The contribution deals with the backfitting of Dukovany NPP for design extension conditions with emphasis on severe accidents. This NPP, located in Czech Republic, is equipped with 4 units of VVER-440/V213 reactors. During the last decades significant attention was paid to backfitting of these units. New safety systems aimed at mitigating the consequences of severe accidents have been designed and have been (are being) implemented. At first, special features of VVER-440/V213 reactor and containment equipped with bubbler condenser pressure suppression system are briefly described. The systematic process of plant backfitting for design extension condition is summarized further. Attention is paid to description of crucial severe accident management strategies that have been adopted and design and implementation of new safety systems that are necessary to fulfil these strategies. Than the requirements of Czech regulatory body (SUJB) set for practical elimination of early or large radioactive releases are discussed. Finally, the performances of severe accident management strategies are illustrated using the specific case of blackout scenario. For this purpose, an analysis was performed using ASTEC code developed by the French IRSN. The results of analysis are briefly described and discussed further. From the obtained results it follows that the high-pressure melting scenario is reliably avoided, corium is localised and stabilised inside the reactor vessel and the hydrogen in containment atmosphere is safely controlled by passive autocatalytic recombiners. Thus, stable plant state with preserved containment integrity is reached. Source term that has been estimated under these conditions indicates that the predicted releases are well below the criteria set by SUJB for early or large radioactive releases, and therefore long-term contamination of the environment is avoided.

The contribution deals with the performance of a new containment heat removal system that is intended to be installed in near future at Dukovany NPP in Czech Republic. This NPP is equipped with 4 units of VVER-440/ V213 reactors. During the last two decades these units were systematically backfitted with new safety systems in order to cope with consequences of severe accidents. These safety systems were designed especially for the 4th level of defence in depth and thus they are - as much as possible - independent from the original plant systems. Namely, installation of a new depressurisation line is currently under preparation in order to avoid high-pressure core melting scenarios. In-vessel corium retention via external reactor vessel cooling ("reactor cavity flooding") was adopted as a cornerstone strategy of severe accident management for all VVER-440/V213 reactors operated in central Europe. Regarding the hydrogen issue, large capacity passive autocatalytic recombiners were installed in containment. The prevention of containment pressurisation in the case of loss of ultimate heat sink represents the most important point of remaining safety issues. At first, special features of VVER-440/V213 containment equipped with bubbler condenser pressure suppression system are briefly described in presented paper. Than the mechanism of containment pressurisation due to steaming as a consequence of in-vessel retention and possible solutions for heat removal from the containment are discussed. The final design chosen at Dukovany NPP is based on the heat exchanger and turbo pump located inside the containment that are cooled and driven by external cooling water. External cooling circuit will be connected during the accident via special inlet/outlet nozzles installed at containment boundary. Main advantage of such containment cooling system is full independence on other plant systems (including essential service water), the fact that radioactive coolant is not taken beyond the containment boundary and all active components of the system (mobile pump, isolation valves) are located outside the containment and, thus, are easily accessible. The performance of this containment heat removal system during severe accidents were analysed using the ASTEC code and the most important results are presented. From the obtained results it follows that there is sufficient time margin for assembling the external cooling loop. Thus, the containment cooling can be assured long before the containment ultimate pressure is reached or before the coolant level in containment sump drops below critical level, when the external reactor cooling can be lost. After start of the heat removal system, the system is capable to prevent further containment pressurisation, increase the coolant level in containment sump and decrease the pressure to atmospheric or even sub-atmospheric level.

An in-vessel retention (IVR) concept was proposed for standard WER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of WER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points:1) *Estimation of thermal loads acting on the inner reactor surface;2) *Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield;3) *Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel subjected to thermal load from corium, and RELAP5 code for assessment of external reactor vessel cooling. Finally, implementation of IVR system in NPPs in Czech Republic and Slovakia is shortly described. (C) 2017 Elsevier B.V. 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.

Since the beginning of ASTEC development by IRSN and GRS the code was widely applied to VVER reactors. In this paper, at first specific features of VVER-440/V213 reactor design that are important from the modelling point of view are briefly described. Then the validation of ASTEC code with focus on its applicability to VVER reactors is briefly summarised and the results obtained with the ASTEC V2.0-rev1 version for the ISP-33 PACTEL natural circulation experiment are presented. In the next section the application of ASTEC V2.0-rev1 code in upgrade of VVER-440/V213 NPPs to cope with consequences of severe accidents is described. This upgrade includes adoption of in-vessel retention via external reactor vessel cooling and installation of large capacity passive autocatalytic recombiners. Results of analysis with focus on corium localisation and stabilisation inside reactor vessel, hydrogen control in confinement and prevention of long-term confinement pressurisation are presented.

