Conference Agenda

This study, motivated by previous analyses from TEPSYS, investigates the impact of different cooling conditions on the distribution of non-condensable gases in the reactor building (R/B) of Fukushima Daiichi Unit 3 (1F3) during a severe accident. To understand this, experiments were conducted in the CIGMA facility, a large-scale test vessel that replicates the R/B structure. Steam and helium were continuously injected at the top of the vessel for 10,000 seconds to simulate steam and hydrogen leakage. CC-SJ-01, with a cooling temperature of 50°C, serves as the base case for comparison. In the present study, parametric investigations were performed under the same cooling conditions, focusing on the effects of increasing the partition aperture from one hole to nine holes (250 mm diameter each) and modifying the steam-to-helium mass ratio from 100:1 to 75:1. Results showed that the aperture change had little effect on helium distribution, with the highest concentration observed in the middle region, similar to CC-SJ-01. However, with the 75:1 steam-to-helium ratio, the highest helium concentration shifted to the upper region. The Shapiro ternary diagram revealed that a higher steam-to-helium ratio intersects the detonation limit in the middle region, while a lower ratio intersects it in the upper region. These findings are essential for understanding non-condensable gas behavior in severe accidents, aiding in the development of safety measures for nuclear reactor designs.

This paper presents a comparison between EDF MAAP 5.04 and EPRI MAAPv6.00 codes in simulating postulated Severe Accident (SA) scenarios in a generic integral PWR characterized by a submerged containment and about 60 MWe. This code comparison has been performed based on the Design 1 of the Horizon Euratom project SASPAM-SA (Safety Analysis of SMR with PAssive Mitigation strategies - SA).

The Modular Accident Analysis Program (MAAP) is a deterministic code developed by EPRI that can simulate the response of light water moderated nuclear power plants during accidental transients for Probabilistic Risk Analysis (PRA) applications. It can also simulate severe accident sequences, including actions taken as part of the Severe Accident Management Guidelines (SAMGs).

EPRI MAAP 5.04 does not natively enable to model iPWR transients: this code has been adapted by EDF (EDF MAAP 5.04) to make it compatible with the simulation of SAs transients for the SASPAM-SA Design 1. Conversely EPRI MAAPv6.00, the latest version of MAAP enables to natively model iPWR designs and the Design 1 of the SASPAM-SA project. EPRI MAAPv6.00 especially embeds new developments related to the In-Vessel Retention (IVR) evaluations, and support plate modeling that are similar to those implemented by EDF in the EDF MAAP5.04 version.

Comparisons performed between EDF MAAP5.04 and EPRI MAAP6 include accident progression from initial events to long-term in-vessel retention of the corium. Both Design Basis Accidents (DBAs) and SAs were considered. Particular attention was also paid to the solution adopted to reproduce the strong vessel-containment interaction typical of SMRs.

For nuclear disaster prevention, reviews of the Emergency Action Level during severe accidents are being examined based on lessons learned from accidents at Tokyo Electric Power Company Fukushima Daiichi Nuclear Power Station. In the examinations, it is important to consider various severe accident (SA) progressions, including very slow accident scenarios, and behaviors of fission product (FP) release. In this study, SA progression analyses of a typical Japanese 4-loop PWR were performed using the integrated severe accident analysis code MELCOR2.2, in the purpose of obtaining the knowledge utilized for the examinations of nuclear disaster prevention. In the analyses, the evaluation model of a typical Japanese 4-loop PWR was used, considering plant configurations, geometries and structural materials, countermeasure equipment and procedures against SA. SA progressions were compared among loss-of-coolant accidents with the different break sizes, such as guillotine-break, 6 inches-break and 2 inches-break of a hot-leg pipe. The results of the MELCOR 2.2 analyses showed that the speed of SA progression and the amount of FP released to the environment differed depending on the break size. It was also found that the FP releases increased in the late phase of SA progression, and their mechanism depended on the break size.

The Accident Source Term Evaluation Code (ASTEC) developed by IRSN has become a leading tool for the simulation of severe accidents in nuclear facilities. This mechanistic computer code models the entire accident sequence from initiating events to release of the source term outside the containment, including core degradation, containment behavior and fission product transport. Recent enhancements have significantly extended ASTEC's applications to Small Modular Reactors (SMRs) and their passive safety systems, Advanced Modular Reactors (AMRs), Accident Tolerant Fuels (ATFs), spent fuel pool accidents, potential incidents in fusion facilities such as ITER, and severe accident scenarios in fuel cycle facilities. ASTEC's capabilities are now extended to different reactor types, including Western PWRs, Russian VVERs, BWRs and CANDUs. It plays a crucial role in safety analyses, source term evaluations and the development of severe accident management procedures. The code is increasingly being adopted by research organizations, safety authorities and industrial companies for applications in existing and new reactor designs. ASTEC supports probabilistic safety assessments, emergency preparedness and interpretation of experimental programs. The flexibility of the software has facilitated its integration into the new European project ASSAS, which focuses on the use of artificial intelligence for severe accident simulation. This paper provides an overview of the new applications of ASTEC in nuclear reactor simulation and related R&D activities. It highlights the importance of the code in improving nuclear safety assessments and its integration into international projects on advanced nuclear technologies, including European initiatives focused on SMRs and passive safety systems.

Steam explosions represent a critical challenge in the management of severe nuclear reactor accidents, particularly in Nordic Boiling Water Reactors (BWRs), where unique operational and environmental conditions affect accident progression. This study focuses on analyzing steam explosions during severe accidents in Nordic BWRs using a coupled MELCOR-TEXAS computational framework. MELCOR, a widely used code for severe accident analysis, provides detailed thermal-hydraulic and fission product behavior modeling, while TEXAS specializes in simulating fuel-coolant interactions and steam explosion dynamics. By coupling these codes, we achieve a comprehensive simulation environment to evaluate steam explosion scenarios with a higher level of accuracy.

The research investigates core melt progression, melt relocation to the lower plenum, and the conditions leading to steam explosions upon interaction with coolant water. Key parameters assessed include melt jet breakup, premixing dynamics, vapor film stability, and pressure wave generation. The study emphasizes factors influenced by Nordic BWR design features, such as high-density containment structures and emergency core cooling systems. The coupled MELCOR-TEXAS model enables a detailed examination of pressure loads on containment structures, critical for understanding potential damage thresholds.

Results from these simulations enhance our understanding of steam explosion risks specific to Nordic BWRs and support improvements in accident management and containment design. Findings from this work aim to inform safety guidelines and regulatory standards, contributing to robust safety measures in the context of Nordic nuclear facilities and advancing preparedness for severe accident scenarios.

An essential safety measure of advanced water-cooled nuclear power plant designs is the use of passive safety systems (on safety level 3) for the control of design basis accidents (e.g. cooling condensers) or of special devices on safety level 4 for the prevention and mitigation of severe accidents (e.g. ex-vessel core catcher concepts). GRS is developing the code package AC² for simulation of safety relevant phenomena and processes from the initiating event up to the release of fission products to the environment. AC² consists of ATHLET for the simulation of the reactor cooling system, CD for the core degradation phenomena and fission product behaviour in the coolant circuit and COCOSYS for the simulation of all phenomena describing the containment thermal-hydraulic state and potential fission product release to the environment in case of severe accidents.

