核电厂管线中的温度振荡现象研究

在核电厂中,如何更好地了解和预防由于温度振荡而导致的管线热疲劳,对于确保核电厂的安全和可靠运行具有重要意义。本文以核电厂安注系统某支管为研究对象,运用计算流体力学软件,结合二次开发,采用修正的k-ε模型,模拟了阀门渗漏冷水进入含有高温水的支管后所发生的温度振荡现象,并与实验测量进行了对比。数值模拟的结果和实验基本吻合,并全面地反映了整个管线中的温度振荡现象,为更好地监控管线热疲劳提供了参考依据。     

堆芯紧急冷却安注热混合实验研究

以核电厂压水堆中失水事故(LOCA)堆芯紧急安注系统(ECCS)启动后安注接管与冷管段的T型管处冷、热流体混合为研究对象,进行安注管和主管道内过冷水-高温冷却剂的热混合特性实验以及过冷水-汽水混合物直接接触冷凝特性实验,通过缩比尺寸实验对热混合相关现象进行研究。结果表明,单相热混合实验管内温度场随不同射流流型成一定分布;两相热混合工况安注后冷凝量随主管蒸汽量变化而成线性分布,并总结实验数据形成适用于本实验直接接触冷凝相关关系式。     

T型管直接接触冷凝特性实验研究

在反应堆发生LOCA时,一回路系统压力降低,产生大量的蒸汽,安注水注入冷腿后可能会发生冷凝现象。为研究冷凝现象,通过开展T型管冷凝实验,在主管通纯蒸汽、支管通过冷水的情况下,研究了不同蒸汽流量和不同安注水流量下的冷凝量。结果表明:冷凝量存在一定的限制,即主管内蒸汽无法全部被冷凝。基于实验结果提出了一个冷凝效率与热力学比系数R_T之间的模型。     

基于DVI管失水事故试验的CATHARE程序模拟评价

采用CATHARE程序对直接注入(DVI)管失水事故(LOCA)试验进行了数值模拟。研究发现:DVI管LOCA中系统卸压、非能动安注、堆芯冷却等主要过程和物理现象得到了较好的模拟。一回路系统压力、堆芯补水箱(CMT)安注流量、安注箱(ACC)安注流量、内置换料水箱(IRWST)安注流量以及堆芯流体温度等参数的计算结果和试验数据符合较好。研究结果表明,CATHARE程序可以用于失水事故下非能动安注系统瞬态特性模拟分析。     

T型管热混合温度波动特性研究

在反应堆发生失水事故(LOCA)时,一回路系统压力降低,产生大量蒸汽,堆芯应急冷却系统(ECCS)启动后,安注水注入冷腿后在T型管处与蒸汽发生热混合,温度会出现明显波动,同时伴随有一定的回流。本文以T型管中冷热流体混合为研究对象,开展了安注过冷水与冷腿中的饱和蒸汽热混合实验。研究内容主要为过冷水与饱和蒸汽在水平T型管发生热混合之后的水跃和回流现象,基于动量分析的方法,分析了不同流型对热混合后温度分布的影响,提出了两相流动量比关系式用于分析T型管内温度波动特性。     

核电厂中小LOCA事故下PSA成功准则研究

为了分析核电厂冷却剂丧失事故（LOCA）的瞬态响应，用于支持核电厂概率安全分析（PSA）成功准则的研究。本文以压水堆核电厂为研究对象，利用系统分析程序建立了电厂模型，研究了堆芯补水箱、安注箱、余热排出热交换器和ADS阀门的失效组合及操作员动作时间、破口尺寸等的敏感性，得出如下结论：在小LOCA事故下，如果3个ADS-4阀门能够开启（自动或安注信号产生后30 min手动开启）且1条IRWST注入管线可用或者1个ADS-4阀门开启（自动开启或安注信号产生后30 min手动开启）且安注信号产生后30 min手动启动一台正常余热排出系统（RNS）泵，则能够维持堆芯冷却；在中等LOCA事故下，至少一个CMT或ACC投入运行，3个ADS-4阀门开启（自动或安注信号产生后20 min手动开启）且1条IRWST注入管线可用或者1个ADS-4阀门开启（自动或安注信号产生后20 min手动开启）且在安注信号产生后20 min内启动一台RNS泵，则能够维持堆芯冷却。     

