核电厂安全壳内置换料水箱过滤系统过滤性能及阻力特性研究

对核电厂安全壳内置换料水箱（IRWST）过滤系统过滤性能及压损特性进行了分析研究。该研究借助试验验证和数值模拟分析方式展开,主要包括下游效应（碎片浓度）试验研究和过滤系统压损分析2部分。结果表明,在事故工况下,地坑滤网下游碎片浓度为368 ppm（1 ppm=1 mg/L）、安全注入系统（RIS）地坑滤网和安全壳喷淋系统（EHR）地坑滤网的压损分别为3.533 kPa和3.631 kPa,上述结果分别满足了过滤系统碎片浓度小于480 ppm和压损小于5.6 kPa的系统功能要求。      

核电厂安全壳地坑滤网堵塞问题的机理分析

破口失水事故工况下,大量碎片可能随着泄漏的冷却剂和喷淋液迁移到安全壳地坑滤网处,并逐渐堆积形成碎片床,不断增大流体通过滤网的阻力,降低应急堆芯冷却系统或安全壳喷淋系统泵的净正吸入压头裕量并导致堆芯、安全壳丧失冷却,从而威胁核电厂安全。本文对核电厂发生假想破口失水事故后碎片的产生、迁移,以及在安全壳地坑滤网处堆积成碎片床,并造成地坑滤网堵塞的机理进行分析说明。     

核电厂长期冷却阶段燃料组件表面化学产物沉积分析

核电厂失水事故后长期冷却阶段,化学产物可能在堆芯燃料棒表面析出,增加燃料组件表面沉积的碎片量,导致堆芯流道堵塞、包壳温度升高等问题,影响应急堆芯冷却系统长期冷却再循环能力。根据核电厂失水事故后安全壳内材料特定的溶解特性及国际通用的堆内下游效应分析方法,开展核电厂长期冷却阶段潜在的化学产物在堆芯燃料组件表面的析出特性分析。结果表明,失水事故后核电厂安全壳内材料会释放铝、硅、钙等元素并在燃料组件表面析出硅酸铝钠和硼酸钙沉淀,硅酸铝钠和硼酸钙在长期冷却阶段的最大沉积厚度分别为0.0124mm和0.0518mm,化学产物在燃料组件表面的沉积在核电厂堆芯长期冷却性能评估中应予以考虑。     

IVR条件下外部冷却流道变形的影响因素研究

当发生堆芯熔化事故时,压力容器外部冷却是保持压力容器完整性及实现熔融物堆内滞留(In-Vessel Retention,简称IVR)的一项重要策略。在高温熔融物的热载荷和内部压力的共同作用下,压力容器外壁面和保温层之间的冷却流道可能发生变形,造成冷却能力的降低,进而威胁到压力容器的完整性。因此,有必要分析IVR条件下压力容器冷却流道变形的影响因素。结果表明,热膨胀是造成冷却流道变形的主要因素。在IVR策略成功的前提下,内压和热流密度对流道变形的影响有限。     

安全壳地坑过滤器压损数值模拟与试验研究

为了研究核电站冷却剂丧失事故后地坑过滤器的压损特性,本文提出了一种数值模拟与试验结合的方法。首先,对过滤器滤筒部分进行满负载试验;然后,运用Fluent软件对过滤器汇流槽部分流场进行模拟。从而得出地坑过滤器在满负载时的总体压损。结果表明:地坑过滤器的总体压损满足安全注入系统的压损要求;过滤器汇流槽流道截面变化,尤其是突缩或突扩是压损的主要原因。     

压水堆安全壳地坑滤网化学效应WCAP分析方法

针对核电厂地坑滤网安全性能问题,美国核管理委员会(NRC)先后出台了一系列RG1.82"失水事故后长期再循环冷却的水源"管理导则的修订版,用以指导地坑滤网堵塞研究。冷却剂失水事故(LOCA事故)后在安全壳喷淋液和安全壳地坑介质的化学环境会导致安全壳内的各种碎渣中化学元素的溶解,并且随着安全壳地坑介质温度的降低形成沉淀析出,所析出的沉淀会在安全壳地坑滤网表面物理碎渣床上形成二次沉积,从而造成滤网压损性能的进一步恶化,此即为安全壳地坑滤网的化学效应。本文介绍压水堆安全壳地坑滤网化学效应的试验分析方法。     

