圆管通道底部再淹没过程实验研究

为验证和优化再淹没模型,通过实验研究了圆管通道内再淹没阶段流动换热特性,获得了不同工况下壁面温度的变化规律,实验工况范围为：入口冷却剂流速3~15 cm/s、入口过冷度15~75 ℃、初始壁面峰值温度340~600 ℃、实验压力0.2~0.4 MPa、加热功率1.3~2.3 kW/m。分析了初始壁温、冷却剂入口温度、入口流速及加热功率对骤冷时刻与骤冷温度的影响。结果表明,骤冷时刻与骤冷温度均随初始壁温、冷却剂入口温度以及加热功率的增加而增加,随入口冷却剂流速的增加而减小。     

Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of droplets after collision with a wet grid spacer depending on flow conditions.     

Flooding limited CHF in a vertical 3 × 3 rod bundle with non-uniform axial heat flux

KAERI has performed an experimental study on the critical heat flux (CHF) under zero flow conditions with a non-uniformly heated 3 × 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.50 to 14.96 MPa and inlet water subcooling enthalpies from 68 to 352 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 × 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a good parametric trend. The CHFs occur in the upper region of the heated section, but the locations of the detected CHFs move gradually in a downward direction with the increase of the system pressure. Even though the effects of the inlet water subcooling enthalpies and system pressure of the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations.     

Experiment and numerical simulation of bubbly two-phase flow across horizontal and inclined rod bundles

Experimental and numerical analyses were carried out on vertically upward air-water bubbly two-phase flow behavior in both horizontal and inclined rod bundles with either in-line or staggered array. The inclination angle of the rod bundle varied from 0 to 60° with respect to the horizontal. The measured phase distributions indicated non-uniform characteristics, particularly in the direction of the rod axis when the rods were inclined. The mechanisms for this non-uniform phase distribution is supposed to be due to: (1) Bubble segregation phenomenon which depends on the bubble size and shape; (2) bubble entrainment by the large scale secondary flow induced by the pressure gradient in the horizontal direction which crosses the rod bundle; (3) effects of bubble entrapment by vortices generated in the wake behind the rods which travel upward along the rod axis; and (4) effect of bubble entrainment by local flows sliding up along the front surface of the rods. The liquid velocity and turbulence distributions were also measured and discussed. In these speculations, the mechanisms for bubble bouncing at the curved rod surface and turbulence production induced by a bubble were discussed, based on visual observations. Finally, the bubble behaviors in vertically upward bubbly two-phase flow across horizontal rod bundle were analyzed based on a particle tracking method (one-way coupling). The predicted bubble trajectories clearly indicated the bubble entrapment by vortices in the wake region.     

矩形窄缝通道底部再淹没实验研究

为了解矩形窄缝通道在失水事故（LOCA）下底部再淹没过程中的热工水力特性,在不同实验条件下开展再淹没实验研究。矩形窄缝通道由2块因科镍合金焊接而成,本研究根据温度变化曲线分析底部再淹没过程,计算并对比不同实验工况下的骤冷前沿的推进速度（骤冷速度）,以及研究实验参数对再淹没过程的影响。实验结果表明,底部再淹没骤冷速度随着系统压力增大、进口流速增大、初始壁面温度降低以及冷却水过冷度的增大而增大。对比分析底部与联合再淹没工况,结果表明流量相同的情况下,底部再淹没的骤冷速度大于联合再淹没。本文研究为板状燃料元件反应堆事故预防以及事故缓解等研究奠定了基础。      

