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CESAM project

​Last update on September 2018


The goal of the Code for European Severe Accident Management (CESAM) project is to enhance the ASTEC software system, the European reference for the study and management of core melt accidents (qualified as severe accidents) for all types of second- and third-generation nuclear power plants (Gen.II and Gen.III NPPs). CESAM was launched in April 2013 under the European Commission's Seventh Framework Programme for Research and Development (FP7) and concluded in March 2017. Coordinated by GRS (Germany) with a major contribution from IRSN, the project brought together 18 European partners and one Indian partner.

 

 

Context and objectives

 

The Fukushima accident in March 2011, which led to the core melt of three reactors at the same site, confirmed that it was necessary to have state-of-the-art tools, validated with experimental data, for numerical simulation of core melt accidents at nuclear reactors and thus contribute to better assess the mitigation resources to implement to limit the environmental consequences. The use of these tools to support emergency response situations relies in particular on improving their performance and coupling them with tools for calculating the dispersion of radioactivity in the environment. The objective of the CESAM project was to further extend the ASTEC software capabilities for core melt modeling to meet these challenges.

 
 

Project outline

 
The CESAM project started immediately after the European SARNET Phase 2 project (FP7) coordinated by IRSN in 2009-2013 [ref-1]. It began with ASTEC V2.0 and continued with V2.1 in 2015 [ref-2]. The main conclusions of this former work thus constituted a valuable input to the CESAM project, notably the synthesis of validations carried out by the SARNET partners with ASTEC V2.0 as well as the identification of the limits of applicability of the ASTEC V2.0 core degradation models to certain types of water-cooled reactors other than pressurized water reactors (PWRs), mainly boiling water reactors (BWR).
 
While waiting for models dealing with the vessel degradation phase to become available in ASTEC V2.1 to cover BWRs, it should be noted that IRSN has successfully used ASTEC V2.0, notably in the framework of the OECD BSAF project, to begin analyzing the accidents affecting the three Fukushima-Daiichi reactors, all of the BWR type, as early as 2012. These initial simulations thus made it possible to better understand the course of these accidents prior to refining the analyses using new models specifically adapted to BWRs.
 
The models used by the ASTEC V2.0 software to represent the phenomena that occur during a core melt accident were first analyzed from two angles: in the light of knowledge about the predominant phenomena in the course of the Fukushima-Daiichi accident and from lessons learned from validating this version with data from about 150 experiments (analytical and integral experiments carried out worldwide in some fifty different experimental devices).
 
Based on this analysis, IRSN has updated ASTEC to allow adequately taking into account both delayed supply of water and actual BWR geometry, as well as to improve modeling (see below) of fission product transport, core flooding, sheath oxidation, molten core-concrete interaction (MCCI) and pH calculation in the sump. The project partners, including IRSN, validated these numerous improvements.
 
In addition, ASTEC data sets describing the generic characteristics of the main types of Gen.II NPPs in operation in Europe have been developed for use with version 2.1. They were developed by the best experts in each of the fields within the CESAM project.
 
Finally, these various ASTEC reference data sets were then used by the partners for benchmarking with other reference software (such as MELCOR and MAAP) on reactor applications, while paying particular attention to the analysis of resources for mitigating severe accidents.
 
 
Project results [ref-3]
 
The essential final product is the new major version 2.1 of ASTEC [ref-4]. It incorporates improvements by removing several technological obstacles [ref-1] and makes it possible to apply ASTEC to several reactor types, including BWRs, heavy water reactors (PHWRs) and, in part, spent fuel pools.
 

Among the most significant developments in terms of functionality:

  • the possibility of simulating all accident sequences involving a delayed injection of water into the vessel, even if the core is already severely degraded;
  • the provision of new types of objects (internal canisters or channel boxes, sub-channels, cross-shaped control rods) to represent the actual geometry of the BWR cores. The possibility of now being able to model non-axisymmetric cores is also of interest for PHWRs (such as e.g. CANDU NPPs);
  • the treatment, which has now been harmonized, of the transport and chemistry of fission products and aerosols in the reactor coolant system and containment.