堆芯过滤装置的设计改进

岭澳核电站3号机组在热态功能试验期间发现多组堆芯过滤装置损坏。本文通过对堆芯过滤装置的失效原因进行分析,针对性地提出了改进方案,并通过机械试验的方式模拟堆芯过滤装置滤网脱落或断裂的失效过程,从而验证了分析结果的准确性以及新型堆芯过滤装置性能的优越性。     

一种非能动余热排出系统设计方案

非能动余热排出系统依靠本身的自然循环特性,应能够在较长时间内提供对堆芯的冷却,保证反应堆的安全。提出一种非能动空气冷却余热排出系统(PRHRS)方案,利用应急冷却水箱作为中间缓冲设备,既可以满足事故初期快速冷却的要求,又能保证非能动余热排出系统在相当长一段时间内的可靠运行。基于自然循环系统特性对所设计的PRHRS系统进行设计计算,并使用RELAP5程序对全厂断电事故下反应堆停堆后PRHRS投入运行的过程进行仿真,以验证设计的合理性。反应堆热工水力动态特性的结果表明,该系统可通过自然循环排出堆芯余热,保证堆芯安全。     

核电厂安全壳地坑滤网堵塞问题的纠正措施研究

核电厂发生破口失水事故后,当应急堆芯冷却系统或安全壳喷淋系统处于再循环模式运行时,碎片堵塞对安全壳地坑滤网的性能存在着潜在的影响,而且碎片迁移过程中的堵塞可能会对再循环模式需要的流道造成不利的影响。本文将从碎片产生、碎片输运和地坑滤网设计等方面,论述说明针对地坑滤网堵塞问题可能采取的纠正措施。     

非能动核电厂全厂断电事故自然循环现象研究

为识别全厂断电事故下非能动核电厂的主要热工水力现象,对AP1000核电厂全厂断电工况下的事故序列和自然循环现象进行了研究。通过建立AP1000的节点模型,进行了全厂断电事故序列的模拟,并划分了事故阶段,分析了非能动堆芯冷却系统中堆芯补水箱(Core Makeup Tank, CMT)投入失效和安全壳内置换料水箱(In-containment refueling water storage tank, IRWST)参数异常对事故自然循环过程的影响,研究结果表明:全厂断电事故下,非能动核电厂的堆芯衰变热由多个单相自然循环过程导出,其中堆芯与非能动余热排出热交换器(Passive Residual Heat Removal Heat Exchanger, PRHR HX)之间的自然循环对堆芯衰变热的导出具有显著影响。根据热阱的不同和系统参数变化的特点,事故序列可划分为主回路自然循环、非能动堆芯冷却系统(Passive Core Cooling System, PXS)自然循环和长期冷却三个阶段;CMT投入、IRWST水箱参数对PXS自然循环过程存在重要影响。     

基于HFETR的快堆燃料短棒辐照试验方案设计与分析

综合考虑辐照试验指标与燃料试验安全、高通量工程试验堆（HFETR）运行要求、试验段压差波动等因素,基于HFETR开展了快堆燃料短棒辐照试验方案设计与分析,确定了铅铋合金层厚度、冷却水流道结构、阻力塞结构、冷却水流量等关键参数,获得了热棒包壳最高温度为(490±60)℃的高线功率密度辐照试验方案。试验结果表明,热棒最大线功率密度为68~85 kW/m时,包壳与燃料芯体温度满足辐照试验要求且留有余量;在200~300 kPa堆芯压差范围内,相同压差下试验段流量的计算流体力学（CFX）计算值比试验值偏小9% ~11%;试验段外侧窄缝流道的流量份额为7.3%,显著低于该流道的流通面积份额,满足线功率密度为85 kW/m时燃料短棒的冷却要求。本文提出的辐照试验方案可为快堆燃料棒的高线功率密度辐照试验提供参考。      

Experimental study of debris head loss through a pressurized water reactor recirculation sump screen after LOCA

A loss of coolant accident (LOCA) in a PWR (pressurized water reactor) would generate debris from thermal insulation and other materials in the vicinity of the break. A fraction of the LOCA-generated debris and pre-LOCA debris would be transported into the sump and accumulated on the sump screens resulting in adverse blockage effects that include decrease in net positive suction head (NPSH) margin. The NUREG/CR-6224 head loss correlation has been widely used to estimate debris-induced head loss across sump screens, but it has been recommended to be used for debris thicknesses less than 4 in. In order to extend the limit, head loss data are obtained over a wider range of bed thicknesses. Experimental results show that the NUREG/CR-6224 correlation conservatively predicts the head loss across NUKON™ debris beds with theoretical thicknesses up to 6 in. Head loss measurements in mixed debris beds show that fibrous debris with calcium-silicate and/or fine particulates such as coatings mainly deteriorates the NPSH.     

