P. Chatelard

The ASTEC Integral Code for Severe Accident Simulation

Abstract For several years the French Institut de Radioprotection et de Sûreté Nucléire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH have been jointly developing a system of calculation codes—the integral Accident Source Term Evaluation Code (ASTEC)—to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment, i.e., the source term. ASTEC has progressively reached a larger European dimension through projects of the European Commission Framework Programme. In particular, in the frame of the European Severe Accident Research NETwork of Excellence (SARNET), jointly executed research activities were performed with the ultimate objectives of providing physical models for integration into ASTEC and making the code the European reference. This effort will go on in the frame of the SARNET2 next network. The ASTEC models are today at the state of the art, except for reflooding of a degraded core. Many applications have been performed by IRSN for significant safety studies, including the probabilistic safety analysis level 2 on a French pressurized water reactor. The first version V2.0 of the new ASTEC series, released in spring 2009, will allow simulation of the European Pressurized Reactor (EPR) and will include advanced core degradation models. Then, ASTEC will remain the repository of knowledge gained from international research and development. Other long-term objectives are on one hand extension of the scope of application to boiling water reactors and CANada Deuterium Uranium (CANDU) reactors, to accidents in the ITER Fusion facility, and to Very High Temperature Reactor (VHTR) Generation IV reactors, and on the other hand to the use for emergency response tools and for severe accident simulators.

Core Loss during a Severe Accident (COLOSS).

KFKI Atomic Energy Research Institute (AEKI), Hungary Electricité de France (EDF), France Ente per le Nuove Tecnologie, l’Energia e l’Ambiente (ENEA) Italy Framatome ANP, France Forschungszentrum Karlsruhe GmbH (FZK), Germany European Commission – JRC/IE, International European Commission – JRC/ITU, International Paul Scherrer Institut (PSI), Switzerland Framatome ANP Gmbh, Germany SKODA-UJP Praha a.s., Czech Republic Universidad Politécnica de Madrid (UPM), Spain Ruhr-Universität Bochum (RUB), Germany Universität Stuttgart (IKE), Germany University Lappeenranta, Finland

Validation Status of the ASTEC Integral Code for Severe Accident Simulation

Abstract The French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and the German Gesellschaft für 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 V1 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.

The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

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.

Assessment of ICARE/CATHARE V1 Severe Accident Code

The ICARE/CATHARE code system has been developed by the French “Institut de Radioprotection et de Surete Nucleaire” (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the “International ICARE/CATHARE Users’ Club”, under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a “circuit” experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the standalone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version (“Users’ Guidelines”). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head.Copyright

Improvement of Core Modeling in ICARE/CATHARE: Application to the Calculation of a Six-Inch-Break LOCA Leading to a Severely Degraded Situation

Abstract In order to analyze the course of a hypothetical severe accident, the French “Institut de Radioprotection et de Sûreté Nucléaire” 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.

The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR

IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE[[ellipsis]]). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and hydrogen is considered as well as turbulence, condensation and heat transfer in the containment walls. Passive autocatalytic recombiners are also modelled. Hydrogen combustion is afterwards analysed solving the compressible Euler equations coupled with combustion models. Examples of on-going applications of these codes to the EPR safety analysis are presented to illustrate their potentialities.Copyright