Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code

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.

Monte Carlo Simulations for a Preliminary Design of TRIGA IPR-R1 PGAA Facility

“…For all calculations estimated in each isotope, the relative code error was < 5%, which is acceptable [15][16][17]. The detection limit is the lowest amount of an isotope that can be identified in the sample, i.e., the Detection Limit (g) 58 …”

“…To set up a preliminary model of the PGAA method at the IPR-R1 reactor and estimate the calculations at PGAA facility, in this work was used the models of previous studies [12,[15][16][17] that were validated through the MCNP5 code.…”

“…All calculations estimated in this study through MCNP5 [15][16][17] were carried out with the 5. For the calculations it was used the cluster CENAPAG-MG [6,[12][13][14]18].…”

“…In the position chosen to insert the sample into the PGAA facility, a thermal neutron flux of 5.97 × 10 According to MCNP manual [15][16][17], a relative error < 10% is acceptable. For all neutron fluence rates estimated by MCNP5, the relative errors were below 10%.…”

“…Along the years, the TRIGA IPR-R1 reactor has been studied and modeled using the MCNP (Monte Carlo N-Particle) code. Several studies using the MCNP code have developed theoretical methods to calculate neutron flux in the sample's irradiation devices of the IPR-R1 reactor (central thimble, rotary rack, and two pneumatic tubes) and compare their results with experimental data in order to validate these models [15][16][17].…”