密度锁内分层传热特性的初步探讨

通过可视化观察方法,对3种不同实验管内的流体分层传热特性进行实验研究,同时,建立传热计算模型,对密度锁内的传热机理进行分析.结果表明:密度锁内的分层工质自上而下分为:混合层、界面层、导热层.混合层内的传热以对流为主,其余两部分的传热以导热为主.对不同管径的研究表明,密度锁内的蜂窝通道能有效地抑制扰动作用,减小混合层的厚度,降低通过密度锁的热量传递.     

密度锁内分层界面的作用分析

进行了冷态和热态实验,结合理论分析研究了密度锁的抗扰动机理。冷态时,流体流动产生的压力差引起了密度锁中各通道流体的大环流运动;热态时,密度锁内分层界面的形成对大环流运动起到了阻碍作用。结果表明:由于压力差作用,密度锁内分层界面向一侧倾斜产生重位差;当压力差与重位差相平衡时,大环流运动被阻止,流体达到分层平衡。     

密度锁启动特性实验研究及计算验证

以基于密度锁的非能动余热排出系统（PRHRS）为研究背景,实验验证了密度锁自平衡启动方案的可行性。结果表明：主回路流量接近平衡流量启动PRHRS时,密度锁内冷热流体温度分界面将在不平衡力作用下向上或向下移动,减小或增大了余热排出回路重位压差,使密度锁内冷热流体温度分界面在新的位置达到受力平衡,最终实现密度锁的自平衡启动,以及余热排出回路与主回路的隔离。依据一维连续性方程、能量方程及动量方程建立数学模型,用Matlab语言编程,对密度锁启动过程进行了数值模拟分析,证明了密度锁自平衡启动方案合理、有效。计算值与实验值符合较好,用该程序可较好地模拟密度锁自平衡启动过程中系统的瞬态运行特性。     

密度锁启动过程中传热特性的实验研究

分别在无扰动以及扰动大小不同的实验情况下,对密度锁启动过程中的传热特性进行了实验研究.研究结果表明,密度锁能抑制热量下传.根据传热情况,密度锁内过渡区可分为强分层、弱分层和无分层3种情况.强分层和弱分层情况下,密度锁都能有效地阻止热量下传,且强分层向下传热量更小.而在无分层情况下热量连续不断地从密度锁传出,导致密度锁失效.     

密度锁界面下方稳态温度场的实验研究及分析

本文通过实验模拟了密度锁内分层界面下方流体稳态温度场分布,发现这部分流体的稳态温度分布曲线存在一个升温结束点,它是导热层与恒温层的分界,只有将升温结束点控制在密度锁内才能有效地抑制传热.本文建立了一个能够描述该部分流体稳态温度分布的数学模型,通过该模型可以给出升温结束点的位置.它为优化密度锁高度提供了理论依据.     

密度锁栅格结构对分层稳定特性影响

对密度锁内流场进行理论分析和实验观察,研究了栅格结构对密度锁分层稳定特性的影响.结果表明:在无分层时,密度锁内为涡旋流动;有分层时,涡旋被限制在分层上方,涡旋与分层相互作用,直至达到平衡;无栅格时,密度锁存在多个平衡点,分层容易从一个平衡点过渡到下个平衡点,很难稳定;有栅格时,分层能较快地达到平衡点,且由于栅格增强了能...     

径向分层碎片床内流动特性研究

为研究非均质结构碎片床内的流动特性,采用两种尺寸颗粒构建了具有径向分层结构的颗粒堆积碎片床,为了对比分析,同时构建了均质结构颗粒堆积碎片床。实验研究了流体在不同堆积结构床内的流动阻力特性,并通过数值模拟揭示了流体在分层床分层界面处的流量再分配现象。研究结果表明,当流体自下而上通过碎片床时,对于均质结构颗粒堆积床,流体呈现一维流动特性;对于具有不同渗透率的径向分层床,除大部分流体自下而上通过分层床外,还存在部分流体从低渗透率层流向高渗透率层,呈现二维流动特性,且绝大部分横流仅发生在分层床的初始部分。     

