The working group on the analysis and management of accidents (WGAMA): A historical review of major contributions

The Working Group on the Analysis and Management of Accidents (WGAMA) was created on December 31st, 1999 to assess and strengthen the technical basis needed for the prevention, mitigation and management of potential accidents in NPP and to facilitate international convergence on safety issues and accident management analyses and strategies. WGAMA addresses reactor coolant system thermal-hydraulics, in-vessel behaviour of degraded cores and in-vessel protection, containment behaviour and containment protection, and fission product (FP) release, transport, deposition and retention, for both current and advanced reactors. As a result, WGAMA contributions in thermal-hydraulics, computational fluid-dynamics (CFD) and severe accidents along the first two decades of the 21st century have been outstanding and are summarized in this paper. Beyond any doubt, the Fukushima-Daiichi accident heavily impacted WGAMA activities and the substantial outcomes produced in the accident aftermath are neatly identified in the paper. Beyond specific events, most importantly, around 50 technical reports have become reference material in the different fields covered by the group and they are gathered altogether in the reference section of the paper; a common outstanding feature in most of these reports is the recommendations included for further research, some of which have eventually given rise to some of the projects conducted or underway within the OECD framework. Far from declining, ongoing WGAMA activities are numerous and a number of them is already planned to be launched in the near future; a short mention to them is also included in this paper

Thermodynamic study of the CsOH(s,l) vaporization by high temperature mass spectrometry

The present study deals with thermodynamic data for the gaseous phase of the Cs-O-H system as studied by high temperature mass spectrometry using the Knudsen cell effusion method. The vapour phase is analyzed and is composed of the monomer CsOH(g), the dimer Cs2O2H2(g), and a small amount of trimer. The vaporization behavior of CsOH(s or l) is analyzed in relation with different physico-chemical phenomena that interfere with the Knudsen method, like creeping and surface diffusion along the walls of the effusion orifice. Besides, the ionization processes are complex and render the interpretation of the mass spectrometric results difficult. Thus, calibration procedures have been carefully analyzed in order to evaluate reliably the uncertainties. The two main independent reactions that lead to thermodynamic data are the following:Cs2O2H2(g) = 2CsOH(g) with ΔdissHring operator(298 K) = 146.6 ± 7.3 kJ · mol-1 (3rd law method).CsOH(s,l) = CsOH(g) with ΔsublHring operator(298 K) = 163.3 ± 6.5 kJ · mol-1 (3rd law method). © 2007 Elsevier Ltd. All rights reserved

Investigation on boron carbide oxidation for nuclear reactor safety: Experiments in highly oxidising conditions

The oxidation kinetics of boron carbide pellets were investigated in steam/argon mixtures in the temperature range 1200-1800 °C for steam partial pressures between 0.2 and 0.8 bar and total flows (steam + argon) between 2.5 and 10 g/min resulting in gas velocities from 1.01 to 5.34 m/s. A kinetic model for boron carbide pellet oxidation depending on temperature, steam partial pressure and flow velocity is obtained. The activation energy of the oxidation process was determined to be 163 ± 8 kJ/mol. The strong influence of temperature and steam partial pressure on the boron carbide oxidation kinetics is confirmed. The obtained data suggest the coexistence of two kinetic regimes, one at 1200 °C and the other at 1400-1800 °C, with different dependence on steam partial pressure. © 2007 Elsevier B.V. All rights reserved

Editorial

International audience[No abstract available

Past and future research at IRSN on corium progression and related mitigation strategies in a severe accident

International audienceReactor core degradation and in-vessel and ex-vessel corium behavior have been major research topics for the last three decades to which Institut de Radioprotection et de Súreté Nucléaire (IRSN) strongly contributed by the coordination of or the contribution to large research programs and through the development and validation of the severe accident (SA) ASTEC code. In recent years, the balance of research efforts has trended toward analyses of pros and cons and assessments of mitigation measures. The outcomes of risk significance analysis [including fuel-coolant interaction (FCI), hydrogen combustion, and molten core-concrete interaction (MCCI) risks] performed in France and corium behavior research are described. The focus these days is on (1) in-vessel melt retention (IVMR) strategies for future reactor concepts and the need to establish the reliability of such strategies when implemented in existing reactors and (2) in-containment corium cooling for existing reactors. This paper summarizes the main achievements and remaining issues related to understanding and modeling of (1) reflooding of a degraded core where, despite substantial knowledge gained through research programs, additional efforts are required to establish the efficiency of such a measure and the associated risks for largely degraded cores; (2) corium behavior in the reactor pressure vessel (RPV) lower head where, despite the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MASCA program results, efforts remain necessary to predict RPV thermal loadings resulting from corium layer evolution and RPV resilience with and without IVMR measures (internal and/or external cooling); (3) FCI for which, despite the OECD/NEA SERENA program results, the knowledge is not sufficient to assess with confidence the induced risk of containment failure; and (4) MCCI, where the knowledge on corium cooling in the containment by top and/or bottom water flooding is insufficient to formulate conclusions regarding the efficiency of such measures. Of particular interest for top flooding are the water ingress and corium eruption processes. Specifically for top flooding, respective impacts of water ingress and corium eruption processes remain to be quantified in reactor conditions. In support of these activities, substantial efforts are also being conducted at IRSN to constantly improve and validate nuclear material property databases that are key tools for corium behavior analysis. This paper describes ongoing and future research programs performed at IRSN or internationally with IRSN coordination or participation to tackle the remaining issues and summarizes expected progress in modeling for SA codes, in risk analysis and in SA management