Neutron Production Cross-Section and Geant4 Calculations of the Structural Fusion Material 59Co for (α,xn) and (γ,xn) Reactions

Abstract

Recent advances in technology and computer sciences give us simplicity to investigate phenomena of nuclear physics. Scientists have improved various nuclear reaction codes such as Talys, Alice/ASH, Cem95, Empire, Geant4 and Fluka. These programs give us chance to calculate crucial quantity like reaction cross-section, energy spectrum of out-going particles, stopping power and penetrating distances in target material. The stopping power of alpha in 59Co material is acquired as it has helpful applications of shielding and choosing the proper thickness of the target. Evaluation of the reaction cross-section data is very important to various applications such as improvement structural material, reactor design and radioisotopes production. In this study, we calculated the neutron production cross-sections of 59Co using TALYS 1.6 and EMPIRE 3.1 codes for different level density models. Penetrating distances and stopping powers have been calculated for the alpha particles taking into consideration all possible reactions in 59Co using GEANT4 simulation program. The obtained (α,xn) and (γ,xn) reactions (x = 1, 2, 3) cross-section values have been compared with the each other and against the experimental nuclear reaction data existing in EXFOR database.