During a loss-of-coolant accident in a pressurized water reactor (PWR), steam of varying quality is released from the primary circuit into the equipment compartments of the containment, followed by the release of a hydrogen-steam mixture during the core degradation phase. In the case of long-lasting accidents, findings of detailed code analyses indicate an enrichment of hydrogen in lower peripheral containment compartments in the reference PWR plant under investigation. During the late accident phase with ex-vessel molten core-concrete interaction, even in the case of an operating passive autocatalytic recombiner system, this poses a threat for local hydrogen combustion later on. Such hydrogen phenomena are not expected and have not been widely studied up to now. Therefore, corresponding experiments have been performed at the THAI test facility operated by Becker Technologies.One of these tests had been precalculated with the COntainment COde SYStem (COCOSYS) as part of the Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) code system AC(2) and has been used to validate the code. The 60-m(3) THAI test vessel has been divided into an inner compartment that has been connected to the surrounding vessel, simulating the upper and peripheral containment part, by very small flow openings at the bottom representing the clearance between door frames and door leaves and one opening at the top representing typical openings by burst disks.The paper discusses both the experimental findings of a test series on the potential enrichment of hydrogen in lower containment compartments and the COCOSYS calculations demonstrating the applicability of the code under complex flow conditions including stratification phenomena.

Pool scrubbing is a major topic in water cooled nuclear reactor technology as it is one of the means for mitigating the source-term to the environment during a severe accident. Pool scrubbing phenomena include coupled interactions between bubble hydrodynamics, aerosols and gaseous radionuclides retention mechanisms under a broad range of thermal-hydraulic conditions as per accident scenarios. Modeling pool scrubbing in some relevant accident scenarios has shown to be affected by substantial uncertainties.In this context, IPRESCA (Integration of Pool scrubbing Research to Enhance Source-term CAlculations) project aims to promote a better integration of inter-national research activities related to pool scrubbing by providing support in experimental research to broaden the current knowledge and database, and by sup-porting analytical research to facilitate systematic validation and model enhancement of the existing pool scrubbing codes based on different modelling approaches, e.g. lumped parameter, computational fluid dynamics, mechanistic. The project consortium includes>30 organisations from 15 countries involving research in-stitutes, universities, TSOs, and industry. For IPRESCA activities, partners join the project with in-kind contributions. IPRESCA operates under NUGENIA Technical Area 2/SARNET (Severe Accident) - Sub Technical Area 2.4 (Source-term).The present paper provides an introduction and overview of the IPRESCA project, including main outcomes and key ongoing and planned activities. New insights obtained from code benchmarks and consolidation of experimental database are discussed. Application of these experimental ("new data") and analytical results shall support in improvement and further development of pool scrubbing models towards reduction in uncertainties for source term calculations.

Phase 2 of the OECD/NEA Project "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF)" was established in mid-2015. The objectives have been similar to Phase 1 of the project but with an extended analysis period of 3 weeks from the occurrence of the earthquake, a major focus on fission product (FP) behaviour and releases to the environment and the comparison to various data including radiological data and results of backwards calculations of the source term. Nine organizations of six countries (Ciemat Spain; IAE, JAEA and NRA Japan; CEA, IRSN France; IBRAE Russia; KAERI Korea; NRC/DOE/SNL U.S.A.; VTT Finland) submitted results of their calculated severe accident scenarios for Unit 1 at the Fukushima Daiichi site using different severe accident codes (ASTEC, MAAP, MELCOR, SAMPSON, SOCRAT, THALES-KICHE). This paper describes the findings of the comparison of the participants results for Unit 1 against each other and against plant data, the evaluation of the accident progression and the final status inside the reactors. Special focus is on RPV status, melt release and FP behaviour and release. Unit specific aspects will be highlighted and points of consensus as well as remaining uncertainties and data needs will be summarised. The results for Units 2 and 3 are presented next in separated papers.

The Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF), which started in 2012 and continued until 2018, was one of the earliest responses to the accident at Fukushima Daiichi. The project, divided into two phases, addressed the investigation of the accident at Units 1, 2, and 3 by severe accident (SA) codes until 500 h, focusing on thermal hydraulics, core relocation, molten corium concrete interaction (MCCI), and fission product release and transport. The objectives of the BSAF were to make up plausible scenarios based primarily on SA forensic analysis, support the decommissioning, and inform SA code modeling. The analysis and comparison among the institutes have brought up vital insights regarding the accident progression, identifying periods of core meltdown and relocation and reactor pressure vessel (RPV) and primary containment vessel (PCV) leakage/failure through the comparison of pressure, water level, and containment atmosphere monitoring system (CAMS) signatures. The combination of code results and inspections (muon radiography, PCV inspection) has provided a picture of the current status of the debris distribution and plant status. All units present a large relocation of core materials and all of them present ex-vessel debris with Unit 1 and Unit 3 showing evidence of undergoing MCCI. Uncertainties have been identified, in particular on the time and magnitude of events such as corium relocation in the RPV and into the cavity floor and RPV and PCV rupture events. Main uncertainties resulting from the project are the large and continuous MCCI progression predicted by basically all the SA codes and the leak pathways from the RPV to the PCV and the PCV to the reactor building and environment. The BSAF project represents a pioneering exercise that has set the basis and provided lessons learned not only for code improvement but also for the development of new related projects to investigate in detail further aspects of the Fukushima Daiichi accident.

The system code package AC(2) by GRS for safety analyses of nuclear reactors from normal operation to severe accidents has been updated with a new release. We briefly describe the main modules of AC(2 ) 2019: ATHLET 3.2, ATHLET-CD 3.2 and COCOSYS 3.0 and selected improvements in these codes. We illustrate the improved capabilities of AC(2) with selected examples. The post-test-calculation of the flooding pool of the INKA test facility demonstrated the improved ATHLET 3D model with explicit mixture level model. For ATHLET-CD, we apply the AIDA module with improvements in wall ablation and heat transfer models to a generic AP1000 lower plenum filled with molten corium. And we apply the new SAFT fission product transport module simulating three release paths through steam generator U-tubes. For COCOSYS, we present uncertainty analysis results for two MCCI experiments which justify using an elevated concrete decomposition temperature of 1800 K. Finally, we illustrate the potential of coupling ATHLET to CFD codes for safety analysis with results of a post-test calculation of the ROCOM PKLIIIT1.1 test on 3D mixing phenomena in an RPV.

Timely evolution of radioactive airborne aerosols, the mass of particles deposited on structures or transported with condensing water into the sump, and any influence of safety systems on the aerosol distribution will have a crucial influence on the potential aerosol source term into the environment. Different phenomena affect the main aerosol processes, and important ones besides gravimetrical and diffusive settling require experimental investigations to improve and validate modeling assumptions. Experimental investigations in the THAI (Thermal-hydraulics, Hydrogen, Aerosol, Iodine) test facility have been performed to investigate (a) insoluble silver aerosol wash-down behavior from vertical steel and horizontal decontamination paint-coated surfaces by condensing steam (test AW-3, supported by a laboratory-scale wash-down test series), and (b) the depletion of the airborne aerosol concentration by wash-out due to the use of a nuclear power plant typical water spray system (test AW-4). The paper discusses experimental findings of both tests supported by analytical analyses using the containment code system COCOSYS developed by Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS). To validate and further improve the new model Abwaschmodell fur unlosliche Aerosole (AULA) in COCOSYS, used for the wash-down of insoluble aerosols from containment typical structures, the AW-3 laboratory tests related to the AW-3 test were used. Building upon these results, the AW-3 wash-down test is simulated. The results of the calculation for the AW-3 test show that the wash-down of insoluble silver particles at least qualitatively resembles the experimental results, though generally the washed-down aerosol mass is lower compared to the experiment. In test AW-4, it was in question if the modeling of aerosol wash-out with spray systems is adequately treated by assuming monodisperse spray droplets or if a droplet distribution has to be applied. Posttest calculation of AW-4 indicates that the wash-out of CsI aerosols by spray systems can be captured qualitatively. However, it is also shown that the calculated wash-out rate is too large and the depletion of the CsI aerosols during the dry phase is underestimated.

The transition from Generation 2 to Generation 3/3+ and 4 reactors, as well as the development of small modular reactors (SMR), place new demands on computational programs designed to simulate conditions of normal operation, operational occurrences, design basis accidents and severe accidents. On the one hand, most passive safety systems of advanced and innovative plants operate at low pressures even down to vacuum conditions and the driving forces are low compared to active systems. On the other hand, the containment is no longer just a barrier to retain radioactive material in the event of leakage of the cooling system, but it is an important link in the passive cooling chain. This requires an expansion and improvement of the existing simulation programs for the cooling circuit and containment, as well as the realization of a coupling between these simulation programs. The new AC(2) program package combines the proven simulation codes ATHLET/ATHLETCD and COCOSYS in one software suite to hit this target. The individual components of the suite are continuously extended and validated for their application to novel safety systems. This makes it possible to simulate the entire spectrum of accidents for Generation 3/3+, 4 and light water cooled SMR systems with just one program package. This publication gives an overview of the current state of development of AC(2) and its individual modules.

