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Портал Begell Электронная Бибилиотека e-Книги Журналы Справочники и Сборники статей Коллекции
Computational Thermal Sciences: An International Journal
ESCI SJR: 0.249 SNIP: 0.434 CiteScore™: 0.7

ISSN Печать: 1940-2503
ISSN Онлайн: 1940-2554

Computational Thermal Sciences: An International Journal

DOI: 10.1615/ComputThermalScien.2018019423
pages 243-254

EVALUATION OF THE NEUTRONIC FEEDBACK EFFECTS IN LOSS OF COOLANT ACCIDENT SIMULATION OF THE IPR-R1 TRIGA REACTOR

Patricia Amélia de Lima Reis
Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Campus UFMG, PCA 1, 31270-901, Belo Horizonte, MG, Brasil Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brasil
Antonella Lombardi Costa
Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Campus UFMG, PCA 1, 31270-901, Belo Horizonte, MG, Brasil Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brasil
Claubia Pereira
Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Campus UFMG, PCA 1, 31270-901, Belo Horizonte, MG, Brasil Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brasil
Maria Auxiliadora Fortini Veloso
Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Av. Antonio Carlos, 6627, Campus UFMG, PCA 1, 31270-901, Belo Horizonte, MG, Brasil Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brasil
Rafael Herrero Miró
Departamento de Ingenieria Quimica y Nuclear, Universidad Politécnica de Valencia, Camino de Vera, 14, 46022, Valencia, Spain
Gumersindo Verdú Martin
Departamento de Ingenieria Quimica y Nuclear, Universidad Politécnica de Valencia, Camino de Vera, 14, 46022, Valencia, Spain

Краткое описание

The safety analysis of research reactors includes simulations of selected cases classified by the International Atomic Energy Agency (IAEA), since the simulations are performed using validated nodalizations and internationally recognized, accepted, and validated best estimate codes. The thermal hydraulic analysis is considered as an essential aspect in the study of safety of nuclear reactors, since it can predict proper working conditions, steady state and transient, thereby ensuring the safe operation of a nuclear reactor. A RELAP5 model verified for the IPR-R1 TRIGA research reactor was used here to perform transient studies. A loss of coolant accident (LOCA) event was simulated. The obtained results demonstrate that, to more realistically simulate this type of transient, it is necessary to consider also the neutronic feedback effects in the thermal hydraulic calculations.

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