竖直热壁冷却下降液膜瞬变传热及相界面波动

采用数值方法对恒热流密度竖直热壁表面自由下降液膜冷却流动过程进行瞬态研究,建立了简化为二维问题的物理模型和数学模型,选用RNG k-ε模型求解汽液两相湍流流动,基于VOF多相流模型结合自编程用户定义函数求解相变过程中的传热传质。得到了不同因素影响下,冷却降膜厚度在热壁不同位置的分布规律。结果显示,冷却降膜的流动过程中,其表面存在不同程度及类型的波动。热扰动会降低液膜相界面流动稳定性,提高液膜流量能够抑制液膜表面波动,两种影响因素对于液膜的流动稳定性是相互抑制的。核态沸腾发生时,液膜内汽泡的运动、变化会对液膜相界面波动产生显著影响。瞬态计算结果显示,不同工况下,降膜冷却传热过程中表面传热系数具有相同的变化趋势。液膜的热稳定性受到入口液流量及壁面热流密度的耦合影响。根据汽液两相流动及传热传质特性,分析了发生以上现象的原因。     

含不凝气水平管内膜状冷凝的传热特性

为深入研究含不凝气水平管内膜状冷凝的传热特性,建立了含不凝气的水平管内膜状冷凝传热的数学模型。模型考虑了汽液两相流间的剪切应力引起的轴向压力梯度及不凝气在相界面处的扩散传质热阻,推导出液膜换热系数、液膜厚度、轴向压力梯度、冷凝液层角度及厚度和冷凝界面温度的理论关系式。通过对纯蒸汽冷凝液膜特性数学模型计算结果间的比较,验证了数学模型的通用性和准确性。研究结果表明:汽液两相流间剪切应力和不凝气对沿管长方向蒸汽冷凝温度的降低有显著影响,要通过增加管长来抵消其对传热性能降低的影响;进口蒸汽中不凝气浓度的降低,有利于减少传热管管长。     

蒸发冷却技术在环网控制装置上应用的可行性研究

针对环网控制换流阀,充分考虑其冷却需求和运行环境,开展了蒸发冷却技术的应用研究工作,提出了贴壁式自循环蒸发冷却技术方案;针对阀柜的功率模块单元设计蒸发冷却液盒单元,并在阀体额定工况下对液盒的换热能力、控温水平进行实验考核;同时建立液盒内气液两相流模型,对功率模块单元冷却过程中的热场进行了仿真计算;实验测试和仿真计算结果显示贴壁式蒸发冷却技术方案在环网控制装置上应用是可行的,且具有换热效率高、温度分布均匀的优势。     

蒸发冷却汽轮发电机定子绝缘结构的模拟试验及分析

本文针对蒸发冷却汽轮发电机的发展现状，介绍汽轮发电机定子绝缘改进的可行性方案以及汽、液、固相绝缘配合的实验模型和数据，为蒸发冷却汽轮发电机向更大容量发展、探索新的绝缘体系、充分发挥蒸发冷却的效益提供有价值的技术参考。     

全浸式液汽相变冷却方式贴片电阻失效机理

随着电力电子技术的发展,开关电源作为供电装置广泛应用于通信、能源、航空航天等领域,高功率密度、高可靠性是其发展方向。因过热问题引发故障继而导致可靠性降低成为开关电源功率密度提升的瓶颈。全浸式液汽相变冷却技术冷却效率高、安全可靠,是实现开关电源高效散热的新途径。对于全浸式液汽相变冷却开关电源,相变冷却的工作介质与电源器件直接接触,电源中与高电压相连的贴片电阻发生了阻值上升甚至断路现象,影响开关电源的稳定运行。该文通过宏观信号监测和微观材料分析手段,得到了工作介质环境下贴片电阻失效的外部影响因素及内在机理,明确了可在工作介质环境下稳定工作的贴片电阻结构特征。研究成果可为全浸式液汽相变冷却开关电源提供器件选型及设计指导,对提高全浸式液汽相变冷却开关电源运行可靠性,完善液汽相变冷却技术应用于电力电子装备的理论体系具有重要的理论意义及应用价值。     

