一维模型。基于单通道和多通道两种不同建模方法,研究不同的初始温度、加热功率、水箱水位工况下,水箱内的温度、沸腾时间等参数的变化。结果表明,RELAP5单通道模型与多通道模型计算结果低于实验值,COSINE的单通道模型与多通道模型计算结果高于实验值,两种软件的计算精度相当。RELAP5计算模型的沸腾时间整体上晚于实验时间,COSINE计算模型的沸腾时间整体上早于实验时间,采用多通道模型后,每个工况达到沸腾的时间均短于单通道模型,表明采用多通道建模方法后,模型整体的换热能力提高,缩短了模型整体沸腾所需的时间。在系统安全分析的建模过程中,可根据水箱内温度、整体沸腾时间对安全保守性的影响,确定具体的建模策略。     

Downcomer and tie plate countercurrent flow in the Upper Plenum Test Facility (UPTF)

The limitation of vertical steam-water countercurrent flow, called flooding, is important for the operation of Emergency Core Cooling (ECC) Systems in Nuclear Reactors. The ECC water injection flooding behavior is scale dependent and the reactor size behavior cannot be extrapolated from small scale data.A new flooding correlation is presented, based on the classic flooding equation where the effects of gravity, interphase momentum exchange, and instability of the gas/liquid interface are considered. Development of a new correlation was necessary in order to correlate the reactor scale downcomer and upper tie plate countercurrent flow data gained in the Upper Plenum Test Facility (UPTF). The new correlation is an extension of the well-known Wallis-type and Kutateladze-type correlations. Each of the three correlations is valid for the experimental facilities on a certain scale. The range of applicability for each of the three correlations is defined for the case of downcomer and tie plate countercurrent flow.     

核电厂事故后安全壳内置换料水箱碎片传输性能分析

为避免事故后安全壳内置换料水箱（IRWST）内滤网堵塞,保证IRWST下游泵的安全运行,需对IRWST内碎片传输效果进行精细评估。针对某核电厂双环池型IRWST,采用计算流体动力学（CFD）方法对其流场进行了模拟,通过高速区和高湍动能区体积比定量评价事故后碎片传输效果。结果表明,事故后各工况下IRWST内碎片传输比均未超过滤网的设计值,保证了事故后滤网及相连系统的安全性;只有内环滤网A投运时,滤网的负载最大;影响事故后碎片传输效果的主要因素是流场的高速区。针对IRWST的现有布置空间,提出了增大外环搅混管线管径的优化方案,可以显著降低事故后IRWST内碎片传输比,提升事故后核电厂的安全性。      

CFD simulation of steam jet-induced thermal mixing in subcooled water pool

This study examined the IRWST thermal mixing phenomena induced by a steam jet in a subcooled water pool. Due to the limitation of the current CFD code to simulate condensation, the steam condensation region model was developed to evaluate the thermal mixing phenomena. Within this region, all the steam was condensed into water, and the steam mass and energy inputs were treated as the source. This calculation was treated using single-phase CFD methods. The benchmark calculation for a thermal mixing experiment in the water tank was performed to develop an optimized 3D evaluation methodology of the thermal-hydraulic behavior in APR1400 IRWST. Steam discharge through the sparger and condensation phenomenon was modeled with the choking flow and thermal mixing model in the quenching tank using CFX11.Three types of thermal mixing experiments, local phenomena test, thermal mixing tests in cylindrical water pool and annulus water pool, were designed to provide data representative of the behavior of the prototype for CFD simulations of the thermal-hydraulic behavior in IRWST. A comparison of the calculated and experimentally measured temperature profiles showed some disagreement particularly around the sparger. The main reason for this disagreement was caused by the difference in the test and simulating conditions at the tank wall. However, moving away from the sparger, the trends of the temperature rise became similar to that in the experiment. Despite these problems, this model is the best way of evaluating the thermal mixing phenomena caused by a steam jet in a subcooled water pool.     