PTS瞬态流动与换热数值模拟研究

为了研究压水堆因“直接安注”冷水注入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,同时进行了常压瞬态传热实验研究。针对下降环腔折算流速0.5 m/s,安注流速10m/s的典型工况,研究了安注水开启后下降环腔内的瞬态流动换热特性,数值模拟与实验结果吻合良好。考察了压力容器安注接管出口区环形焊缝区及堆芯段筒体中子强辐照区所承受的热冲击状况,基于稳态流动研究了下降环腔内流体混合特性及流动机理,为热冲击分析提供参考。     

池式铅基堆热分层现象研究综述

铅基堆具有系统简单紧凑、安全性高等优点，已成为第四代核能系统的主要发展方向。铅基堆发生事故停堆时，堆芯功率骤降、驱动泵停闭，堆芯出口冷却剂温度急剧下降且流速降低，无法冲入热池顶部与高温流体进行混合换热，只能聚集在热池底部，导致热池中发生热分层现象。热分层现象会影响堆芯的余热排出能力，并会造成反应堆容器及内构件热疲劳。本文阐述了铅基堆热分层产生的机理与危害，调研并总结了铅基堆热分层现象国内外研究进展和存在的问题，最后从理论研究、实验研究和数值模拟研究方面提出热分层的未来研究方向。     

波动管小破口失水事故实验研究

在模块化小型反应堆非能动安全系统综合模拟实验装置上进行波动管小破口尺寸失水事故实验,研究波动管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。模块化小型反应堆发生失水事故后,压力平衡管和安注管线内流体的密度差可以驱动堆芯补水箱(CMT)内的冷流体注入反应堆压力容器,压力平衡管裸露后CMT安注流量出现波动;安注箱(ACC)的安注对事故初期的堆芯冷却效果显著;经自动卸压系统卸压后,内置换料水箱(IRWST)可以对堆芯进行持续稳定的安注和冷却。研究结果表明:波动管小破口失水事故中,非能动安注系统可以对堆芯进行有效注水,并带走堆芯衰变热量。     

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。     

Thermal stratification and reactor piping integrity

Thermal stratification, cycling, and striping phenomena have drawn much attention recently because of the incidents at several nuclear plants that raised significant safety concerns. The concerns due to these phenomena relate to thermal fatigue in branch pipes connected to the main coolant piping. Nuclear utility industry is addressing the issue with the aim to understand the mechanisms that lead to fatigue in nominally stagnant piping systems near the reactor coolant piping. Two key results from this effort are described in this paper. First, tests to investigate the interaction between the main coolant piping and the stagnant attached lines by turbulence penetration are described and a working correlation is obtained. Turbulence penetration into unisolable lines, or the transport of turbulence into stagnant piping from the reactor coolant system (RCS) line, represents a mechanism for carrying hot RCS water into regions filled with cold water. The possibility of stratification of the two fluids (and the resultant thermal stresses) is the reason for developing an understanding of the turbulence penetration process. Secondly, results of an evaluation to develop a loading definition for thermal striping are included. Based on this testing several important conclusions relating to fatigue in nominally static reactor coolant systems are reached.     

On the relevance of low side flows for thermal loads in T-junctions

The mixing of coolant streams of different temperatures in pipe junctions leads to temperature fluctuations that may cause thermal fatigue in the pipe wall. Numerous T-junction experiments are known from literature, which were performed to study the nature of thermal loads in the pipe walls occurring during the mixing of hot and cold liquid. It is common to all known experiments that the experimental boundary conditions are set to reflect cases, in which the flow velocities in both main and side branches of the T-junctions are of the same order of magnitude. In the present experiments, carried out using wire-mesh sensors, it was observed that very low flow velocities in the side branch compared to the main pipe may lead to conditions potentially severe for thermal fatigue due to the low frequency of the temperature fluctuations occurring. The T-junction presented here consists of a perpendicular connection of two pipes of 50 mm inner diameter. The straight and the side branches are supplied with water of different electrical conductivities, to enable performing generic, isothermal tests on turbulent mixing with the idea to model the temperature fluctuations in thermal mixing processes. A pair of wire-mesh sensors, each with a grid of 16 × 16 measuring points, are used to record conductivity distributions in the downstream of the T-junction as well as directly at the junction in both branches. At very low flow rates in the side branch, a characteristic entrainment of liquid from the main branch into the side branch was found. Typically the entrainment flow in the side branch results in relatively high fluctuations at the low-frequency range. While the sensor in the main flow shows fluctuations with a power spectrum similar in character to mixing experiments with comparable flow velocities in both branches of the T-junction. The phenomenon of entrainment of water from the main branch into the side branch against the main flow direction vanishes at a certain critical velocity in the side branch.     

