scholarly journals THE RADIOACTIVITY ESTIMATION OF THE IRRADIATED 13 MEV CYCLOTRON’S CONCRETE SHIELD

2020 ◽  
Vol 22 (1) ◽  
pp. 17
Author(s):  
Isdandy Rezki Febrianto ◽  
Puradwi Ismu Wahyono ◽  
Suharni Suharni
Keyword(s):  
Half Life ◽  
Flux Distribution ◽  
Gamma Ray ◽  
Computer Code ◽  
Gamma Dose ◽  
Fast Neutrons ◽  
Neutron Shielding ◽  
Radiation Workers ◽  
High Neutron ◽  
Induced Radioactivity

The Center for Accelerator Science and Technology (PSTA) planned to install K500 concrete shield in its 13 MeV cyclotron facility (DECY-13). However, fast neutrons that are generated by this cyclotron could activate materials of the concrete. It may harm the radiation workers. In this work, we conducted simulations using ORIGEN2 and PHITS computer code to estimate the formed radioactivity and the neutron flux distribution in the DECY-13 cyclotron's concrete shield. Based on the simulation, the induced radioactivity is 2.3478 × 109 Bq, while its gamma dose rate is 22.09 µSv/m2h. The most contributed isotopes are Th-233, Ho-166, Al-28, Mn-56 and Si-31. This dose is quite high. Neutron fluxes in the rear of the simulated concrete shield are also still prominent. Accordingly, it is necessary to attach neutron shielding materials which do not generate high-intensity gamma-ray. The formed radioactivity is high; but it appears from the short half-life isotopes such as Th-233, Ho-166, Al-28, Mn-56 and Si-31. Its activity will diminish quickly after the cyclotron is off. Hence, it will be safe for radiation workers.Keywords: Radioactivity, Concrete Shield, 13 MeV Cyclotron, Neutron Irradiation, DECY-13, PHITS

scholarly journals Investigations of gamma ray and fast neutron shielding properties of tellurite glasses with different oxide compositions

2016 ◽  
Vol 94 (11) ◽  
pp. 1133-1137 ◽  
Author(s):  
M.I. Sayyed
Keyword(s):  
Photon Energy ◽  
Gamma Ray ◽  
Effective Atomic Number ◽  
Mass Attenuation Coefficient ◽  
Radiation Shielding ◽  
Fast Neutrons ◽  
Tellurite Glasses ◽  
Neutron Shielding ◽  
Shielding Properties ◽  
Mass Attenuation

In the present paper, the basic radiation parameters of tellurite glasses with different forming oxides (B2O3, BaO, K2O, V2O5, WO3, and ZnO) have been studied over a wide photon energy range from 1 keV to 100 GeV, using WinXCom program. These parameters are the mass attenuation coefficient (μ/ρ), effective atomic number (Zeff), and electron density (Ne,eff). In addition, the macroscopic effective removal cross sections (ΣR) for fast neutrons have been calculated. The dependence of different parameters on incident photon energy and chemical content has been discussed. Among the selected glass systems TeO2–WO3 and TeO2–B2O3 showed superior shielding properties for gamma-ray and neutrons, respectively. It is shown that the selected glass systems have higher values of the mass attenuation coefficients (μ/ρ) than concretes and 0.7SiO2:0.3BaO glass. These results indicate that tellurite glasses can be used as radiation shielding materials.

scholarly journals Shielding efficiency of metal hydrides and borohydrides in fusion reactors

2016 ◽  
Vol 31 (4) ◽  
pp. 339-348
Author(s):  
Vishvanath Singh ◽  
Nagappa Badiger ◽  
Leif Gerward
Keyword(s):  
Gamma Ray ◽  
Metal Hydrides ◽  
Fast Neutrons ◽  
Fusion Reactors ◽  
Neutron Shielding ◽  
Attenuation Coefficients ◽  
Removal Cross Section ◽  
Suitable Combination ◽  
Effective Neutron ◽  
Buildup Factors

Mass attenuation coefficients, mean free paths and exposure buildup factors have been used to characterize the shielding efficiency of metal hydrides and borohydrides, with high density of hydrogen. Gamma ray exposure buildup factors were computed using five-parameter geometric progression fitting at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neutron removal cross-section. It is shown that ZrH2 and VH2 are very good shielding materials for gamma rays and fast neutrons due to their suitable combination of low- and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors.