Two-phase natural circulation flow of air and water in a reactor cavity model under an external vessel cooling during a severe accident

As a part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of the Advanced Power Reactor (APR) 1400, a Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale (HERMES-HALF) experiment has been performed by using the non-heating method of an air injection. This large-scale experiment uses a half-height and half-sector model of the APR1400. This experiment has been analyzed to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is very similar to the experimental results of the HERMES-HALF, in general. Increases in the water inlet area and the water level in the reactor cavity lead to an increase in the water circulation mass flow rate. The effects of an air injection mass flow rate and the water outlet area on the water circulation mass flow rate are dependent on the water inlet area size. As the water outlet moves to a lower position, the water circulation mass flow rate increases slowly.     

承压热冲击下压水堆压力容器壁面换热特性的数值模拟研究

为了研究压水堆因安注冷水直接注入反应堆压力容器下降环腔而导致的承压热冲击（PTS）热工水力问题，基于1∶10比例模型，应用计算流体力学商用软件FLUENT5．4进行了紊流流动换热的数值模拟分析，同时进行了常压传热实验研究。针对下降环腔折算流速0．5m/s，安注流速10m/s的典型工况，研究了压力容器下降环腔的壁面换热特性。通过分析下降环腔内的流动及混合特性，从流动机理上解释了压力容器内壁上准重接触点附近换热强烈的现象，并指出壁面换热强弱与近壁流体紊流脉动动能密切相关，为热冲击分析提供参考。     

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。      

Investigation of Sweepout Mechanism and Critical Void Height in Annular Downcomer

The APR1400 (Advanced Power Reactor 1,400 MWe) has adopted the direct vessel injection (DVI) in lieu of the conventional cold leg injection for its emergency core cooling system (ECCS). In this reactor, sweepout from the water surface by gas (vapor or air) flow plays an important role in analyzing the mass and momentum transfer in the reactor downcomer of multidimensional geometry during a loss-of-coolant accident (LOCA) by decreasing the water level in the downcomer. The core water level will tend to decrease rapidly if a considerable amount of the entrained water stream and droplets bypasses through the break. The amount of entrained water is mostly determined by the interacting gas flow rate, the geometric condition, and the interfacial area between the gas and the water. The sweepout is observed to take place in three rather distinct steps: the beginning of undulation, the full wave and the wave peak (droplet separation). In view of these observations we investigated the relation between the gas flow rate and the amount of bypass as a function of time. The current experimental results shed light on the flow mechanism and the semi-empirical relations for the three-dimensional sweepout in a large-diameter annulus such as the reactor downcomer. A physico-numerical model is being developed to predict the multidimensional bypass flow rate resulting from the sweepout and entrainment in the downcomer.     

安全壳内置换料水箱内自然对流现象试验研究

以第3代核电技术中广泛采用的安全壳内置换料水箱（IRWST）为对象,通过比例分析获得了自然对流现象的相似准则,设计了缩比试验装置,对事故条件下IRWST内的自然对流现象进行了试验研究,分析了IRWST内自然对流的演变规律及初始条件的影响。结果表明：相似格拉晓夫数、相似雷诺数和相似普朗特数是IRWST自然对流现象试验装置设计应遵循的相似准则;加热初期,IRWST内以轴向上升羽流为主,随冷热分层的形成,流体的轴向上升运动被抑制,转变为以IRWST中下部区域的径向横流为主;不同初始条件下IRWST内自然对流的演变规律基本一致,但流场演变过程的快慢、流体速度的大小不同。     

MELCOR 1.8.4 sensitivity analysis of the severe accident evolution during the APR 1400 LOCA

Unlike most other systems in which the emergency core cooling (ECC) water is injected into the cold-legs, the Advanced Power Reactor (APR) 1400 employs a concept of a direct vessel injection (DVI) to reduce the bypass effects of the ECC water via a break during a design basis LOCA. For this, the DVI piping is designed so that the ECC water taken from an in-containment refueling storage tank (IRWST) directly flows into the reactor pressure vessel (RPV) down-comer. The main objective of this paper is to provide the MELCOR 1.8.4 sensitivity analysis results on the evolution of the severe accidents that can be expected during the APR 1400 LOCA and the insights gained from the analysis. For this purpose, the present sensitivity analysis mainly focuses on: (1) the impact of the foregoing engineering features (i.e., DVI and IRWST) in mitigating a severe core degradation and (2) the APR 1400-specific impacts of different break locations and sizes, and an operation of the containment spray systems on the timings of the key thermal-hydraulic responses, the severe degradation of the core, and the evolution of the core materials. No significant accident management strategy that plays a great role in mitigating a further progression of severe accidents has been taken into account in the present analysis. As a result, the present analysis results can be taken as the technical basis for assessing the effectiveness of a potential severe accident management.     

