A hypothetical severe accident in a nuclear power plant has the potential for causing severe core damage, including a meltdown. To prevent or in the case of an already formed debris bed to limit the in-vessel core degradation, the basic severe accident management strategies consider the in-vessel reflooding to ensure the debris bed coolability. The purpose of our research was to understand the key processes and conditions related to the in-vessel debris bed coolability in the bottom reflooding conditions. Recently, experimental tests in the PEARL facility (IRSN, France) were performed to highlight the behaviour of the steam and water flow in a hot porous medium and to provide experimental data to validate 2-D and 3-D models for the debris bed reflooding. Our aim was to analyse chosen PEARL experiments performed at the atmospheric pressure. The objective was to analyse the importance of the uncertainties in the initial and boundary conditions on the simulation results and to assess the heat transfer modelling approaches. Simulations were performed using the MC3D code (IRSN, France). In general, the performed simulations are in good agreement with the experiments. The general features, in particular the water preferential entrainment in the bypass are recovered and the analysis of calculation gives further information on the mechanisms. In particular, the mechanism of water deviation in the bypass (2-D behaviour) is described. The hypothesis of water dragged by steam coming from the debris bed region cannot be supported. However, the simulation results are indicating a noticeable impact of the actual conditions as the water temperature and the initial support bed and bypass temperature. The simulations, varying the porosity of the test section, showed that this impact affects the flow configuration and is important for cases with the 2-D configuration. The reflooding capabilities in this configuration may depend strongly on the characteristics of the debris bed. Changes in the heat transfer modelling do not have greater effect on the simulation results.

A nuclear severe accident progression may lead to the formation of a corium debris bed either in the vessel lower head (in-vessel debris bed) or in the vessel pit (ex-vessel debris bed). For safety analyses it is essential to know if a debris bed is coolable or not, i.e. whether a given water mass flow rate poured into the debris bed - either from its top or from its bottom - will be sufficient to evacuate the residual heat and stop the accident progression. The IRSN code, mostly used for fuel-coolant interaction studies, has been modified with the addition of new friction laws for diphasic flows in porous media. The validation of the code in the case of debris coolability concerns the friction in isothermal configuration in cold and hot situations, the evaluation of critical heat flux and the bottom and top reflooding of debris beds. The results obtained with MOD are in good agreement with the experimental data and are estimated satisfactory regarding to the nuclear safety issues. (C) 2017 Elsevier B.V. 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 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.