径向分层碎片床内流动特性研究

为研究非均质结构碎片床内的流动特性,采用两种尺寸颗粒构建了具有径向分层结构的颗粒堆积碎片床,为了对比分析,同时构建了均质结构颗粒堆积碎片床。实验研究了流体在不同堆积结构床内的流动阻力特性,并通过数值模拟揭示了流体在分层床分层界面处的流量再分配现象。研究结果表明,当流体自下而上通过碎片床时,对于均质结构颗粒堆积床,流体呈现一维流动特性;对于具有不同渗透率的径向分层床,除大部分流体自下而上通过分层床外,还存在部分流体从低渗透率层流向高渗透率层,呈现二维流动特性,且绝大部分横流仅发生在分层床的初始部分。     

密度锁内无扰动时稳态温度场分析

实验模拟了密度锁内无扰动时稳态温度场分布。结果发现,稳态温度分布曲线上存在一个温度分层结束点;它是导热层与恒温层的分界,只有当温度分层结束点在密度锁内才能有效地抑制传热。应用半无限大平板导热模型、一维等截面直肋稳态导热模型和Fluent流体计算软件对无扰动时稳态温度场分布进行了理论计算。结果表明,半无限大平板导热模型是计算密度锁内无扰动时稳态温度场分布和温度分层结束点位置的最佳方法。     

密度锁内分区模型研究

通过实验研究流速对密度锁内温度场和分层的影响,并建立了分区模型。研究结果表明:密度锁可分为混合区、分层区和恒温区,其中分层区又可分为强分层与弱分层,分层界面则位于混合区与分层区之间。此外,本文还将密度锁内温度场分为5类,其中第2类温度场最好,是密度锁正常工作时的最佳选择。     

密度锁自平衡启动特性的实验研究

以应用密度锁的非能动余热排出系统为背景,结合实验研究和理论分析,对密度锁自平衡启动的运行特性及可行性进行研究。结果表明：自平衡启动可分为关阀预热和开阀平衡两个阶段,在这两个阶段内能分别实现密度锁“关闭”所必需满足的两个条件,达到隔离非能动余热排出回路与主回路的目的。在关阀预热阶段,密度锁借助其特殊结构,有效地控制了传热方式转变点的位置,促进了热/冷流体交界面的形成;在开阀平衡阶段,借助其水力平衡的自稳定特性,有效地控制了热/冷流体交界面的移动,建立了密度锁内的水力平衡。所得结果充分证明了自平衡启动方法的可行性。     

Dynamic Response of Hot/Cold Liquid Interfaces to Pump Speed Perturbations in a Thermal-Hydraulic Loop Simulating a PIUS-type Reactor

The operation of a PIUS-type reactor requires controlling the reactor pump speed to keep stationary the hot/cold liquid interfaces between the reactor coolant and cold borated water. The dynamic response of the interface location to pump speed perturbations is analyzed for an experimental loop simulating a PIUS-type reactor. The transfer function between the pump speed and the interface location is obtained by perturbing and Laplace-transforming the one-dimensional fluid momentum equations. The analytical results agree well with experimental data taken from the same facility. It is shown that the magnitude of the phase lag in the response of the interface location, which needs to be considered in designing a pump speed controller, primarily depends on the fluid inertia in the loop, the density lock flow area, and the density difference between the simulated reactor coolant and borated water.     

Prediction method for thermal stratification in a reactor vessel

The prediction method for thermal stratification phenomena in a fast breeder reactor is described. The focus of attention is placed on the applicability of water test results to predict thermal stratification phenomena in a real plant. The basic feature of thermal stratification was examined in a cylindrical plenum, using water and sodium as test fluids. The similitude relationship between a small-scale test and a real plant is discussed in order to understand the experimental results. The scale-model experiments for LMFBRs (liquid metal-cooled fast breeder reactors) were also performed to see the effects of a reactor configuration and reactor-trip operation condition. Then the magnitudes of the temperature gradient and the ascending speed of stratified interface in the hot plenum of LMFBRs were predicted, based on the results of the water scale-model.     

