AbstractAbstract
[en] In order to evaluate the reactivity coefficients of the TRIGA IAN-R1 core, a simplified core configuration with no control rods and no internal irradiation channels was calculated. The cross-sections set were recalculated running a Wims code for each temperature of fuel, water and the water density. The effective reactivity was calculated using Citation code with a conceptual model and an X-Y-Z calculation in order to avoid buckling recalculations. For the conceptual model of the TRIGA IAN-R1 core, a value of -7.37 pcm/Celcius degrade was obtained for the fuel temperature coefficient; 3.67 pcm/Celsius degrade and -4.28 pcm/Celsius degrade for the temperature coefficient of the moderator and -95.5 pcm/% for the void coefficient.
Original Title
Calculo de coeficientes de reactividad del reactor nuclear de investigacion TRIGA IAN-R1
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Secondary Subject
Record Type
Journal Article
Journal
Revista de Investigaciones y Aplicaciones Nucleares; ISSN 2590-7468;
; v. 2; p. 31-34

Country of publication
ANALOG SYSTEMS, COMPUTER CODES, DEVELOPING COUNTRIES, DIMENSIONLESS NUMBERS, ENERGY SOURCES, ENRICHED URANIUM REACTORS, FUELS, FUNCTIONAL MODELS, IRRADIATION, IRRADIATION REACTORS, ISOTOPE PRODUCTION REACTORS, LATIN AMERICA, MATERIALS, POOL TYPE REACTORS, REACTIVITY COEFFICIENTS, REACTOR COMPONENTS, REACTOR MATERIALS, REACTORS, RESEARCH AND TEST REACTORS, RESEARCH REACTORS, SIMULATION, SIMULATORS, SOUTH AMERICA, THERMAL REACTORS, TRAINING REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS
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AbstractAbstract
[en] With cooperation of the International Atomic Energy Agency (IAEA), thermal-hydraulic calculations were carried out for conversion of the IAN-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. To establish thermal-hydraulic calculation and analysis research in Colombia, this program was carried out and included training, acquisition of hardware, software and natural convection flow calculations for the TRIGA IAN-R1 research reactor operating at 100 kW. The purpose of the study is to validate the steady state thermal hydraulic analysis that has been carried out by means of the NATCON code. This paper presents the results of the maximum axial temperature distribution for fuel, clad, and coolant. In addition, the Bernath critical heat flux with pool water temperature as a parameter is presented.
Original Title
Calculos de temperatura y flujo de calor para el canal de maxima potencia del reactor de investigacion TRIGA IAN-R1
Primary Subject
Source
Available on line: https:www.sgc.gov.co
Record Type
Journal Article
Journal
Revista de Investigaciones y Aplicaciones Nucleares; ISSN 2590-7468;
; v. 3; p. 36-39

Country of publication
DEVELOPING COUNTRIES, ENRICHED URANIUM REACTORS, FLUID MECHANICS, HYDRAULICS, INTERNATIONAL ORGANIZATIONS, IRRADIATION REACTORS, ISOTOPE PRODUCTION REACTORS, LATIN AMERICA, MECHANICS, NATIONAL ORGANIZATIONS, POOL TYPE REACTORS, REACTORS, RESEARCH AND TEST REACTORS, RESEARCH REACTORS, SOUTH AMERICA, THERMAL REACTORS, TRAINING REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS
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