乏燃料水池非能动冷却热管传热性能研究

大尺度分离式热管具有无需电力驱动、换热效率高的特点,可用于断电事故后乏燃料水池非能动冷却,能有效提高核电厂的安全性能。针对大尺度分离式热管的传热特性开展实验研究,获得热管蒸发段外侧加热水流速0.007~0.02 m/s,加热水温度50~90℃,冷凝段冷却空气速度0.5~2.5 m/s参数范围下换热量、蒸发段平均换热系数、工作温度、工作压力以及循环流量随冷凝段风速、蒸发段热源进口温度和速度的变化规律。结果表明,大尺度热管的最大换热量达到20.1 kW。参数的敏感性分析表明,热源温度和热源流速对热管的循环流量、换热量具有显著的影响。冷凝段外冷却空气速度超过1.5 m/s后,其对分离式热管的影响相对较小。     

不同管束布置方式下含有空气的蒸汽自然对流冷凝换热特性分析

为研究非能动安全壳冷却系统（PCCS）热交换器管束布置对自然对流条件下含有空气的蒸汽冷凝换热特性的影响,采用气体组分输运方程和冷凝模型耦合,对单管、单排到五排管束通道内冷凝换热过程进行数值研究。研究发现,管束区内存在由于管间高浓度空气层干扰使冷凝换热能力减弱的“抑制效应”,以及由于水蒸气壁面冷凝导致气体横向流动使壁面冷凝能力强化的“抽吸效应”。对不同管束结构下2种效应对冷凝换热的影响进行分析,结果表明,随着管束排数的增加,2种效应对冷凝换热的影响逐渐增强,导致冷凝管周向局部冷凝换热能力不均匀性增加,其中五排管束周向局部冷凝换热系数（HTC）最大值为单管的2.3倍,最小值仅为单管的44.7%。在双排、三排和四排管束中,正四边形布置管束的冷凝换热能力优于正三角形布置,而五排管束中,正三角形布置的冷凝换热能力更强。本研究可对PCCS热交换器管束布置优化提供参考。      

以R134a为工质的乏燃料池非能动冷却热管实验研究

随着分离式热管不断被提出用于核电站非能动余热排出方案中,开展针对大尺度分离式热管的换热性能的实验研究变得日益迫切。为此,本文开展了以R134a为工质的304不锈钢材质的分离式热管传热特性实验研究,获得了热管整体换热性能、蒸发段内部温度分布特性,以及热源温度和冷凝段外风速对热管工作温度、换热量、换热系数和循环流量的影响。热管蒸发段内R134a经历过冷、两相和过热状态,其中两相区域较长,达6.6 m,因而具有较好的换热能力,在所研究的工况下换热量最高达21 kW。参数敏感性分析表明,热源入口温度和冷凝段风速的增大能促进热管的换热性能,特别是热源入口温度的影响更显著。冷凝段风速较小时,其对换热量的影响较为显著,然而随空气速度的增加,影响降低。此外,依据试验数据拟合得到了换热量与冷热源温差的经验关系式,能在工程应用中快速预测热管的性能。     

基于分离式热管换热器的非能动安全壳热量导出系统实验研究

非能动安全壳冷却系统（PCS）是核电厂用以预防和缓解严重事故的重要系统，分离式热管换热器作为一种高效的热交换设备，是其优先设计选项之一。本文介绍了基于分离式热管换热器的PCS原理实验台架的比例设计方法、实验系统和实验结果，分析了热管换热器在特定工况下的换热特性及功率极限，并论证了基于分离式热管换热器的PCS的设计可行性。结果表明：分离式热管单位热端面积换热量可达61 kW，有应用于PCS的潜力；热管的换热性能随冷热端温差的降低而降低，随真空度的提高而升高。     