Thermal stress analysis for fatigue damage evaluation at a mixing tee

Fatigue cracks have been found at mixing tees where fluids of different temperature flow in. In this study, the thermal stress at a mixing tee was calculated by the finite element method using temperature transients obtained by a fluid dynamics simulation. The simulation target was an experiment for a mixing tee, in which cold water flowed into the main pipe from a branch pipe. The cold water flowed along the main pipe wall and caused a cold spot, at which the membrane stress was relatively large. Based on the evaluated thermal stress, the magnitude of the fatigue damage was assessed according to the linear damage accumulation rule and the rain-flow procedure. Precise distributions of the thermal stress and fatigue damage could be identified. Relatively large axial stress occurred downstream from the branch pipe due to the cold spot. The variation ranges of thermal stress and fatigue damage became large near the position 20° from the symmetry line in the circumferential direction. The position of the cold spot changed slowly in the circumferential direction, and this was the main cause of the fatigue damage. The fatigue damage was investigated for various differences in the temperature between the main and branch pipes. Since the magnitude of accumulated damage increased abruptly when the temperature difference exceeded the value corresponding to the fatigue limit, it was suggested that the stress amplitude should be suppressed less than the fatigue limit. In the thermal stress analysis for fatigue damage assessment, it was found that the detailed three-dimensional structural analysis was not required. Namely, for the current case, a one-dimensional simplified analysis could be used for evaluating the fatigue damage without adopting the stress enhancement factor K_t quoted in the JSME guideline. The results also suggested that, for a precise assessment of the fatigue damage at a mixing tee, the effect of multi-axial stress on the fatigue life together with the mean stress effect should be taken into account.     

蒸汽发生器给水环管热分层强度影响因素及缓减措施研究

为研究低功率运行工况下压水堆核电厂蒸汽发生器给水环管热分层强度的影响因素和缓减措施,采用计算流体力学(CFD)方法,对给水环管内部流体的热分层现象进行了数值模拟。研究了功率提升过程中给水运行方式及给水环管结构对热分层强度的影响,从给水环管布置及结构两个方面提出了两种改进结构。结果表明:提高给水流量提升速率对热分层强度几乎没有影响,给水弯头弯曲半径增大能有效减弱给水弯头处的热分层强度,给水弯头向下布置能明显减弱给水环管内的热分层强度,给水管倾斜角度增大能有效减弱给水弯头、给水管的热分层强度,同时使得给水环的热分层强度增强。将给水弯头向下布置且给水管水平段改为多级阶梯渐缩-渐扩结构、多级球形渐缩-渐扩结构后,给水环管内的热分层现象能得到不同程度的缓解,因此提出的两种改进结构是较理想的改进结构。     

CFD simulation of Fortum PTS experiment

Detailed simulation of the thermal stresses of the reactor pressure vessel (RPV) wall in case of pressurized thermal shock (PTS) requires the simulation of the thermal mixing of cold high-pressure safety injection (HPI) water injected to the cold leg and flowing further to the downcomer. The simulation of the complex mixing phenomena including, e.g., stratification in the cold leg and buoyancy driven plume in the downcomer is a great challenge for CFD methods and requires careful validation of the used modelling methods.The selected experiment of Fortum mixing test facility modelling the Loviisa VVER-440 NPP has been used for the validation of CFD methods for thermal mixing phenomena related to PTS. The experimental data includes local temperature values measured in the cold leg and downcomer. Conclusions have been made on the applicability of used CFD method to thermal mixing simulations in case with stratification in the cold leg and buoyant plume in the downcomer.     

压水堆稳压器波动管热分层的分析研究

热分层是管道水平管段中相对滞止或缓慢流动的冷、热流体因缺少混合而产生的不均匀温度分布现象.通过稳压器波动管热分层现象产生的原因和机理分析,并对稳压器波动管热分层现象进行数值模拟,建立了不同稳压器内部不同截面的热分层瞬态.     

压力容器外部冷却系统冷却水池温度场的实验研究及分析

压力容器外部非能动冷却系统采用换料水池作为冷却水源。在浮升力驱动的自然循环流动作用下,冷却水池内会逐渐出现热分层现象。本实验基于先进压水堆压力容器外部冷却系统模拟装置REPEC实验回路,通过测量实验系统内冷却水箱的温度场空间分布,对冷却水池的热分层与混合现象、发展规律和主要影响因素进行了实验分析。结果表明：实验水箱内温度场分布差异主要表现在高度方向;循环流量是影响热分层的重要参数,而水箱工质初始温度的影响非常微弱;针对本实验的无量纲一维瞬态温度场方程分析表明,水箱内温度场的发展规律主要受对流传热控制。