No abstract

No abstract

In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport) code. The sensitivity analyses included small differences of the core and the rods dimensions and different levels of model detailing. Four models were simulated and neutronic parameters such as effective multiplication factor (keff), reactivity (ρ), and thermal and total neutron flux in central thimble in some different conditions of the reactor operation were analysed. The simulated models presented good agreement between them, as well as in comparison with available experimental data. In this way, the sensitivity analyses demonstrated that simulations of the TRIGA IPR-R1 reactor can be performed using any one of the four investigated MCNP models to obtain the referenced neutronic parameters.

BackgroundThis work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed.FindingsThe proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room’s level where shall be located the rack containing the set sample/detector/shielding. The evaluation of the thermal and epithermal neutron flux in the sample position was done considering the experimental data obtained from a vertical neutron guide, already existent in the reactor, and the simulated model for the facility.MethodsThe experimental determination of the neutron flux was obtained through the standard procedure of using Au monitors in different positions of the vertical tube. In order to validate both, this experiment and calculations of the simulated model, the flux was also determined in different positions in the core used for sample irradiation. The model of the system was developed using the Monte Carlo code MCNP5.ConclusionThe preliminary results suggest the possibility of obtaining a beam with minimum thermal flux of magnitude 106 cm-2 s-1, which confirm the technical feasibility of the installation of PGAA at the TRIGA IPR-R1 reactor. This beam would open new possibilities for enhancing the applications using the reactor.

Abstract:In this paper, the evaluation of a preliminary design of a PGAA (prompt gamma activation analysis) facility based on Monte Carlo simulations is presented and discussed. The implementation of the PGAA method at the CDTN (nuclear technology development centre) would increase the applications of the TRIGA reactor. The preliminary design is based on a quasi vertical hollow cylinder (called neutron extractor) in the reactor pool to extract the neutron flux. This study evaluates the neutron flux in the upper position of the cylinder in the suggested position of the samples to be analyzed by the PGAA. The calculations of the radioactive capture reaction rates and of the detection limits for some isotopes were performed. Through all these calculations, the feasibility of the application of the PGAA method at the IPR-R1 installations was evaluated. According to the obtained results, it can be concluded that is possible to apply the PGAA method at the IPR-R1 reactor, even with its design restrictions.

No abstract