Many thermohydraulic issues about the safety of light water reactors are related to complicated two-phase flow phenomena. In these phenomena, computational fluid dynamics (CFD) analysis using the volume of fluid (VOF) method causes numerical diffusion generated by the first-order upwind scheme used in the convection term of the volume fraction equation. Thus, in this study, we focused on an interface compression (IC) method for such a VOF approach; this technique prevents numerical diffusion issues and maintains boundedness and conservation with negative diffusion. First, on a sufficiently high mesh resolution and without the IC method, the validation process was considered by comparing the amplitude growth of the interfacial wave between a two-dimensional gas sheet and a quiescent liquid using the linear theory. The disturbance growth rates were consistent with the linear theory, and the validation process was considered appropriate. Then, this validation process confirmed the effects of the IC method on numerical diffusion, and we derived the optimum value of the IC coefficient, which is the parameter that controls the numerical diffusion.

In the case of a severe accident, natural convection plays an important role in the atmosphere mixing of nuclear reactor containments. In this case, the natural convection might not in the steady-state condition. Hence, instead of steady-state simulation, the transient simulation should be performed to understand natural convection in the accident scenario within a nuclear reactor containment. The present study, therefore, was aimed at the transient 3-D numerical simulations of natural convection of air around a cylindrical containment with unsteady thermal boundary conditions (BCs) at the vessel wall. For this purpose, the experiment series was done in the CIGMA facility at Japan Atomic Energy Agency (JAEA). The upper vessel or both the upper vessel and the middle jacket was cooled by subcooled water, while the lower vessel was thermally insulated. A 3-D model was simulated with OpenFOAM^®, applying the unsteady Reynolds-averaged Navier–Stokes equations (URANS) model. Different turbulence models were studied, such as the standard k-ε, standard k-ω, k-ω shear stress transport (SST), and low-Reynolds-k-ε Launder–Sharma. The results of the four turbulence models were compared versus the results of experimental data. The k-ω SST showed a better prediction compared to other turbulence models. Additionally, the accuracy of the predicted temperature and pressure were improved when the heat conduction on the internal structure, i.e., flat bar, was considered in the simulation. Otherwise, the predictions on both temperature and pressure were underestimated compared with the experimental results. Hence, the conjugate heat transfer in the internal structure inside the containment vessel must be modeled accurately.

A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Dai-ichi nuclear power plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants had been limited by radiation, analytical investigation for the plant was required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named “HOTCB”, based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperatures up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi nuclear power plant (NPP) as examples to show the usefulness of the code.

Our previous study investigated the rewetting behavior of dryout fuel surface during transients beyond anticipated operational occurrences for BWRs, which indicated the rewetting velocity was significantly affected by the precursory cooling defined as cooling immediately before rewetting. This study further investigated the previous experiments by conducting additional experimental and numerical heat conduction analyses to characterize the precursory cooling. For the characterization, the precursory cooling was first defined quantitatively based on evaluated heat transfer rates; the rewetting velocity was investigated as a function of the cladding temperature immediately before the onset of the precursory cooling. The results indicated that the propagation velocity appeared to be limited by the maximum heat transfer rate near the rewetting front. This limitation was consistent with results of the heat conduction analysis using heat transfer models for the precursory cooling expressed as a function of distance from the rewetting front, the maximum wetting temperature, and the heat transfer coefficients in the wetted region. This paper also discusses uncertainties in the evaluation of transient heat flux from the measured surface temperature, and technical issues requiring further investigation.

In the case of a severe accident, natural convection plays an important role in the atmosphere mixing of nuclear reactor containments. In the previous study, to simulate the natural convection in the accident scenario within a nuclear reactor containment, the steady thermal boundary conditions (BCs) were prescribed on either cooled or heated wall. The present study, therefore, aims at the transient 3D numerical simulations of natural convection of air around a cylindrical containment with unsteady thermal BCs at the vessel wall. For that purpose, the experiment series was done in the CIGMA facility at Japan Atomic Energy Agency (JAEA). The upper vessel or both the upper vessel and the middle jacket was cooled by subcooled water, while the lower vessel was thermally insulated. A 3D model was simulated with OpenFOAM®, applying the Unsteady Reynolds-averaged Navier–Stokes equations (URANS) model. Different turbulence models were studied, such as the standard k-ε, standard k-ω, k-ω Shear Stress Transport (SST) and, low-Reynolds-k-ε Launder-Sharma. The results of the four turbulence models were compared versus the results of experimental data. The k-ω SST showed a better prediction compare to other turbulence models. Also, the accuracy of the predicted temperature and pressure were improved when the heat conduction on the internal structure, i.e., flat bar, was considered in the simulation. Otherwise, the predictions on both temperature and pressure were underestimated compared with the experimental results. Hence, the conjugate heat transfer in the internal structure inside the containment vessel must be modeled accurately.

A heat transfer coefficient (HTC) model was developed for the prediction of post-boiling transition (post-BT) behavior that might occur during anticipated operational occurrences (AOOs) for boiling water reactors (BWRs). The model development was based on measurements of heat transfer coefficient, liquid droplet deposition rate, and droplet concentration in our experiments conducted at high pressure. The model focused on the heat transfer near the rewetting front where the cooling by droplet deposition significantly affects the propagation behavior of a liquid film. The correlation by Sugawara was validated for the prediction of the deposition by using the experimental data. The model was also expressed as a function of the distance from the rewetting front to use in analytical models for the rewetting propagation. Both expressions of the present model successfully predicted our experimental data simulating the BWR thermal-hydraulic conditions.