The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
| Autor: | Valery F. Strizhov, I. A. Klimonov, I. G. Kudashov, S. A. Frolov, A. E. Kutlimetov, A. A. Sorokin, V. I. Chukhno, A. A. Butov, V. S. Zhdanov, N. A. Mosunova, E. V. Usov |
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| Rok vydání: | 2019 |
| Předmět: |
Fission products
Liquid metal Neutron transport Materials science Nuclear engineering Energy Engineering and Power Technology 02 engineering and technology 01 natural sciences 010305 fluids & plasmas law.invention Coolant 020401 chemical engineering Nuclear Energy and Engineering Liquid metal cooled reactor Nuclear reactor core law 0103 physical sciences Heat transfer Nuclear power plant 0204 chemical engineering |
| Zdroj: | Thermal Engineering. 66:293-301 |
| ISSN: | 1555-6301 0040-6015 |
| Popis: | The article describes the basic models laid down in the second version of the EUCLID/V2 integrated code developed for carrying out end-to-end analysis of severe accidents in liquid metal cooled reactors. Brief information about the basic analogs of the code is given. Unlike the first version of the code, its second version includes additional tools for analyzing design-basis and beyond-design-basis accidents involving fuel pin, fuel assembly, and reactor core failures. To this end, the code is supplemented with additional modules using which it is possible to calculate fuel rod tightness failure as a consequence of its melting, escape of fission products into the coolant, their transport over the circuit, and release into the nuclear power plant rooms. The code also incorporates modules for calculating the core failure processes. Special attention is paid to the physical models for calculating the core materials' melting processes, motion of the produced melt, its interaction with the coolant and with other materials, and propagation of fission materials. For calculating the core failure processes, a multicomponent 3D model has been implemented. The methods used for calculating heat transfer and friction between the components are based on well-proven analytical and empirical relations for determining the heat transfer and friction coefficients. The coefficients presented in the article also depend on the obtained multicomponent flow motion regime and the type of components (metal and ceramics). The algorithms governing joint operation of the thermomechanical, thermal-hydraulic, neutronics, and the fuel rod thermal failure module are described. Emphasis is placed on data exchange methods in the course of an accident in the reactor. The approaches used for calculating the transport of fission products in the coolant and in the NPP rooms are presented. |
| Databáze: | OpenAIRE |
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