C型换热器管外流体两相自然循环数值模拟

建立了简化的C型换热器管外流体CFD分析模型,模拟了反应堆安全壳内置换料水箱（IRWST）中典型气液两相自然循环特性。首先用公开发表文献中的试验数据对计算方法进行校验,计算中采用的湍流模型、壁面沸腾模型等能较好地捕捉主流流体升温特性、两相自然循环特性。结果表明：C型换热器增加了管外流体流场分布的不均匀性,提高了冷、热流体间的搅混强度,有助于降低管外流体温度差,增加大容积水池内的自然循环能力;但由于壁面对气泡的阻滞作用,换热器弯管及水平管局部区域空泡份额最大,发生了气泡聚集。计算结果可为非能动余热排出换热器的设计提供支持。     

基于CFD方法的非能动余热排出系统数值模拟

本文应用FLUENT软件对APl000的非能动余热排出热交换器和换料水箱进行了数值模拟，分析了不同c型传热管数量和冷却剂入口温度对热交换器换热性能和换料水箱内热分层、自然循环现象的影响。分析表明，总体通流面积不变，随着传热管数量增加，热交换器出口温度变小，水箱水温整体提升，热分层现象显著，自然循环趋势明显；质量流量不变，随着冷却剂入口温度的增加，入口流速增加，热交换器出口温度变大，但降温幅度也变大，水箱平均水温升高，热分层范围扩大，自然循环流速加快。     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

Safety of Super LWR, (I) Safety System Design

This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis.     

非能动余热排出换热器换热能力数值分析

基于多孔介质模型,对AP1000非能动余热排出换热器(PRHR-HX)运行初始阶段进行了数值模拟。一回路的入口温度及流量采用RELAP5的计算结果,并以此作为CFD计算的边界条件。采用多孔介质模型处理C型管束区,添加管束区分布阻力。通过商业CFD软件FLUENT计算得到安全壳内置换料水箱(IRWST)侧冷却剂的三维温度及速度分布,通过用户自定义函数UDF完成一回路侧与IRWST侧的耦合换热计算,获得一回路温度分布及换热量。计算结果表明,随着IRWST内冷却剂温度升高,换热器热负荷降低,并出现明显的热分层现象,同时证明采用多孔介质模型与耦合换热计算是分析PRHR/IRWST系统瞬态热工水力特性的有效方法。     

基于RELAP5的LOCA喷放阶段下降段内CCFL特性研究

反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL）,不利于应急冷却水顺利进入堆芯,极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为;通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性,并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明,由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性,随着蒸汽流量增大,在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大,较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系;在靠近破口的环路内注入的冷却水更难到达下腔室,而在远离破口环路的冷却水容易进入到下腔室;过热的蒸汽在流动过程中被冷却水冷却发生凝结现象,导致出口蒸汽流量小于进口蒸汽流量,且随着进口蒸汽流量的增大,凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。      

压水堆核电站安全注入试验期间执行机构拒动和误动的干预对策分析

安全注入试验是压水堆核电厂热试期间涉及范围最广、风险最高的试验。试验程序要求在热停平台通过快速开启蒸汽排放阀模拟二回路破口触发安注信号,验证反应堆跳闸,安全壳隔离,安注执行机构动作,并对开盖冷试期间调整的安注流量进行再次验证。安注信号一旦触发将导致22个系统共计234个设备真实动作,一回路被注入含硼水。任何在线错误、设备缺陷或操作失误都可能导致试验失败,甚至可能导致一回路设备损坏;同时因安全注入试验将导致核电站主回路产生一次瞬态,对一回路设备冲击极大,所以安全注入试验必须保证一次成功。为了保证试验的真实性及完整性,提高试验的一次成功率,控制试验的风险,本研究针对以往项目执行该试验时存在的一回路水位过高及设备误动或拒动的难题,对试验方案进行了优化创新。该方案成功运用于阳江3号机安全注入试验,一定程度上解决了稳压器水位过高及设备误动、拒动的难题,获得了机组安全可控且试验顺利高效的效果,达到了同行领先水平。