A key consideration in developing integral simulation codes such as AC² is determining the appropriate level of detail needed to accurately represent all essential processes while ensuring the calculation time remains manageable. This aspect is addressed in this paper with a focus on selected passive systems/devices that found application in Generation III+ reactor concepts, such as the Russian type WWER-1200 light water reactor: Passive heat removal systems and ex-vessel core catcher devices. AC² provides new modelling features for their simulation based on an improved coupling between ATHLET/CD and COCOSYS. The basics of the new COCOSYS modelling concepts for passive systems are described and the paper shows their combined performance in a single calculation for a WWER-1200.

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.

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).

Severe accidents (SA) in nuclear power plants (NPPs) pose critical safety challenges due to the potential release of radioactive aerosols, particularly various iodine species. This study aims to develop advanced filtration technology to reduce the release of fission particles into the atmosphere during SA. Experimental tests were performed to measure the electrostatic precipitator’s (ESP) reduction efficiencies using caesium iodide (CsI, 5 g/l) particles, generated via the droplet-to-particle method. Varying electric voltages were applied across the ESP electrodes depending on the total flow rate to optimize the reduction efficiency. After optimising the operating parameters, the ESP achieved a mass and number filtration efficiency of ~90% for CsI particles, effectively capturing particles in the size range of 0.04–0.5 µm. These findings demonstrate the potential of optimised ESP technology in significantly enhancing NPP safety by effectively capturing radioactive aerosols during SA scenarios.

Estimates from the measurements after Fukushima accidents suggest that substantial part of the
total radioiodine release to environment was formed by volatile compounds, I2 and organic iodides.
The computational efforts to guess on the iodine behavior at an accident are very difficult to carry
out. For example, serious attempts to do such calculations integrally -like those performed in the
OECD BSAF-2 Project for Fukushima- were all limited to modeling of just aerosol forms of iodine.
As “almost anything” can influence the volatile iodine behavior at an accident, we now would like to
know which processes and parameters are really crucial and what might be possibly abandoned. This
cannot be resolved through thousands of runs of one particular containment iodine code (with a
different code, the results will be different). In this work we use engineering estimates, iodine codes,
and methods from the PRA.
We try to identify the processes and circumstances which do matter at the prototypical conditions.
We employ some dedicated Fukushima analyses, both the already published ones and also our own,
new calculations. Secondly, selected experiments and their simulations are analyzed, including
PHEBUS FTP-1 test which is an illustrative example of how the huge uncertainties in the iodine
modeling can influence the interpretation of an experiment.
The findings clearly indicate that the modeling of the volatile iodine formation and behavior is
needed for the basic understanding of what is happening with iodine. However, the focus on really
relevant processes and parameters is necessary. Otherwise we are lost in the plethora of
phenomena, the effects of many are easily smeared out given the significant uncertainties linked to
the iodine analysis.

During nuclear power plant accidents, substantial radioactive materials accumulated in the reactor core may release into the containment or environment, posing radiological hazards. Notably, 60–70% of the core-accumulated iodine inventory is released, with iodine-based aerosols and vapor representing dominant forms of fission products during such events. As a critical severe accident mitigation measure, nuclear power plants employ containment spray systems to depressurize the containment atmosphere and remove suspended radioactive fission products. Investigating the removal characteristics during spray processes is of significant importance for understanding the elimination mechanisms of radioactive iodine-based fission products under accident conditions. This study conducted a series of spray removal experiments using independently developed aerosol and iodine behavior experimental platforms. Results demonstrate that the removal processes of aerosols and iodine vapor generally follow an exponential decay pattern. The spray system rapidly removes aerosols and iodine vapor, with aerosol removal efficiency increasing proportionally to spray flow rate. Smaller droplet sizes resulted in higher removal efficiencies. Additionally, this work analyzed the iodine removal dynamics during spray processes and the speciation of iodine within the containment sump.

As the deployment of Small Modular Reactors (SMRs), particularly integral Pressurized Water Reactors (iPWRs) advances, understanding the behavior of radioactive releases under severe accident (SA) conditions is critical to ensure safety. This study investigates the source term during a selected set of SA scenarios, involving LOCA and non-LOCA type events, occurring in a submerged containment type of iPWR and evaluates the implications for Emergency Planning Zones (EPZs). Unlike traditional reactors, iPWR SMRs feature integrated primary systems, reduced reactor size, and passive safety mechanisms, which impact source term behavior and, subsequently, EPZ requirements. MELCOR is used in this study to develop conservative source terms estimates and MACCS is used to calculate radiological consequences. This research examines potential release pathways, the performance of passive safety systems, and containment responses. The study quantifies key parameters such as fission product release rates, containment retention effectiveness, and potential environmental impact, focusing on how these factors influence EPZ size and scope. It is observed that, with the Swedish dose criterion of the current regulatory framework, iPWRs do not necessitate a precautionary action zone (PAZ), but an urgent protective action planning zone (UPZ). Even in the most severe accident scenario, the UPZ is around 16.1km. Comparative analysis with conventional reactors identifies enhanced containment capabilities of iPWR SMRs, suggesting the potential for smaller EPZs.

Accurate assessment of radioactive source terms following severe accidents of nuclear power plant is critical for off-site consequence analysis and calculation of radioactive release frequencies. Currently, worldwide research institutions have conducted plenty of studies on uncertainty and sensitivities of nuclear power plant radioactive source terms based on Best Estimate Plus Uncertainty (BEPU) analysis method; however, studies on the impacts caused by actions from the Severe Accident Management Guidelines (SAMG) are limited. This paper first develops a procedure for uncertainty and sensitivity analysis of radioactive source terms for nuclear power plant (NPP) based on BEPU method, especially considering inclusion of SAMG actions. Secondly, integrated severe accident analysis code MAAP was used to build a model of China typical two-loop nuclear power plant and 200 MAAP input cases using Latin hypercube sampling method were generated through own developed sampling code. Then 200 cases were run by MAAP and statistical evaluated by self written code, and uncertainty and sensitivity analysis was performed. Uncertainty analysis results indicate that all selected parameters have certain influences on release source terms in containment intact release category; and sensitivity analysis shows that the impact of SAMG actions may far larger than those from severe accident phenomena or model uncertainties. This study highlights the importance of selecting accident sequences for release categories with particular attention to SAMG action uncertainties during source term calculation. The research provides valuable references for selecting representative accident sequences and conducting uncertainty/sensitivity analysis of radioactive releases following severe accidents of nuclear power plants.

Severe accident analyses in nuclear power plants are highly complex and models applied in simulation codes are often derived based on limited amount of available experimental data. Evaluating the accuracy and uncertainty of such models provides valuable information for safety analyses, as well as further development needs and helps the priorisation of experimental resources. Consequently, there has been a growing interest in the recent years in uncertainty quantification and BEPU analyses in the context of severe accident analyses.

