Validation and verification of FLUKA for neutron shielding problems

Dondolo, Petrus

Validation and verification of FLUKA for neutron shielding problems

Master Thesis

2022

Abstract

Monte Carlo-based radiation transport codes provide an opportunity to simulate situations with various levels of activation and different induced nuclides. However, to test their reliability, it is important to verify the simulation codes by comparing them with experimental data. In this study, validation of simulation models with experiments was performed with the purpose of determining the reliability of the simulation/experimental results. Concrete is the most generally used shield material as it is inexpensive and adjustable for any construction design. Radiation shielding properties of concrete may vary depending on the concrete composites. In this thesis, the fluences (i.e. the flux integrated over time) of neutrons impinging on the shielding nuclear material were studied using FLUKA Monte Carlo package. The rectangular blocks of shielding nuclear materials such as concrete ingredients: cement, sand and water were irradiated with a beam of 14 MeV neutrons and the shielding properties of these materials were investigated using FLUKA Monte Carlo simulation code. The simulation set-up replicates the experimental measurements performed within the nuclear laboratory in the Department of Physics at the University of Cape Town. The comparison of the effective removal cross-section shows a good agreement between experiments and FLUKA. The results from these two approaches show general agreement for sand and cement, but show some minor deviations for water and concrete. The source of these deviations is discussed, along with potential solutions. FLUKA has been well benchmarked and validated against other Monte Carlo codes. The discrepancies obtained on water and concrete may have occurred from the material properties in the input file. Comparisons of results are presented and the discrepancies and agreements between the two methods are discussed for these target materials. The effective removal cross section of a concrete mix was measured by simulation to be 0.1038 +/- 0.0005 cm-1 and by experiment to be 0.1230 +/- 0.0002 cm-1 of 14 MeV neutrons. This illustrates a broad agreement between experiment and simulation in the case of concrete ingredients. Validation and comparison of measured and simulated neutron irradiation on concrete ingredients shows good agreement, supporting the use of FLUKA for estimating the neutron transmission into the shielding material.