An experimental investigation on penetration of buoyancy-driven cold liquid into a vertical channel with hot, forced-flow

An experimental investigation on penetration of cold liquid into a modeled subassembly channel under mixed convection, from the hot plenum of a prototypical fast breeder reactor (FBR) design has been conducted. The penetration phenomenon occurs under certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system (DRACS) for decay heat removal and can influence the natural circulation head that determines the core flow rate and consequently the core cooling rate. In the present experiment, a simplified test section simulating the upper plenum and a subassembly channel was constructed wherein both temperature and velocity measurements were made to investigate the penetration of cold water into a vertical channel with upward flowing hot water. Upon comparing temperature and velocity data we found an overall similarity in their respective spatio-temporal distributions as expected. The data suggested correlations describing the onset of flow penetration and penetration depth into the channel, expressed as a combination of Reynolds and Grashof numbers for a given Prandtl number. Flow penetration was identified based on the temperature and velocity data. This work suggests a dimensionless grouping based on an order of magnitude analysis that satisfactorily casts the experimental data while noting the subtle differences with other investigations.     

组合阀步降流道三维流场数值分析

运用计算流体力学程序CFX对组合阀步降流道内三维流场进行了分析和计算,流道水力学特性计算结果与组合阀步降流道流动阻力实验结果符合良好。在此基础上,对流道内流速和压力分布进行了研究,结果表明:阀芯流道侧壁入口上端流体流速达到峰值,回零水腔内流体的流速分布均匀,流道内压力损失最大的部位处于入口流道竖管段与横管段之间,可通过增大管径或加缓变过渡段的方法来减少该部位的流动阻力损失。     

Assessment of passive safety system performance under gravity driven cooling system drain line break accident

A generation III+ Boiling Water Reactor (BWR) which relies on natural circulation has evolved from earlier BWR designs by incorporating passive safety features to improve safety and performance. Natural circulation allows the elimination of emergency injection pump and no operator action or alternating current (AC) power supply. The generation III+ BWR's passive safety systems include the Automatic Depressurization System (ADS), the Suppression Pool (SP), the Standby Liquid Control System (SLCS), the Gravity Driven Cooling System (GDCS), the Isolation Condenser System (ICS) and the Passive Containment Cooling System (PCCS). The ADS is actuated to rapidly depressurize the reactor leading to the GDCS injection. The large amount of water in the SP condenses steam from the reactor. The SLCS provides makeup water to the reactor. The GDCS injects water into the reactor by gravity head and provides cooling to the core. The ICS and the PCCS are used to remove the decay heat from the reactor. The objective of this paper is to analyze the response of passive safety systems under the Loss of Coolant Accident (LOCA). A GDCS Drain Line Break (GDLB) test has been conducted in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) which is scaled to represent the generation III+ BWR. The main results of PUMA GDLB test were that the reactor coolant level was well above the Top of Active Fuel (TAF) and the reactor containment pressure has remained below the design pressure. In particular, the containment maximum pressure (266 kPa) was 36% lower than the safety limit (414 kPa). The minimum collapsed water level (1.496 m) before the GDCS injection was 8% lower than the TAF (1.623 m) but it was ensured that two-phase water level was higher than the TAF with no core uncovery.     

Downcomer boiling phenomena during the reflood phase of a large-break LOCA for the APR1400

Downcomer boiling phenomena in a conventional pressurized water reactor has an important effect on the transient behavior of a postulated large-break LOCA (LBLOCA), because it can degrade the hydraulic head of the coolant in the downcomer and consequently affect the reflood flow rate for a core cooling. To investigate the thermal hydraulic behavior in the downcomer region, a test program for a downcomer boiling (DOBO) is being progressed for the reflood phase of a postulated LBLOCA. Test facility was designed as a one side heated rectangular channel which adopts a full-pressure, full-height, and full-size downcomer-gap approach, but with the circumferential length reduced 47.08-fold. The test was performed by dividing it into two-phases: (I) visual observation and acquisition of the global two-phase flow parameters and (II(a)) measurement of the local bubble flow parameters on the measuring planes along five elevations. In the present paper, the test results of Phase-I and a part of Phase-II(a) were introduced.     

