Gamma and neutron attenuation characteristics of bricks containing zinc extraction residue as a novel shielding material

2021 ◽  
Vol 139 ◽  
pp. 103878
Author(s):  
Osman Gencel ◽  
Ahmet Bozkurt ◽  
Erol Kam ◽  
Ali Yaras ◽  
Ertugrul Erdogmus ◽  
...  
Keyword(s):  
Zinc Extraction ◽  
Neutron Attenuation ◽  
Attenuation Characteristics ◽  
Shielding Material

Effect of heat treatment on neutron attenuation characteristics of high density concretes (HDC)

2016 ◽  
Vol 93 ◽  
pp. 76-83 ◽  
Author(s):  
Sivathanu Pillai C. ◽  
Santhakumar A.R. ◽  
Chandrasekaran S. ◽  
Viswanathan S. ◽  
Mathiyarasu R. ◽  
...  
Keyword(s):  
Heat Treatment ◽  
High Density ◽  
Neutron Attenuation ◽  
Attenuation Characteristics

scholarly journals Analysis of Barite Concrete as a Potential Neutron Radiation Shielding Material for BNCT Facilities

2019 ◽  
Vol 36 (1) ◽  
pp. 19-21
Author(s):  
Rini Murtafi'atin ◽  
Widarto Widarto ◽  
Susilo Susilo ◽  
Ngurah Made Dharma Putra
Keyword(s):  
Attenuation Coefficient ◽  
Germanium Detector ◽  
Neutron Radiation ◽  
Radiation Shielding ◽  
Radiation Level ◽  
Boron Neutron Capture ◽  
Neutron Attenuation ◽  
Density Analysis ◽  
Shielding Material

This research aimed to determine the potential of barite concrete as a neutron radiation shielding material in the development of boron neutron capture therapy, by obtaining its neutron attenuation coefficient. Barite concrete samples were supplied by the Center of Accelerator Science and Technology in Yogyakarta, Indonesia. The experiment consisted of two parts, namely density analysis and determination of the neutron attenuation coefficient. For the latter, plutonium-beryllium was used as the neutron source, while a high purity germanium detector was used to measure the neutron radiation level. The results showed that barite concrete with a 2130 kg.m–3 density had a neutron attenuation coefficient of 0.0871 cm–3.

Comparison of Lead, Copper and Aluminium as Gamma Radiation Shielding Material through Experimental Measurements and Simulation Using MCNP Version 4c

2018 ◽  
Vol 9 (08) ◽  
pp. 20193-20206 ◽  
Author(s):  
Md. Akhlak Bin Aziz ◽  
Md. Faisal Rahman ◽  
Md. Mahidul Haque Prodhan
Keyword(s):  
Gamma Radiation ◽  
Sodium Iodide ◽  
Experimental Approach ◽  
Radiation Shielding ◽  
Test Material ◽  
Simulation Techniques ◽  
Radiation Sources ◽  
Simulation Monte Carlo ◽  
Shielding Material ◽  
Code Version

The paper compares  Lead, Copper and Aluminium as gamma radiation shielding material using both experimental and simulation techniques. Cs- 137 (662KeV), Na-22 (511KeV) and Na- 22(1274KeV) were used as gamma radiation sources and a sodium iodide (NaI) detector was used to detect the radiation. Variations were noted for detected gamma count rates by changing shielding material thickness. In the experimental approach, thickness was varied by placing sheets of a particular test material one by one. For simulation, Monte Carlo n- Particle (MCNP) code version 4c was used and the geometry of the whole experimental setup was plotted in it. The results were then compared for each test material and it was found that lead is the best shielding material for gamma radiation followed by copper and aluminium.

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