Neutron Shielding Properties Of Concrete With Boron And Boron Containing Mineral

Abstract

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Keywords

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Neutron shielding properties of concrete with boron and boron containing mineral

Abstract

OBJECTIVE: Concrete is a material which widely used as a neutron shielding and in building construction such as nuclear power stations, particle accelerators and medical hospitals. Concrete is very significant for neutron shielding, because of concrete is contain of some elements (hydrogen, iron etc.) to moderate the fast neutrons which are very penetrative. Boron increases the neutron shielding effectiveness of concretes, since the boron isotope 10B has a high capture cross-section for thermal neutrons (over 3800 barns). Boron can be added to concrete different ways such as addition of boron to the water used in concrete or addition of boron containing natural minerals.METHODS: Neutron shielding capabilities of the sample can be described by total neutron macroscopic cross-section (t). It is the sum of the cross-sections for all the neutron-interaction processes such as the elastic and inelastic scattering reactions and neutron capture reactions ((n,α),(n,γ)). RESULTS: In this study, the effect of addition boron and colemanite on the total macroscopic cross section of Portland concrete was investigated. Colemanite is one of the most important boron minerals and Turkey has the largest colemanite reserves in the world. Its compact formula is Ca2B6O115(H2O). Also, colemanite can be used for shielding fast and thermal neutrons, since it includes both of hydrate and boron.CONCLUSION: In experiments, 241Am-Be neutron source with 74 GBq activity were used in. Average neutron energy of this source is approximately 4.5 MeV. BF3 detector with diameter 2.54 cm and length of 28 cm was used for counting neutrons. Also, Monte Carlo simulations were done for comparison of macroscopic cross section experimental results. Besides total macroscopic cross sections, absorbed doses and deposited energies by low energy neutron interactions were calculated using MCNP4C2 Monte Carlo code. The results have been compared with the standard shielding material of paraffin. Also, half-value layer (HVL) and tenth-value layer (TVL) were calculated and compared.

Keywords

• Neutron Shielding, MCNP Monte Carlo Code, Total Macroscopic Cross Section, HVL, TVL

References

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