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.     

Cooling mechanism during transient reflooding of a reactor fuel bundle after loss of coolant

Cooling efficiency during transient reflooding under loss of normal coolant conditions has been examined with a 7 × 7 simulated fuel rod bundle and jet pump bypass. The bundle contains 49 electrically heated rods with 3600 mm heated length and a pseudo cosine axial power distribution. Water is injected into the lower plenum and the superheated bundle is reflooded from the bottom with some flow diverted to the simulated jet pump bypass. The results show that effective cooling can be maintained.     

Onset Criterion for Liquid Entrainment in Reflooding Phase of LOCA

Reflooding tests were conducted in a rod bundle geometry at the maximum pres- sure of 12 MPa to investigate thermal-hydraulic behavior during a small break loss-of-coolant accident (SBLOCA) in a nuclear reactor. The test conditions ranged 0.6 ～ 12 MPa for pressure, up to 920 K for rod surface temperature, up to 20 cm/s for bundle inlet flow velocity and up to 2 kW/m for linear power input. The principal objective of this paper is to investigate the onset condition for liquid entrainment by steam flow in the relatively high pressure reflooding phase. Experimental results showed a tendency that the liquid entrainment decreased with increase in pressure when the other parameters such as an inlet flow rate and rod temperature were fixed. A new correlation for the onset criterion for liquid entrainment was derived from the experimental results and an analysis of a force balance for a liquid droplet. Effects of pressure on liquid entrainment in the reflooding phase were made clear from the experimental and analytical results.     

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.     

Experimental comparisons for bottom reflooding in tight LWHCR and standard LWR geometries

The NEPTUN test facility is currently being used for reflooding experiments in tight hexagonal geometry, representative of light water high conversion reactors (LWHCRs). Results of parametric studies, based on over sixty forced-feed bottom reflooding experiments carried out with the NEPTUN-III (p / d = 1.13) test bundle, show that flooding rate is the most important, single thermal-hydraulics parameter.Direct comparisons with earlier NEPTUN experiments in standard LWR geometry indicate — on the basis of pressure difference considerations — that much smaller flooding rates may be expected to occur in tight LWHCR cores. The corresponding NEPTUN-III experiments show long-lasting rod surface temperature excursions with relatively high maximum temperatures being reached, and some of the more detailed experimental data collected is used to explain this behaviour. In spite of the above, rewetting of the tight-LWHCR geometry bundle was found to occur in all experiments with reasonably LWHCR-representative values for the various thermal-hydraulics parameters.     

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。     

基于RELAP5的高温棒束通道再淹没数值模拟研究

基于ABB Atom 3×3棒束再淹没实验,运用RELAP5建立其实验装置的定流量再淹没计算模型,通过与实验结果做比对验证模拟的有效性,研究在高、低两种注水流量下从底部再淹没高温棒束通道时的不同骤冷现象,分析期间的流动形态、传热特性,液位进程,先驱冷却效果差异等。模拟结果表明:低流量下主液位落后于骤冷前沿,高流量下骤冷前沿明显落后于主液位;通过对比发现在高流量下的高液位为高温壁面带来更强的先驱冷却,使壁面温度更快的降到再湿温度,而低流量下几乎匀速上升的液位变化进程对前沿下游的高温壁面冷却较慢,需要更长的时间才能降到再湿温度。这些分析将为研究此模型下的重力注水打下坚实的基础。     

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。     

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。      

Modifications to COBRA-TF to model dispersed flow film boiling with two flow,four field Eulerian–Eulerian approach – Part 1

In the case of a postulated loss of coolant accident (LOCA) in a nuclear reactor, an accurate prediction of clad temperature is needed to determine the safety margins. During the reflood phase of the LOCA, when the local void fraction is greater than 80% with the wall temperature above minimum film boiling temperature (T_min), the heat transfer process is dispersed flow film boiling (DFFB). This study has been performed to model DFFB in the reflood phase of a LOCA in a pressurized water reactor (PWR) rod bundle. The COBRA-TF computer code is utilized, since it has a detailed reflood package which takes into account the effect of spacer grids on the local heat transfer. The COBRA-TF code has also been improved to include a four field Eulerian–Eulerian modeling for the two-phase dispersed flow film boiling heat transfer regime. The modifications include adding a small droplet field to COBRA-TF as the fourth field. In addition, the spacer grid models of COBRA-TF have been revised and modified. In the first part of the paper, the results of the code predictions are presented by comparing the experimental data from rod bundle heat transfer (RBHT) experiments with the results of code simulations performed with original and modified code. Measurements and calculations for the heater rod, vapor temperatures and quench front progression have been compared and the results are described in detail. The results of the analysis performed with the modified code indicate the improvement in code predictions for the rod surface temperature, vapor temperature and quench front behavior. The results also indicate the need for improvement in the entrainment and interfacial drag models for the drop fields. The effects of spacer grids on the heat transfer, the models improved and developed for spacer grids and the results of the code calculations with these models are described in the part 2 of the paper.     

Critical heat flux prediction in vertical bottom-closed rod bundles

This paper presents the analysis of experimental data and calculational relationships for heat the transfer crisis in LWR rod bundle with closed bottom. A new relationship for critical heat flux prediction in the rod bundle with closed bottom based on the improved drift model is described. The comparison of critical heat flux values given by different correlations (including Groeneveld's algorithm used in RELAP5/MOD3.1 Code) and those obtained from the tests in the wide range of regime and geometric parameters is presented.     

