Development of Al2O3-Bi2O3-B2O3 Glasses for Neutron Shielding Material

2016 ◽  
Vol 675-676 ◽  
pp. 430-433
Author(s):  
Suparat Tuscharoen ◽  
Smit Insiripong ◽  
T. Korkut ◽  
Jakrapong Kaewkhao
Keyword(s):  
Monte Carlo ◽  
Physical Properties ◽  
Chemical Formula ◽  
Melt Quenching ◽  
Neutron Shielding ◽  
Equivalent Dose Rate ◽  
Monte Carlo Techniques ◽  
Positive Effects ◽  
Shielding Properties ◽  
Shielding Material

A glass system with chemical formula xB2O3:20Bi2O3:(100-x) Al2O3 (x = 55, 60, 65, 70, 75 and 80 mol%) is prepared by melt quenching technique and were investigated the physical and neutron shielding properties. The physical properties were investigated by density, molar volume and discussed with different Al2O3 contents. The neutron shielding property was investigated by Monte Carlo techniques (FLUKA and GEANT4 codes) and neutron equivalent dose rate measurements. As a result, neutron shielding capacity of glass samples decrease with increased Al2O3 content, so increased B2O3 content is a result of positive effects on neutron shielding.

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.

Neutron shielding material design based on Monte Carlo simulation

2012 ◽  
Vol 24 (12) ◽  
pp. 3006-3010 ◽  
Author(s):  
陈飞达 Chen Feida ◽  
汤晓斌 Tang Xiaobin ◽  
王鹏 Wang Peng ◽  
陈达 Chen Da
Keyword(s):  
Monte Carlo Simulation ◽  
Monte Carlo ◽  
Material Design ◽  
Neutron Shielding ◽  
Shielding Material

Preparation, shielding properties and mechanism of a novel neutron shielding material made from natural Szaibelyite resource

2018 ◽  
Vol 106 ◽  
pp. 140-145 ◽  
Author(s):  
Mengge Dong ◽  
Xiangxin Xue ◽  
Zhefu Li ◽  
He Yang ◽  
M.I. Sayyed ◽  
...  
Keyword(s):  
Neutron Shielding ◽  
Shielding Properties ◽  
Shielding Material

Radiation Shielding Properties of Bi2O3-Na2O-B2O3 of 662 keV

2016 ◽  
Vol 702 ◽  
pp. 77-82
Author(s):  
Wasu Cheewasukhanont ◽  
Kitipun Boonin ◽  
Pruittipol Limkitjaroenporn ◽  
Jakrapong Kaewkhao
Keyword(s):  
Physical Properties ◽  
Radiation Source ◽  
Glass Sample ◽  
Mass Attenuation Coefficient ◽  
Radiation Shielding ◽  
Sodium Borate ◽  
Borate Glasses ◽  
Melt Quenching ◽  
Shielding Properties ◽  
Bismuth Sodium

The bismuth sodium borate glasses in formula xBi2O3:20Na2O:(80-x)B2O3 (x = 0, 5, 10, 15, 20, 25, 30, 35, 40, 45, and 50 mol %) have been prepared by melt-quenching technique. The glass sample at x=0 mol% show the colorless and show the stronger yellow at the higher Bi2O3 content. The glasses’s colors are corespondening with the cutoff wavelength’s results. The shielding properties were measured at 662 keV by Cs-137 radiation source. The obtained results show that the mass attenuation coefficient increased with the increasing of Bi2O3 concentration. Half Value Layer (HVL) of glasses were determined and compared with the some standard shielding materials. The optical and physical properties were also investigated.

Investigation of neutron shielding properties depending on number of boron atoms for colemanite, ulexite and tincal ores by experiments and FLUKA Monte Carlo simulations

2012 ◽  
Vol 70 (1) ◽  
pp. 341-345 ◽  
Author(s):  
Turgay Korkut ◽  
Abdulhalik Karabulut ◽  
Gökhan Budak ◽  
Bünyamin Aygün ◽  
Osman Gencel ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Simulations ◽  
Neutron Shielding ◽  
Shielding Properties

Preparation of an Ultra-Thin, Highly Filled, Neutron-Shielding Material

2015 ◽  
Vol 815 ◽  
pp. 616-621
Author(s):  
Xin Peng Hou ◽  
Shu Feng Li ◽  
Yan Chen ◽  
Di Xu ◽  
Wei Tang
Keyword(s):  
Material Properties ◽  
Chemical Properties ◽  
Neutron Shielding ◽  
Base Materials ◽  
Functional Additives ◽  
Physical Chemical ◽  
Radiation Crosslinking ◽  
Ethylene Octene Copolymer ◽  
Shielding Properties ◽  
Shielding Material