 

Among the most significant developments in physical models:

  • integrating a new model of reflooding of a degraded core, specifically designed to be applicable to the geometries of porous media;
  • improving the oxidation model of Zircaloy cladding when it is exposed to a mixed air/vapor atmosphere, while taking nitriding phenomena into account;
  • improving corium behavior models, once relocated in the bottom of the vessel, with in particular the challenge of being able to better model the conditions representing transients for which there is an external vessel cooling circuit (in-vessel melt retention (IVMR) strategy);
  • integrating new corium cooling models with top water in the molten corium-concrete interaction (MCCI) phase, relating to corium ejection and water ingression;
  • integrating a dedicated model for calculating pH in the containment sumps as well as various improvements to the physicochemical behavior models of iodine in the RCS as well as the containment.

 

Furthermore, significant progress has been made in numeric performance to reduce calculation times and more generally increase software reliability.

 

IRSN has extended the ASTEC software to diagnosis applications in emergency situations. The work was divided into two complementary areas. On the one hand, version 2.1 has been coupled with tools for assessing the atmospheric dispersion of radioactive releases outside the containment, and on the other, a method, based on the implementation of Bayesian networks, has been developed to refine tools to trace possible scenarios of radioactivity releases from measurements taken during an accident.

 

This work is now continuing mainly within the European FASTNET project that IRSN is coordinating as part of the Horizon-2020 Framework Programme.

 

Once integrated in version 2.1, the new physical models were validated by the project partners. In particular, this validation work concerned the following issues:

  • reflooding of degraded cores;
  • retention of corium in the vessel in the presence of a cooling circuit outside the vessel;
  • possibility of cooling the corium and debris relocated outside the vessel during the MCCI phase;
  • mitigation of the risk of hydrogen combustion in the containment using passive autocatalytic recombiners;
  • trapping of fission products and aerosols in the containment by pool scrubbing;
  • efficiency of various venting/filtration systems implemented in the containment.

 

The other essential end product of the project is the creation of a library of generic data sets describing the characteristics of the main Gen.II NPP types in operation in Europe. These data sets are referred to as “generic” in the sense that they each model a given reactor type but not a particular unit in operation. They can therefore be freely distributed to all organizations using ASTEC (they are now included in the standard ASTEC V2.1 delivery package), thus offering an ideal base so that each organization can adapt the data sets to the plant unit concerned. In concrete terms, the “generic” data set library resulting from the CESAM project covers four main groups of Gen.II NPPs:

  • two types of Western design PWRs (900 MWe and 1300 MWe);
  • two types of Russian-designed PWRs (VVER-440 and VVER-1000);
  • one type of BWR (BWR-4 with Mark I containment);
  • a CANDU heavy water reactor (PHWR 220 MWe).

 

 

Outlook

 

As an extension to CESAM, IRSN is coordinating a new project called ASCOM, which begins in October 2018 as part of NUGENIA’s Technical Area 2, “Severe Accidents-SARNET”. The project, which received the NUGENIA label in late 2017, will make it possible to consolidate the ASTEC developments made during the CESAM project and to develop new functionalities as the partners' needs evolve. The extension of the “generic” data set library will also be continued. These new data sets will primarily concern Gen.III NPPs (AP1000 and VVER-1200), and possibly spent fuel pools and small modular reactors.

 

 

References

 

[1]   J.P. Van Dorsselaere, A. Auvinen, D. Beraha, P. Chatelard, L.E. Herranz, C. Journeau, W. Klein-Hessling, I. Kljenak, A. Miassoedov, S. Paci, R. Zeyen, "Recent severe accidents research: Synthesis of the major outcomes from the SARNET network", Nuclear Engineering and Design, vol.291 (Sept.2015), p.19-34

 

[2]   P. Chatelard, S. Belon, L. Bosland, L. Carénini, O. Coindreau, F. Cousin, C. Marchetto, H. Nowack, L. Piar., "Main modelling features of the ASTEC V2.1 major version", Annals of Nuclear Energy, Vol.93 (July 2016), pp. 83–93.

 

[3]   H. Nowack, P. Chatelard, L. Chailan, S. Hermsmeyer, V.H. Sanchez, L.E. Herranz, "CESAM– Code for European Severe Accident Management, EURATOM project on ASTEC improvement", Annals of Nuclear Energy, Volume 116, June 2018, Pages 128-136

 

[4]   L. Chailan, A. Bentaïb, P. Chatelard, "Overview of ASTEC code and models for Evaluation of Severe Accidents in Water Cooled Reactors", IAEA Technical Meeting on the Status and Evaluation of Severe Accident Simulation Codes for Water Cooled Reactors, Vienna (Austria), October, 9-12, 2017



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