Significant efforts are being continuously put for many years into the assessment of the severe accident integral code ASTEC developed by IRSN, through comparison with the results of most of the experiments developed internationally or through benchmarks with other severe accident simulation codes. For this assessment process, the IRSN code developers are supported by international partners, notably in the frame of the recent SNETP-NUGENIA ASCOM collaborative project.This paper relates to the 3rd major version of the ASTEC V2 series, V2.2, that was released in 2021 to the ASTEC community. It aims at providing an overview of the ASTEC V2.2 validation by comparison to experimental data. After a reminder of the ASTEC validation strategy, the ASTEC V2.2 validation matrix is depicted, including more than 300 experimental tests conducted at various scales in more than 50 different facilities worldwide. Then some V2.2 results are discussed for a few representative applications. These calculation examples are selected in a way to cover diverse aspects of severe accident phenomenology in order to provide a good picture of the ASTEC V2.2 modelling status for both in-vessel and ex-vessel processes. Finally, the main lessons drawn from this quite large validation task are summarized, along with an evaluation of the current physical modelling relevance and how it relates to the current state-of-the-art. Based on those outcomes, the ASTEC V2.2 validity domain is specified and some prospects for further improvements are put forward.

The European severe accident integral code ASTEC, developed by IRSN, aims at simulating the progression and consequences of severe accidents (SA) in a water-cooled nuclear power plant (NPP). In particular, in France, the current ASTEC V2 series version are and will be in the next five years intensively used to perform deterministic and probabilistic safety analyses addressing the lifetime extension of the French operating fleet and the start-up of the EPR. This version is also used for emergency preparedness and response purposes. The benchmarking activities with MELCOR and MAAP on the Fukushima Daiichi accidents within the frame of the BSAF OCDE project evidenced significant discrepancies between SA codes likely to affect SAMG assessment. To deepen SA codes benchmarking, so-called crosswalk activities are conducted to identify needed SA models enhancement for consolidation of SAMG assessment. These discrepancies are related to the complex modelling of the fuel degradation and relocation that includes coupled thermodynamic, thermal mechanic and thermal hydraulic processes and that is currently revisited in a coordinated way by the leading SA code development teams (USNRC/SNL, EPRI, IAE and IRSN). The elicitation of the SA modelling, its validation through dedicated and assessed experimental databases and its benchmarking on large sets of reactor configurations are the sound bases of SA codes. They will stay a subject of research and development at IRSN and they will be strengthened for ASTEC V2.1 by the users’ community through the NUGENIA ASCOM project coordinated by IRSN that started in autumn 2018. These bases are the pillars for the innovative trajectory of development of the next ASTEC series of versions that just started in 2019, named ASTEC+, with two main objectives: to develop capacities to address any nuclear facility (NPP, spent fuel pool, fuel cycle facilities, material testing reactor, small modular reactor, Gen. IV concepts, nuclear fusion reactors...) and related risks, to streamline and to pool the efforts in the safety evaluation chain: deterministic evaluations plus uncertainties, level 2 Probabilistic Safety Analyses, emergency preparedness and response, SA desktop simulator. Meeting these objectives at short term implies to reconsider globally the SA code development approach that was designed more than two decades ago to drastically improve the extendibility, the reusability, the verifiability and the ease of use of the code. IRSN has thus designed a new approach based on solutions that demonstrated their efficiency in various engineering contexts The 9 European Review Meeting on Severe Accident Research (ERMSAR2019)Log Number: 096 Clarion Congress Hotel, Prague, Czech Republic, March 18-20, 2019 2/23 (object development, agile methods, advanced algorithms, high fidelity informed low fidelity models, surrogate models).

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.

ASTEC is an integral code for the prediction of Severe Accidents in Nuclear Power Plants. As such, it has to cover all physical processes that could occur during accident progression, and to produce results within an acceptable time. The present paper is about the newest capabilities of ASTEC V2.1 version with a focus on an original reflooding model devoted to both a rod-like geometry representative of a quasi-intact core and severely degraded situations in the core including debris bed and molten pool with local flow blockages. Indeed, reflooding is an important accident measure to stop the progression of a severe accident and it leads to molten material solidification, debris components formation and additional hydrogen production. In order to predict accurately phenomena occurring during the reflooding phase, an original model developed initially in the ICARE/CATHARE V2 IRSN code has been improved and implemented in ASTEC V2.1. The first part of this paper will present the reflooding model, the second one is devoted to its validation on the PERICLES experimental facility and the last one will concern an ASTEC V2.1 calculation on a power plant calculation. The scenario retained and the geometry chosen for this power plant application is the TMI-2 real scenario.