This paper investigates a postulated cold leg break combined with a station blackout scenario in a generic German PWR using the AC2 code package, developed by GRS. Both the reactor cooling circuit and containment are modelled in detail and the scenario is analysed starting from normal operational conditions up to the evaluation of the source term into the environment.

Based on the best estimate simulation an uncertainty and sensitivity analysis is performed. The selected uncertain input parameters are propagated through the model and the 95/95 uncertainty ranges are determined based on Wilks` findings. Furthermore, the Spearman Rank Correlation Coefficient is derived as sensitivity index. This allows to characterize both the variation in model responses and identify important uncertain parameters.

In 2019 was launched the Horizon-2020 MUSA project with the aim of reviewing the uncertainty sources as well as defining Uncertainty Quantification methodologies for assessing Severe Accidents (SA) scenarios. Energy Software Ltd. (ENSO) contributed to the Working Package 5 “reactor applications” with the simulation of a long-term Station Black Out (SBO) occurring in a Gen-II PWR. The Uncertainty Analysis (UA) was carried out with RELAP/SCDAPSIM/MOD3.4 (RS3.4) and its IUA2.0 module, using the Wald multi-variable form of the Wilks’ equation. A group of 20 input parameters and 19 Figures of Merit (FOM) were selected to assess the uncertainty propagated to the Source Term (ST) released to the containment in an in-vessel simulation. To improve the ST capabilities of the code, a Fission Product Transport model was implemented into RS3.4. The relevant outcome of ENSO contribution was the estimation of tolerance regions for the Fission Products, Gases and Debris materials released at the vessel failure. Such outputs can be then used as initial conditions and/or probability density functions (PDF) for ex-vessel calculations performed with containment codes. The sensitivity analysis was conducted for the original sample used in the UA, and for an increased sample size, to evaluate the consistency of the correlation coefficients and resulting PDFs. The results showed rather large standard deviations for some of the output parameters because of cliff edge phenomena in the material slumped to the lower plenum. Such results suggested the use of one-sided tolerance limits to set the initial conditions for posterior containment UA.

In 2019, the International Atomic Energy Agency (IAEA) launched the five-year Cooperative Research Project (CRP) I31033 to advance the understanding and characterization of sources of uncertainty and investigate their effects on the key figures-of-merit (FOMs) of the severe accident code predictions in water-cooled reactors (WCRs). Energy Software Ltd. (ENSO) contributed with an assessment of the uncertainty propagation in a long-term SBO scenario postulated for a CAREM-like integral PWR (iPWR). The study aimed at demonstrating the RELAP/SCDAPSIM/MOD3.5 (RS3.5) capability to carry out a BEPU calculation of a Severe Accident scenario in a single sequence from operational conditions to Reactor Pressure Vessel (RPV) creep rupture. The uncertainty analysis was conducted with the IUA2.0 module integrated into RS3.5 code, using the input-propagation methodology with a statistical description of the uncertainty proposed by Wilks. A group of 20 input parameters and 10 Figures of Merit (FOM) were selected for the assessment. The input parameters included boundary and initial conditions, material properties and code correlations, and the FOMs were related to the time of the main events and the fission product releases. To support the results, the Pearson, Spearman and Kendall correlation coefficients were analyzed for the selected input parameters and FOMs by using scalar values tables of the significance level. The relevant conclusions of the assessment are first, the importance of using the relative time for FOMs during core damage progression, and second, that including the Kendall formulation is advisable because it seems less dependent to singular data.

In an effort to address the inherent uncertainties in severe accident codes used in nuclear accident analysis, Energy Software S.L. (ENSO) has embarked on an innovative project aimed at developing a stepwise methodology for analyzing these uncertainties. The results obtained from two international projects, IAEA CRP I31033 and HORIZON-2020 MUSA, have provided ENSO with a solid foundation to identify and quantify the sources of uncertainty in severe accident analyses. However, they also revealed significant limitations, such as excessive computation time, multiple simulation errors, and truncation effects.

The proposed approach is stepwise, applying the Wilks/Wald method in two consecutive phases: the first linked to the "in-vessel" phase, where the accident prior to vessel failure is analyzed, and the second focused on the "ex-vessel" phase, examining subsequent events in the containment. With this methodology, ENSO plans to develop a model of a Generation II four-loop Westinghouse PWR reactor to simulate the "ex-vessel" phase of a low-pressure station blackout (SBO) accident using the MELCOR code. For the "in-vessel" phase, a previously created model with RELAP/SCDAPSIM will be used. The project also encompasses the development of pre- and post-processing tools with Python for uncertainty analysis with MELCOR, and the selection of input parameters based on probability distributions to apply the Wilks/Wald methodology.

Implementing the stepwise methodology will allow the identification and quantification of the sources of uncertainty at different stages of the accident, providing critical information for decision-making in emergency situations and for designing future research in this field.

In the process of severe accidents of the pressurized water reactor (PWR), the high temperature on the primary side of the steam generator and the high pressure difference between the primary and secondary sides can pose a high risk of creep rupture for the heat transfer tubes. Once the heat transfer tubes rupture, they lead to a bypass of the containment shell, causing radioactive materials to be directly released into the environment, and creating significant safety issues. The present study uses software to investigate the changes in parameters on the primary and secondary sides of the reactor under severe accident conditions, and calculates the rupture probability of the steam generator heat transfer tubes based on these parameter changes. MELCOR is used to simulate the station black out event (SBO), which has great impact on the heat transfer tubes. The main objective of the study is to clarify the changes in pressure and temperature on both sides of the heat transfer tubes under this accident condition, and to calculate the creep failure probability. The influence of microscopic and large cracks on the failure time of the heat transfer tubes is calculated. Considering that the surge line has the same failure risk, this study also shows the probability of the heat transfer tubes failing before the surge line.

In the frame of the French ANR Post-Fukushima ICE program (“Interaction between Corium and Water”), a series of Fuel Coolant Interaction (FCI)/Ex- vessel Steam Explosion (EVSE) integral tests have been performed at the KROTOS facility of the Severe Accident platform PLINIUS, located at CEA-Cadarache. Complementary, thermodynamic and thermophysical properties of prototypic corium chosen for integral tests KROTOS have been measured on VITI facility (CEA-Cadarache) and ATTILHA facility (CEA-Saclay). Post-test analyses on corium steam exploded debris have been performed through high-resolution X-ray diffraction measurements done on the MARS beamline at the SOLEIL synchrotron radiation source. The experimental research of ICE program was focused on fragmentation, dispersion of corium jets and formation of debris beds mechanisms, steam explosion energetics, corium oxidation and solidification mechanisms. The up-grade of KROTOS facility and new configurations for VITI and ATTHILA facilities to answer to the ICE program scientific objectives will be presented. Three integral KROTOS tests will be described and the knowledge gained for FCI/SE modelling will be discussed. A special focus will be done in the assessment of thermodynamic and thermophysical corium properties measurements and modelling. The very new results obtained concerning the corium final solid state and the cationic composition fluctuation that occurs in the U_1-xZr_xO_2-y solid solution will be presented.

