Investigations of Neutron Radiation Shielding Properties for a New Composite Material Based on Heavy Concrete and Basalt Fiber


basalt fiber, heavy composite concrete, radiation protection, Serpent code, neutron transport simulation

How to Cite

Romanenko, I., Holiuk, M., Nosovsky, A., Vlasenko, T., & Gulik, V. (2018). Investigations of Neutron Radiation Shielding Properties for a New Composite Material Based on Heavy Concrete and Basalt Fiber. Nuclear and Radiation Safety, (3(79), 42-47.


The paper presents a new composite material for radiation shielding properties. This material is based on super-heavy concrete reinforced with basalt fiber, which could be used in biological protection systems for neutron radiation sources. The simulation of the neutron transport in the presented material was performed using the Monte Carlo Serpent code. Two types of heavy concretes were considered in the present paper: 1) with ordinary rubble coarse aggregate and 2) with barite coarse aggregate. For each type of concrete, the basalt fiber with dosage from 1 kg/m3 to 50 kg/m3 was added. The current transmission rates were obtained as a result of the neutron-physical modelling for neutron transport from source to detector through the proposed concrete samples with different thicknesses. The obtained modelling results were analyzed from the viewpoint of effectiveness of the radiation shielding properties. Also, elastic and capture microscopic cross-sections were considered for some isotopes and as a result, some aspects of the radiation shielding properties were clarified. The concrete with ordinary rubble coarse aggregate has better radiation shielding properties in case of low concrete thicknesses due to better neutron scattering on light nucleuses. In contrast to this, the concrete with barite coarse aggregate has better radiation shielding properties in case of high concrete thicknesses due to better neutron absorption. It is shown that the addition of basalt fiber to concrete not only improves its mechanical properties and reduces the number and size of microcracks, but also increases the ability to protect against neutron flux. The proposed composite material could be recommended for use with the following neutron sources: (D, T) neutron generators, plasma focus devices, fusion reactors and fast reactors. This research was carried out with the financial support of the IAEA, within the terms and conditions of the Research Contract 20638 in the framework of the Coordinated Research Project (CRP) “Accelerator Driven Systems (ADS) Applications and use of Low-Enriched Uranium in ADS (T33002)” within the project ‘The Two-Zone Subcritical Systems with Fast and Thermal Neutron Spectra for Transmutation of Minor Actinides and Long-Lived Fission Products’.


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