Conference Agenda

The objective of this study was to evaluate condensation heat flux q_c from steam and air mixture on a vertical flat plate, which is one of boundary conditions in CFD (computational fluid dynamics) analysis for thermal hydraulic behavior in a containment vessel during accident conditions. In our previous study, we carried out numerical simulation of wall condensation from saturated steam and air mixture on a vertical flat plate with the cooling height of 6 m by using the CFD code FLUENT, and evaluated effects of mixture velocity u_in on q_c. In this paper, we evaluated effects of steam mass fraction Y_s,in on q_c from saturated steam and air mixture on the vertical flat plate with the conditions of u_in = 0.53-3.2 m/s and Y_s,in = 0.226-0.68 by using the FLUENT code. The boundary condition for q_c, which was defined in the viscous sublayer, was used, and the size of the computational cell was 0.1 mm for the cells in contact with the condensation wall (where the dimensionless distance was y⁺ = 0.12-0.56). The q_c,CFD values computed with FLUENT were well expressed by an existing q_c correlation for forced convection (FC) condensation, but was a little larger than the q_c,cal values computed with an existing q_c correlation for natural convection (FC) condensation. The u_in value at the transition from FC to NC condensation became large for large Y_s,in due to large density difference.

The condenser is an important component in the secondary circuit of a nuclear power system. In marine applications, vibration and noise reduction are critical design goals of the condenser, with the heat transfer tube—its central element—playing a vital role in ensuring the reliability and longevity of the equipment. Two key physical phenomena occur within the condenser heat transfer tubes: condensation heat transfer and fluid-induced vibration. These phenomena are interdependent and continuously coupled. To investigate the vibration characteristics of heat transfer tubes during the condensation phase change, a visual experimental setup was designed. This setup allows for the observation of vibration behaviors and the analysis of how various parameters influence heat transfer performance. Modal tests revealed that temperature significantly impacts the natural frequency of the heat transfer tube, with the natural frequency decreasing as temperature increases. Dynamic tests demonstrated that the changes in volume and pressure due to condensation phase change are particularly significant at low steam flow rates. As steam velocity increases, the effect of condensation diminishes, and fluid shock becomes the dominant factor. Regarding heat transfer, an increase in the heat transfer rate leads to a higher vibration amplitude, while the vibration frequency decreases. These findings provide experimental and theoretical basis for optimizing condenser performance.

The application of a two-phase loop thermosiphon is evident in Passive Containment Cooling Systems designed to remove core decay heat following a Loss-of-Coolant Accident. It is characterized by two dominant effects, which are boiling and condensation. The phenomenon in practical applications usually includes the effects of noncondensable gas, however, only pure vapor cases are considered in the present study for simplicity.

The two fluid model (TFM) is widely applied in various code to analyze two-phase flow behavior in nuclear field. For the implementation of subcooled nucleate boiling, we use the RPI heat flux partitioning model in near wall region because it has been well described in literature and favoured in many modern CFD codes. Meanwhile, condensation rate in subcooled bulk is calculated by assuming zero resistance in vapor side via specifying an infinite value of vapor heat transfer coefficient. On the contrary to boiling model, only a limited number of film condensation models in TFM approach are present, especially for large-scale volume like the containment structure. Instead of resolving the thin film thickness, a subgrid liquid film model is implemented in wall-adjacent cells. Phase change rate at the cooling wall is computed by solving governing equations coupled with an additional mechanistic liquid momentum equation. In bulk region, a same approach as in boiling model is applied as we consider zero resistance condition for liquid side. Subsequently, the subcooled flow boiling and film condensation models are incorporated and specified in the corresponding domains to model a two-phase thermosiphon.

During a severe accident, the potential leaking of fission products (FPs) from a nuclear facility to the atmosphere represents a significant nuclear safety challenge. Accurate estimation of these releases is important to conduct an appropriate risk assessment and implement the necessary measures. To achieve this, IRSN conducts research dealing with the mitigation of FPs transported in a carrier gas injected through a liquid pool. This process, referred as 'pool scrubbing', can occur in several accident scenarios in light water reactors (PWRs, BWRs), including Filtered Containment Venting Systems (FCVS) or with the Steam Generator Tube Rupture (SGTR), as well as in nuclear-powered submarines or in new Small Modular Reactors (SMRs) using pressurised water.

Thus, experiments are currently being conducted to characterize bubble hydrodynamics and trapping of iodine compounds (decontamination factor measurements) on dedicated facilities, involving demineralized water and saline solution for different carrier gas injection flowrates at ambient conditions. In this context, advances results have been previously obtained on the hydrodynamics that occur along the pool [1] and on the retention of CsI aerosols [2] and volatile I₂. Building on this, a new study has been initiated to investigate the impact of a saline solution on these phenomena and to develop a more sophisticated bubble plume model.

First results suggest that saltwater generates smaller bubbles (except near the nozzle) and improves the retention of I₂ compared to clear water. Ultimately, these works will be used to enhance the pool scrubbing modelling implemented in the ASTEC integral code, developed by IRSN.

With the recent research on the miniaturization of nuclear power plants, it has become important to also miniaturize and modularize components, such as heat exchangers. Particularly, the steam generators used in newly developed SMRs (Small Modular Reactors) have adopted helical tubes instead of conventional U-tube designs to achieve compactness and enhance thermal efficiency. In PWRs (Pressurized Water Reactors), a loss of coolant accident (LOCA) leads to decrease the pressure and temperature of the primary side, causing the pressurized water to transform into two-phase steam. This steam flows downward from the upper shell side of the steam generator and condenses through the interaction with feedwater on the tube side. This study was conducted to evaluate the condensation heat transfer performance on the shell side under these conditions and to investigate the condensation heat transfer mechanisms. For this purpose, a test section was developed to assess the condensation heat transfer on the helical tube, and visualization experiments were performed to evaluate the behavior and thickness of the condensate film formed on the helical tube. Additionally, qualitative analysis of the condensation heat transfer mechanisms occurring on the helical tube was conducted based on the observed condensate film thickness from the visualization experiments and the measured heat transfer performance

Modern nuclear power plants often rely on passive systems, especially during severe accident scenarios. Systems like that utilize natural circulation and convection, which can challenge the capabilities of existing safety analysis codes. The forces governing the natural circulation phenomena are usually weak, and thereby, the natural circulation patterns can be easily disturbed. In accident analyses this may happen by physical phenomena or modeling/numerical issues. Therefore, the verification and validation (V&V) of these system codes are crucial for credible safety analysis to ensure the safety of current reactors and for licensing new designs, including Small Modular Reactors.

In this study, the authors analyze containment thermal-hydraulic phenomena using the MELCOR 2.2 code, drawing on selected experiments conducted at the Paul Scherrer Institute's PANDA facility. The focus is investigating natural convection and fluid circulation in containment-like geometries under accident conditions. As part of the preliminary analysis, the HYMERES HP6 experiments were chosen. In the tests, steam and He (to mimic H2) are injected into one of four PANDA vessels, to analyze circulation patterns between vessels, as well as atmosphere stratifications. The HP6 tests allow to explore various phenomena, including stratification, circulation, condensation, and the influence of non-condensable gases.

The results, which include sensitivity analyses, help identify containment phenomena that are well-represented by the MELCOR 2.2 code, as well as those that present challenges for accurate modeling. This work aims to contribute to potential future recommendations for improving the modeling of these complex phenomena.