Advanced Technology Fuels (ATF) cladding materials have become a key research focus worldwide, particularly at KIT, due to their promising potential to enhance reactor safety under accident conditions. FeCrAl and Cr-coated Zirconium alloys are designed to reduce hydrogen generation at least at the beginning of severe accidents (SAs). Their application in Small Modular Reactors (SMRs) is particularly relevant, as SMRs are emerging as a safer alternative to traditional Nuclear Power Plants (NPPs) due to their reduced core inventory and advanced safety systems. Within the framework of the EU SASPAM-SA project, this study focuses on the analysis of hypothetical SA scenarios involving a generic integral Pressurized Water Reactor (iPWR) with natural circulation, using the integral code ASTEC v3.1.2, ASNR all rights reserved, 2024. This code models thermohydraulic and physicochemical phenomena, allowing a detailed assessment of the accident progression from the initial event to the potential release of the radioactive material to the environment. This study specifically examines the performance of Zircaloy-4 and ATF cladding materials, focusing on their influence on the core degradation and hydrogen generation. The results show distinct hydrogen release kinetics for ATF materials compared to Zircaloy, emphasizing the impact of cladding properties on hydrogen production and safety margins during SAs.

The optimization of post-accident management in case of a severe accident (SA) is complex, particularly in what concerns handling of nuclear materials. After a SA, most of nuclear materials remain within the nuclear power plant (NPP) units in a solidified state, usually referred to as corium. The dismantling phase entails cutting such corium chunks into manageable pieces without causing an unnecessary radioactive remobilization to the gas phase that might result in further source term to the environment. This is currently the stage to be faced shortly in the Fukushima Daiichi site.

The OECD/FACE (Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation) project is dedicating significant resources to finding the best process for dismantling the site. Achieving this goal requires exploring different techniques and protection measures. This work is an exploratory analysis on how effectively an overlying water layer could absorb particulate material generated during laser cutting, in preparation for retrieving fuel debris from the affected units. Using the SPARC-Jet code, an in-house extension of SPARC-90 (Suppression Pool Aerosol Removal Code), the influence of uncertain boundary conditions on water retention efficiency has been studied. The focus was on factors such as carrier gas flow rate, particle size and concentration, pool temperature, and water depth. Preliminary results suggest that injection spot diameter and injected gas mass flow rate lead to higher Decontamination Factor (DF) values. However, from the cleaning efficiency standpoint, variables such as water temperature or depth should not be a concern, as their effect is very minor.

This work is being carried out in the context of severe accidents mitigation in sodium-cooled fast reactors (SFRs). The corium (set of molten core materials) formed may be transferred through discharge tubes to the lower part of the reactor vessel, towards a core-catcher. However, this corium could reach the core-catcher in the form of a hot jet (3000K), which could lead to local ablation of the core-catcher. This risk must therefore be taken into account to ensure that the core-catcher retains its integrity during corium relocation phase.

In the present work, the effect of the core-catcher geometry on its ablation process by a hot jet is investigated experimentally. Experiments were conducted on HAnSoLO setup, with simulating materials (transparent ice /jet of water). The experimental conditions were determined to be as representative as possible of those of a nuclear reactor. The geometric features of the core-catcher which are studied are its inclination and roughness. It has been observed that these two parameters significantly influence the ablation phenomenon, and in some cases can increase the ablation rate.

Sodium Cooled Fast Reactor (SFR) is one of the most promising Gen-IV concept for earliest deployment, owing to vast operating experience worldwide. Currently operating SFRs adapted partial core meltdown as a design basis for Core Cather (CC). However, the safety criteria for Gen-IV demands demonstration of safe mitigation of whole core accident and accordingly the CC design shall consider in-vessel retention and long-term cooling of the degraded core. Whole core retention would impose considerably higher heat flux and the CC need to withstand higher thermomechanical loads. To fulfil this requirement, development of an advanced core catcher has been taken up at IGCAR, India. The main objective is to develop and qualify a refractory protective layer for the CC, which is compatible with sodium and can withstand severe thermal transients expected during corium relocation.

Based on several tests in-house, refractory magnesia was identified as a candidate material for protective lining on a stainless-steel substrate. Dedicated experiments were conducted with magnesia test specimens to study i) long-term sodium compatibility, and ii) resistance to thermal shock under simulated accident conditions. Based on the microstructure and phase analysis, the sodium compatibility was assessed whereas degradation of the specimens was determined from the destructive/ non-destructive tests before and after the experiments. Results showed the magnesia specimens to have excellent sodium compatibility and good resistance to thermal shock, indicating the magnesia lined CC to be a potential option as advanced CC for future SFRs. Design concept, experimental methods and important results are discussed in the paper.

Based on both the internal-trigger and external-trigger experiments conducted by Shanghai University of Electric Power, air entrainment is proved to be a significant factor that affects the triggering on the surface of molten droplets during fuel-coolant interaction (FCI). In this study, based on the Rayleigh equation, the mass ratio of steam to entrained air, and the disturbance amplitude and interface pressure difference at the vapor-liquid interface under different working conditions are calculated. The relationship between the air volume, the disturbance amplitude and the trigger time of the molten droplet, and the relationship between the interface pressure difference and the trigger strength of the molten droplet surface are thus analyzed. It is revealed that the air entrainment can stir the disturbance amplitude, thereby reducing the trigger time of the molten droplet. The variation of the trigger strength of the molten droplet surface is consistent with that of the vapor-liquid interface pressure difference. In order to further exhibit and verify the phenomenon, the breakup of the steam envelope with air entrainment is simulated by Moving Particle Semi-Implicit (MPS) method, and it is quantitatively estimated that under the air mass ratio of 50% and 90%, the air entrainment may reduce the trigger time of the steam envelope by nearly 1.4ms.

An ex-vessel core catcher for advanced light water reactors is being developed to stabilize molten corium outside the reactor vessel and prevent molten corium concrete interaction (MCCI) during severe accidents. The core catcher includes a sacrificial concrete (SC) layer, carbon steel body, protective material, and an external cooling channel. Its key function is to collect molten corium and remove heat, while preventing re-criticality and excessive hydrogen generation. During severe accidents, molten corium discharge and some parts of a reactor pressure vessel (RPV) are expected to impact directly on the core catcher body. Given their high momentum and large internal energy, it might degrade the integrity of the structural bodies, affecting safety performance of the core catcher. To investigate these scenarios, we developed a smoothed particle hydrodynamics (SPH) framework to analyze the interaction among the molten corium, reactor pressure vessel, and core catcher. As a fully Lagrangian particle-based method, SPH is suitable for handling free surface of corium flows and structural topology changes in the RPV and core catcher body. The explicit incompressible SPH (EISPH) model simulates corium behavior with an implicit viscosity solver for computational efficiency. A finite multiplicative model based on Total Lagrangian SPH (TLSPH) handles large-strain elastoplasticity in the structures. The coupled EISPH-TLSPH model is used to investigate mainly corium relocation behaviors and structural integrity under several potential accident scenarios. We expect that the proposed model will be useful for accurate safety analyses of the accident mitigation strategy using the ex-vessel core catcher.