5×5花瓣形燃料棒束组件内单相流动与换热特性数值模拟研究

相较于传统圆柱形燃料棒,花瓣形燃料棒具有安全裕量高等优点,研究其在压水堆运行工况下的热工水力特性具有重要意义。本文通过STAR-CCM+对5×5花瓣形燃料棒束组件进行数值模拟研究,计算并分析了组件内二次流速度、温度、换热系数等关键热工参数,获得了入口流速、螺旋节距对组件内部流动与换热特性的影响规律。计算结果表明：花瓣形燃料棒的螺旋结构可增强冷却剂的横向流动,同一高度上燃料棒表面温度分布具有周期性,增大入口流速可增强燃料棒的表面换热,消除温度分布的不均匀性。此外,螺旋节距大于750 mm,燃料棒换热性能与无扭转的燃料棒相差不大,甚至更低。     

棒束通道内过冷沸腾起始点的实验研究

本文以去离子水为实验介质,在进口温度80～100 ℃、质量流速0～100 kg/(m²•s)、热流密度0～80 kW/m²的条件下对棒束通道内的过冷沸腾起始点（ONB）进行了实验研究。分析了部分热工参数和棒束特殊的几何结构对ONB的影响,通过引入雷诺数,对棒束通道内ONB的数据进行非线性回归分析,得到适用于棒束通道ONB的经验关系式。结果表明：新拟合得到的关系式能较准确地预测棒束通道内ONB的热流密度,其预测值的相对误差为14.75%。     

AP1000中ADS-4液体夹带模型研究

研究了由AP1000核电厂一回路热管段和ADS-4管道组成的大尺寸支管T型管液体夹带的实验和理论模型,主要包括起始夹带和稳态夹带模型的研究。通过实验和理论分析,建立了与实验数据符合良好的起始夹带模型和稳态夹带模型。通过分析实验段入口长度对夹带的影响,确定了合适的入口段长度。不同液体流量下的研究表明,ADS-4管道中,液体流量对起始夹带和稳态夹带的影响可忽略。     

环形燃料棒束再淹没行为实验研究

环形燃料是一种可在维持或提升安全裕度的前提下大幅提高反应堆经济效益的新型压水堆燃料,由于其双面冷却的特点,环形燃料在LOCA再淹没阶段的热工水力行为与传统实心燃料存在显著差异。现有关于环形燃料再淹没行为的实验研究鲜有报道。本研究基于自主设计的高温环形电加热棒建立了环形棒束再淹没实验装置,开展了3×3环形棒束底部再淹没实验研究,探究了环形棒束再淹没典型物理过程及不同工况下再淹没关键参数的变化规律。结果表明,环形棒束再淹没物理过程与传统实心棒束类似,且内外通道的骤冷前沿推进和传热模式变化趋于同步。在同一时刻下,环形棒内外壁面间存在温度梯度。骤冷前沿推进速度随再淹没速度和过冷度的增大而增大,随峰值包壳温度和线功率密度的增大而减小。此外,定位格架在低流速、低过冷度与高壁温工况下能显著提升下游的骤冷前沿推进速度。     

AP1000中ADS-4液体夹带实验研究

以AP1000核电厂中自动泄压管线(ADS-4)与热管段形成的T型结构为研究对象,开展缩小比例的T型管夹带实验。实验结果表明:大尺寸支管的夹带与小尺寸支管的夹带有明显差别。分层流情况下的夹带研究中发现两种夹带机理;在较低气相Froude数及较低气腔高度时,容易产生间歇流夹带;在较高气相Froude数情况下,往往出现环状流夹带。此外,实验研究发现,大尺寸支管中回流现象显著。支管形状对起始夹带有重要影响,而液体横流似乎并不影响起始夹带。     

环形通道内再淹没过程流动传热现象实验研究

通过可视化实验手段观察了环形通道内再淹没过程两相流动现象,分析总结了再淹没骤冷前沿推进过程中流型和传热机理的演化规律;通过不同工况下两相流动现象的对比,研究了是否加热和入口质量流速对再淹没过程流型和传热过程的影响规律。研究结果表明,在本参数范围内,实验中加热棒是否存在内释热对两相流动现象的影响不显著;而入口质量流速明显影响再淹没流动传热过程,入口质量流速越大,骤冷前沿附近汽化越剧烈,液膜中汽泡含量增加,更容易发生传热机制的转变。      

