Development and experimental validation of a Monte Carlo simulation model for the Triga Mark II reactor

In recent years, many computer codes, based on Monte Carlo methods or deterministic calculations, have been developed to separately analyze different aspects regarding nuclear reactors. Nuclear reactors are very complex systems, which require an integrated analysis of all the variables which are intrinsically correlated: neutron fluxes, reaction rates, neutron moderation and absorption, thermal and power distributions, heat generation and transfer, criticality coefficients, fuel burnup, etc. For this reason, one of the main challenges in the analysis of nuclear reactors is the coupling of neutronics and thermal-hydraulics simulation codes, with the purpose of achieving a good modeling and comprehension of the mechanisms which rule the transient phases and the dynamic behavior of the reactor. This is very important to guarantee the control of the chain reaction, for a safe operation of the reactor. In developing simulation tools, benchmark analyses are needed to prove the reliability of the simulations. The experimental measurements conceived to be compared with the results coming out from the simulations are really precious and can provide useful information to improve the description of the physics phenomena in the simulation models. My PhD research activity was held in this framework, as part of the research project Analysis of Reactor COre (ARCO, promoted by INFN) whose task was the development of modern, flexible and integrated tools for the analysis of nuclear reactors, relying on the experimental data collected at the research reactor TRIGA Mark II, installed at the Applied Nuclear Energy Laboratory (LENA) at the University of Pavia. In this way, once the effectiveness and the reliability of these tools for modeling an experimental reactor have been demonstrated, these could be applied to develop new generation systems. In this thesis, I present the complete neutronic characterization of the TRIGA Mark II reactor, which was analyzed in different operating conditions through experimental measurements and the development of a Monte Carlo simulation tool (relied on the MCNP code) able to take into account the ever increasing complexity of the conditions to be simulated. First of all, after giving an overview of some theoretical concepts which are fundamental for the nuclear reactor analysis, a model that reconstructs the first working period of the TRIGA Mark II reactor, in which the “fresh” fuel was not heavily contaminated with fission reaction products, is described. In particular, all the geometries and the materials are described in the MCNP simulation model with good detail, in order to reconstruct the reactor criticality and all the effects on the neutron distributions. The very good results obtained from the simulations of the reactor at low power condition -in which the fuel elements can be considered to be in thermal equilibrium with the water around them- are then used to implement a model for simulating the full power condition (250kW), in which the effects arising from the temperature increase in the fuel-moderator must be taken into account. The MCNP simulation model was exploited to evaluate the reactor power distribution and a dedicated experimental campaign was performed to measure the water temperature within the reactor core. In this way, through a thermal-hydraulic calculation tool, it has been possible to determine the temperature distribution within the fuel elements and to include the description of the thermal effects in the MCNP simulation model. Thereafter, since the neutron flux is a crucial parameter affecting the reaction rates and thus the fuel burnup, its energy and space distributions are analyzed presenting the results of several neutron activation measurements. Particularly, the neutron flux was firstly measured in the reactor's irradiation facilities through the neutron activation of many different isotopes. Hence, in order to analyze the energy flux spectra, I implemented an analysis tool, based on Bayesian statistics, which allows to combine the experimental data from the different activated isotopes and reconstruct a multi-group flux spectrum. Subsequently, the spatial neutron flux distribution within the core was measured by activating several aluminum-cobalt samples in different core positions, thus allowing the determination of the integral and fast flux distributions from the analysis of cobalt and aluminum, respectively. Finally, I present the results of the fuel burnup calculations, that were performed for simulating the current core configuration after a 48 years-long operation. The good accuracy that was reached in the simulation of the neutron fluxes, as confirmed by the experimental measurements, has allowed to evaluate the burnup of each fuel element from the knowledge of the operating hours and the different positions occupied in the core over the years. In this way, it has been possible to exploit the MCNP simulation model to determine a new optimized core configuration which could ensure, at the same time, a higher reactivity and the use of less fuel elements. This configuration was realized in September 2013 and the experimental results confirm the high quality of the work done. The results of this Ph.D. thesis highlight that it is possible to implement analysis tools -ranging from Monte Carlo simulations to the fuel burnup time evolution software, from neutron activation measurements to the Bayesian statistical analysis of flux spectra, and from temperature measurements to thermal-hydraulic models-, which can be appropriately exploited to describe and comprehend the complex mechanisms ruling the operation of a nuclear reactor. Particularly, it was demonstrated the effectiveness and the reliability of these tools in the case of an experimental reactor, where it was possible to collect many precious data to perform benchmark analyses. Therefore, for as these tools have been developed and implemented, they can be used to analyze other reactors and, possibly, to project and develop new generation systems, which will allow to decrease the production of high-level nuclear waste and to exploit the nuclear fuel with improved efficiency.