This paper presents the analysis of the effect of Molten Corium Stratification (MCS) during Molten Corium Concrete Interaction (MCCI) using developed the PUMBAA (Prediction modUle of MCCIs with Basemat Attack and Ablation) code under the severe accident analysis platform called SAFARI (Safety Analysis Code For Severe Accident Risk Identification) currently being developed in Korea. PUMBAA built upon CORQUECH developed by Argonne National Laboratory in US, as a reference code focuses on the MCCI phenomena, aiming to extend its ability to analyze MCCIs phenomena interlaced with Molten Corium-Water Interactions (MCWIs) and Containment Thermal Hydraulics (CTHs) along with Severe Accident Management (SAM) actions during severe accidents. In this study, the development and performance of PUMBAA’s corium stratification analysis capability are introduced and validated against the MOCKA 3.1 experiment. Part of the MOCKA series, this experiment simulates a dry 2D cylindrical siliceous concrete cavity, where continuous corium pouring induces various physical phenomena. The paper first describes the modeling of corium stratification in PUMBAA and key physical phenomena in the MOCKA 3.1 experiment. It then presents the validation process and result analysis. The results show that the PUMBAA code, through its corium stratification analysis, effectively accounted for the physical phenomena of the stratified oxide and metal layers observed in the MOCKA 3.1 experiment, accurately predicting the trends in the experimental data.

The melt jet breakup is an important phenomenon for assessment of the ex-vessel phase severe accident, which is highly related to the debris bed coolability. The violent two-phase boiling is accompanied by the melt jet breakup phenomenon due to the high temperature of the melt, and it can affect the jet breakup behavior. The effect of the vapor generation intensity, which represents the two-phase mixing zone behavior, was investigated by controlling both the melt and water temperatures. The melt jet breakup length was observed by visualization using high speed cameras. Based on the experimental observations, the effect of the vapor generation intensity was confirmed. As the vapor generation intensity increases, the jet breakup length became longer. Therefore, the parameter for the vapor generation intensity was suggested to develop the new jet breakup length correlation including the vapor generation intensity parameter so that the existing correlations (Saito correlation and Epstein & Fauske correlation) could be integrated into single correlation.

Hualong One is the first independently developed million-kilowatt-class pressurized water reactor nuclear power plant in China, which meets the design standards of the third generation nuclear power technology. The Cavity Injection and Cooling System (CIS) for the reactor vessel is one of the measures to mitigate severe accidents in Hualong One, and its unique combination of active and passive technologies can effectively prevent the rupture of the reactor pressure vessel and achieve the retention of molten debris within the reactor. Based on the layout features of the CIS system in Hualong One and the design experience of the first unit, this paper proposes design optimization solutions and improvement measures from the perspective of layout design for subsequent projects, improving the compact arrangement of pumps and valves in the plant, and providing valuable design experience for future PWR nuclear power plant design in China.

In a postulated station blackout accident, the fuel channels inside of a Canada Deuterium Uranium (CANDU) reactor would dry out, heat up, and collapse to the calandria vessel bottom. Due to decay heat generation, the debris bed would continue to heat, compact, and melt, forming a molten corium pool. Here, the transient heating and melting of a compacted debris bed is simulated using unsteady Reyolds-averaged Navier-Stokes based computational fluid dynamics.

A time-varying decay heat is used with the starting conditions representing post moderator dry-out. A source-based enthalpy-porosity phase change model is employed to capture the non-eutectic melting process, accounting for a 500K difference between the solidus and liquidus temperatures of the corium. The developing molten region, characterized by a maximum modified Rayleigh number around 10¹², is modelled with the k-ω turbulence model. Turbulence is allowed to develop from an imposed very low level in the melting region consisting of a growing liquid pocket and a partially molten layer, as the temperature locally exceeds the solidus temperature.

Heat flow through the vessel wall to the surrounding water is modelled with a conjugate boundary, and a convection-radiation boundary is applied to the corium top surface. Verification and validation cases are done based on previous studies using molten-salt corium simulants.

The evolution of the molten corium velocity and temperature fields, the unmolten crust thickness, as well as their impact on the exiting heat flux are presented and analyzed. These findings assist the in-vessel retention studies of CANDU reactors and inform future modelling efforts.

Hydrogen distribution in the reactor building of Fukushima Daiichi NPP Unit 3 during the accident was analyzed using the CFD code BAROC. The main feature of BAROC is that it solves the pressure Poisson equation based on energy conservation. This makes the calculation stable and fast, even under sudden changes of fluid conditions. The total number of spatial meshes with 50 cm cubic was approximately 750,000. Transient from the hydrogen inflow to 18 hours later were analyzed in 14 parallel using an Intel Gold 5218 2.3GHz CPU, and the computational time was almost real time. Initial conditions were assumed to be that the building was filled with air at room temperature and atmospheric pressure. Hydrogen was assumed to have entered the building from a shield plug on the 5^th top floor of the building. The analysis assumed that 75 tons of steam and 650 kg of hydrogen would both enter the building. Although the blowout panel in the 5^th floor was not opened at the accident, it was assumed that there was a certain amount of leakage because the building was not a leak-tight structure. Analyses were performed for 3 cases with 2%, 3.3%, and 6.6% leakage of the blowout area. The analysis showed that the hydrogen concentration in the 5^th floor was within the flammable range regardless of the leak area. It was also found that when the amount of inflow hydrogen was increased to 1,300 kg, the hydrogen concentration was within the detonation range in the 5^th floor.

Direct Air Capture (DAC) is a technology that separates and captures CO₂ contained in trace amounts in the atmosphere. It is the only existing CO₂ capture process with a negative net emission and is receiving attention as an active CO₂ removal technology (Carbon Dioxide Removal (CDR)) to achieve net zero. Currently, DAC has low technological maturity overall, and the large amount of air intake requirement and the large amount of heat energy consumption for regeneration are pointed out as major bottlenecks in technology commercialization. Linking with nuclear power, a carbon-free power/heat energy source, is one of the effective strategies to resolve the technological bottleneck of DAC, but the research and development for this is still in the basic development stage. In this paper, at first, it is explored the concept development research of the nuclear power-DAC combined system currently in progress and evaluate the development level. In addition, it will review the technical feasibility of the DAC system, development of the system, the current level of technological development, and propose technology candidates for linking with nuclear power.

In the Republic of Korea, the development of the innovative Small Modular Reactor (i-SMR) is ongoing. The i-SMR will be equipped with the following three passive safety systems to replace the active safety systems of existing commercial nuclear power plants: Passive Emergency Core Cooling System (PECCS), Passive Auxiliary Feedwater System (PAFS), and Passive Containment Cooling System (PCCS). The PECCS performs the core makeup/cooling function, the PAFS removes residual heat with theby steam generator (SG) cooling, and the PCCS conducts the heat removal from the containment vessel (CV) atmosphere. Since the passive safety system can carry out safety functions by natural forces without continuous power supply or any operator action, it is expected to dramatically improve the safety of nuclear power plants (NPP). In this paper, we present the design status of the passive safety systems in the i-SMR and the performance analysis results using MARS-KS.