棒束通道内气液两相流流型的实验研究

常温、常压条件下,在7×7矩形截面棒束通道内进行了垂直向上气液两相流动实验,气液两相折算速度的变化范围分别为0.04~14 m/s和0.238~1.860 m/s。实验中用高速摄像仪对流型进行记录,观察到了泡状流、泡状-搅混流、搅混流和搅混-环状流4种流型,发现搅混流是主要流型,并对Hewitt流型图的界限进行了修正。分析实验数据发现,摩擦压降在泡状流和搅混流区域的变化是相反的。根据实验数据,参考前人的研究得到棒束通道中泡状流向搅混流转变的边界。     

Reflooding experiments with LWR-type fuel rod simulators in the QUENCH facility

The QUENCH experiments at the Karlsruhe Research Center are carried out to investigate the hydrogen generated during reflooding of an uncovered Light Water Reactor (LWR) core. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in PWRs (Pressurized Water Reactors) with respect to material and dimensions. Pellets are made of zirconia to simulate UO₂. After a transient phase with a heating rate of 0.5–1 K s⁻¹ water of approx. 395 K is admitted from the bottom when the test bundle has reached its pre-determined temperature. Except for the flooding (quenching) phase, the QUENCH test phases are conducted in an argon/steam atmosphere at 3 g s⁻¹ each. The results of the first two experiments, QUENCH-01 (with pre-oxidation of 300 μm oxide layer thickness at the cladding outside surface) and QUENCH-02 (reference test without pre-oxidation), are compared in the paper. The pre-oxidized LWR fuel rod simulators of QUENCH-01 were quenched from a maximum temperature of 1870 K. In the second bundle experiment, QUENCH-02, quenching started at 2500 K. Pre-oxidation apparently prevented a temperature escalation in the QUENCH-01 test bundle, while the QUENCH-02 test bundle experienced a temperature excursion which started in the transient phase and lasted throughout the flooding phase. The different behavior of the two experiments is also reflected in hydrogen generation. While the bulk of H₂ was produced during pre-oxidation of test QUENCH-01 (30 g), the largest amount, i.e. 170 g, of hydrogen was generated during the reflooding phase of test QUENCH-02, at a maximum production rate of 2.5 g s⁻¹ as compared to 0.08 g s⁻¹ in test QUENCH-01. Similarities between the two experiments exist in the thermo-hydraulics during the quench phase, e.g. in the cooling behavior, the quench temperatures, and quench velocities.     

The top-down reflooding model in the CATHARE code

A top-down reflooding model was developed for the French best-estimate thermal hydraulic code cathare. The paper presents the current state of development of this model. Based on a literature survey and on compatibility considerations with respect to the existing cathare bottom reflooding package, a falling film top-down reflooding model was developed and implemented into cathare version 1.3U. Following a brief review of previous work, the paper describes the most important features of the model. The model was validated with the Winfrith single-tube top-down reflooding experiment and with the REWET-II simultaneous bottom and top-down reflooding experiment in a rod bundle geometry. The results demonstrate the ability of the new package to describe the falling film rewetting phenomena and the main parametric trends both in a simple analytical experimental set-up and in a much more complex rod bundle reflooding experiment.     

Void Fraction in Simulated PWR Fuel Bundle during Reflood Phase

In order to evaluate void fraction in a bundle geometry during the reflood phase, reflooding experiment with a 4×4 simulated fuel array was conducted.

As the result, it was found that the effects of the clad temperature and the power of the heater rods are small and the effects of the pressure and the inlet flow rate are large on the relationship between the superficial steam velocity and the void fraction in a bundle geometry during the reflood phase. It was, also, found that there is no distinct difference of the void fractions caused by the different flow patterns in the wet clad region and in the dry clad region in a bundle geometry during the reflood phase, when compared at the same superficial steam velocities.