非能动余热排出换热器运行初始阶段换热特性研究

以非能动余热排出换热器运行初始阶段二次侧水箱水的升温过程为原型,通过实验研究了高位水箱内竖直换热管束在主流水温达到饱和前的换热特性。结果表明,换热管束运行初期热量依靠水的单相自然对流带走,水箱竖直方向上出现温度分层,换热量随主流的升温而下降。随着主流欠热度的减小,从管束上端开始换热机理逐渐向欠热沸腾转变;之后,主流水温逐渐达到饱和,沸腾成为换热的主要手段。在实验研究基础上,利用Churchill&Chu公式从管外平均换热系数中分离出自然对流换热系数,分析了不同阶段自然对流和欠热沸腾在管外换热系数中所占的比例。本文的研究对非能动余热排出换热器的设计有一定的指导意义。     

AP1000乏燃料池非能动冷却系统事故后冷却能力分析

AP1000乏燃料池冷却系统采用了先进的非能动设计理念,事故后以池水升温-沸腾的方式带走衰变热,并通过持续的非能动安全补水保证乏燃料安全。对AP1000乏燃料池冷却系统的事故后冷却能力进行分析发现,在核电厂正常换料工况和应急整堆芯卸载工况下,安全水源重力注水能保证事故后72 h内乏燃料安全;在核电厂正常整堆芯换料过程中应等待约56 h,以保证非能动安全壳冷却水箱可为乏燃料池补水,确保堆芯和乏燃料池安全。长期补水可以通过预留的安全接口持续进行。补水手段事故后有效,仅需操纵员有限干预。相对传统乏燃料池冷却系统设计,AP1000能更好地应对冷却丧失的事件。     

竖直通道内下降液膜蒸发换热的数值研究

下降液膜蒸发换热是CAP1400型反应堆非能动安全壳采用的重要换热机理,准确计算下降液膜蒸发换热量对非能动安全壳换热性能的评价有至关重要的影响。本文利用ANSYS FLUENT软件二次开发,实现了两种下降液膜蒸发换热模型的构建,并将两种模型计算结果与实验结果进行了对比分析。计算结果表明：两种模型均可较为准确地计算壁面下降液膜的蒸发换热系数;模型1的计算结果更加可靠,但在靠近壁面处需非常精细的网格;模型2在壁面处可使用较粗网格,但计算结果对对流换热系数的依赖较大。     

中低雷诺数条件下单侧加热矩形通道内混合对流换热的实验研究

非能动安全壳冷却系统是先进压水堆非能动安全系统的重要组成部分,其中空气对流换热的能力较差,对安全影响较大,因此本文主要研究了在大尺寸垂直单侧加热矩形通道内空气自下而上流动时的混合对流换热,用于模拟核电厂非能动安全壳冷却系统的换热情况。研究结果表明在较小雷诺数条件下自然对流的影响不能忽略且自然对流会占据主导作用;随着空气流量的增加,强迫对流换热的作用越来越明显。当前学者所用经验关系式都不能很好地体现出自然对流在混合对流中起的作用,因此本文还通过实验数据拟合了一个新的计算混合对流换热的关系式,该公式在一定雷诺数范围内与实验值能很好地符合。     

有内热源的液池与管内流体的耦合传热

对均匀内热源溶液池内的自然对流及冷却盘管内强制对流的耦合传热过程进行了分析,建立了相应的三维物理数学模型,并采用有限容积法进行了数值求解.针对不同容积热源强度以及不同管内流速下溶池内自然对流与冷却管内流体的耦合传热过程进行了数值计算.计算得到的管内外流体对流换热系数以及溶液池与管内流体的总传热系数与理论值以及实验测试结果吻合较好,表明了本文所建立数学模型的正确性.研究结果揭示了溶液池内容积热源强度及管内流速对总传热系数有显著的影响.     

