Modelling of low-pressure subcooled flow boiling using the RELAP5 code

A new model for upward vertical subcooled flow boiling at low pressure has been proposed. The model considers the most relevant closure relationships of one-dimensional thermal-hydraulic codes that are important for accurate prediction of vapour contents in the channel: wall evaporation model, condensation model, flow regime transition criterion and drift-flux model. The new model was incorporated in the current version of the RELAP5 code, MOD3.2.2 Gamma. The modified code was validated against a number of published low-pressure subcooled boiling experiments, and in contrast to the current code, shows good agreement with experimental data. The presented analysis also leads to a better understanding of the basic mechanisms of subcooled flow boiling at low pressure.     

Inter-subassembly heat transfer of sodium cooled fast reactors: Validation of the NETFLOW code

This paper describes the prediction of temperature at the exit of subassemblies of a sodium cooled fast reactor using the NETFLOW code. Until present time, this plant dynamics calculation code is expected as a tool of nuclear education, and has been validated using data obtained at facilities or reactors cooled with water or sodium. A natural circulation test was conducted in the experimental fast reactor ‘Joyo’ with a 100 MW irradiation core. Also a turbine trip test was conducted in the prototype fast breeder reactor ‘Monju’. These tests were chosen to validate a model to calculate inter-subassembly heat transfer consisting of heat conduction and heat transfer by inter-wrapper flow. Based on the calculation for the natural circulation test in primary and secondary loops of ‘Joyo’, the model to calculate the heat transfer in radial direction of the inter-subassemblies simulated reasonable sodium temperature behaviors at the exit of subassemblies. Good agreement was also obtained in prediction of temperatures at the exit of the ‘Monju’ subassemblies. Through these validations, it was shown that the one-dimensional plant dynamics code NETFLOW could trace temperatures at the exit of the subassemblies of fast reactors with the inter-subassembly heat transfer model.     

竖直圆管内低压过冷沸腾相分布特性实验研究

实验采用双探头光学探针对内径24 mm竖直圆管内低压过冷沸腾局部空泡份额、界面面积浓度及汽泡尺寸等局部相界面参数径向分布特性进行了研究。实验结果表明：竖直圆管内过冷沸腾相分布形态呈现轴对称特性,随着热流密度的增大,相分布形态出现近壁峰值并逐渐向中间峰值分布形态的发展,较高热流密度工况下出现轴心峰值分布;随着质量流速的增加,局部空泡份额减小,并出现中间峰值向近壁峰值分布形态的转变;随着压力的增大,局部相界面参数减小。     

Analysis of subcooled boiling flow with one-group interfacial area transport equation and bubble lift-off model

To enhance the multi-dimensional analysis capability for a subcooled boiling two-phase flow, the one-group interfacial area transport equation was improved with a source term for the bubble lift-off. It included the bubble lift-off diameter model and the lift-off frequency reduction factor model. The bubble lift-off diameter model took into account the bubble's sliding on a heated wall after its departure from a nucleate site, and the lift-off frequency reduction factor was derived by considering the coalescences of the sliding bubbles. To implement the model, EAGLE (elaborated analysis of gas-liquid evolution) code was developed for a multi-dimensional analysis of two-phase flow. The developed model and EAGLE code were validated with the experimental data of SUBO (subcooled boiling) and SNU (Seoul National University) test, where the subcooled boiling phenomena in a vertical annulus channel were observed. Locally measured two-phase flow parameters included a void fraction, interfacial area concentration, and bubble velocity. The results of the computational analysis revealed that the interfacial area transport equation with the bubble lift-off model showed a good agreement with the experimental results of SUBO and SNU. It demonstrates that the source term for the wall nucleation by considering a bubble sliding and lift-off mechanism enhanced the prediction capability for the multi-dimensional behavior of void fraction or interfacial area concentration in the subcooled boiling flow. From the point of view of the bubble velocity, the modeling of an increased turbulence induced by boiling bubbles at the heated wall enhanced the prediction capability of the code.     

矩形窄通道内汽泡滑移与冷凝前期生长模型研究

矩形窄流道内汽泡生长会直接改变相界面浓度,从而影响流道的传热传质性能。为获得适用于窄流道内不同类型的汽泡生长模型,基于通体可视的实验本体,开展壁面沸腾流动换热实验。基于传热能量方程,研究过冷沸腾中汽泡滑移与冷凝前期两种情况下汽泡生长模型。实验结果表明汽泡呈现两种形式的生长,即汽泡滑移生长以及冷凝前期生长。建立了两种情况下的汽泡生长模型,实验数据验证模型误差在20%以内。因此,本研究能为沸腾两相数值模拟提供更加精细化的汽泡生长模型,从而提高汽泡行为的预测精度。     