scholarly journals Enhancement of the Shielding Capability of Soda–Lime Glasses with Sb2O3 Dopant: A Potential Material for Radiation Safety in Nuclear Installations

2020 ◽  
Vol 11 (1) ◽  
pp. 326
Author(s):  
M.I. Sayyed ◽  
K.A. Mahmoud ◽  
O.L. Tashlykov ◽  
Mayeen Uddin Khandaker ◽  
M.R.I. Faruque
Keyword(s):  
Attenuation Coefficient ◽  
Elastic Moduli ◽  
Gamma Ray ◽  
Radiation Safety ◽  
Soda Lime ◽  
Mass Attenuation Coefficient ◽  
Transmission Factor ◽  
Linear Attenuation Coefficient ◽  
Neutron Shielding ◽  
Removal Cross Section

Elastic moduli were theoretically computed using the Makishima–Mackenzie model for SiO2–Na2O–CaO glasses doped with Sb2O3 contents. The calculated elastic moduli (Young’s, bulk, shear, and longitudinal modulus) were observed to increase with an increase in the Sb2O3 contents. The microhardness showed an increase, while Poisson’s ratio decreased with the rise of the Sb2O3 contents. In addition, gamma-ray and neutron shielding parameters were evaluated for the investigated glasses. The linear attenuation coefficient (LAC) was simulated using the Monte Carlo N-particle transport code (MCNP-5). Other parameters, such as the mass attenuation coefficient (MAC), transmission factor (TF), and half-value layer, were calculated based on the simulated LAC. The addition of Sb2O3 content was observed to enhance the investigated glasses’ shielding parameters, where the highest LAC was achieved for the SCNSb10 glass with 10 mol% Sb2O3 and decreased from 0.441 to 0.154 cm−1 at gamma energies between 0.248 and 1.406 MeV. Furthermore, the fast neutron effective removal cross-section (∑R) was computed theoretically. The calculated results showed that the highest ∑R was equal to 0.0341 cm2g−1 and was obtained for the SCNSb0 glass, which had no Sb2O3 content, while the lowest ∑R was equal to 0.0286 cm2 g−1 for the SCNSb10 glass sample. The present work was carried out to examine the advantages of the soda–lime glasses with different Sb2O3 contents in several photon shielding applications, especially for radiation safety in nuclear installations.

Development of SiO2 based doped with LiF, Cr2O3, CoO4 and B2O3 glasses for gamma and fast neutron shielding

2020 ◽  
Vol 0 (0) ◽  
Author(s):  
Bünyamin Aygün ◽  
Erdem Şakar ◽  
Abdulhalik Karabulut ◽  
Bünyamin Alım ◽  
Mohammed I. Sayyed ◽  
...  
Keyword(s):  
Monte Carlo Simulation ◽  
Monte Carlo ◽  
Fast Neutron ◽  
Cross Sections ◽  
Gamma Ray ◽  
Absorbed Dose ◽  
Neutron Shielding ◽  
Shielding Properties ◽  
Effective Neutron ◽  
Simulation Codes

AbstractIn this study, the fast neutron and gamma-ray absorption capacities of the new glasses have been investigated, which are obtained by doping CoO,CdWO4,Bi2O3, Cr2O3, ZnO, LiF,B2O3 and PbO compounds to SiO2 based glasses. GEANT4 and FLUKA Monte Carlo simulation codes have been used in the planning of the samples. The glasses were produced using a well-known melt-quenching technique. The effective neutron removal cross-sections, mean free paths, half-value layer, and transmission numbers of the fabricated glasses have been calculated through both GEANT4 and FLUKA Monte Carlo simulation codes. Experimental neutron absorbed dose measurements have been carried out. It was found that GS4 glass has the best neutron protection capacity among the produced glasses. In addition to neutron shielding properties, the gamma-ray attenuation capacities, were calculated using newly developed Phy-X/PSD software. The gamma-ray shielding properties of GS1 and GS2 are found to be equivalent to Pb-based glass.

scholarly journals Conceptual Shield Design for Boron Neutron Capture Therapy Facility Using Monte Carlo N-Particle Extended Simulator with Kartini Research Reactor as Neutron Source