Optimization of zigzag flow channels of a printed circuit heat exchanger for nuclear power plant application

Printed circuit heat exchanger (PCHE) is recently considered as a recuperator for the high-temperature gas cooled reactor. In this study, shape optimization of zigzag flow channels in a PCHE has been performed to enhance heat transfer performance and reduce the friction loss based on three-dimensional Reynolds-averaged Navier–Stokes analysis with the Shear Stress Transport Turbulence model. A multi-objective genetic algorithm is used for the multi-objective optimization. Two non-dimensional objective functions related to heat transfer performance and friction loss are employed. The shape of a flow channel is defined by two geometric design variables, viz. the cold channel angle and the ellipse aspect ratio of the cold channel. The experimental points within the design space are selected using Latin hypercube sampling as the design of the experiment. The response surface approximation model is used to approximate the Pareto-optimal front. Five optimal designs on the Pareto-optimal front have been selected using k-means clustering. The flow and heat transfer characteristics, as well as the objective function values, of these designs have been compared with those of the reference design.     

压水堆核电厂安全壳内碎片源的踏勘方法介绍

安全壳地坑是许多压水堆核电厂设计为在失水事故后为堆芯冷却和安全壳排热提供再循环水的专设安全设施。安全壳内的潜在碎片源在事故中可能堵塞安全壳内的地坑滤网,从而造成安全壳地坑性能下降。为了评价安全壳地坑在破口事故后能否满足设计要求,首先应确定潜在碎片源的类型以及它们在安全壳内的位置。安全壳内现场踏勘就是寻找与定位碎片源的有效方法,并能够提供一些进行安全壳地坑性能分析的必要信息。介绍了压水堆核电厂安全壳内碎片源的一些踏勘方法。     

石墨慢化通道式熔盐堆有效缓发中子份额计算方法研究

有效缓发中子份额(βeff)是研究反应堆动力学特性的关键参数。在液态燃料熔盐堆(MSR)中,燃料流动引起缓发中子先驱核(DNP)在堆内的再分布,并使部分DNP在堆外回路衰变,从而导致βeff的计算方法与固态燃料反应堆不同。为评估石墨慢化通道式熔盐堆内燃料流动引起的反应性损失,研究缓发中子随燃料的流动行为,同时为堆设计和安全分析提供依据,分别基于解析方法和数值方法推导了计算βeff的数学模型,计算了熔盐实验堆(MSRE)在额定工况下的DNP损失份额和堆内DNP浓度分布,并分析了燃料在堆外流动时间和入口流量对βeff的影响。结果表明：两种方法均可对DNP行为提供合理描述;固定燃料在堆外流动时间,βeff随入口流量的增加而减小;固定入口流量,βeff随燃料在堆外流动时间的增加而减小,80 s后趋于稳定。     

Effect of inlet geometry on hot-leg U-bend two-phase natural circulation in a loop with a large diameter pipe

In relation to nuclear reactor accident and safety studies, experiments on hot-leg U-bend two-phase natural circulation in a loop with a relatively large diameter pipe (10.2 cm ID) was performed for understanding the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR. The loop design was based on the scaling criteria developed under this program and a horizontal section was inserted between the gas injector and the hot leg in order to investigate the effect of the vapor phase inlet section on the flow regimes and flow interruption. The loop was operated either in a natural circulation mode or in a forced circulation mode using nitrogen gas and water. Various tests were carried out to establish the basic mechanism of the flow termination as well as to obtain essential information on scale effects of various parameters such as the loop frictional resistance, thermal center, and pipe diameter. The void distribution in a hot leg, flow regime and natural circulation rate were measured in detail for various conditions. The termination of the natural circulation occurred when there was insufficient hydrostatic head in the downcomer side. The superficial gas velocity at the flow termination could be predicted well by the simple model derived from a force balance between the frictional pressure drop along the loop and the hydrostatic head difference. The bubbly-to-slug flow transition was found to be dependent on axial locations. It turned out that the inlet geometry affected the flow regime at the inlet of the hot leg, namely the void distribution in the hot leg.