Mechanistic modeling for ring-type boiling water reactor fuel spacer design (3) run-off effect and model formulation

The fuel spacer is one of the components of a fuel rod bundle and its role is to maintain an appropriate rod-to-rod clearance. The fuel spacer influences the liquid film flow distribution in the fuel rod bundle, so that the spacer geometry has a strong effect on thermal hydraulic characteristics of BWR such as critical power and pressure drop in the fuel bundle. In this paper, liquid film flow characteristics were experimentally investigated in a circular channel with ring-type spacers using air and water as test fluids in order to be compared with the analytical results introducing the mechanistic spacer model. The spacer model was composed of three effects; the drift flow effect, the narrow channel effect and the run-off effect. The drift flow effect and the narrow channel effect were discussed in the previous works and the run-off effect is done in this paper. This paper shows the formulation around the spacer with use of the three effects. The proposed model explains well the experimental results of liquid film flow rates and thickness carried out in reference to the spacer thickness, the gap between the channel wall and the spacer.     

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。      

Slug to annular flow transition during boiloff in a rod bundle under high-pressure conditions

High-pressure boiloff experiments in a wide range of bundle powers by using the Two-Phase Flow Test Facility (TPTF) were conducted. Two kinds of boiloff patterns were observed in these experiments. One is the boiloff pattern in a low bundle power, in which the dryout points of rods locate at a certain elevation in the bundle because the mixture level controls the dryout points. The other is the boiloff pattern in a high bundle power, in which the clear mixture level can not be observed and the dryout points of rods locate in a wide range of vertical directions. The vertical scatter of the dryout points is considered to be due to the break of the thin water film on the heater rods under the annular flow pattern.A simple model to predict the slug to annular flow transition in the rod bundle is proposed. In the model, the slug to annular flow transition takes place when the interferences of the water films on the neighboring rods cease. The model appeares to give good predictions of the previous flow transition experiment conducted in a rod bundle. The slug-annular transition below the dryout points was predicted with the present model in the high power boiloff experiments of TPTF. No slug-annular transition below the dryout points is predicted with the present model in the low power boiloff experiments.     

基于双区域模型的钠冷快堆组件子通道分析程序的开发与验证

子通道分析程序是钠冷快堆堆芯热工水力设计和安全分析的重要工具。本文为计算和分析钠冷快堆组件在径向均匀与倾斜功率分布工况下的热工水力特性,利用双区域绕丝交混模型开发了一款适用于钠冷快堆组件分析的子通道程序SPLICA,并与FFM2A 19棒束实验数据与WARD 61棒束实验数据进行了对比验证。由于本文开发的子通道分析程序SPLICA使用了详细的绕丝交混模型,与经过二次开发后的COBRA程序的计算结果相比,对于FFM2A实验SPLICA程序计算得到的结果与实验结果符合得更好。这两个实验数据的验证结果证明了本文开发的子通道分析程序的准确性以及对高流量工况和低流量工况均具有良好的适用性。本程序能为钠冷快堆组件热工水力分析提供有效的设计和研究手段。     

基于丝网传感器的5×5棒束通道空泡分布测量研究

为研究压水反应堆燃料组件棒束通道内的两相分布规律,设计并制造了适用于棒束通道的丝网传感器模块,开展了5×5棒束通道内空气-水泡状流的空泡分布测量实验,分析了棒束通道内空泡份额的分布规律及气泡尺寸对空泡分布的影响。实验结果表明,发生横升力方向反转的小气泡在壁面附近聚集、大尺寸气泡则聚集在子通道中心;常温常压下发生横升力方向反转的临界气泡直径在4~6 mm之间,证明了横升力模型在棒束通道中的适用性。      

Development of a Transient Boiling Transition Analysis Method Based on a Film Flow Model

A new single-channel, transient boiling transition (BT) prediction method based on a film flow model has been developed for a core thermal-hydraulic code. This method could predict onset and location of dryout and rewetting under transient conditions mechanically based on the dryout criterion and with consideration of the spacer effect. The developed method was applied to analysis of steady-state and transient BT experiments using BWR fuel bundle mockups for verification. Comparisons between calculated results and experimental data showed that the developed method tended to predict occurrence of rewetting earlier, however, onset time of BT and maximum rod surface temperature were well predicted within 0.6 s and 20°C, respectively. Moreover, it was confirmed that consideration of the spacer effect on liquid film flow rate on the rod surface was required to predict dryout phenomena accurately under transient conditions.     

Reflooding Experiments on a Vertical Surface with Radiation-Induced Surface Activation

In this study, reflooding experiments were performed on a vertical rod surface before and after γ-ray irradiation to see the effect of Radiation-Induced Surface Activation (RISA) on the quenching speed. The test section was an annular channel with a concentric inner rod made of stainless steel SUS304 and an outer tube made of quartz glass. The inner rod was irradiated by ⁶⁰Co γ-rays with predetermined radiation intensity and period to improve the surface wettability based on the radiation-induced surface activation phenomenon. Prior to the reflooding experiments, the contact angle of a droplet on the inner rod surface was measured using a CCD camera. It was indicated that the irradiated surface wettability was clearly improved and the quenching speed was enhanced after γ-ray irradiation.     

A simple forced diversion model for study of thermal-hydraulic transients in LMFBR subassembly

The generalized simple, transient, integral energy balances based on the average properties for the fuel and cladding have been used in our new multichannel thermal-hydraulic model for calculating the transient behavior of coolant in the rod bundle. This model was developed to provide a simple useful tool for analyzing the flow and thermal transients in a rod bundle with reasonable accuracy, and to understand the fundamental characteristics of flow in the rod bundle under both normal and abnormal condition of reactor-core operation.