Experimental Study on Thermal Stratification in a Reactor Vessel of Innovative Sodium-Cooled Fast Reactor—Mitigation Approach of Temperature Gradient across Stratification Interface—

An innovative sodium-cooled fast reactor has been investigated as part of the fast reactor cycle technology development project (FaCT). Thermal stratification after a scram is one of the main thermal loads of a reactor vessel (R/V). R/V has an upper inner structure (UIS), which consists of perforated horizontal plates and control rod guide tubes, and has a slit in the radial direction for fuel handling. The UIS slit causes an asymmetric flow pattern in R/V. A water experiment using a 1/10-scale model was carried out. A steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. This means that the jet through the UIS slit entrains the bottom of the stratification interface. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at the dipped plates where a fuel handling machine was inserted during a fuel exchange operation, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 21% smaller than that in the case of the higher plug position. This reduction of the temperature gradient was sufficient to maintain the structural integrity of the R/V wall against the thermal stratification.     

管型对波动管热分层现象的影响分析

为分析评价压水堆核电厂稳压器波动管管型对热分层现象的影响,提出采用螺纹管来减弱热分层的措施。利用计算流体力学（CFD）分析方法,对升温、升压阶段波动管原型和改进模型的热分层现象进行数值模拟,得到两种模型不同波动流速下沿波动管轴线方向的截面最大温差分布以及流场分布。对比分析结果表明：波动管结构由光管改为螺纹管后流场紊动加强并出现涡流,冷热流体间的混合增强,与原型相比可使波动管的截面温差减小约1/3,从而有效地减弱热分层的影响。     

Atmospheric-Pressure Small-Scale Thermal-Hydraulic Experiment of a PIUS-Type Reactor

An experimental small-scale low-pressure setup of a PIUS (Process Inherent Ultimate Safety)-type reactor was used for the examination of the stability during normal operation such as startup and load following operation and of the safety during accidents such as loss-of-feed- water and pump runaway. Automatic feedback pump control system based on differential pressure at lower honeycomb density lock was quite effective to maintain the stratified interface between primary and pool water in the honeycomb density lock during normal operation. The process inherent ultimate safety characteristics of the PlUS-type reactor was confirmed with pump-trip scram at the pump speed limit for the various simulated accidents such as a loss-of- feedwater and pump runaway.     

Validation of a Computational Simulation Method for Evaluating Thermal Stratification in the Reactor Vessel Upper Plenum of Fast Reactors

Validation of a numerical simulation method is carried out for thermal stratification phenomena in the reactor vessel upper plenum of advanced sodium-cooled fast reactors. The study mainly focuses on the fundamental applicability of commercial computational fluid dynamics (CFD) codes as well as an inhouse code to the evaluation of thermal stratification behavior including the simulation methods such as spatial mesh distribution and RANS-type turbulence models in the analyses. Two kinds of thermal stratification tests are used in the validation, which is done for relatively simple- and conventional-type upper plenum geometries with water and sodium as working fluids. Quantitative comparison between the simulation and test results clarifies that when used with a high-order discretization scheme of the convection term, the investigated CFD codes are applicable to evaluations of the basic behaviors of thermal stratification and especially the vertical temperature gradient of the stratification interface, which is important from the viewpoint of structural integrity. No remarkable difference is seen in the simulation results obtained using different RANS turbulence models, namely, the standard kε model, the RNG k-ε model, and the Reynolds stress model. It is further confirmed in a numerical experiment that the distribution of two or more meshes within the stratification interface will lead to accurate simulation of the interface temperature gradient with less than 10% error.     

Structural evaluation of a piping system subjected to thermal stratification

Piping systems in nuclear power plants are often designed for pressure, mechanical loads originating from deadweight and seismic events and operating thermal transients using the equations in the ASME Boiler and Pressure Vessel Code, Section III. In the last few decades a number of failures in piping have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. A simplified method has been developed in this work to estimate the stresses caused by the circumferential temperature distribution from thermal stratification. It has been proposed that the equation for the peak stress in the ASME Section III piping code include an additional term for thermal stratification.