This paper extends the evaluation of an adaptive algorithm for estimating decay heat in transient scenarios with steadily decreasing reactor power. The previously introduced approach offers a low-cost alternative with simpler implementation to traditional methods that typically rely on extensive nuclide tracking or standardized procedures. The adaptive algorithm utilizes precomputed decay heat curves from SCRAM scenarios, and enables real-time decay heat estimation during simulation while dynamically adjusting to varying power levels without requiring detailed nuclide tracking.
This study applies the adaptive algorithm to additional non-SCRAM shutdown scenarios in a sodium-cooled fast reactor, demonstrating its flexibility in handling diverse transient conditions. Verification is performed by comparing results with Monte Carlo depletion calculations using Serpent, confirming high accuracy. Furthermore, the model is successfully applied to FFTF LOFWOS Test #13, showing good agreement with experimental decay heat data. These results demonstrate the potential of the adaptive method as an accurate decay heat estimation approach for real-time transient simulations of advanced reactors.

The water-based Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory is a large-scale test facility built to study passive decay heat removal performance of one Reactor Cavity Cooling System (RCCS) concept for advanced nuclear reactors. The inventory fill within the primary water tank is known to have an impact on the natural circulation thermal hydraulics and two-phase phenomena development, including instabilities such as startup oscillations. Six inventory levels, from 20% to 80% initial fill, were examined. The facility was heated from room temperature to an input power corresponding to 2.1 MW_t full-scale (51.6 kW_t NSTF-scale) and operated at saturation conditions for at least four hours, uninterrupted. While minor changes to integral thermal hydraulic characteristics and instabilities were observed, ultimate heat removal performance was not impacted by decreasing inventory. Following the last 20% fill test, a depletion scenario was performed where the facility continued operating at saturation conditions with boil-off until reaching critical levels where natural circulation flow stagnated. Comparisons were made to similar, previous inventory parametric series and depletion tests. The height of the tank inlet has an impact on the development and suppression of the natural circulation instabilities as well as natural circulation stagnation conditions. However, the lower tank inlet resulted in significant increase in available inventory prior to stagnation in a boiloff scenario. Additionally, no short-circuiting effects between the hot and cold legs were observed as a result of the lower tank inlet.

The objective of the present study is to investigate the effects of varying decay heat loads and tank inlet elevation on the performance of a large-scale water-based Reactor Cavity Cooling System (RCCS). The water-based Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne represents a ½ axial scale and 12.5° sector slice of the full-scale Framatome 625 MWt SC-HTGR RCCS concept. A power parametric series with prototypic decay heat loads of 1.4, 1.75, 2.1, and 2.4 MWt (34.4, 43.0, 51.6, and 58.5 kWt scaled) with the lower tank inlet configuration was first discussed. System performance metrics at the two-phase quasi-steady state were compared, mainly the system flow, steam generation rate, system pressure, and liquid and structure temperatures. The effect of the decay heat power level on the boiling front progression was also investigated. It was found that lower decay heat loads would cause delayed boiling front progression. Results were then compared to previous data sets collected at identical testing condition but with chimney pipnig configured at the mid tank elevation configuration. Overall, the tank inlet elevation did not significantly influence the system two-phase quasi-steady-state performance at the studied decay heat levels. However, the tank inlet elevation was found to have significant impacts on the two-phase flow instability characteristics and the boiling front propagation. At the lowest power of 1.4 MWt with the lower tank inlet configuration, boiling never propagated from the tank to the upper chimney, while boiling propagation to the upper chimney was observed in the mid tank inlet configuration.

The lead-cooled fast reactor (LFR) is one of the promising Gen-IV designs which is being actively developed by several commercial vendors. One of the safety aspects of interest for LFR designs is the potential burst of a fuel pin.

During a fuel pin failure event, a jet of gaseous fission products is ejected into the coolant subchannels adjacent to the failed pin. The jet has a relatively high momentum due to the pre-pressurized nature of fuel pins. The question of interest is whether the gas jet and subsequent gas bubble formation in coolant subchannels could potentially thermally blanket adjacent fuel pins, leading to them failing. Hence, a potential chain failure propagation across the core is imaginable.

The aim of the present work is to experimentally investigate the bubble formation, location and behavior.

The experimental setup used for the investigation consists of a liquid metal loop equipped with high-resolution measurement techniques. First, experiments were conducted using a single sub-channel test section, combined with high-speed imaging. This experimental campaign was used to gain novel first general insights into the behavior of a buoyant gas jet in low Prandtl, high-density liquid. A second test section was then built, allowing for a multichannel observation of the phenomenon. The paper will include the results of the two experimental campaigns and the conclusions that could be drawn concerning the potential occurrence of fuel pin failure propagation in LFR cores.

After the SGTR (steam generator tube rupture) accident in the LFRs, the high-pressure water on the secondary side enters the primary side and is heated to generate a large number of bubbles, which may hinder the flow of LBE in the reactor core, cause heat transfer deterioration, and threaten the nuclear safety. The behavior of bubbles in the fluid phase is obviously affected by the size of the flow channel. However, there have been relatively few systematic studies on the influence of channel size on bubble distribution characteristics in LBE. The drift flux model is one of the most successful models to predict the distribution of void fraction in gas-liquid two-phase flow. In this paper, based on the upward flow experiment of gas-LBE two-phase flow in circular pipes with various hydraulic diameters, the phase distribution parameters were measured, and the influence of channel size on the parameters of drift flow model was studied.

After the occurrence of a steam generator tube rupture (SGTR) accident in lead-bismuth fast reactors, the interface evolution and transport of bubbles can lead to bubble aggregation, thereby affecting core safety. The interfacial area transport equation (IATE) is an important method for predicting IAC and has significant applications in system analysis codes. In this study, a nitrogen-liquid lead-bismuth metal two-phase flow experiment was conducted in a vertical circular tube channel, and the local interfacial area concentration (IAC) was measured. The measurement data reflect the radial distribution and axial development characteristics of IAC and reveal the evolution and transport characteristics of interfaces. Additionally, this study reviewed the available IAC prediction models including IAC correlations and IATE. However, most of these prediction models have been not developed for the gas-liquid metal two-phase flow, the experiment database was used to verify the applicability of these models in the liquid lead-bismuth metal fluid. The verification shows that the IAC correlations cannot give good predictions of the IAC in liquid lead-bismuth two-phase flow, while the IATE could have a better prediction result, but there is still some difference from the experimental measurement IAC.

During a steam generator tube rupture (SGTR) accident in a lead-cooled fast reactor (LFR), vapor entering the core can induce power excursion and threaten reactor safety. Accurately predicting bubble transport in LFR during SGTR is crucial for its safety assessment. This paper uses a neural network (NN) to predict the bubble distribution within the Europe Lead Cooling System primary system during SGTR accidents. The NN-based model uses one-hot encoding to accommodate heterogeneous inputs and implements a modified Softmax function to avoid non-physical outputs. The method of deep ensembles then quantifies the prediction model uncertainties. The prediction model can accurately predict bubble distributions at three different locations. A relatively large ensemble size is required to converge the ensemble mean, while the convergence of ensemble standard deviation may suffer from outlier samples. Ensemble predictions at different locations tended to be negatively correlated, which usually became weak near extreme values (0 and 1).

