[en] The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m³, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal neutron flux with rhodium like target. The values of such detectors are proportional to the thermal neutron flux and, therefore, the values during the measurements can be directly used to calculate the spatial distribution of thermal neutron flux. Aluminium-covered and cadmium-covered gold foils were employed to determine the absolute value of thermal neutron flux throughout the reactor core. The neutron flux was determined by measurement of the activity of the samples irradiated. In the irradiation channels in core position 4C, and 3D, two gold foils, one in aluminium cover while the other in cadmium cover were irradiated. The foils in polyethylene sample holder were forwarded to the operating reactor by means of the pneumatic rabbit system joining the irradiation channels 4C, 3D. The gold foils were irradiated during 1.5 hours to 20 kW power level. The sample intensities were taken by a NaI Canberra counter. The calorimetric experiment used in this work for establishing the thermal reactor power it is a quick and precise method for being implemented as a calibration method for IAN-R1 TRIGA reactor. The experimental procedures using a self-powered detector together with the measurement of the activity irradiated foils, enable determination of thermal neutron flux throughout the reactor core of the IAN-R1 TRIGA reactor

[en] With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)