The best-estimate and multi-physics system code package AC is developed by the Gesellschaft für Anlagenund Reaktorsicherheit (GRS) gGmbH for the simulation of all relevant phenomena during normal operation, anticipated operational occurrences, design basis accidents, and severe accidents in nuclear power plants (NPPs). AC consists of the thermal-hydraulic code ATHLET, its extension for severe accidents in the coolant circuit ATHLET-CD and the containment code COCOSYS. Development for these codes was started more than 40 years ago and, until 2016, they were developed separately. Recognizing the increasing need for multi-physics simulations of the integral reactor system behaviour, like the interaction between the cooling circuit and the new containment, the demands posed by passive safety systems in advanced (Gen III/III+) reactors and small modular reactors (SMRs), and new models required for innovative (Gen IV) reactor designs, GRS decided to combine its system codes into the integral code package AC. This requires enhanced coupling strategies for the codes, which is a current focus activity of GRS. In addition, the codes need further improvements for new innovative components (e.g. bayonet, plate or helical coil steam generators), new working fluids (e.g. supercritical water, liquid metals, gases, molten salts), and heat transfer models needed for containment concepts with an infinite passive containment cooling to an ultimate heat sink, which could be either air or water or also large water pools. Further development priorities are coupled multi-physics calculations (e.g. considering 3D fluid dynamics or 3D neutronics and/or structural mechanics) and the qualification of the code system for passive (safety) systems with small driving forces. The development of AC was and is mainly funded by the German Federal Ministry of Economic Affairs and Energy (BMWi) to maintain and enhance nuclear safety competencies in Germany, to provide an independent evidence tool chain, and to collaborate internationally on nuclear safety issues. AC is part of GRS’ nuclear simulation chain and is used not only for safety research but also for evidence in nuclear licensing and supervisory procedures, e.g. for independent confirmatory calculations. Numerous national as well international organisations (e.g. universities and research centres) support GRS in the development and validation of the code package. AC can be provided to interested parties like universities, research centres, TSO and regulators upon request and free of license fees. In this way, AC contributes to the worldwide improvement of the nuclear safety standards and is especially important for developing and emerging countries, which do not have the financial means and/or the necessary know how for this purpose. The current release is AC 2019, which is available since June 2019 and includes numerous new features. Besides extensive model improvements, there are an improved coupling interface to CFD codes, and a new numerical tool kit (NuT) to speed up the calculation of complex ATHLET/CD models. The new release also includes effective and powerful software tools for preand postprocessing (e.g. ATLAS). The new release and its features were presented to the AC user community at an international user meeting in Garching from November 19–21, 2018. During the meeting, the users reported on their experiences with the AC code system. Additionally, there were several round table discussions on further user needs for both the individual programs ATHLET, ATHLET-CD and COCOSYS as well as the coupled code system AC. Outstanding contributions of this user meeting are presented in this edition of the independent Journal for Nuclear Engineering Kerntechnik and thus made available to the wider scientific community. It should be noted that Kerntechnik is well established inter alia in the Chinese research environment, e.g. special issues on the Sino-German Symposium Series on Fundamentals of Advanced Nuclear Safety Technology (SG-FANS). China is currently a hot spot of reactor development and construction. For this reason, a scientific exchange especially with Chinese researchers is of particular interest for the AC development and validation team. EDITORIAL

