Conference Agenda

CINEMA (Code for INtegrated severe accident Evaluation and Management), a severe accident analysis code developed in South Korea, consists of several modules enabling independent analysis of complex severe accident phenomena. Previous validations studies, including simulations of the Three Mile Island (TMI) accident and a comparison with the MAAP, demonstrated that CINEMA simulated both thermal hydraulic behavior and accident progression well. Further comparative analyses with other well-validated codes that cover both accident sequences and fission product (FP) behavior can enhance understanding of CINEMA’s simulation characteristics. This study compares the results of MELCOR and CINEMA’s analyses of accident progression and FP behavior during SBLOCA (Small Break Loss Of Coolant Accident) for OPR1000. Both codes showed broadly consistent sequences of major events, and CINEMA’s detailed NSSS nodalization showed specific fluid flows. Regarding FP transport, both codes predicted the inert gas Xe mostly remained suspended in the containment building. For Cs and CsI, both codes assessed that these species did not exist in an airborne with in the containment building. However, MELCOR predicted that the release fraction to the containment building could vary with break size. In CINEMA, Cs and CsI were deposited more within the reactor coolant system than in the containment building. The FP behavior was influenced by the flow direction in the RCS and modeling such as chemisorption, pool deposition, and re-evaporation.

This paper presents an experimental study on the pressure drop characteristics in fixed beds packed with sand particles, with the goal of improving the accuracy of pressure drop predictions. Single- and two-phase flow tests were conducted using a custom-designed, adiabatic test facility specifically built to investigate the frictional behavior of flow through porous media. The facility allows for precise control and measurement of flow conditions, providing robust data for analysis. Using the effective diameter derived from single-phase flow tests in sand-packed beds, two-phase flow experiments were performed, and various prediction models were validated by comparing the measured pressure drop data against calculations from different analytical models. The results demonstrate that for two-phase flow in beds packed with smaller sand particles, the measured pressure drops increase steadily with fluid flow rate. In contrast, for beds with larger, coarser sand particles, the pressure drops exhibit an initial decrease followed by an increase as flow rate rises a down-up tendency. Notably, only models that account for interfacial drag effects successfully predicted this behavior. However, despite this, the prediction models showed significant deviations from the experimentally observed data, highlighting the complexity of accurately modeling two-phase flow in porous media. These findings suggest the need for further refinement of predictive models to better capture the intricate behavior of two-phase flow in such systems.

Various types of non-light water reactors are currently engaged in the U.S. licensing process. Because of inherent differences compared with well-established large light water reactors, appropriate assessment tools are needed. Specifically, source term analysis, which determines environmental dose impacts from potential accident scenarios, is a crucial part of design and licensing. The U.S. Nuclear Regulatory Commission has emphasized the importance of mechanistic source term analysis for advanced reactor deployments. To align with these needs, Argonne National Laboratory has developed the Simplified Radionuclide Transport (SRT) source term analysis code for metal fuel Sodium-cooled Fast Reactors (SFRs) and microreactors. SRT conducts time-dependent radionuclide transport and retention in SFRs for core and ex-core radionuclide source accident sequences. The main objective of SRT is to provide rapid sensitivity and uncertainty analyses, incorporating parametric uncertainties and summarizing probabilistic results. As part of the code validation process, a study focused on the bubble scrubbing module was performed using an experiment recently carried out by the University of Wisconsin-Madison. Based on the analysis, the modeling approach in SRT provides accurate results for small and large aerosols, while slight underprediction of radionuclide aerosol removal are observed for medium sized aerosols. However, the deviation is minor, considering the highly uncertain phenomenon and range of results, and is in the conservative direction. In addition, uncertainty information derived from the experiments is further implemented, reflecting the actual span of parameters, which leads to enhanced agreement with code predictions. The results demonstrate that SRT provides reasonable predictions for the bubble scrubbing process in sodium pool.

The Modular Accident Analysis Program (MAAP), developed by EPRI, allows users to analyze simulated nuclear plant accident scenarios. The code predicts plant responses to severe accidents by evaluating the core, reactor vessel, and containment conditions, and tracks the transport of energy and mass, including water, hydrogen, aerosols, and radioactive species.

The latest version, MAAPv6.00, is being developed in C++ to incorporate modern, state-of-the-art approaches. EDF has contributed to the MAAPv6.00 update to support new designs like Small Modular Reactors (SMRs) that rely on the In-Vessel Strategy for severe accident management. Among EDF updates is the advanced modeling of the corium pool in the Lower Head. This modeling approach includes features such as the kinetics of stratification, which tracks the progressive formation of stratified layers in the pool. This can result in the Focusing Effect, where heat flux concentrates on a narrow lateral surface, potentially exceeding the Critical Heat Flux (CHF) and leading to vessel failure.

Additionally, the distributed modeling of the core support plate enhances heat transfer from the corium pool to the support plate. The sequential melting of the support plate can increase the mass of the metal layer in the corium pool, thereby mitigating the Focusing Effect.

This paper provides a comprehensive description of EDF recent modeling in MAAPv6.00 and presents a use case demonstrating their practical application for severe accident assessments.

The open-source package ‘containmentFOAM’ was developed to provide highly resolved insights, supporting the assessment of the effectiveness of safety measures and possible combustion loads challenging the containment integrity. It comprises a CFD solver and model library developed for and tailored to the expected phenomenology in a large dry PWR containment, as well as tools for input creation and solution monitoring.

This paper aims to summarize the progress made after its first introduction on NURETH-19 (2019). The package was continuously refactored and is currently available as an add-on to OpenFOAM^®-11. Major advancements in the physical modeling capabilities are related to radiative heat transport in participating media, aerosol, and fog transport as well as two-phase flows. Besides, the functional mockup interface (FMI) was implemented, allowing for a flexible integration of system models, packaged as functional mockup units (FMU). Along with the application-oriented validation, best practices were derived and an efficient sensitivity and uncertainty quantification method, based on deterministic sampling, was developed. Concluding, the paper will summarize ongoing applications as well as the strategy for further development.

The evaluation of source terms, which determines the species and quantity of radioactive materials released during a severe accident, is essential for timely safety assessment and the formulation of mitigation strategies. During severe accidents, significant releases of fission products undergo complex integrated phenomena, which inherently introduce substantial uncertainties in the evaluation of source terms. In particular, the behavior of various radionuclides with complex physical and chemical properties significantly increases the uncertainty in the analysis results.

Without quantifying these uncertainties, the reliability of predictions regarding radioactive material release during an accident is compromised, resulting in inaccurate mitigation strategies and safety assessments. Therefore, identifying and quantifying uncertainty factors is fundamental for reliable predictions of source term releases and the development of effective response strategies.

To quantitatively assess these uncertainties, evaluations based on reliable experimental data or actual accident scenarios are required. These experiments offer direct observations of the release behavior of fission products under various accident conditions and provide reference points for accident results.

In this study, the MELCOR code was utilized to benchmark the PHEBUS FPT-1 experiment, assessing core degradation behavior and source term release. Based on these results, key uncertainty variables affecting source term release were identified, and uncertainty analysis was conducted. Additionally, sensitivity analysis was performed to quantify the impact of each variable on the results.