Safety margin of the in-vessel retention strategy is given by the difference between thermal load acting on inner reactor surface and coolability limit (in terms of critical heat flux) on outer reactor surface. In order to study the two-phase flow in external reactor vessel cooling loop and heat transfer from curved reactor wall, the RESCUE-2 (Representative loops for External System Cooling Understanding Experiments) experimental facility was erected in CEA Cadarache in France. The facility consists of electrically heated simulator of reactor vessel with an ellipsoidal lower head and cooling loop that enables natural circulation of coolant around the reactor wall. In the frame of SARNET 2 project (Severe Accident Research Network of Excellence, 7th EU Framework Programme) several experiments devoted to external reactor vessel cooling phenomena with certain relevance to VVER-440/V213 reactors, were performed on this facility. The heat flux profile generated by electrical heaters in these experiments was based on the results that were obtained by using ASTEC code for this reactor design. The results of two RESCUE-2 experiments are used in this paper for benchmarking of RELAP 5 Mod.3:3 code and ASTEC V2.0 code.

The motivation of the work performed within the work package "Corium and Debris Coolability" of the Severe Accident Research Network of Excellence (SARNET) is to reduce or possibly solve the remaining uncertainties on the efficiency of cooling reactor core structures and materials during severe accidents, either in the core, in the vessel lower head or in the reactor cavity, so as to limit the progression of the accident. This can be achieved either by ensuring corium retention within the reactor pressure vessel or at least by limiting the corium progression and the rate of corium release into the cavity. These issues are to be covered within the scope of accident management for existing reactors and within the scope of design and safety evaluation of future reactors. The specific objectives are to create and enhance the database on debris formation, debris coolability and corium behavior in the lower head, to develop and validate the models and computer codes for simulation of in-vessel debris bed and melt pool behavior, to perform reactor scale analysis for in-vessel corium coolability and to assess the influence of severe accident management measures on in-vessel coolability. The work being performed within this work package comprises experimental and modeling activities with strong cross coupling between the tasks. Substantial knowledge and understanding of governing phenomena concerning coolability of intact rod-like reactor core geometry was obtained in previous projects. Hence the main thrust of experimental and modeling efforts concentrates mainly on the study of formation and cooling of debris beds in order to demonstrate effective cooling modes, cooling rates and coolability limits. Modeling efforts have been aimed at assessing and validating the models in system-level and detailed codes for core degradation, oxidation and debris behavior. The paper describes the work performed up to now and summarizes the main results achieved so far.

Within the SARNET network of excellence in the 6th Framework Programme of the European Commission, the severe accident integral code ASTEC, jointly developed by IRSN (France) and GRS (Germany), has been validated against international experiments to evaluate the suitability and capability of new or improved models implemented in successive code versions up to V1.3rev2, delivered in December 2007. This paper focuses on the code applications concerning circuit thermal-hydraulics and core degradation to integral and separate-effect experiments: for the CESAR thermal-hydraulic module, BETHSY 9.1 b, PACTEL ISP 33 and T2.1, PMK2-SBLOCA, LOFT-LP-FP-2; for the DIVA core degradation module, CORA-13 and -W2, QUENCH-11 and -13, LOFT-LP-FP-2, Phebus FPT-4, FARO L14 and L28, LIVE-L1, OLHF-1, FOREVER EC2. Besides, the TMI-2 accident has been analyzed using the CESAR and DIVA modules in a coupling mode.The emphasis was put on the following new or improved models: i.e. in CESAR, reflooding of an intact core, condensation in the pressurizer, sub-critical break flow correlation, and new pressurizer spray model: in DIVA, corium behaviour in the lower head and lower head mechanical failure.For thermal-hydraulics in the circuits, good results have been obtained with ASTEC on the three integral experiments that cover various thermal-hydraulic flow regimes: LOFT-LP-FP-2 in Western PWR geometry and the two PACTEL experiments in VVER-440 geometry. These good results have been confirmed by the validation done on several BETHSY integral tests.For core degradation, the ASTEC results are good for early-phase models of core heat-up, oxidation and hydrogen production (before any quenching phase) on different CORA and QUENCH experiments and on LOFT-LP-FP-2. For the in-vessel late-phase, the results can be considered as good regarding debris bed melting (Phebus FPT-4), corium fragmentation at slump into vessel lower plenum (FARO), molten pool behaviour in lower plenum (LIVE-L1), and vessel lower head mechanics (OLHF-1 and FOREVER EC2). Furthermore, the first two phases of the TMI-2 accident before core reflooding are very well calculated by ASTEC. The main remaining modelling weaknesses concern the reflooding of a degraded core and the corresponding hydrogen production. The implementation of detailed magma 2D relocation models in the new series of ASTEC V2 versions (the first one being released mid-2009) will allow a more realistic simulation of late phase phenomena up to the failure of the lower head. (C) 2009 Elsevier Ltd. All rights reserved.