高温热管内钠蒸发冷凝特性分子动力学研究

高温钠热管是热管堆中进行非能动热量传输的核心部件。为深入理解热管内工质钠的蒸发机理及气液交界的传热传质特性，用分子动力学软件LAMMPS模拟了600 K下钠的蒸发，统计了质量调节系数，定为0.388 7。随后变更壁温，打破体系内热质输运平衡，进行非平衡态模拟，观察液膜变化，求解气液交界处的净蒸发通量和换热系数。结果表明，9～10 ns后，底部的液膜厚度、气液交界处的净蒸发通量及换热系数分别在0.1～0.52 nm、0.03～0.07 kg/(m~2·s)、2.2～3.9 kW/(m~2·K)范围波动，此时上部液膜厚度在6 nm左右，其气液交界的净蒸发通量在10^-4量级，换热系数为0.028 kW/(m~2·K),至末期降为0.003 5 kW/(m~2·K)。本文为钠热管启动阶段的数值模拟提供了参考。     

Thermal-hydraulic design and transient analysis of passive cooling system for CPR1000 spent fuel storage pool

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool(SFP). Our design can effectively manage the SFP temperature not to exceed80 C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout(SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively.The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 C after losing the heat sink.     

Experimental study on the sub-atmospheric loop heat pipe passive cooling system for spent fuel pool

As one kind of the natural circulation cooling system, loop heat pipe is promising in improving the safety of the nuclear power station since it is passive and has no electricity driven components. A novel heat pipe cooling system is designed for passively removing the residual heat released by the spent fuel stored in the spent fuel pool (SFP) under the accidental conditions such as the station blackout. This system is characterized by its large-diameter and long-length evaporator. Its working fluid is water and it's sub-atmospheric. To test such system's heat transfer performance and get to know its thermo-fluid dynamics, a test facility for a simplified heat pipe made of one evaporator tube and one condenser has been developed. The heat transfer rate of the simplified heat pipe is obtained in a wide range of conditions covering the potential working conditions in spent fuel pool. Moreover, it's found that heat pipe with such a large-diameter and long-length evaporator is vulnerable to be unstable. The periodic state mode is more likely to happen when the heat source temperature, the air velocity or the volumetric filling ratio is low. Furthermore, the effects of hot water temperature, the air velocity and the filling ratio of the water in the circulation system have been analyzed.     

AP1000外部灾害情形下乏燃料池缓解策略研究

日本福岛核事故后,乏燃料池（SFP）在事故中的安全性得到广泛的关注。AP1000乏燃料池冷却系统（SFS）是一非安全相关的系统,不需在事故后运行以缓解设计基准事故。但乏燃料池在超设计基准事故或外部灾害事件（包括自然灾害和人为事件）下的安全性一直是核电厂设计的重点。本工作结合美国核能研究所（NEI）给出的扩大损害的缓解导则（EDMG）提出了针对AP1000外部灾害情形下的SFP缓解策略（包括内部策略和外部策略）,并对策略进行了评估。本工作结论有助于AP1000 SFP EDMG的建立,对AP1000核电厂的设计、建造、运行管理和事故管理均有很强的参考价值。     

Experimental Study on Heat Removal Characteristics for Water Wall Type Passive Containment Cooling System

To evaluate the heat removal capability of a water wall type cooling system, which is one passive containment cooling system (PCCS), the thermal hydraulic behavior in the suppression pool (S/P) and the outer pool (O/P, flat plate water wall) have been investigated experimentally. The following results were obtained. (1) A thermal stratification boundary, which separates the pools into the upper high temperature and lower low temperature regions, was formed just below the vent tube outlet. (2) Convection heat transfer characteristics in the S/P and O/P along the primary containment vessel (PCV) wall had no significant differences and were those of natural convection. Correlation of the natural convection heat transfer up to the Ra number of 2×10¹⁴ was obtained. (3) Vertical variations of local condensation heat transfer coefficients under a noncondensable gas presence were within ±10% of the average value for the 4.7 m heat transfer length. An experimental correlation for the average condensation heat transfer coefficients was obtained as a function of steam and noncondensable gas mass ratio. (4) An analytical model to evaluate the system performance of the water wall type PCCS was verified. (5) A baffle plate concept to mitigate thermal stratification at the vent outlet and to enlarge the high temperature region in the S/P was considered as a means to improve heat release capability. Thermal hydraulics with a baffle plate were examined, and effectiveness of the baffle plate to improve the heat release capability was confirmed.     