The steam generator heat transfer tube rupture (SGTR) accident can lead to violent interactions between lead-bismuth liquid metal (LBE) and water in lead-cooled fast reactors, which can seriously threaten the safety of the core. In this paper, high parameter experiments and refined numerical simulations are used to investigate the lead-bismuth liquid metal-water interaction mechanism. Thermocouples and pressure sensors were used to capture the fluctuation behavior in temperature and pressure in the experiments. This complex and opaque internal interaction is modeled by constructing dynamic boundary conditions of multiphase and multiphysics processes. We demonstrated the existence of three stepwise sequential interaction mechanisms. Moreover, special phenomena such as vapor film wrapping around the core of the jet and secondary penetration have been discovered. his study provides new insights into the interaction between LBE and water and offers important reference for developing mitigation strategies for SGTR.

This study investigates the optimization characteristics of a secondary-side passive residual heat removal system (PRHRS) for enhancing inherent safety performance in lead-bismuth cooled fast reactors (LFRs). Using the fully implicit NUSOL-LMR code with fluid-structure coupling, analyses demonstrate that the PRHRS activates secondary-side natural circulation during unprotected transient overpower (UTOP) and unprotected loss of heat sink (ULOHS) accidents, reducing core temperature rise by 12.3% (p<0.01) while maintaining fuel temperatures 231 K below melting thresholds. The system synergizes with inherent reactivity feedback (coolant density feedback: −1.22 pcm/°C, Doppler feedback: −0.663 pcm/°C) to suppress coolant solidification risks. Under UTOP conditions, PRHRS caps peak cladding temperatures at 1,147 K (52% below safety limits), whereas during ULOHS, it sustains decay heat removal via secondary-side passive flow (2.3% rated capacity). Results conclusively show that integrating the passive system significantly enhances inherent safety under extreme accidents, substantially mitigating potential risks. These findings provide critical insights for optimizing safety designs in forced-circulation LFRs.

During a severe accident in a light water nuclear reactor, large amounts of hydrogen could be generated and released into the containment during reactor core degradation. Additional burnable gases (H₂ and CO) may be released into the containment in case of molten corium/concrete interaction. As observed during the Fukushima accidents, H₂ and CO combustion could cause high pressure peaks that could challenge the reactor containments.To prevent this risk, most of the mitigation strategies adopted in European countries are based on the implementation of Passive Autocatalytic Recombiners (PARs). Nevertheless, studies indicate that, despite the installation of PARs, it is difficult to prevent, at all times, the formation of a combustible mixture potentially leading to local flame acceleration. To better understand the phenomena associated with the combustion hazard and to address the issues highlighted after the Fukushima events, such as the explosion hazard inside the venting systems. The AMHYCO project aims to propose innovative enhancements in the way combustible gases are managed in case of a severe accident in operating reactors.

As first step, a critical review of the available literature had been performed with the objective to form the basis for the project regarding (1) PAR efficiency under ex-vessel conditions, (2) existing PWR Emergency Operating Procedures (EOPs) and SAMGs regarding containment risk management (3) H2/CO combustion and the available engineering correlations for combustion risk estimation, (4) equipment and instrumentation surveillance under severe accident conditions. This paper provides a survey on the available literature related to the four topics mentioned above.

Most countries with nuclear power plants have implemented measurement systems to assess hydrogen concentration by extracting air from the containment building during severe accidents. This method samples the atmosphere to determine hydrogen concentration, rather than depending on sensors within the containment, to preserve sensor integrity and ensure accurate readings. However, uncertainties may emerge. Firstly, steam condensation during sampling can change the gas composition ratio. Secondly, the recorded time for hydrogen concentration includes a delay from sampling, which can be compared to the time taken for direct pressure and temperature measurements inside the containment. This time lag may influence flammability predictions based on thermal-hydraulic conditions.
We are evaluating the uncertainties associated with hydrogen concentration measurements. The sampling-type measurement system consists of an accident environmental control module, a hydrogen concentration sensing module, and a data processing module. The control module extracts mixed gases from an experimental setup that simulates an accident environment, filters them through a water tank, and returns them to the experimental vessel. The pool scrubbing method eliminates steam and aerosols. Part of the mixed gases passing through the water tank is entering the sensing module with a consistent sampling flow rate. The data processing module employs an algorithm to adjust the output values from the hydrogen sensor, accurately reflecting the true hydrogen concentration in the accident environment. The accuracy of hydrogen concentration monitoring in a severe accident environment is assessed based on the test results obtained from the experimental matrix.

The generation of combustible gases (H₂ and CO) during a severe accident (SA) and their potential accumulation in the containment atmosphere could threaten the containment integrity and/or safety components in case of uncontrolled combustion. The AMHYCO project (2020-2025), funded by the European Commission, aims to enhance the understanding of H₂/CO combustion risk within the containment of a nuclear power plant, particularly in the late phase of a severe accident, to revise the management of combustible gas risk.

This work, performed in the frame of AMHYCO, explores the impact of passive autocatalytic recombiners (PARs) performance on SA progression, and particularly their interaction with other safety systems (i.e., sprays and fan-coolers). Two Western PWR scenarios (a double-ended guillotine LOCA and an SBO) were simulated with the MELCOR 2.2 code. In the LOCA scenario, steam concentration is strongly reduced shortly after the initiating event by the automatic spray actuation. The suppression of steam promotes the formation of flammable gas mixtures in the ex-vessel phase. Parametric cases showed that cooling systems' unavailability or deactivation could reduce combustion risk. Contrarily, the SBO accident initially evolves at high pressure with a high steam content in the containment. In this sequence, a late spray operation significantly affects the gas mixture's flammability. In both sequences, the oxygen depletion by the PARs operation leads to containment inertization in the late phase of the accident. As a future step, CIEMAT will launch a calculation campaign to assess how uncertainties may impact the insights gained through best-estimate analyses.

The mitigation of the hydrogen risk with passive auto-catalytic recombiners (PARs) is state-of-the-art in nuclear power plants with water-cooled reactors. In the ex-vessel phase of a severe accident, the operation of PARs faces several challenges. While hydrogen is continuously released from the interaction between molten corium and concrete, carbon monoxide is also produced, along with other gases. Inside the PAR, hydrogen and carbon monoxide compete for the available oxygen, which is continuously consumed. As a consequence, the performance of the PAR in terms of recombination rates and overall efficiency decreases. In order to enable a realistic assessment of the availability and performance of the measures to control combustible gases, numerical models developed for PAR operation during the in-vessel phase need to be enhanced towards these boundary conditions.
The REKO-DIREKT code was developed to describe the operational behavior of PARs with a mechanistic approach. This model describes heat and mass transfer processes inside the catalyst section in detail in order to simulate PAR performance and conversion efficiency. Based on an experimental program which was part of the European AMHYCO project, REKO-DIREKT was further developed to cope with the ex-vessel conditions of a severe accident.
The paper describes aspects of the code development work and gives examples for the code validation based on the existing database.