Interpretation of the results of the CORA-33 dry core boiling water reactor test

All boiling water reactor (BWR) degraded core experiments performed prior to CORA-33 were conducted under ‘wet’ core degradation conditions, in which water remains within the core and continuous steaming feeds metal-steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ‘dry’ core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ‘dry’ core severe accident scenarios and to resolve partially phenomenological uncertainties concerning the behavior of relocating metallic melts that drain into the lower regions of a ‘dry’ BWR core (the ex-reactor experiments at Sandia National Laboratories will further address these uncertainties). CORA-33 was conducted on 1 October 1992, in the CORA test facility at Karlsruhe. A review of the CORA-33 data indicates that the objectives were achieved; i.e. core degradation occurred at a core heat-up rate (characterized by the absence of any temperature escalation caused by oxidation) and a test section axial temperature profile (at incipient structural melting) that are prototypic of full-core nuclear power plant simulations under ‘dry’ core conditions. Simulations of the CORA-33 test at Oak Ridge National Laboratory (ORNL) have required the modification of existing control blade-canister materials interaction models to include the eutectic melting of the stainless steel-zircaloy interaction products and the heat of mixing of stainless steel and zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the post-test analyses carried out at ORNL based on the experimental data and the post-test examination of the test bundle at Karlsruhe. The implications of these results with respect to degraded core modelling and the associated safety issues are also discussed.     

Development and validation of a transition boiling model for RELAP5/MOD3 reflood simulation

The current version of the RELAP5/MOD3.1 code significantly underpredicts the transition boiling heat transfer during reflooding of hot fuel rods. In order to extend the code’s range of application for LOCA and degraded core analyses, a new transition boiling model has been developed, assessed and implemented. The model is based entirely on local state variables calculated by the code (wall and fluid temperatures, pressure, void fraction, mass flux and static quality) and does not rely on other history parameters, such as quench position or CHF and minimum film boiling temperatures. A number of separate-effect and bundle experiments are analyzed with the modified version of the code, and the predictions are compared with the ones obtained by the current version and with available experimental data. In all cases, the predictions of the improved model better fit the measured data. The shape of the new temperature curves is more physically and conceptually sound than the one calculated by the current version of the code.     

蒸汽发生器传热管一、二次侧耦合换热及管外过冷沸腾数值研究

采用两流体欧拉数学模型,结合气相和液相之间的界面传热、传质和动量交换封闭模型以及RPI壁面沸腾模型,利用ANSYS CFX 12.0对蒸汽发生器局部传热管束二次侧的过冷沸腾进行数值研究。数值研究结果与单管内过冷沸腾实验数据对比验证符合良好。结果表明,采用壁面沸腾模型能准确预测沸腾起始点的位置,同时梅花孔板的存在对二次侧流动换热特性影响显著。     

CFD modelling of subcooled boiling—Concept, validation and application to fuel assembly design

The paper describes actual Computational Fluid Dynamics (CFD) approaches to subcooled boiling and investigates their capability to contribute to fuel assembly design. In a prototype version of the CFD code CFX a wall-boiling model is implemented based on a wall heat flux partition algorithm. It can be shown, that the wall boiling model is able to calculate the cross sectional averaged vapour volume fraction of vertical heated tubes tests with good agreement to published experimental data. The most sensitive parameters of the model are identified. Needs for more detailed experiments are established which are necessary to support further model development. The model is applied for investigation of the phenomena inside a hot channel of a fuel assembly. Here the essential phenomenon is the critical heat flux. Although subcooled boiling represents only a preliminary state towards the critical heat flux occurrence, essential parameters like swirl, cross flow between adjacent channels and concentration regions of bubbles can be determined. By calculating the temperature of the rod surface the critical regions can be identified which may later on lead to departure from nucleate boiling and possible damage of the fuel pin. The application of up-to-date CFD with a subcooled boiling model for the simulation of a hot channel enables the comparison and the evaluation of different geometrical designs of the spacer grids of a fuel rod bundle.     

Low-quality and subcooled film boiling of water at elevated pressures

Stable film boiling heat transfer data have been obtained in an 8.9 mm ID tube at pressures from 2 to 9 MPa. These data were obtained at low-quality and subcooled conditions, over a mass flux range of 0.11 to 2.75 Mg m⁻² s⁻¹. Excessive film boiling surface temperatures were avoided by using the hot patch technique. Contrary to the high-quality data, the low-quality data showed a decrease in heat transfer coefficient with an increase in quality. The film boiling data were compared with existing film boiling correlations. None of these were found to be satisfactory.     

Spectra of D-T neutrons transmitted through rare earth mixed concretes

Spectra of D-T neutrons transmitted through different densities of concrete containing rare-earth oxides are measured using a neutron generator. The NE-213 Liquid Scintillation Spectrometer is used in these measurements. The count spectra have been unfolded using the code ‘DUST’. The experimental results are compared with the one dimensional transport code ‘ASFIT’ and also the Monte-Carlo code ‘MORSE-CG’. The computations agree reasonably well with the experimental measurements. The characteristic features of the transmitted neutron spectra are explained based on the available cross-section data.     

