We present the main features of code system AC² developed by GRS for the mechanistic integral simulation of severe accidents. AC² uses ATHLET-CD, an extension of ATHLET, for relevant processes in the cooling circuit or spent fuel pool and COCOSYS (COntainment COde SYStem) for processes in the containment. We briefly explain the main models and approaches used in these codes for the core degradation and melt relocation, lower plenum processes, including reactor pressure vessel failure and ex-vessel cooling, molten corium–concrete interactions, iodine chemistry, hydrogen deflagration, and source term prediction. Moreover, we briefly discuss code validation matrices and give examples from recent validation calculations. This shows that ATHLET-CD and COCOSYS are well suited to simulate accident progression in light-water reactor designs, including water–water energetic reactor. Finally, we give a brief outlook on current lines of development for both codes.

The Fukushima-Daiichi accidents in 2011 underlined the importance of severe accident management (SAM), including external events, in nuclear power plants (NPP) and the need of implementing efficient mitigation strategies. To this end, the Euratom work programmes for 2012 and 2013 was focused on nuclear safety, in particular on the management of a possible severe accident at the European level. Relying upon the outcomes of the successful Euratom SARNET and SARNET2 projects, new projects were launched addressing the highest priority issues, aimed at reducing the uncertainties still affecting the main phenomena. Among them, PASSAM and IVMR project led by IRSN, ALISA and SAFEST projects led by KIT, CESAM led by GRS and sCO2-HeRO lead by the University of Duisburg-Essen. The aim of the present paper is to give an overview on the main outcomes of these projects.

Abstract The GRS program package AC2 with its codes ATHLET/ATHLET-CD and COCOSYS aims for the reliable computational simulation of significant phenomena occurring during normal operation, design basis accidents, and severe accidents in the cooling circuit and containment of a nuclear power plant. To keep the modelling at the state-of-the-art, continuous development and validation is required. This is accomplished through participation in several national and international experimental research programs, where AC2 or one of its codes are assessed against both separate effect tests and integral tests. This paper exemplifies the status of validation and application of COCOSYS by means of calculations of iodine chemistry and molten corium/concrete interaction after reactor pressure vessel rupture. Further, calculations using the external 3D module CoPool coupled to COCOSYS on thermal stratification in large water pools are discussed. The examples given demonstrate the progress of the COCOSYS development and the capability to simulate phenomena in the containment during incidents and accidents with good results. Future applications comprise the entire spectrum of incidents and accidents for Generation III/III+ systems with just one program package.

The CESAM FP7 project (Van Dorsselaere et al., 2015) of EURATOM has been conducted from April 2013 until March 2017 in the aftermath of the Fukushima Dai-ichi accidents. Nineteen international partners from Europe and India, including the European Joint Research Centre, have participated under the coordination of GRS and with a strong involvement of 1RSN that were both ASTEC code developers. The Project objectives were: to understand all relevant phenomena during the Fukushima Dai-ichi accidents and their importance for Severe Accident Management (SAM) measures; and to improve the ASTEC computer code to simulate plant behaviour throughout accident sequences including SAM measures. The starting point was the analysis of current SAM measures implemented in European nuclear power plants. To achieve these goals, simulations of relevant experiments that allow a solid validation of the ASTEC code against single and separate effect tests have been conducted. Covered validation topics in the CESAM project have been grouped in 9 different areas among which are re -flooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pool behaviour. Furthermore, modelling improvements have been implemented in the current ASTEC V2.1 series for the estimation of source term consequences in the environment and the prediction of plant status in emergency centres. Finally, ASTEC reference input decks have been created for all reactor types operated in Europe today as well as for spent fuel pools. These reference input decks generically describe plant types like PWR, WER, PHWR, and BWR without defining proprietary data of a special plant and they account for the best recommendations from code developers and users. In addition, a generic input deck for a spent fuel pool was elaborated. These input decks can be used as basis by all (and especially new) ASTEC users in order to understand code basic requirements and model features and to implement the specificities of their own NPP type. Based on these generic inputs, benchmark calculations have been performed with other codes (such as MELCOR, MAAP, ATHLET-CD, COCOSYS...) with a focus on applicability of ASTEC models to currently implemented SAM measures. This article provides a final summary of the CESAM project. Therefore, an overview of the improved modelling capabilities of the recent ASTEC V2.1 version is given followed by the validation status of ASTEC V2.1 as concluded after CESAM. Further, plant applications performed by CESAM partners will be summarized with a special focus on simulation of SAM measures in various NPP types, and insights gained on SAM measures will be derived. (C) 2018 Elsevier Ltd. All rights reserved.

