Conference Agenda

Small Modular Reactors (SMRs) are a promising solution for carbon-free energy. With advanced passive systems SMRs show reduced possibility of transition to severe accidents (SA) compared to large reactors. Nevertheless, SA in extreme conditions is ineliminable due to inherent characteristics of fuel assemblies in pressurized water reactors (PWRs), necessitating a comprehensive SA analysis. Limitations in acquiring SA data and uncertainties in physical models require that uncertainty analysis be conducted to understand potential outcomes.

This study investigates the thermal-hydraulic behavior of SA progression in the reactor and metal containment vessel (MCV) integrity under SA conditions. i-SMR, an SMR under development in the Republic of Korea, is adopted as a reactor type. Using CINEMA, a system code developed in Korea, SA scenario was simulated. With the initial event set as a Loss of Coolant Accident (LOCA), a common SA sequence. Parameters, including minimum and maximum oxidation temperatures and steam heat transfer coefficients, were varied within specified ranges using Latin Hypercube Sampling. Key Figures of Merit (FOMs) related to MCV integrity, including core uncover timing, corium mass, and hydrogen production were analyzed. Results indicate that MCV integrity is maintained across these variations, supporting i-SMR’s ability to protect containment under extreme conditions.

In case of a severe accident in a PWR-type nuclear power plant, maintaining the structural integrity of the containment building – representing the last barrier against the release of radioactive material into the surrounding environment – is of utmost importance. Therefore, addressing the various challenges that the containment may be facing during an accident is a crucial part of reactor safety research. Reliable prediction of thermal hydraulic processes and phenomena inside the containment are key to optimizing accident management measures, such as e.g., the hydrogen mitigation strategy or filtered venting systems.

In a collaborative effort of Ruhr-Universität Bochum (RUB) and Forschungszentrum Jülich GmbH (FZJ), multiple simulations of postulated accident transients in a generic Konvoi-type PWR (1300 MW_el / 70.000 m³ free containment volume) with a simplified nodalization and junction structure were carried out using the lumped parameter Containment Code System (COCOSYS) developed by GRS gGmbH. Both a medium break loss-of-coolant accident (MBLOCA) and a station blackout (SBO) scenario were investigated. Unmitigated reference calculations are used for comparative assessment with the respective mitigated cases putting special emphasis on the development of the gas composition inside the containment aiming at enhancing the general understanding of the H₂/CO combustion risk, particularly in the late phase of a severe accident.

Consequently, this paper gives a detailed overview of the simulations performed and includes a comprehensive discussion of the results. The work presented here was conducted within the framework of the European AMHYCO project (Euratom 20192020, GA No 945057).

Spray systems represent a critical safety feature of nuclear power plants. The performance and efficiency of such spray systems depend upon various parameters and boundary conditions. During a severe accident, the spray systems interact with the containment atmosphere, which may contain aerosols and iodine, thereby influencing the radiological source term. To assess thermohydraulic conditions, spray was injected into the THAI vessel, which was equipped with various measurement systems depending on the test objective. The cooling efficiency was investigated by injecting spray via both a nozzle and boreholes. The depletion of cesium iodide aerosol concentration was investigated using a spray system with a polydisperse droplet size spectrum. The removal of molecular iodine was examined with fresh and recirculating water spray at varying pH and iodine contents. The nozzle and boreholes tests revealed that the cooling efficiency is enhanced with an increase in drop height and a reduction in droplet size. The efficiency of the reduction of aerosol concentration by spray was found to be higher for larger particles than for smaller ones, as indicated by a shift in the particle size distribution towards smaller particle sizes. The efficiency of iodine removal by spraying deionized water is significantly higher than that of iodine-containing water from the sump at higher pH levels. In conclusion, this work presents a comprehensive set of experimental data that enhances the understanding and knowledge of the behavior of spray systems and its interaction with the containment atmosphere under accident conditions and can be used for code validation.

Small modular reactors (SMRs) present unique safety challenges and opportunities due to their compact, integral designs and reliance on passive safety systems. This study investigates the impact of containment and reactor pool modeling on the progression of severe accidents (SAs) in a SMR using ASTEC 3.1. The ASTEC code, recognized as the European reference tool for SA analysis, was applied to model a SMR featuring a thermal power of 160 MW, a submerged containment design, and fully passive cooling systems. However, the lumped-parameter approach and simplified subcooling models in ASTEC present challenges in accurately reproducing key phenomena in this SMR. The code must effectively capture passive heat transfer mechanisms under subcooling conditions, the intricate dynamics of natural circulation flows within large-pool volumes, and the in-vessel retention (IVR) process to ensure a consistent representation of SA evolution.

The study emphasizes the influence of modeling approaches for the containment and reactor pool on ASTEC results, specifically in terms of containment pressure, core degradation progression, hydrogen production, and fission product release. It discusses efforts to mitigate these limitations, highlighting the need for refined nodalization and validation through experimental data or comparison with best-estimate codes.

This work contributes to the broader effort to enhance the predictive capabilities of SA codes in replicating the behavior of passive safety systems, thereby ensuring the robustness of safety assessments for next-generation nuclear technologies.

The aim of this work is to investigate the behavior of the spray system within nuclear reactor containments. This mitigation system is often modeled at the system scale using 0-D modules, which provide conservative estimates of the gaseous environment conditions in the containment. The purpose of the WISDOM spray model within the CATHARE system code is to offer a more precise representation of the droplet phenomenology during accidental scenarios. To evaluate the impact of the various input parameters on the spray phenomenology in containment, sensitivity analyses on the WISDOM model are conducted. The objective is to assess the main parameters of interest related to the containment spray system, and to identify which parameters have the greatest influence on thermal exchanges between the droplets and the containment's gaseous environment. A comparison between the code predictions and experimental data is also presented, along with discussions of data from the CARAIDAS, TOSQAN and MISTRA experimental facilities.

The release of hydrogen into the containment during a severe accident in a nuclear power plant can lead to undesirable consequences, such as the deflagration or detonation of a combustible hydrogen-air mixture, posing a risk to containment integrity. During a severe accident, such as a Loss of Coolant Accident (LOCA), hydrogen production arises not only from the strong exothermic metal-steam oxidation in the fuel cladding but also from additional sources, including molten corium-concrete interactions and carbon monoxide release due to the reduction of carbonates in the concrete.

Pressurized Water Reactors (PWR) are designed with a large internal volume to mix and dilute the combustible gases that may be produced during a severe accident, intended to keep the gas concentrations below the combustion limit. Atmospheric stratification can, however, result in poor mixing of the combustible gases released, and regions with a combustible gas mixture. To understand and assess the hydrogen risk in a PWR containment, it is important to accurately model the process of atmospheric mixing with accident simulation codes.

System thermal hydraulic (STH) or lumped parameter codes are known to have inherent limitations in representing 3-dimensional mixing phenomena compared to CFD codes. Therefore, a comparison is made between the STH code SPECTRA and the CFD code FLUENT for the atmospheric mixing in the dome of a generic PWR containment during a LOCA. This work has been performed within the framework of the AMHYCO project (Euratom 2019-2020, GA-No-945057).