We develop an ultra-thin, highly filled, neutron-shielding material. This material exhibits a desirable neutron-shielding performance, and also has certain advantageous mechanical properties and uses. We study the physical properties of shielding materials with different polyolefins as base materials, and investigate the neutron-shielding performance of boron-containing and lithium-containing shielding materials. We furthermore report on the effect of additive amounts of functional additives on shielding properties and physical-chemical properties. We additionally study the effect of radiation crosslinking technology on shielding material properties. We show that, using ethylene-octene copolymer (POE) modified low-density polyethylene (PE-LD), the additive amounts of boron carbide (B4C) and nano-montmorillonite (OMMT) are 60–70% and 4%, respectively. The optimal radiation dose is 160 kGy, and the shielding materials exhibit good neutron-shielding performance and mechanical strength.

Bayesian Estimation of DSGE Models

2015 ◽  
Author(s):  
Edward P. Herbst ◽  
Frank Schorfheide
Keyword(s):  
Monte Carlo ◽  
Sequential Monte Carlo ◽  
Likelihood Function ◽  
Academic Research ◽  
Dynamic Stochastic General Equilibrium ◽  
Computational Techniques ◽  
Dsge Models ◽  
Monte Carlo Techniques ◽  
Dynamic Stochastic ◽  
Theoretical Foundations

Dynamic stochastic general equilibrium (DSGE) models have become one of the workhorses of modern macroeconomics and are extensively used for academic research as well as forecasting and policy analysis at central banks. This book introduces readers to state-of-the-art computational techniques used in the Bayesian analysis of DSGE models. The book covers Markov chain Monte Carlo techniques for linearized DSGE models, novel sequential Monte Carlo methods that can be used for parameter inference, and the estimation of nonlinear DSGE models based on particle filter approximations of the likelihood function. The theoretical foundations of the algorithms are discussed in depth, and detailed empirical applications and numerical illustrations are provided. The book also gives invaluable advice on how to tailor these algorithms to specific applications and assess the accuracy and reliability of the computations. The book is essential reading for graduate students, academic researchers, and practitioners at policy institutions.

Neutron dose evaluation of Elekta Linac at two energies (10 & 18 MV) by MCNP code and comparison with experimental measurements

2014 ◽  
Vol 6 (1) ◽  
pp. 1006-1015
Author(s):  
Negin Shagholi ◽  
Hassan Ali ◽  
Mahdi Sadeghi ◽  
Arjang Shahvar ◽  
Hoda Darestani ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Measurement Data ◽  
Calculated Data ◽  
High Energy ◽  
Neutron Dose ◽  
Dose Assessment ◽  
Photon Flux ◽  
Dose Equivalent ◽  
Monte Carlo Techniques ◽  
Neutron Dose Equivalent

Medical linear accelerators, besides the clinically high energy electron and photon beams, produce other secondary particles such as neutrons which escalate the delivered dose. In this study the neutron dose at 10 and 18MV Elekta linac was obtained by using TLD600 and TLD700 as well as Monte Carlo simulation. For neutron dose assessment in 2020 cm2 field, TLDs were calibrated at first. Gamma calibration was performed with 10 and 18 MV linac and neutron calibration was done with 241Am-Be neutron source. For simulation, MCNPX code was used then calculated neutron dose equivalent was compared with measurement data. Neutron dose equivalent at 18 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 3.3, 4, 5 and 6 cm. Neutron dose at depths of less than 3.3cm was zero and maximized at the depth of 4 cm (44.39 mSvGy-1), whereas calculation resulted  in the maximum of 2.32 mSvGy-1 at the same depth. Neutron dose at 10 MV was measured by using TLDs on the phantom surface and depths of 1, 2, 2.5, 3.3, 4 and 5 cm. No photoneutron dose was observed at depths of less than 3.3cm and the maximum was at 4cm equal to 5.44mSvGy-1, however, the calculated data showed the maximum of 0.077mSvGy-1 at the same depth. The comparison between measured photo neutron dose and calculated data along the beam axis in different depths, shows that the measurement data were much more than the calculated data, so it seems that TLD600 and TLD700 pairs are not suitable dosimeters for neutron dosimetry in linac central axis due to high photon flux, whereas MCNPX Monte Carlo techniques still remain a valuable tool for photonuclear dose studies.

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