The Japan Nuclear Regulation Authority recognizes that the hydrogen explosion at Fukushima Daiichi Nuclear Power Plant Unit No. 3 originated not on the operating floor but on the lower level of the reactor building. This study aims to obtain knowledge on hydrogen behavior by examining analytical conditions under severe accident scenarios to develop an evaluation method for the retention and diffusion behavior of hydrogen leaked into the lower level of the reactor building, which represents a typical BWR plant in Japan. Based on plant walk-down data and the results of safety analysis evaluation of the actual plant, the dimensional shape and heat transfer characteristics of the floor area where hydrogen may leak, and the fluid characteristics of the leaking gas, were organized. An analytical model was developed using representative parameters as basic conditions. Sensitivity analysis of various parameters showed that the height from the leak point to the ceiling and the horizontal distance to the ceiling cavity were highly sensitive to the hydrogen concentration in the ceiling cavity. The hydrogen concentration increased as the vertical distance from the leak location to the ceiling decreased, and as the horizontal distance to the ceiling cavity decreased. By contrast, other parameters, such as the temperature of the leaking gas, had little effect on the hydrogen concentration in the ceiling cavity. The results of the sensitivity analysis indicate that these are the three main factors that increase the hydrogen concentration in the ceiling cavity.

During a severe accident, hydrogen distribution in a containment building and characteristics of hydrogen depletion by PARs differ depending on the thermal-hydraulic behaviors occurring in the containment. Various pressure control systems are installed in the containment building to prevent overpressure in severe accident conditions. Representative systems include a spray, a fan-cooler (RCFC: reactor containment fan-cooler), a filtered containment venting system (FCVS), and a passive containment cooling system (PCCS). The containment pressure control system ensures the integrity of the containment building by maintaining the containment pressure lower than the design pressure in a severe accident condition. However, during the operation of this pressure control system, the effectiveness of the hydrogen control system and the hydrogen safety in the containment building must be ensured. This study intends to experimentally evaluate the hydrogen removal characteristics of a PAR when pressure control systems such as RCFC, and PCCS are operating. The following were obtained from the experiment. In the PAR-PCCS experiments, the hydrogen removal rates of the PAR show a similar value to the correlation even during the PCCS operation, so it seems that the PCCS has little effect on the PAR operations. It is judged that the operation of the RCFC does not hurt the removal of hydrogen from the PAR through the evaluation experiment of the PAR performance according to the operation of the fan cooler.

The analysis of severe accidents in nuclear power plants is critical due to their potentially catastrophic impacts on public safety and the environment, underscoring the need for severe accident analysis codes like MELCOR. However, MELCOR faces two major challenges: (1) difficulty in solving multi-physics problems and (2) instability caused by complex computational schemes. To address these issues, this study investigates the feasibility of a physics informed neural network (PINN)-based MELCOR code by designing and evaluating a module for the CVH/FL package. MELCOR's governing equations were first implemented as a Python-based module for a thorough understanding, followed by the development of a PINN-based module applied to simplified 2-tank and 3-tank gravity problems. While both models approximated height and velocity well across most regions, discrepancies emerged when the height of the last tank approached the pipe. Notably, the PINN module struggles to accurately predict physical phenomena, particularly in scenarios involving singularities. We believe that our benchmarking study of PINN modules against the MELCOR CVH/FL package is very useful for examining its feasibility in severe accident analysis.

Accurate prediction of thermodynamic properties near the critical point is crucial for safety analysis in nuclear reactors, especially during severe accidents involving steam explosions. Existing methods face challenges in this region due to rapid and nonlinear changes in physical properties, leading to numerical instability and unreliable results. To address these limitations, we developed a deep learning-based standalone model that predicts physical properties near the critical point with high accuracy and computational efficiency. Utilizing data from the International Association for the Properties of Water and Steam (IAPWS), the model is trained to take specific internal energy and specific volume as inputs and outputs the corresponding pressure and temperature. The neural network employs a multilayer perceptron architecture with Leaky ReLU activation functions and is optimized using the mean squared error loss function and the Adam optimizer. Hyperparameter tuning, including adjustments to batch size and learning rate, was performed to enhance model performance. The developed model successfully captures the complex thermodynamic behavior near the critical point, overcoming the deficiencies of previous approaches. Integration of this model into the SCAR (Steam Explosion Code for Associated Risk) module, which is currently under development, enhances its predictive capabilities, providing more reliable inputs for severe accident analysis. This work demonstrates the potential of deep learning approaches in improving thermodynamic property predictions and paves the way for their application in other areas of nuclear safety analysis.

This study aims to reinforce safety measures for pressurized water reactors, ensuring more effective mitigation of severe accidents. Through uncertainty and sensitivity analyses, the research evaluates the effectiveness of reactor coolant system and steam generator injection strategies during an Extended Station Blackout scenario in the OPR-1000 nuclear reactor. Uncertainty analysis focuses on both code-related uncertainty parameters and the human reliability of executing time, critical factors that influence accident mitigation. Additionally, sensitivity analysis is performed to examine the injection rate of the Multi-barrier Accident Coping Strategies equipment, providing insights into the optimization of mobile equipment performance. The research evaluates the effects of these variables on key outcomes, including core cooling, reactor coolant system depressurization, and integrity of the reactor vessel. Utilizing the MAAP5 code, the study provides relevant data to enhance Severe Accident Management Guidance and improve accident management strategies.

As an integrated SMR, IP200 has the advantages of compact structure and high safety, and can be applied to floating nuclear power platforms through certain improved designs. The inherent characteristics and safety facility design of IP200 make its accident sequence slightly different from that of land-based PWR. The complete accident process from the initiation of the accident to the early occurrence of the reactor phenomenon, and then to the IVR and even the reactor reaction after the pressure vessel damage in the late stage of the serious accident, as well as the thermal and hydraulic effects of the safety facility input, are worth further research. A complete and detailed simulation model including the main coolant system and safety facilities of severe accident is established based on the mechanical severe accident analysis program Melcor and the integrated PWR thermal model, and the floating nuclear power platform IP200 is taken as the research object. The research results indicate the complete accident development sequence, key physical response characteristics of the core, and response characteristics of thermal and hydraulic parameters inside and outside the floating nuclear power platform under the condition of DVI pipeline rupture accidents before and after the failure of safety facilities, verifying the effectiveness of safety facility design.

One of the challenges in severe accident evaluation of Generation IV Sodium-cooled Fast Reactors (SFR) is the eutectic reaction between boron carbide (B4C) and stainless steel (SS), leading to boron migration in a molten pool within the core, which increases neutron absorption. To investigate this phenomenon, high-resolution radiative heating was employed to observe boron migration, eutectic behaviour, and melt structure. Experiments replicating control rod designs were conducted using B4C pellets in SS tubes at temperatures up to 1372°C. Two melting mechanisms were identified: SS separating from the B4C pellet and forming a melt drop, and B4C pellets fracturing due to thermal stress The use of visualisation techniques allowed for the detection of the eutectic onset, and the resulting eutectic melt was further analysed using material characterisation techniques. X-ray diffraction (XRD) and X-ray photoelectron spectroscopy (XPS) confirmed the formation of metal borides and metal carbides, attributed to the high chromium, iron, and carbon content. The current paper's findings confirm the relocation of the B4C-SS eutectic mixture and the formation of diverse boride phases, conditions likely to occur under extreme reactor conditions.