Отправить статью по E-mail The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
- Авторы: Butov A.A.1, Zhdanov V.S.1, Klimonov I.A.1, Kudashov I.G.1, Kutlimetov A.E.1, Mosunova N.A.2, Strizhov V.F.2, Sorokin A.A.2, Frolov S.A.1, Usov E.V.1, Chukhno V.I.1
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Учреждения:
- Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
- Nuclear Safety Institute, Russian Academy of Sciences (IBRAE)
- Выпуск: Том 66, № 5 (2019)
- Страницы: 293-301
- Раздел: Nuclear Power Plants
- URL: https://bakhtiniada.ru/0040-6015/article/view/173245
- DOI: https://doi.org/10.1134/S0040601519050021
- ID: 173245
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Аннотация
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.
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Об авторах
A. Butov
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
V. Zhdanov
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
I. Klimonov
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
I. Kudashov
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
A. Kutlimetov
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
N. Mosunova
Nuclear Safety Institute, Russian Academy of Sciences (IBRAE)
Email: usovev@gmail.com
Россия, Moscow, 115191
V. Strizhov
Nuclear Safety Institute, Russian Academy of Sciences (IBRAE)
Email: usovev@gmail.com
Россия, Moscow, 115191
A. Sorokin
Nuclear Safety Institute, Russian Academy of Sciences (IBRAE)
Email: usovev@gmail.com
Россия, Moscow, 115191
S. Frolov
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
E. Usov
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Автор, ответственный за переписку.
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
V. Chukhno
Novosibirsk Branch, Nuclear Safety Institute, Russian Academy of Sciences
Email: usovev@gmail.com
Россия, Novosibirsk, 630090
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