The OECD-IAEA Paks Fuel Project was developed to support the understanding of fuel behaviour in accident conditions on the basis of analyses of the Paks-2 incident. Numerical simulation of the most relevant aspects of the event and comparison of the calculation results with the available data from the incident was carried out between 2006 and 2007. A database was compiled to provide input for the code calculations. The activities covered the following three areas: (a) Thermal hydraulic calculations described the cooling conditions possibly established during the incident. (b) Simulation of fuel behaviour described the oxidation and degradation mechanisms of the fuel assemblies. (c) The release of fission products from the failed fuel rods was estimated and compared to available measured data.The applied used codes captured the most important events of the Paks-2 incident and the calculated results improved the understanding of the causes and mechanisms of fuel failure. The numerical analyses showed that the by-pass flow leading to insufficient cooling amounted to 75-90% of the inlet flow rate, the maximum temperature in the tank was between 1200 and 1400 degrees C, the degree of zirconium oxidation reached 4-12% and the mass of produced hydrogen was between 3 and 13 kg. (C) 2009 Elsevier B.V. All rights reserved.

This paper summarizes the work done in the SARNET European Network of Excellence on Severe Accidents (6th Framework Programme of the European Commission) on the capability of the ASTEC code to simulate in-vessel corium retention (IVR). This code, jointly developed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fur Anlagen und Reaktorsicherheit mbH (GRS) for simulation of severe accidents, is now considered as the European reference simulation tool.First, the DIVA module of ASTEC code is briefly introduced. This module treats the core degradation and corium thermal behaviour, when relocated in the reactor lower head. Former ASTEC V1.2 version assumed a predefined stratified molten pool configuration with a metallic layer on the top of the volumetrically heated oxide pool. In order to reflect the results of the MASCA project, improved models that enable modelling of more general corium pool configurations were implemented by the CEA (France) into the DIVA module of the ASTEC V1.3 code.In parallel, the CEA was working on ASTEC modelling of the external reactor vessel cooling (ERVC). The capability of the ASTEC CESAR circuit thermal-hydraulics to simulate the ERVC was tested. The conclusions were that the CESAR module is capable of simulating this system although some numerical and physical instabilities can occur. Developments were then made on the coupling between both DIVA and CESAR modules in close collaboration with IRSN. In specific conditions, code oscillations remain and an analysis was made to reduce the numerical part of these oscillations. A comparison of CESAR results of the SULTAN experiments (CEA) showed an agreement on the pressure differences.The ASTEC V1.2 code version was applied to IVR simulation for VVER-440/V213 reactors assuming defined corium mass, composition and decay heat. The external cooling of reactor wall was simulated by applying imposed coolant temperature and heat transfer coefficient (HTC). The obtained results (pool temperatures, heat flux distribution, reactor wall ablation) were compared with available predictions of other codes. The agreement was correct, in particular on the shape and depth of ablation, as well as the maximum heat flux in case of a thick metallic layer, while ASTEC calculated a lower maximum heat flux for a thin metallic layer. (C) 2009 Elsevier B.V. All rights reserved.

Within SARNET, the corium topic covers all the behaviors of corium from early phase of core degradation to in or ex-vessel corium recovery with the exception of corium interaction with water, direct containment heating and fission product release. The corium topic regroups in three work packages the critical mass of competence required to improve significantly the corium behavior knowledge. The spirit of the SARNET networking is to share the knowledge, the facilities and the simulation tools for severe accidents, so to reach a better efficiency and to rationalize the R&D effort at European level. Extensive benchmarking has been launched in most of the areas of research. These benchmarks were mainly dedicated to the recalculation of experiments, while, in the next periods, a larger focus will be given to integral experiments or reactor applications. Eventually, all the knowledge will be accumulated in the ASTEC severe accident simulation code through physical model improvements and extension of validation database. This paper summarizes the progress that has been achieved in the frame of the networking activities. A special focus is placed on the melt pool and debris coolability and corium-concrete interaction, in which, the effects due to multidimensional geometries and heterogeneities has been shown, during SARNET, to play a crucial role and for which further research is still needed.