Prediction of temperature and water level in a spent fuel pit during loss of all AC power supplies

A prediction method for water temperature in a spent fuel pit of a pressurized water reactor (PWR) has been developed to calculate the increase in water temperature during the shutdown of cooling systems. In this study, the prediction method was extended to calculate the water level in a spent fuel pit during loss of all AC power supplies, and predicted results were compared with measured values of spent fuel pools in the Fukushima Daiichi Nuclear Power Station. The calculations gave reasonable results, but overestimated the decreasing rate of the water level and the water temperature. This indicated that decay heat was overestimated and evaporation heat transfer from the water surface was underestimated. Results of calculations with 80% decay heat and 155% (Unit 4 pool) or 230% (Unit 2 pool) evaporation heat flux were in good agreement with measured values. The data-fitted evaporation heat fluxes agreed rather well with the evaporation heat transfer correlation proposed by Fujii et al.     

不凝性气体对高温锂热管传热特性的影响研究

热管作为一种具有高热导率的传热装置,工作核心在于其内部工作流体的蒸发和冷凝。若热管工作过程中气腔内存在不凝性气体,主流区中蒸气和不凝性气体在对流运动的作用下将一起移动到气-液分界面,不凝性气体的存在阻碍了工作流体在气-液交界面处的正常冷凝。本文基于热阻网络法添加了不凝性气体区域传热模型,研究了不凝性气体对高温锂热管稳态传热特性的影响。结果表明,热管达到稳态时不凝性气体的存在缩短了热管的有效传热长度,破坏了热管的等温性和良好的传热效率。此外随着不凝性气体体积份额的增大,不凝性气体区域温度降低幅度越大;随着热管蒸发段输入功率的增大,热管正常工作区域整体温度越高,相同质量的不凝性气体占据的体积份额越小,热管壁面温度出现明显温度梯度降低的位置随着功率升高而向下游移动。     

恰希玛核电厂乏燃料自然循环冷却分析

利用ＲＥＴＲＡＮ－０２程序对乏燃料自然循环冷却进行了分析。计算结果表明，恰希玛核电厂乏燃料池冷却系统失效后，只要在１９个小时内修复冷却系统，不会出现大量放射性物质外泄。     

Analyzing the alternative shutdown cooling behaviors for Chinshan Nuclear Power Plant using CFD simulation

The Chinshan Nuclear Power Plant (CSNPP) is a GE-designed BWR4 plant, having two identical units with rated core thermal power of 1804 MWt each unit. Several alternative shutdown cooling methods driven by natural or mixed convection has been proposed by the plant for studying the core cooling capability when the Residual Heat Removal (RHR) systems are not available or the refueling tasks, such as the In Vessel Visual Inspection (IVVI) work etc., is being proceeded. One of the examples is to connect a pipe from the outlet of the new spent fuel heat exchanger to the reactor cavity. The design of the alternatives shall ensure that the maximum core fluid temperature is limited below the boiling temperature of water. In this study, a Computational Fluid Dynamics (CFD) model is developed to analyze the natural convection phenomena during the shutdown period. Through a series of assumption, modeling and meshing processes, a calculation domain with approximate four million meshes including the RPV, reactor cavity and spent fuel pool, have been solved in this study. The analysis results showed that the passive alternative shutdown cooling system could provide sufficient heat removal capability to maintain the maximum core fluid temperature below boiling temperature. The results also indicated that the alternative shutdown cooling system could be served as an appropriate solution for CSNPP when the RHR is inoperable.     

换热器类型在核电厂中的应用研究

管壳式换热器是目前压水堆核电厂中普遍采用的换热器。在保证核电厂安全性的基础上,还需要进一步提高其经济性,因此选择反应堆水池和乏燃料水池冷却和处理系统的冷却水热交换器为研究对象进行计算比较,然后利用各种形式换热器的传热计算公式,并用Fortran语言编制程序进行计算,根据计算结果绘图比较不同形式换热器的传热系数及传热面积。通过比较分析得出在核电厂中,板式换热器具有结构紧凑、质量轻、换热效率高等可能的优点。