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 SARNET network (Severe Accident Research NETwork of excellence), co-funded by the European Commission from 2004 to 2013, has allowed to significantly improve the knowledge on severe accidents and to disseminate it through courses and ERMSAR conferences. The major investigated topics, involving more than 250 researchers from 22 countries, were in- and ex-vessel corium/debris coolability, molten-core-concrete-interaction, steam explosion, hydrogen combustion and mitigation in containment, impact of oxidising conditions on source term, and iodine chemistry. The ranking of the high priority issues was updated to account for the results of recent international research and for the impact of Fukushima nuclear accidents in Japan. In addition, the ASTEC integral code was further developed to capitalize the new knowledge. The network has reached self-sustainability by integration in mid-2013 into the NUGENIA Association. The main activities and outcomes of the network are presented. (C) 2015 Published by Elsevier B.V.

Significant efforts are put into the assessment of the severe accident integral code ASTEC, jointly developed since several years by IRSN and GRS, either through comparison with results of the most important international experiments or through benchmarks with other severe accident simulation codes on plant applications. These efforts are done in first priority by the code developers' organisations, IRSN and GRS, and also by numerous partners, in particular in the frame of the SARNET European network.The first version of the new series ASTEC V2 had been released in July 2009 to SARNET partners. Two subsequent V2.0 code revisions, including several modelling improvements, have been then released to the same partners, respectively in 2010 and 2011.This paper summarises first the approach of ASTEC validation vs. experiments, along with a description of the validation matrix, and presents then a few examples of applications of the ASTEC V2.0-rev1 version carried out in 2011 by the SARNET users. These calculation examples are selected in a way to cover diverse aspects of severe accident phenomenology, i.e. to cover both in-vessel and ex-vessel processes, in order to provide a good picture of the current ASTEC V2 capabilities. Finally, the main lessons drawn from this joint validation task are summarised, along with an evaluation of the current physical modelling relevance and thus an identification of the ASTEC V2.0 validity domain. (C) 2013 Elsevier B.V. All rights reserved.

Among the 43 organisations which joined the SARNET2 FP7 project from 2009 to 2013, 31 have been involved in the activities on the ASTEC code. This paper presents a synthesis of the main achievements that have been obtained on the ASTEC V2 integral code, jointly developed by IRSN (France) and GRS (Germany), on development, validation vs. experimental data and applications at full scale conditions for both Gen.II and Gen.III plants. As to code development, while the current V2,0 series of ASTEC versions was continuously improved (elaboration and release by IRSN and GRS of three successive V2.0 revisions), IRSN and GRS have also intensively continued in parallel the elaboration of the second ASTEC V2 major version (version V2.1) to be delivered end of 2014. Regarding code validation vs. experiments, the partners have assessed the V2.0 version and subsequent revisions vs. more than 50 experiments; this extended assessment notably confirmed that most models are today close to the State of the Art, while it also corroborated the yet known key-topics on which modelling efforts should focus in priority. As to plant applications, the comparison of ASTEC results with other codes allows concluding on a globally good agreement for in-vessel and ex-vessel severe accident progression. As to ASTEC adaptations to BWR and PHWR, significant achievements have been obtained through the elaboration and integration in the future V2.1 version of dedicated core degradation models, notably to account for multi coolant flows. (C) 2014 Elsevier Ltd. All rights reserved.

The severe accident integral code ASTEC, jointly developed since almost 20 years by IRSN and GRS, simulates the behaviour of a whole nuclear power plant under severe accident conditions, including severe accident management by engineering systems and procedures. Since 2004, the ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out in the frame of the SARNET European network of excellence.The first version of the new series ASTEC V2 was released in 2009 to about 30 organizations worldwide and in particular to SARNET partners. With respect to the previous V1 series, this new V2 series includes advanced core degradation models (issued from the ICARE2 IRSN mechanistic code) and necessary extensions to be applicable to Gen. III reactor designs, notably a description of the core catcher component to simulate severe accidents transients applied to the EPR reactor. Besides these two key-evolutions, most of the other physical modules have also been improved and ASTEC V2 is now coupled to the SUNSET statistical tool to make easier the uncertainty and sensitivity analyses. The ASTEC models are today at the state of the art (in particular fission product models with respect to source term evaluation), except for quenching of a severely damage core.Beyond the need to develop an adequate model for the reflooding of a degraded core, the main other mean-term objectives are to further progress on the on-going extension of the scope of application to BWR and CANDU reactors, to spent fuel pool accidents as well as to accidents in both the ITER Fusion facility and Gen. IV reactors (in priority on sodium-cooled fast reactors) while making ASTEC evolving towards a severe accident simulator constitutes the main long-term objective.This paper presents the status of the ASTEC V2 versions, focussing on the description of V2.0 models for water-cooled nuclear plants. (C) 2013 Elsevier B.V. All rights reserved.

Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP). After a first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…), source term issues (mainly iodine behaviour). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

ASTEC and ICARE/CATHARE computer codes, developed by IRSN (France) (the former with GRS, Germany), are used in RRC KI (Russia) for the analyses of accident transients on VVER-type NPPs. The latest versions of the codes were continuously improved and validated to provide a better understanding of the main processes during hypothetical severe accidents on VVERs.This paper describes modelling improvements for VVERs carried out recently in the ICARE common part of the above codes. These actions concern the important models of fuel rod cladding mechanical behaviour and oxidation in steam at high and very high temperatures. The existing models were improved basing on the experience in the field and latest literature data sources for Zr + 1%Nb material used for manufacture of VVERs fuel rod claddings.Best-fitted correlations for the Zr alloy oxidation through a broad temperature range were established, along with recommendations on model application in clad geometry and starvation conditions. A model for the creep velocity was chosen for the clad mechanical model and some cladding burst criteria were established as a function of temperature.After verification of modelling improvements on Separate Effect Tests, validation was carried out on integral bundle tests such as QUENCH, CODEX-CT, PARAMETER-SF (the application to the CORA-VVER experiments is not described in the present paper) and on the Paks-2 cleaning tank incident. The comparison of updated code results with experimental data demonstrated very good numerical predictions, which increases the level of code applicability to VVER-type materials. (C) 2010 Elsevier B.V. All rights reserved.

The French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fur Anlagen und Reaktorsicherheit mbH (GRS) have been jointly developing for several years a system of calculation codes (or "integral" code), ASTEC (Accident Source Term Evaluation Code), to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor from the initiating event through the possible radiological release of fission products out of the containment, the so-called "source term." Very intensive validation work has been performed in recent years by IRSN and GRS on the VI versions by comparison of code calculations with results of more than 160 international experiments. Complementary validation was performed by 30 partners of the SARNET European Network of Excellence in the 6th Framework Programme of the European Commission, where ASTEC is considered the European reference code. The global status of validation is good for most phenomena, as shown by several examples that are described in this paper, and even very good on fission product behavior. The main need for modeling improvement concerns reflooding of a degraded core, due to the lack in ASTEC V1 of any dedicated model, and intensive efforts will focus on this topic in the next years. Molten core concrete interaction models are at the state of the art, but new experiments under way in the international frame and a better understanding of physical mechanisms are necessary to make further progress. Version V2.0 of the new ASTEC series, released mid-2009, takes benefit of the previous very intensive validation of the ICARE2 IRSN mechanistic code since its core degradation models have now been implemented. Validation will continue in the SARNET network from 2009 to 2013.

In order to analyze the course of a hypothetical severe accident, the French "Institut de Radioprotection et de Surete Nucleaire" in the last decade has developed computer codes that have been extensively used for supporting the Level 2 Probabilistic Safety Assessment (PSA2) and, in general, for the safety analysis of French pressurized water reactors (PWRs).In particular, the computer code ICARE/CATHARE V1 is a tool that has been widely validated and intensively used within the framework of the PSA2 of the 900-MW(electric) French PWR. This code has been tested on many accident scenarios, and the results obtained have been considered to be satisfactory and reliable up to the end of the early degradation phase. But, severe accidents in PWRs are characterized by a continuous evolution of the core geometry due to chemical reactions, melting, and mechanical failure of the rods and other structures. These local variations of the porosity and other parameters lead to multidimensional flows and heat transfers. So, the lack of a multidimensional two-phase thermal-hydraulic model appeared to be prejudicial to achieve best-estimate reactor studies with ICARE/CATHARE V1 in the case of large core blockages and/or in the case of large cavity appearance. In accordance, a full multidimensional modeling (covering both the fluid flow and the corium behavior) was developed and introduced in a new ICARE/CATHARE version referenced as V2, which includes two options for the thermal-hydraulic modeling: either one-dimensional (1D) or two-dimensional (2D).The first part of this paper demonstrates that without activating the new V2 models, ICARE/CATHARE V2 (1D) is able to reproduce the results obtained with ICARE/CATHARE V1 on the basis of a 6-in-break loss-of-coolant accident. Then, in order to illustrate some of the new V2 modeling improvements, the last part is focused on the results obtained with ICARE/CATHARE V2(2D), and a preliminary comparison is made with ICARE/CATHARE V2(1D).This 1D-2D comparison points out in particular the important role that could be played in the course of a severe accident by the multidimensional flow pattern.

Severe accidents in PWRs are characterized by a continuously changing geometry of the core due to chemical reactions, melting and mechanical failure of the rods and other structures. These local variatio ns of the porosity and other parameters lead to multidimensional flows and heat transfers. In this paper , a comprehensive set of multi-dimensional models describing heat transfers, thermal-hydraulics and m elt relocation in a reactor vessel is presented. Those models are suitable for the core description during a seve re accident transient. A series of applications at the reactor scale shows the benefits of using such model.

This document presents an overview of the version V2.0 of the integral code ASTEC, jointly developed by IRSN and GRS for source term evaluation during severe accidents in water-cooled reactors.