Analysis of ITU TRIGA Mark II research reactor using Monte Carlo method

TÜRKMEN M., Çolak Ü.

PROGRESS IN NUCLEAR ENERGY, vol.77, pp.152-159, 2014 (SCI-Expanded) identifier identifier

  • Publication Type: Article / Article
  • Volume: 77
  • Publication Date: 2014
  • Doi Number: 10.1016/j.pnucene.2014.06.015
  • Journal Indexes: Science Citation Index Expanded (SCI-EXPANDED), Scopus
  • Page Numbers: pp.152-159
  • Hacettepe University Affiliated: Yes


Research reactors include many complicated components with various shapes and sizes. Such complex parts also observed in TRIGA core are modelled by the researchers as simplified physical geometries when a particle transport computer code is used to analyse the reactors. These models are used to gain information on possible modifications in the reactors with no cost except a certain computational time demand. Besides, they can be used to understand the fabrication uncertainties of the core components and the methodologies used in the design process. The main objective of this study is to make a detailed three-dimensional full-core model of ITU (Istanbul Technical University) TRIGA Mark II research reactor for the use of Monte Carlo method and making a comparison of the simulation with the experimental observations. In case of lacking of experimental values reported, Final Safety Analysis Report values are used as reference. Furthermore, it is aimed to observe possible influences of using various neutron cross-section libraries (ENDF/Bs and JEFFs) onto the simulation results. The Monte Carlo simulations are carried out by using MCNP5 radiation transport code. All unsteady conditions are ignored, assuming the reactor operates at cold-zero power under the steady-state condition. For comparison, effective core multiplication factor (k(eff)) and effective delayed neutron fraction (beta(eff)) are computed. Reactivity worth ($) of control rods with rod position is presented. Pin power distribution within the fuel elements, axial power peaking distribution along the fuel length and normalized distribution of fast/thermal neutron flux throughout the core are analysed. The simulation results show that MCNP5 model of the reactor is properly established with sufficient detail in such a way that all simulation results are in an excellent agreement with the experimental data (or FSAR values). Results also show that the model yields more or less the same value even different neutron libraries are used. (C) 2014 Elsevier Ltd. All rights reserved.