2020 ◽  
Vol 35 (3) ◽  
pp. 177-181
Author(s):  
Afifah Hana Tsurayya ◽  
Azzam Zukhrofani Iman ◽  
R. Yosi Aprian Sari ◽  
Arief Fauzi ◽  
Gede Sutresna Wijaya
Keyword(s):  
Monte Carlo ◽  
Radiation Dose ◽  
Dose Rate ◽  
Neutron Source ◽  
Gamma Ray ◽  
Outer Part ◽  
Radiation Shielding ◽  
Radiation Dose Rate ◽  
Boron Neutron Capture ◽  
Radiation Workers

The research aims to measure the radiation dose rate over the radiation shielding which is made of paraffin and aluminium and to determine the best shield material for the safety of radiation workers. The examination used MCNP (Monte Carlo N-Particle) simulator to model the BNCT neutron source and the shield. The shield should reduce radiation to less than the dose limit of 10.42 µSv/h, which is assumed to be the most conservative limit when the duration of workers is 1920 h. The first design resulted in a radiation dose rate which was still greater than the limit. Therefore, optimization was done by adding the lead on the outer part of the shield. After optimization by adding the lead with certain layers, the radiation dose rate decreased, with the largest dose being 57.60 µSv/h. Some locations over the limit could be overcome by other radiation protection aspects such as distance and time. The paraffin blocks were covered by aluminium to keep the shield structure. The lead was used to absorb the gamma ray which resulted from the interaction between the neutrons and aluminium.

scholarly journals Design and Optimisation of a Three Layers Thermal Neutron, Fast Neutron and Gamma-Ray Imaging System

2020 ◽  
Vol 225 ◽  
pp. 07002
Author(s):  
H. Al Hamrashdi ◽  
S. D. Monk ◽  
D. Cheneler
Keyword(s):  
Spatial Resolution ◽  
Gamma Ray ◽  
Imaging System ◽  
Detection System ◽  
Fast Neutrons ◽  
Silicon Photomultiplier ◽  
Current Configuration ◽  
Lithium Glass ◽  
Gamma Ray Imaging ◽  
Small Form Factor

The design and configuration of a multi-layered imaging system with the ability to detect thermal neutrons, fast neutrons and gamma rays has been developed and its efficacy demonstrated. The work presented here numerically determines the systems efficiency and spatial resolution, using 252Cf and 137Cs as a case study. The novelty of this detection system lies in the use of small form factor detectors in a three-layer design, which utilises neutron elastic scattering and Compton scattering simultaneously. The current configuration consists of 10 mm thick natural lithium glass (GS10) scintillator integrated with a 20 mm thick plastic scintillator (EJ-204) in the first layer, a 15 mm thick lithium glass (GS10) scintillator in the second and a 30 mm thick CsI(Tl) scintillator forming the final layer. Each of these layers is backed with an 8 x 8 silicon photomultiplier diode (SiPM) array. The overall size of the imaging system is 27 mm x 27 mm x 135 mm. MCNPv6.1 and Geant4-10.04 were alternatively used to optimise the overall configuration and to investigate detection modalities. Results show promising performance with high precision source localisation and characterization abilities. Measurements were virtually obtained of two gamma-ray sources within steel enclosures at angles of 15°, 30° and 50° separation in order to test spatial resolution ability of the system. With the current active size of the system and the 8x8 SiPM configuration, the results estimate the spatial resolution to be close to 30°. The ability of the system to characterise and identify sources based on the type and energy of the radiation emitted, has been investigated and results show that for all radiation types the system can identify the source energy within the energy range of typical reported sources in literature.

Structure, optical, gamma-ray and neutron shielding properties of NiO doped B2O3–BaCO3–Li2O3 glass systems

2020 ◽  
Vol 46 (2) ◽  
pp. 1711-1721 ◽  
Author(s):  
M.S. Al-Buriahi ◽  
A.S. Abouhaswa ◽  
H.O. Tekin ◽  
C. Sriwunkum ◽  
F.I. El-Agawany ◽  
...  
Keyword(s):  
Gamma Ray ◽  
Neutron Shielding ◽  
Shielding Properties
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