Review of current Severe Accident Management (SAM) approaches for Nuclear Power Plants in Europe

The Fukushima accidents highlighted that both the in-depth understanding of such sequences and the development or improvement of adequate Severe Accident Management (SAM) measures are essential in order to further increase the safety of the nuclear power plants operated in Europe. To support this effort, the CESAM (Code for European Severe Accident Management) R&D project, coordinated by GRS, started in April 2013 for 4 years in the 7th EC Framework Programme of research and development of the European Commission. It gathers 18 partners from 12 countries: IRSN, AREVA NP SAS and EDF (France), GRS, KIT, USTUTT and RUB (Germany), CIEMAT (Spain), ENEA (Italy), VUJE and IVS (Slovakia), LEI (Lithuania), NUBIKI (Hungary), INRNE (Bulgaria), JSI (Slovenia), VTT (Finland), PSI (Switzerland), BARC (India) plus the European Commission Joint Research Center (JRC). The CESAM project focuses on the improvement of the ASTEC (Accident Source Term Evaluation Code) computer code. ASTEC,, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European R&D on the domain. The project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. In the frame of the CESAM project one of the tasks consisted in the preparation of a report providing an overview of the Severe Accident Management (SAM) approaches in European Nuclear Power Plants to serve as a basis for further ASTEC improvements. This report draws on the experience in several countries from introducing SAMGs and on substantial information that has become available within the EU “stress test”. To disseminate this information to a broader audience, the initial CESAM report has been revised to include only public available information. This work has been done with the agreement and in collaboration with all the CESAM project partners. The result of this work is presented here.JRC.F.5-Nuclear Reactor Safety Assessmen

ASTEC application to in-vessel corium retention

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 Sûreté Nucléaire (IRSN) and the German Gesellschaft für 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. © 2009 Elsevier B.V. All rights reserved

Synthesis of the ASTEC integral code activities in SARNET - Focus on ASTEC V2 plant applications

International audienceAmong 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. © 2014 Elsevier Ltd. All rights reserved

Recent advances in ASTEC validation on circuit thermal-hydraulic and core degradation

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, Phébus 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 (Phébus 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. © 2009 Elsevier Ltd. All rights reserved