Furthermore, the applicability of Cunningham-Yeh's void fraction correlation was investigated under a wide range of conditions anticipated during the reflood phase. The range of conditions under which Cunningham-Yeh's correlation predicts the void fraction within an error band of ±20% were made clear.     

Experimental and numerical simulation of air-water two-phase flow in the rod bundle with grid spacer

1 Introduction Grid spacer is the key part of reactor fuel assem-bly. The presence of spacers in fuel assemblies affectsvarious thermal-hydraulic characteristics of the reactorcore. The grid spacer with fine performance can im-prove thermal-hydraulic performance of the core fuelassembly and enhance the critical heat flux withouttoo much augment of the pressure loss. As a result,the implementation of grid spacer with high thermalperformance provides more thermal margin, then in-creases s…     

脉动流下棒束通道内相位差及瞬态流场研究

事故条件及海洋条件下反应堆处于非稳态工况,堆芯燃料组件内热工水力行为具有瞬变及多因素耦合特性,对反应堆的安全提出更高挑战,因此有必要对燃料组件内瞬态特性进行研究。本文通过测量棒状燃料组件内压降和流量之间延迟时间开展棒束通道脉动流条件下相位差研究,对比了相位差在不同振幅、不同流动状态下的变化特性,并分析了定位格架对脉动流相位差的作用特点。另外,基于粒子图像测速（PIV）技术开展了脉动流条件下棒束通道内流场分布特性研究,对比了相同流量条件下稳态工况与瞬态工况下流场分布差异,分析了主流具备不同加速度时棒束通道内流场分布特征。实验结果表明：定位格架可减小脉动流下棒束通道内相位差;棒束通道内流场演化滞后于主流量变化。实验结果有助于揭示燃料组件在非稳态条件下瞬态特性,并为燃料组件的设计和优化奠定基础。     

六边形排列的7棒束通道内液态钠流动换热特性试验研究

以液态钠作为试验工质,对六边形排列的7棒束通道内液态钠流动换热特性进行了试验研究。试验流速为0~4 m·s^-1,热流密度为0~120 kW·m^-2,系统压力为1.5~200 kPa,对应的雷诺数和佩克莱数分别为4 000~60 000和0~340。深入分析了部分热工参数对7棒束通道内液态钠流动换热特性的影响,通过对7棒束通道内液态钠流动换热的试验数据的非线性拟合,得到适用于7棒束通道内液态钠流动换热的经验关系式。结果表明：拟合得到的摩擦系数关系式和努塞尔数关系式能准确地预测7棒束通道内的试验数据,其预测误差分别小于5%和6%。将获得的努塞尔数关系式与其他研究者的试验数据进行比较,与其他研究者985%的试验数据误差在30%以内,表明获得的关系式适用于7棒束通道内液态钠流动换热。     

Experimental Study of Liquid Sodium Flow and Heat Transfer Characteristics Along Hexagonal 7-rod Bundle Channel

The flow and heat transfer characteristics of single-phase liquid sodium were experimentally investigated in a hexagonal 7-rod bundle channel with the velocity of 0-4 m/s, the heat flux of 0-120 kW/m² and the absolute pressure of 1.5-200 kPa. The corresponding Reynolds number ranges from 4 000 to 60 000, and the Peclet number varies from 0 to 340. The influence of some thermal parameters on the heat transfer characteristics of liquid sodium flow in a hexagonal 7-rod bundle channel was analyzed in depth. Empirical correlations of liquid sodium flow and heat transfer in a hexagonal 7-rod bundle channel were obtained by nonlinear regression analysis for experimental data. The results show that these correlations can accurately predict the friction coefficient and Nu in a hexagonal 7-rod bundle channel. The prediction error for flow and heat transfer is less than 5% and 6%, respectively. The new equation was compared with other results, and the error is within 30%. It is shown that the new empirical correlation is suitable for the flow heat transfer of liquid sodium in a hexagonal 7-rod bundle channel.