A new major version of the European severe accident integral code ASTEC, developed by IRSN with some GRS support, was delivered in November 2015 to the ASTEC worldwide community.Main modelling features of this V2.1 version are summarised in this paper. In particular, the in-vessel coupling technique between the reactor coolant system thermal-hydraulics module and the core degradation module has been strongly re-engineered to remove some well-known weaknesses of the former V2.0 series. The V2.1 version also includes new core degradation models specifically addressing BWR and PHWR reactor types, as well as several other physical modelling improvements, notably on reflooding of severely damaged cores, Zircaloy oxidation under air atmosphere, corium coolability during corium concrete interaction and source term evaluation.Moreover, this V2.1 version constitutes the back-bone of the CESAM FP7 project, which final objective is to further improve ASTEC for use in Severe Accident Management analysis of the Gen.II-III nuclear power plants presently under operation or foreseen in near future in Europe. As part of this European project, IRSN efforts to continuously improve both code numerical robustness and computing performances at plant scale as well as users' tools are being intensified.Besides, ASTEC will continue capitalising the whole knowledge on severe accidents phenomenology by progressively keeping physical models at the state of the art through a regular feed-back from the interpretation of the current and future experimental programs performed in the international frame. (C) 2016 Elsevier Ltd. All rights reserved.

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.

The focus of the ASTEC (Accident Source Term Evaluation Code) development at GRS is primarily on the containment module CPA (Containment Part of ASTEC), whose modelling is to a large extent based on the GRS containment code COCOSYS (COntainment COde SYStem). Validation is usually understood as the approval of the modelling capabilities by calculations of appropriate experiments done by external users different from the code developers. During the development process of ASTEC CPA, bugs and unintended side effects may occur, which leads to changes in the results of the initially conducted validation. Due to the involvement of a considerable number of developers in the coding of ASTEC modules, validation of the code alone, even if executed repeatedly, is not sufficient. Therefore, a regression testing procedure has been implemented in order to ensure that the initially obtained validation results are still valid with succeeding code versions. Within the regression testing procedure, calculations of experiments and plant sequences are performed with the same input deck but applying two different code versions. For every test-case the up-to-date code version is compared to the preceding one on the basis of physical parameters deemed to be characteristic for the test-case under consideration. In the case of post-calculations of experiments also a comparison to experimental data is carried out. Three validation cases from the regression testing procedure are presented within this paper. The very good post-calculation of the HDR E11.1 experiment shows the high quality modelling of thermal-hydraulics in ASTEC CPA. Aerosol behaviour is validated on the BMC VANAM M3 experiment, and the results show also a very good agreement with experimental data. Finally, iodine behaviour is checked in the validation test-case of the THAI IOD-11 experiment. Within this test-case, the comparison of the ASTEC versions V2.0r1 and V2.0r2 shows how an error was detected by the regression testing procedure and why the regression testing is an important part of the validation process. The corrected version V2.0r2 delivers a very good validation result for the iodine behaviour in the post-calculation of the THAI IOD-11 experiment.

