Verification of multi-group cross-section generation code ARES-MACXS with fusion neutron shielding benchmark OKTAVIAN-Fe

2022 ◽  
Vol 166 ◽  
pp. 108739
Author(s):  
Jiaju Hu ◽  
Bin Zhang ◽  
Cong Liu ◽  
Ni Dai ◽  
Yixue Chen
Keyword(s):  
Cross Section ◽  
Neutron Shielding ◽  
Fusion Neutron

scholarly journals Research on Thermal Neutron Shielding Performance of TiB2-Al Composite Materials

2021 ◽  
Author(s):  
Chao Wang ◽  
Zhefu Li ◽  
Mengge Dong ◽  
Lu Zhang ◽  
Jianxing Liu ◽  
...  
Keyword(s):  
Thermal Neutron ◽  
Cross Section ◽  
Neutron Fluence ◽  
Thermal Neutrons ◽  
Neutron Shielding ◽  
Preparation Methods ◽  
Neutron Fluence Rate ◽  
Maximum Value ◽  
Al Composite ◽  
Macroscopic Cross Section

<p>Although the various excellent properties and preparation methods of TiB<sub>2</sub>-based composites have been extensively studied, their neutron shielding properties have not received as much attention. In this article, the neutron shielding performance of the previously prepared TiB<sub>2</sub>-Al composite will be studied. The photo neutron source device was used to carry out neutron irradiation tests on test samples with a thickness of 10 mm. The average thermal neutron shielding rate of TiB<sub>2</sub>-based boron-containing composites is 17.55%, and the shielding rate increases with the increase of BN content. The macroscopic cross-section of thermal neutrons of the composites generally shows a stable trend, and when the BN content is 10%, the thermal neutrons macroscopic cross section reaches the maximum value of 7.58cm<sup>-1</sup>. With the increase of the BN content, the thermal neutron fluence rate shows a gradually decreasing trend.</p>

Study on the Properties of Boron-Containing Ores/Epoxy Composites for Slow Neutron Shielding

2011 ◽  
Vol 201-203 ◽  
pp. 2767-2771 ◽  
Author(s):  
Zhe Fu Li ◽  
Xiang Xin Xue ◽  
Tao Jiang ◽  
He Yang ◽  
Mi Zhou
Keyword(s):  
Cross Section ◽  
Iron Ore ◽  
Low Cost ◽  
Slow Neutron ◽  
Linear Attenuation Coefficient ◽  
Neutron Shielding ◽  
Neutron Absorber ◽  
Macroscopic Cross Section ◽  
Γ Ray ◽  
Iron Ore Concentrate

Three kinds of low cost shielding materials using boron-containing ores and epoxy as starting materials were developed. The three boron-containing ores consist of natural green ore that is specific in China and the two artificial minerals that are boron-containing iron ore concentrate and boron-rich slag. They were got after natural green ore dressing and separation from blast furnace, respectively. The shielding performance of slow neutron and 60Co γ ray among the three shielding materials were compared. The results show that boron-rich slag as neutron absorber performs relative good slow neutron shielding ability, its macroscopic cross section (Σ) is 0.271cm-1 which is about 1.72 times higher than Lead-Boron Polyethylene B201 whose macroscopic cross section (Σ) is 0.158cm-1. Boron-containing iron ore concentrate/epoxy composite performs relative good 60Co γ ray shielding ability and its linear attenuation coefficient (μ) is 0.0772cm-1. The elements boron and iron that exist as compound in ore are the main reasons to affect the slow neutron shielding performance. The iron element that exists as compound in the ore has the main attribution to 60Co γ ray shielding.

scholarly journals Investigation of fast neutron shielding properties of new polyurethane-based composites loaded with B4C, BeO, WO3, ZnO, and Gd2O3 micro-and nanoparticles

