[en] The present work describes the procedure carried out for the preliminary determination of neutron flux parameters for the nuclear research reactor IAN-R1 (RNI IAN-R1) through the not covered triple monitor method in the nucleus peripheral irradiation position. Using this method, the thermal flux value (φ_th), the epithermal neutron flux symmetry factor (α) and the ratio between thermal neutron fluxes with respect to the epithermal neutron flux (f) were estimated. Those parameters were obtained by irradiating zirconium (Zr) monitors and a gold-aluminum alloy monitors (Au-Al 0.1% Au), which were irradiated at the G3 and G4 irradiation positions of the RNI IAN-R1. The following values were found for the parameters estimated at an operating power of 30 kW, φ_th= 2,1 * 10¹¹ cm² s^-1 (variance CV 4%), α = 0,02 (CV 83 %), and f =67 (CV 8 %). The high variance in α could be explained if we consider that the method only uses 3 capture reactions to describe the epithermal neutron spectrum. The variance could be improved by application of multimonitor methods for neutron flux characterization.

[en] This document presents a historical description of the nuclear research reactor IAN-R1. A contextualization is made about the origin of the reactor within the framework of the Atoms for Peace program, including the technical characteristics and the initial configuration of the core, which was replaced by nuclear fuel MTR technology (90 %) to a new fuel type TRIGA (20 %) (acronyms of material testing reactor and training, research, isotopes, general atomics respectively). In the same way, the characteristics of the two modernization that have been made to the instrumentation and control are presented, the first oriented to the installation of three nuclear channels two of wide range and one power channel, renovation of the control console and the installation of the data acquisition system (DAC) cabinet. The second modernization, which corresponds to the new instrumentation and control of the reactor, is oriented to the change of the control console which supports the control and supervision servers, a nuclear channel NP-1000, printer, four screens of the human interface machine HMI, keyboard of the bar handling system and two keyboards for each of the servers. In addition, the DAC was replaced by the instrumentation cabinet, which includes the reactor protection systems, the redundant control system and the supervision system. The instrumentation and control is characterized by the use of the Ethernet standard to achieve inter-connectivity of the systems, programming of the human machine interface (HMI) using open source code Java, and multi platform, logical separation of functions plying concepts of distributed control and modularity, redundancy, unique failure criteria and independence. The use of the reactor is shown, referring to the irradiation facilities available for irradiation of materials to be studied using the neutron activation analysis (NAA) technique. Likewise, irradiation is planned to support the use of the fission fingerprint dating technique, research and support to educational institutions through technical conferences and a visit to the nuclear facility.

[en] In this paper, the parameters thermal to epithermal neutron flux ratio (f) and thermal neutron flux (θ_th) are estimated (assuming α = 0) for the periphery irradiation positions (G3-G4) available at Nuclear Research Reactor IAN-R1 that belongs to the Colombian Geological Survey. This estimation was performed by measuring the induced activity, from neutron capture reactions (n, y) in two reference soils, which were irradiated (n = 4) together with Al-Au monitors. Four reference capture reactions were considered in soils, which involve isotopes of elements with certified mass fraction and 1/v behavior for cross section variation in the range of thermal neutron energy. An average value of 33 (CV = 9%) for f was found, considering 8 positions in the irradiation container; this value is comparable with measurements reported for periphery irradiation channels of TRIGA type reactors. The average thermal neutron flux determined for the 8 position of the container resulted in 1,5 *10¹¹ ncm^-2 s^-1 (CV = 4 %).

[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.

[en] A review of the radionuclides produced in nuclear reactors and their applications as radiotracers in hydrology, agriculture, industry, medicine, the environment and research in general are presented, both in Latin America and those produced in Colombia since 1965 with the Research Nuclear Reactor IAN-R1, so that under that view it is possible to reactivate the production of radionuclides in the country as 198Au, 24Na, 32P and 82Br.

[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.