Passive Autocatalytic Recombiners (PARs) in PWR-Containments remove hydrogen released in case of a severe accident with core damage by catalytic oxidation with the oxygen of the containment atmosphere. The removal of hydrogen in autocatalytic recombiners (PARs) results in temperature levels at the catalytic surfaces up to 900 °C and leads to elevated temperatures up to several hundred degrees of the gas flowing over these surfaces. Under such operating conditions suspended CsI and other iodide particles transported with the convective gas flow through PARs can be converted into volatile iodine thus influencing the iodine source term. Even low conversion rates might lead to a significant influence on the concentration of gaseous iodine in the early phase of an accident where high CsI/I2 ratios of 100:1 or even higher can be expected. In the frame of the German national THAI programme two technical-scale experiments, AER-2 and AER-5, have been performed to investigate the conversion rates CsI → I2 occurring under realistic PAR operating conditions and beyond, and the influence of PARs on CsI aerosol parameters. An original SIEMENS (now AREVA NP) type PAR has been operated under realistic thermalhydraulic conditions. CsI aerosol has been generated by evaporating ultra pure CsI in an inductive furnace and monitored in the THAI vessel, following re-condensation, by low pressure impactors and filters. Experiments with hydrogen concentrations of up to 5 vol% H2 under normal conditions and of up to 10.5 vol% H2 in a steam-inerted atmosphere have been performed. An approved and qualified sampling and detection method has been applied to determine gas borne molecular iodine concentrations. CsI → I2 conversion rates of 3% have been observed as an upper limit under realistic thermalhydraulic conditions. The results are compared to CsI → I2 conversion rates achieved in the French RECI experiments. These experiments investigated the interaction of catalytic material and metal iodides (CsI, CdI) in a lab-scale setup showing conversion rates up to 60%. However, these experiments might deviate from representative thermochemical conditions. A comparison of THAI and RECI results — as far as appropriate concerning aerosol parameters and gas temperatures — shows conversion rates in comparable dimensions. In other cases they differ significantly. The potential contribution of an operating PAR (metal iodide interaction) to the iodine source term together with the restricted knowledge from both THAI and RECI tests, means that further investigations are required.

During a hypothetical severe incident in a nuclear power plant with core meltdown a large part of radioactive material is present as aerosol particles in the reactor containment. In current severe accident containment codes the potential influences of hydrogen combustions on the behaviour of aerosols are not considered. Among other effects dry resuspension can increase the aerosol concentration in the atmosphere. Already deposited aerosol material can be re-released into the containment atmosphere by atmospheric currents induced by hydrogen deflagrations or by other phenomena like steam explosions. The objective is to assess the possible influence of this dry resuspension effect on the radioactive source term. 1 INTRODUCTION During a severe accident with core meltdown in a light water reactor hydrogen is produced and released from the primary cooling circuit into the containment atmosphere. Along with the oxygen present in the atmosphere a flammable mixture of steam, hydrogen, and oxygen may develop. At the same time radioactive fission and activation products are released from the primary cooling circuit into the containment. A large part of these fission products is bound to airborne aerosol particles. For a complete understanding of the aerosol behaviour the balance between the aerosol release, the sedimentation of aerosol into pool water and onto solid surfaces, and the re-release from deposits or water pools into the atmosphere has to be described. We are working on a better understanding of the dry resuspension phenomenon caused by hydrogen deflagrations. For this reason a new resuspension model was developed to describe the conditions found under strong and very transient air currents as expected after hydrogen deflagrations. The time dependent multilayer particle model of Fromentin (1) is used as a basis for the new model. This modified Fromentin model has been built into the GRS containment code system COCOSYS (2). An experimental programme with the resuspension experiments Aer-1, Aer-3, and Aer-4 was performed in the ThAI (Thermalhydraulics Aerosol and Iodine) facility (3, 4) at Becker Technologies in cooperation with AREVA and GRS. The results gained from these experiments have been very valuable for the modelling of the resuspension mechanisms. Along with the modified Fromentin model for resuspension after highly transient air currents source term studies have been performed with COCOSYS. Therefore a generic COCOSYS dataset of a KONVOI type reactor was used to calculate an incident scenario with a possible hydrogen deflagration.