蒸发冷却电机中汽液两相沿程摩擦阻力的研究

该文旨在提高蒸发冷却电机中冷却介质两相流动的沿程摩擦阻力的计算精度。用5种计算方法(2种均相模型计算方法,L-M方法,Chisholm方法和Friedel方法)计算冷却介质两相流动的摩阻放大倍数和沿程摩擦阻力;以李家峡400兆瓦蒸发冷却电机为参照,搭建起两相流实验台,在0.146～0.152MPa压力范围内对冷却介质的两相流动进行了实验研究。在2.59、3.45、4.3g/s3种流量条件下比较实验数据与计算结果后发现,基于分相模型的L-M方法对放大倍数和沿程摩擦阻力的预测结果最准确,最大误差不超过15%,根据冷却介质的流量修正系数C后,计算精度大大提高,误差不超过6%。这说明修正后的L-M方法可以应用于蒸发冷却电机的设计计算,为冷却系统的循环计算奠定了理论基础。     

蒸发冷却介质两相流状态下工频击穿特性的实验研究

一些不燃型氟碳化合物具有非常好的绝缘特性,既可以作为蒸发冷却变压器等电气设备的冷却介质,也可以作为绝缘介质,使得蒸发冷却变压器不燃不爆、安全可靠、经济高效、绿色环保、成本适中,适用于某些特定场合的需求。在作为绝缘介质实际运行过程中,氟碳化合物将以气液两相流状态存在。为研究该工况下介质的绝缘性能,本文参考国家标准研制了实验装置,测量了若干种氟碳化合物介质在两相流状态下的工频击穿电压特性。实验采用3mm间隙的板-板电极,得到了不同工作压力下的工频击穿实验数据,以及击穿电压与工作压力的经验公式。实验结果的统计分析显示两相流状态下工频击穿数据很好地服从于威布尔分布和正态分布,通过计算得到了各工作压力点下击穿电压值的概率分布函数,为蒸发冷却变压器绝缘结构的设计提供了参考。     

分体式真空锅炉汽液两相流动及压降特性实验研究

通过搭建可视化热态实验台,对分体式真空锅炉汽液两相流动及压降特性进行了研究。采用压差法对上升管处的质量含汽率进行测量,进而根据所测量的含汽率对锅筒内的不同压降进行计算分析。结果表明:下锅筒热负荷分布不均、上锅筒蒸汽空间压力波动及其引发的过冷沸腾、工质内部压力波动等均会对锅炉运行过程中的稳定性造成影响。在锅炉内部汽液两相压降中,重位压降是最主要的部分,约为其他形式压降的50~100倍。锅炉内部汽液两相压降、含汽率、循环质量流量以及传热系数等参数之间相互影响,密不可分。在实验结果与理论分析基础上提出了分体式真空锅炉的优化建议。     

汽液两相流自调节液位控制器的应用

应用汽液两相流自调节液位控制器控制连排扩容器水位，解决了锅炉连续排污扩容器水位控制不稳定的问题，实现了二次蒸汽的全部回收，提高了机组的经济性。     

蒸发冷却电机管内气液两相流截面含气率测量的研究

本阐述了蒸发冷却电机管内气液两相流截面含气率测量的重要性及其测量的特点,分析了各种测量方法的优缺点．并对静态气液两相介质进行了射线吸收法的实验研究,认为采用射线吸收法测量蒸发冷却电机管内气液两相流出口处截面含气率是可行的,并且可以获得较高的精度。     

Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code

Experiments with impulse gas injection into model coolants, such as water or the Rose alloy, performed at the Novosibirsk Branch of the Nuclear Safety Institute, Russian Academy of Sciences, are described. The test facility and the experimental conditions are presented in details. The dependence of coolant pressure on the injected gas flow and the time of injection was determined. The purpose of these experiments was to verify the physical models of thermohydraulic codes for calculation of the processes that could occur during the rupture of tubes of a steam generator with heavy liquid metal coolant or during fuel rod failure in water-cooled reactors. The experimental results were used for verification of the HYDRA-IBRAE/LM system thermohydraulic code developed at the Nuclear Safety Institute, Russian Academy of Sciences. The models of gas bubble transportation in a vertical channel that are used in the code are described in detail. A two-phase flow pattern diagram and correlations for prediction of friction of bubbles and slugs as they float up in a vertical channel and of two-phase flow friction factor are presented. Based on the results of simulation of these experiments using the HYDRA-IBRAE/LM code, the arithmetic mean error in predicted pressures was calculated, and the predictions were analyzed considering the uncertainty in the input data, geometry of the test facility, and the error of the empirical correlation. The analysis revealed major factors having a considerable effect on the predictions. The recommendations are given on updating of the experimental results and improvement of the models used in the thermohydraulic code.     

System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment

The system of equations from a two-fluid model is widely used in modeling thermohydraulic processes during accidents in nuclear reactors. The model includes conservation equations governing the balance of mass, momentum, and energy in each phase of the coolant. The features of heat and mass transfer, as well as of mechanical interaction between phases or with the channel wall, are described by a system of closing relations. Properly verified foreign and Russian codes with a comprehensive system of closing relations are available to predict processes in water coolant. As to the sodium coolant, only a few open publications on this subject are known. A complete system of closing relations used in the HYDRA-IBRAE/LM/V1 thermohydraulic code for calculation of sodium boiling in channels of power equipment is presented. The selection of these relations is corroborated on the basis of results of analysis of available publications with an account taken of the processes occurring in liquid sodium. A comparison with approaches outlined in foreign publications is presented. Particular attention has been given to the calculation of the sodium two-phase flow boiling. The flow regime map and a procedure for the calculation of interfacial friction and heat transfer in a sodium flow with account taken of high conductivity of sodium are described in sufficient detail. Correlations are presented for calculation of heat transfer for a single-phase sodium flow, sodium flow boiling, and sodium flow boiling crisis. A method is proposed for prediction of flow boiling crisis initiation.     

Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors

Closing relations describing friction pressure drop during the motion of two-phase flows that are widely applied in thermal-hydraulic codes and in calculations of the parameters characterizing the flow of water coolant in the loops of reactor installations used at nuclear power stations and in other thermal power systems are reviewed. A new formula developed by the authors of this paper is proposed. The above-mentioned relations are implemented in the HYDRA-IBRAE thermal-hydraulic computation code developed at the Nuclear Safety Institute of the Russian Academy of Sciences. A series of verification calculations is carried out for a wide range of pressures, flowrates, and heat fluxes typical for transient and emergency operating conditions of nuclear power stations equipped with VVER reactors. Advantages and shortcomings of different closing relations are revealed, and recommendations for using them in carrying out thermal-hydraulic calculations of coolant flow in the loops of VVER-based nuclear power stations are given.     

核主泵惰转转速计算模型的比较

针对核主泵在惰转过程中还要求有足够的冷却流量,利用核主泵瞬态动量守恒方程进行计算和分析,对模型做了合理简化,得到新的转速模型,验证了该模型精准度。结果表明:该模型完全可用于核主泵瞬态分析,可得到不同转动惯量下核主泵达到半流量时惰转时间的变化规律。     

Some features of the effect the pH value and the physicochemical properties of boric acid have on mass transfer in a VVER reactor’s core

Certain features of the effect of boric acid in the reactor coolant of nuclear power installations equipped with a VVER-440 reactor on mass transfer in the reactor core are considered. It is determined that formation of boric acid polyborate complexes begins under field conditions at a temperature of 300°C when the boric acid concentration is equal to around 0.065 mol/L (4 g/L). Operations for decontaminating the reactor coolant system entail a growth of corrosion product concentration in the coolant, which gives rise to formation of iron borates in the zones where subcooled boiling of coolant takes place and to the effect of axial offset anomalies. A model for simulating variation of pressure drop in a VVER-440 reactor’s core that has invariable parameters during the entire fuel campaign is developed by additionally taking into account the concentrations of boric acid polyborate complexes and the quantity of corrosion products (Fe, Ni) represented by the ratio of their solubilities.     

Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor

Numerical simulation of the ULOF-type accident development in a fast reactor with sodium coolant performed using the COREMELT code indicates that sodium boiling in the active core takes place. The boiling is accompanied by oscillations of the technological parameters of the reactor installation; these oscillations can go on during several tens of seconds. In this case, it is possible that a stable regime of removal of heat from residual energy release is implemented. The model of the two-phase coolant flow applied in the code has an important effect on the numerical results; that is why this model needs experimental verification. For eliminating the development of an accident resulting in destruction of the active core elements, a structural solution is proposed; the essence of it is the application of the sodium void above the reactor active core. The experimental installation was developed and the heat exchange at sodium boiling in the model fuel assembly of the fast reactor in the regimes of natural and forced circulation in the presence of the sodium void and the top end shield was studied. It was demonstrated that, in the presence of the sodium void, it is possible to provide long-term cooling of the fuel assembly for a thermal flux density on the fuel element simulator surface of up to 140 and 170 kW/m² in the natural and forced circulation modes, respectively. The obtained data are used for more precise determination of the numerical model of sodium boiling in the fuel assembly and verification of the COREMELT code.     

Verification of the computer code KORSAR taking into account the effect of nondensables on thermal-hydraulic processes

Calculations to verify the Russian computer code KORSAR were carried out for the B4.1 experimental operating conditions, in which nitrogen was supplied to the reactor coolant (primary) circuit of a reactor plant model, and which were simulated at the PKL III integral test facility. It is shown that dissolution of gases in coolant has an essential effect on the thermal-hydraulic processes during long-term passive removal of heat from the primary to secondary coolant circuit of the reactor plant model under the conditions of natural circulation.     

Dielectric constant as one of the main parameters used to monitor the state of working fluid

It is shown that flow density must be estimated along with measuring the temperature (or pressure) for monitoring the state of two-phase medium. To this end, it is proposed to use a dielectric constant of coolant. A simple single-parameter relation is proposed, the use of which makes it possible to calculate the density of water and steam with the known values of temperature and dielectric constant in a concrete part of heat-transfer equipment.     

汽轮发电机定子强迫蒸发冷却循环流动与传热计算

基于C语言编程软件和气液两相流相关理论,在忽略绕组线棒辐射散热的情况下,对汽轮发电机定子采用强迫蒸发冷却循环时的流动与传热进行了编程计算,得到了汽轮发电机定子的蒸发点、循环流量、压力和温度分布。结果表明,随着绕组线棒热负荷增加,蒸发点提前,循环流量减小,压力平衡器前的含气率增加;当含气率增加到某一临界值时,可能造成危险工况,因此必须通过计算确定临界热负荷。     

The influence of operational and water chemistry parameters on the deposits of corrosion products on fuel assemblies at nuclear power plants with VVER reactors

The phenomenon involving a growth of pressure drop in the reactor core and redistribution of deposits in the reactor core and primary coolant circuit of a nuclear power station equipped with VVER-440 reactors is considered. A model is developed, the physicochemical foundation of which is based on the dependence of corrosion product transfer on the temperature and pH _t value of coolant and on the correlation between the formation rate of corrosion products (Fe) (after subjecting the steam generators to decontamination) and rate with which they are removed from the circuit. The purpose of the simulation carried on the model is to predict the growth of pressure drop on the basis of field data obtained from nuclear power installations and correct the water chemistry (by adjusting the concentrations of KOH, H₂, and NH₃) so as to keep the pressure drop in the reactor at a stable level.