Improvements to COMETHE III-J fuel rod modeling code

Current improvements to the COMETHE fuel performance code focus on pellet-clad axial interaction and Zircaloy cladding failure predictions. Slipping and sticking between pellets and clad as well as trapped stack are evaluated. The main conclusions are that slipping with friction concerns only local effects while axial PCMI is primarily dependent on pellet expansion with a strong ‘strain biaxiality’ effect dictated by the dishing. The notion of locking prior to radial PCMI is also introduced, which explains experimental features not previously understood. Benckmarking of the version of COMETHE against ramp tests has been initiated and will enable assessment of the code capability in Zircaloy clad failure predictions.     

Effect of flow-induced vibration on local flow parameters of two-phase flow

A preliminary study was conducted experimentally in order to investigate the effect of flow-induced vibration on flow structure in two-phase flow. Two kinds of experiments were performed, namely ‘reference’ (no vibration) and ‘vibration’ experiments. In the reference experiment, an experimental loop was fixed tightly by three structural supports, whereas the supports were loosen a little in the vibration experiment. In the vibration experiment vibration was induced by flowing two-phase mixture in the loop. For relatively low superficial liquid velocity, flow-induced vibration promoted the bubble coalescence but liquid turbulence energy enhanced by the vibration might not be enough to break up the bubble. This leaded to the marked increase of Sauter mean diameter, and the marked decrease of interfacial area concentration. Accordingly, flow-induced vibration changed the void fraction profile from ‘wall peak’ to ‘core peak’ or ‘transition’, which increased distribution parameter in the drift-flux model. For high superficial liquid velocity, shear-induced liquid turbulence generated by two-phase flow itself might be dominant for liquid turbulence enhanced by flow-induced vibration. Therefore, the effect of flow-induced vibration on local flow parameters was not marked as compared with that for low superficial liquid velocity. Since it is anticipated that flow structure change due to flow-induced vibration would affect the interfacial area concentration, namely interfacial transfer term, further study may be needed under the condition of controlled flow-induced vibration.     

基于流固共轭传热的两相流动数值模拟及燃料子通道CHF预测研究

为对过冷沸腾两相流动进行准确模拟,并探索临界热流密度（CHF）预测方法,本文基于共轭传热和两相CFD分析的方法,通过流固界面耦合,建立流固共轭传热两相流动耦合求解的数值模型。首先通过典型燃料棒栅元过冷沸腾两相流动的模拟,验证数值模型的正确性。随后对燃料子通道内两相流动进行模拟,并在两相流动模拟的基础上,通过准瞬态的方法,建立与CHF试验过程非常近似的CHF预测方法,将加热壁面的温度飞升作为CHF判定的标准,实现对燃料组件子通道CHF的数值预测。研究表明,本文建立的数值模拟方法,可为燃料组件或其他换热系统的CHF预测奠定基础,为燃料组件的设计提供新的辅助手段。     

压水堆子通道欠热沸腾数值验证及交混翼研究

为改善压水堆交混翼格架在欠热沸腾工况下的热工水力特性,以子通道为研究对象验证了所使用的欠热沸腾数值模型在不同工况下的有效性。基于已验证的数值模型,对含不同偏折角交混翼格架的子通道模型在不同工况下进行了两相流数值模拟,研究交混翼及其偏折角对子通道中两相流动、传热及气泡分布的影响。结果表明：交混翼在增大压降的同时明显强化了冷却剂的交混、降低了近壁面气泡份额、提高了换热效率,且在一定范围内偏折角越大影响越明显。相对较高的气泡份额将导致更大的压力损失、减弱冷却剂的交混、降低传热效率。当交混翼偏折角达25°时,继续增大其偏折角对降低近壁面气泡份额和提高传热效率的作用不再明显,反而造成压降的快速增大,因此建议其偏折角在25°左右。     

蒸汽发生器传热管束过冷沸腾区两相流动数值模拟

基于两流体欧拉数学模型结合RPI壁面沸腾模型,利用大型商用CFD软件ANSYS CFX 12.0对蒸汽发生器传热管束过冷沸腾区一次侧、壁面和二次侧耦合传热过程进行了数值模拟。研究了三叶梅花孔支撑板和不同入口过冷度条件下蒸汽发生器传热管束内的流动沸腾现象,得到一、二次侧流场与温度场,二次侧空泡份额分布,支撑板梅花孔局部的流动状况及不同入口过冷度对蒸汽发生器热工水力特性的影响。数值模拟结果表明,三叶梅花孔支撑板的存在及不同入口过冷度对蒸汽发生器传热管束过冷沸腾区域的热工水力特性影响显著。