%0 Journal Article
%T Preliminary Assessment of the Possible BWR Core/Vessel Damage States for Fukushima Daiichi Station Blackout Scenarios Using RELAP/SCDAPSIM
%A C. M. Allison
%A J. K. Hohorst
%A B. S. Allison
%A D. Konjarek
%A T. Bajs
%A R. Pericas
%A F. Reventos
%A R. Lopez
%J Science and Technology of Nuclear Installations
%D 2012
%I Hindawi Publishing Corporation
%R 10.1155/2012/646327
%X Immediately after the accident at Fukushima Daiichi, Innovative Systems Software and other members of the international SCDAP Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1¨C3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM. The calculations used a detailed RELAP/SCDAPSIM model of the Laguna Verde BWR vessel and related reactor cooling systems. The Laguna Verde models were provided by the Comision Nacional de Seguridad Nuclear y Salvaguardias, the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy Agency on March 21 to support their emergency response team and later to our Japanese members to support their Fukushima Daiichi specific analysis and model development. 1. Introduction Immediately after the accident at Fukushima Daiichi, Innovative Systems Software (ISS) and other members of the international SCDAP Development and Training Program (SDTP) [1, 2] started an assessment of the possible core/vessel damage states of the Fukushima Daiichi Units 1¨C3. The assessment included a brief review of relevant severe accident experiments and a series of detailed calculations using RELAP/SCDAPSIM [3, 4] for a representative BWR vessel and related cooling systems. As described briefly in Section 2, the experimental review presented to the IAEA emergency response team included representative highlights and phenomena identified from separate effects experiments BWR specific and other bundle experiments [5, 6], performed by the Karlsruhe Institute of Technology (KIT), and selected in-pile experiments [7, 8]. The KIT experiments were limited to peak temperatures less than 2600£¿K and thus covered the initial stages of core heat up and melting including the liquefaction and relocation of BWR control blades, structural material, and fuel rod cladding. The in-pile experiments reached higher peak temperatures and included the liquefaction of the fuel and other oxidized cladding materials and the formation of ceramic melts and blockages. As described in Section 3, a combination of RELAP/SCDAPSIM/MOD3.4 and RELAP/SCDAPSIM/MOD3.5 was used to perform the detailed calculations. Both versions use publically available RELAP5/MOD3.2 and MOD3.3 thermal hydraulic models and correlations in combination with the detailed fuel behaviour and severe accident (SCDAP) models and correlations [9, 10]. The RELAP/SCDAPSIM code is designed to predict
%U http://www.hindawi.com/journals/stni/2012/646327/