Recent activities in the field of thermal hydraulic researches

In the field of thermal hydraulics, substantial progress has been made in research on single and two-phase heat transfer. Grid-enhanced convection heat transfer has been studied [1,2]. Moon et al. [1] performed an experimental study in a 6 × 6 rod bundle to investigate the effects of spacer grids on the single-phase convection heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. They suggested more systematic experiments should be performed using various spacer rids with large blockage ratios at low Reynolds numbers, considering an early phase of the re-flood conditions. Miller et al. [2] reported a two-phase dispersed droplet flow investigation of the grid-enhanced heat transfer augmentation using a 7 × 7 rod bundle heat transfer facility. It was found that a second-stage augmentation occurs under wet grid conditions at a distance of 10 diameters downstream of the grid. This second-stage augmentation was not observed under dry-grid conditions, nor was it observed in single-phase steam cooling tests [2]. Schlegel et al. [3] performed extensive experiments in pipes with diameters up to 0.304 m to collect area-averaged void fraction data using electrical impedance void meters for the purpose of remedying an inability of current drift-flux models to accurately predict the void fraction in churn-turbulent flows in large diameter pipes. They obtained a distribution parameter modified for churn-turbulent flows. It has been evaluated through comparison of the void fraction predicted by the drift-flux model and the measured void fraction.

Experimental data bases are important for models’ assessment and verification. Heat transfer and flow experiments using a mercury flow system were carried out by Kinoshita et al. [4] to clarify the validity and predictability of existing experimental correlations. They obtained a result that the heat transfer coefficients agreed well with the Subbotin correlation and analytical results with the STAR-CD code. Conner et al. [5] reported hydraulic benchmark data on Westinghouse PWR mixing vane grids at Texas A&M University. The data acquisition of interest is from an advanced particle image velocimetry (PIV) technique which can attain the high spatial and temporal resolution of the velocity vectors. The data obtained provided a much more thorough benchmark of computational fluid dynamics (CFD) results than were available before. Use of this data can not only help in benchmarking steady-state CFD simulations, but can also be used in benchmarking transient CFD simulations such as large eddy simulation [5].

To assess the safety at nuclear facilities and to respond to emergencies against accidental or intentional release of radioactive materials, a LOcal-scale High-resolution atmospheric DIspersion Model using Large-Eddy simulation (LOHDIM-LES) has been developed by Nakayama et al. [6]. It was extended to turbulent flows and plume dispersion in various building arrays, and successfully simulated the unsteady behaviors of turbulent flows and plume dispersion in urban-type surface geometries. The CUPID code and TAPINS code were also developed for the analyses of transient two-phase flows in nuclear reactor components and transient analysis of an integral reactor, REX-10 [7,8]. Lee and Park [8] compared the calculation results of TAPINS with the experimental data obtained from a series of integral effects tests using a scaled apparatus of REX-10. It was concluded that TAPINS can provide the reasonable prediction on the thermal-hydraulic responses of REX-10 during the transient and accident conditions [8]. Moreover, Onder and Leung [9] evaluated the ASSERT-PV subchannel code using boiling-length-average (BLA) critical heat flux (CHF) values for the CANFLEX bundle at cross-sectional average subcooled conditions.

Severe accident analyses are also carried out. Kawahara et al. [10] proposed a method for identifying the success criteria regarding alternative water injection in long-term station blackout (SBO) of a BWR-5 model plant by summarizing the sensitivity analysis results using RELAP5/SCDAP mod 3.5. They found that preventing core damage was almost equivalent to preventing cladding rupture because the fuel rod heats up to the temperature region of cladding creep rupture and the increased surface area after the cladding rupture triggered conditions at which oxidation and heat-up of the Zircaloy cladding excite each other. They proposed the discriminants for calculating an index of the core condition and its threshold values for preventing core damage [10]. Hydrogen release and dispersion could occur at an early stage of a severe nuclear accident and its potential explosion could seriously threaten the integrity of the containment or auxiliary buildings as it happened in the Fukushima accident. Wilkening and Ammirabile [11] carried out a simulation of helium gas release in the Battelle Model Containment facility using the OpenFOAM code. The SST (shear stress transport) turbulence model was applied to model helium release and dispersion with the code. The overall behavior was captured adequately [11].

There is also some significant progress made in the research on advanced reactor thermal-hydraulics [12,13]. In order to accurately model sodium–water reaction jets in steam generators of fast breeder reactors, Kudoh et al. [12] successfully visualized argon-gas jet behaviors injected into liquid sodium using an endoscope and a glass tube. The size distributions and mean diameters of liquid sodium droplets entrained into the gas jet were obtained in the bubble regime. The droplet size distribution of entrained sodium droplets was found to agree well with the Nukiyama–Tanasawa distribution function. The Santer mean diameter obtained in the study was also found to be well correlated with an empirical equation proposed by Epstein et al. [12]. Liquid lead–bismuth eutectic was considered as one of the most promising coolants for fourth-generation reactors. Recently, Thiele and Anglart [13] conducted numerical modeling of forced-convection heat transfer to lead–bismuth eutectic flowing in vertical annuli. They obtained temperature and velocity distributions in heated annuli using the RANS approach and employing three different turbulent viscosity models. They also carried out extensive sensitivity study. The results showed that the RANS approach and the turbulent viscosity models can be used for prediction of forced convection heat transfer to lead–bismuth eutectic [13].

There are some papers related to Fukushima Daiichi Nuclear Power Plants (F1NPPs) [14,15,16]. In order to obtain an understanding of the serious situation in reactors of F1NPPs, Tanabe [14] carried out analyses based on a measured data investigation and a simple model calculation. The first core melt behavior of the Unit 1, Unit 2, and Unit 3 reactors on 11–15 March 2011 as well as the re-melt behavior in another chaotic period of 19–31 March 2011 were analyzed [14]. Behaviors were obtained such as the core water level, core material temperature, and accumulated hydrogen mass after the beginning of core uncovering. Sibamoto et al. [15] proposed another evaluation method for the analysis of thermal-hydraulics in reactor pressure vessel and primary containment vessel. They developed a code, named “HOTCB”, which was based on mass and heat balance in a lamped parameter system of single-phase fluid or thermally equilibrium two-phase fluid. Analysis results for accident progression of the Unit 1 and Unit 2 reactors were presented as examples showing the usefulness of the code to understand the accident behavior [15]. Hirano et al. [16] reported review and analysis on the progression of the accident at Units 1–3 of F1NPPs. The basic information on the damages to F1NPPs by the earthquake and tsunami was summarized. They also discussed some key issues raised by the accident based on the insights gained from the analysis.