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.