Conference Agenda

The development of small modular reactors responds to a need for more flexible energy generation for a wider range of applications. Within the ELSMOR European project, the integrated E-SMR light-water-reactor concept consisting in a tank containing the core, pressurizer and compact steam generators has been investigated. The tank is in a metal enclosure, itself immersed in a large pool. The aim of this work is to study the pool thermal-hydraulic behavior by means of the 3D module of CATHARE-3.

Within the first phase of calculations, a standalone analysis of the water pool is carried out using the CATHARE-3 3D module. The containment is represented by a fixed temperature boundary condition. It is found that the evolution of the liquid temperature distribution is uniform across the pool as long as the water is in contact with the containment wall. A further refinement of the pool nodalisation does not significantly improve the results.

Within the second phase of calculations, the previous 3D water pool model is coupled with the containment, the mass and energy releases being taken from ASTEC and MAAP severe accident calculations. Two accidental sequences are considered: the first one involves the evacuation of the decay heat through the containment walls, while the second involves the use of a passive condenser. The results obtained by CATHARE are comparable to the tendencies predicted by ASTEC and MAAP. Concerning the 3D pool behavior, a uniform liquid temperature distribution is observed in the first accident, while the second one shows a temperature stratification.

High Power Density Pressurized Water Reactors (HP-PWRs) offer significant advantages in terms of thermal output within compact volumes, making them a promising option for applications such as small modular reactors. However, under high-power operating conditions, the occurrence of high local void fractions within HP-PWR cores presents unique challenges, affecting both the neutronic and thermohydraulic behaviors. This paper introduces a high-fidelity, fine-mesh neutronics-thermohydraulics (N-TH) coupling method to address these challenges for accurately modeling core behavior under high local void fraction conditions. The method incorporates flow distribution calculations (FDC), which significantly improve simulation accuracy by overcoming the limitations of traditional methods that assume uniform flow distribution. Our results show that under two-phase flow conditions, the introduction of FDC significantly alters the void fraction distribution, as well as the fuel and cladding temperatures, compared to traditional methods. Specifically, under 100% full power conditions, the power level of the hottest assembly decreased by approximately 0.8%, the mass flow rate of the hottest channel decreased by 12.87%, the maximum fuel temperature dropped by 0.77 K, and the maximum void fraction increased by 0.144. The impact of FDC is more pronounced in two-phase conditions and minimal under single-phase conditions. This study provides a valuable tool for the design and optimization of HP-PWRs and offers insights into improving reactor power density.

In U-tube steam generators employed in pressurized water reactors, flow-induced vibrations within the upper cross-flow U-section of the bundle are a major cause of fatigue and equipment damage. As it evaporates flowing upward and the steam quality increases, the water-steam mixture on the shell side transitions from bubbly flow to the intermittent and annular flow regimes. The local regime significantly influences the force exerted on the tubes. Consequently, accurate knowledge of the local flow conditions is crucial for assessing flow-induced vibration, but the consistent numerical prediction of the two-phase flow regime remains a major challenge for available CFD methodologies.

In this paper, the morphology-adaptive GEMMA (GEneralized Multifluid Modelling Approach) CFD model is used to predict the flow across a horizontal 7 x 5 cross-flow tube bundle in the bubbly and intermittent regimes. The GEMMA model, implemented in OpenFOAM, is based on the multifluid framework, but partially resolves interfaces over a certain length scale. This enables GEMMA to handle the entire spectrum of interface length scales in all flow regimes, which is traditionally challenging for available CFD approaches. In the intermittent regime, unsteady large gas structures are successfully predicted, and this enables a more accurate estimation of the void fraction inside the bundle. The intermittent nature of the local flow is reflected in the predicted force exerted on the tubes. The use of the GEMMA model results in much more time-fluctuating forces on the tubes, compared to the standard dispersed phase two-fluid model, unable to predict the intermittency of the flow.

The growing integration of renewable energy into electricity markets is driving nuclear power plants to shift from traditional baseload operation to more flexible modes. Flexible operation of nuclear reactors necessitates the evaluation of several technical challenges, including Xe-135 oscillations, a fission product that can significantly impact the reactor's operational stability.

This study focuses on analyzing Xe-135 oscillations triggered by load variations during the flexible operation of nuclear reactors. Flexible operating conditions are implemented in a 3D thermohydraulic-neutronic model of a PWR-KWU reactor core using the coupled code RELAP5/PARCSv3.2. Key parameters, such as the Axial Offset (AO), are examined to assess spatial distortions in power and Xe-135 distribution within the reactor core.

The results of this study highlight how variations in Xe-135 concentration affect the process of load increases and decreases during the flexible operation of PWR-KWU reactors.

The IAEA CRP on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” aims to support and achieve validation of computational fluid dynamics (CFD), subchannel, and system thermal-hydraulics (STH) analysis codes for heavy liquid metal systems. In particular, the Benchmark consists of two reference cases used for model training and a blind case to be reproduced for sake of model validation and accuracy assessment.

Together with stand-alone codes applications, a whole work package is devoted to the analysis of the addressed scenarios adopting multi-scale STH/CFD coupled applications. The CRP participants addressed the common problems adopting different system thermal hydraulics code and CFD codes, also considering different assumptions regarding the boundary conditions and involved phenomena representation. In particular the CFD approach was adopted for the simulation of the Fuel Pin Simulator, which represents a key component of the NACIE-UP loop, while the STH was considered for the remaining sections of the facility. The use of CFD for the FPS should allow for a better representation of the involved heat transfer and friction phenomena as well as the capability to obtain refined predictions of local wall and bulk temperature distributions in transient conditions. On the other hand, the STH approach allows for a relatively small computational cost of the other facility components.

The present paper reports on the results obtained by the CRP participants providing comments and improvement suggestions for the liquid metal loop modelling.