On behalf of the German Federal Ministry of Economics and Technology, Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) participated in the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) project titled Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF). Analysis of the severe accidents (SAs) that happened in the Fukushima Daiichi nuclear power plant (NPP) requires well-qualified methods and codes, e.g., ATHLET-CD and COCOSYS developed and applied at GRS. Coupled ATHLET-CD/COCOSYS analyses for the SA progression during the first days for the similar Units 2 and 3 of Fukushima Daiichi have been provided as the German contribution to the OECD/NEA BSAF project, phase 1. ATHLET-CD is a detailed SA code based on the thermal-hydraulic code ATHLET of GRS to simulate the processes in the reactor circuit before and during core degradation. COCOSYS is focused on the simulation of design basis and SA progression in the containment and the surrounding buildings of the NPP.The focus is on selected results of the SA analyses in the boiling water reactors at the Fukushima Daiichi site especially with regard to the conditions in the torus-shaped wetwell (WW) of the primary containment and specific modeling needs. The GRS results obtained in this OECD/NEA BSAF project, phase 1, are encouraging in terms of capturing essential SA signatures like reactor and containment pressure, reactor water level, and WW temperature history for the first days of the accident in the analyzed Units 2 and 3. A detailed plant model was built up especially with a detailed torus nodalization allowing modeling of relevant phenomena like thermal stratification in the torus water pool and consideration of plant-specific details with regard to local water/steam injections into the torus water pool through safety systems and valves. As a result, the calculated accident progression of the best-estimate analyses for both units follows the accident time line quite closely. This is a prerequisite for reasonable core degradation calculations, as the time window available for the onset of core degradation between known points in time when safety injection stops and mobile pump injection into the reactor starts is small. The analyses are useful to identify areas that require further attention, to define information needs to be gained from the decommissioning, and to define further research needs with regard to experiments and code improvement.

The Great East Japan earthquake occurred on March 11, 2011, at 14: 46, and the subsequent tsunami led Tokyo Electric Power Company's (TEPCO's) Fukushima Daiichi Nuclear Power Station (NPS) beyond a design-basis accident. After the accident, the Japanese government and TEPCO compiled a roadmap toward an early resolution to the accident including, among the main activities, the employment and improvement of existing severe accident (SA) computer codes. In the member countries of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA), SA codes were developed after the accident at Three Mile Island Unit 2 and widely employed to assess NPS status in the postulated SA conditions. Therefore, working plans have been set up with the country members of the OECD/NEA to apply existing SA codes to analyze the accidents at the Fukushima Daiichi NPS Units 1, 2, and 3 and support the decommissioning, constituting an international program named Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF).The objectives of the BSAF project are to analyze the accident progression of Fukushima Daiichi NPS, to raise the understanding of SA phenomena, to contribute to the improvement of the methods and models of the SA codes, and to define the status of the distribution of debris in the reactor pressure vessels and primary containment vessels for decommissioning.The present technical paper summarizes the achievements obtained through a comparison of the results, emphasizing the portions of the accident where all the participants reached a common consensus and identifying still open questions where future work should be directed. Consensus exists on the current condition of Unit 1, where a large fraction of the fuel is assumed to have relocated ex-vessel. On the other hand, larger uncertainties exist for Units 2 and 3, where in-vessel and ex-vessel scenarios produce a reasonable prediction of the accident progression.

Abstract Over 60 technical experts of the reactor safety research division of the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH are developing and validating reliable methods and computer codes – summarized under the term nuclear simulation chain – for the safety-related assessment for all types of nuclear power plants (NPP) and other nuclear facilities considering the current state of science and technology. This nuclear simulation chain has to be able to simulate and assess all relevant physical processes and phenomena for all operating states and (severe) accidents. In the present contribution, the nuclear simulation chain developed and applied by GRS as well as selected examples of its application are presented. The latter demonstrate impressively the width of its scope and its performance. The GRS codes can be passed on request to other (national as well as international) organizations. This contributes to a worldwide increase of the nuclear safety standards. The code transfer is especially important for developing and emerging countries lacking the financial means and/or the necessary know-how for this purpose. At the end of this contribution, the respective course of action is described.

The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R&D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive European R&D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU "stress tests" for NPPs and has been complemented by information provided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BVVR model enhancements, and from new experimental results on severe accident phenomena.

Après l’accident de Fukushima Daiichi, le Comité sur la sûreté des installations nucléaires (CSIN) de l’OCDE a engagé plusieurs activités prioritaires, dont un rapport sur l’éventage filtré de l’enceinte de confinement dans les pays de l’OCDE. 15 pays ont contribué à ce rapport publié par l’Agence pour l’énergie nucléaire de l’OCDE en juin 2014. Les stress tests ont conduit de nombreux pays à envisager la mise en oeuvre de systèmes d’éventage filtré de l’enceinte, renforçant la capacité de réponse en cas d’accident grave. Le rapport du CSIN détaille les exigences de qualification pour les systèmes d’éventage et indique des pistes d’amélioration des circuits existants et pour la conception des systèmes futurs.