2019 ◽  
Vol 25 (4) ◽  
pp. 211-219 ◽  
Author(s):  
Asghar Mesbahi ◽  
Khatibeh Verdipoor ◽  
Farhad Zolfagharpour ◽  
Abdolali Alemi
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Neutron Cross Section ◽  
Radiation Shielding ◽  
Fast Neutrons ◽  
Nano Composites ◽  
Monte Carlo Code ◽  
Neutron Shielding ◽  
Shielding Properties ◽  
Polyurethane Matrix

Abstract The aim of the current research was to study the radiation shielding properties of polyurethane-based shielding materials filled with B4C, BeO, WO3, ZnO, and Gd2O3 particles against fast neutrons. The macroscopic cross sections of composites containing micro- and nanoparticles with a diameter of 10 µm and 50 nm were calculated using MCNPX (2.6.0) Monte Carlo code. The results showed that adding nano-scaled fillers to polyurethane matrix increases attenuation properties of neutron shields compared to micro-scaled fillers for intermediate and fast neutrons. Among the studied composites, WO3 and Gd2O3 nano-composites presented higher neutron cross section compared to others.

scholarly journals Research on Thermal Neutron Shielding Performance of TiB2-Al Composite Materials

2021 ◽  
Author(s):  
Chao Wang ◽  
Zhefu Li ◽  
Mengge Dong ◽  
Lu Zhang ◽  
Jianxing Liu ◽  
...  
Keyword(s):  
Thermal Neutron ◽  
Cross Section ◽  
Neutron Fluence ◽  
Thermal Neutrons ◽  
Neutron Shielding ◽  
Preparation Methods ◽  
Neutron Fluence Rate ◽  
Maximum Value ◽  
Al Composite ◽  
Macroscopic Cross Section

<p>Although the various excellent properties and preparation methods of TiB<sub>2</sub>-based composites have been extensively studied, their neutron shielding properties have not received as much attention. In this article, the neutron shielding performance of the previously prepared TiB<sub>2</sub>-Al composite will be studied. The photo neutron source device was used to carry out neutron irradiation tests on test samples with a thickness of 10 mm. The average thermal neutron shielding rate of TiB<sub>2</sub>-based boron-containing composites is 17.55%, and the shielding rate increases with the increase of BN content. The macroscopic cross-section of thermal neutrons of the composites generally shows a stable trend, and when the BN content is 10%, the thermal neutrons macroscopic cross section reaches the maximum value of 7.58cm<sup>-1</sup>. With the increase of the BN content, the thermal neutron fluence rate shows a gradually decreasing trend.</p>

The Testing and Application of Continuous-Energy Neutron Cross Section Library HENDL-ADS/MC

2013 ◽  
Author(s):  
Chong Chen ◽  
Jun Zou ◽  
Dezheng Xu ◽  
Qin Zeng ◽  
Minghuang Wang
Keyword(s):  
Cross Section ◽  
Neutron Energy ◽  
Fission Products ◽  
Neutron Cross Section ◽  
Nuclear Data ◽  
Neutron Shielding ◽  
Section Data ◽  
Nuclear Data Library ◽  
Data Library ◽  
High Neutron

A point-wise cross-section data library HENDL-ADS/MC (Hybrid Evaluated Nuclear Data Library) has been produced by FDS team to do the nuclear analysis for the ADS system. The HENDL-ADS/MC library contained 408 nuclide cross-section files including actinides, fission products and structural materials for neutron energy up to 150 MeV. The nuclear library also contained several sub-libraries with different temperatures. A series of neutron integral experiments and critical safety benchmarks have been performed to test the availability and reliability of the HENDL-ADS/MC data library. To validate and qualify the reliability of the high neutron energy cross section for HENDL-ADS/MC library further, a series of high neutron shielding experiments have been performed using MCNP. The testing results indicated the accuracy and reliability of HENDL-ADS/MC library.

Preparation and Thermal/Fast Neutron Shielding Properties of Novel Boron Containing Ore Composites

2013 ◽  
Vol 743-744 ◽  
pp. 613-622 ◽  
Author(s):  
Zhe Fu Li ◽  
Xiang Xin Xue ◽  
Pei Ning Duan ◽  
Ang Tian ◽  
Shu Xian Li ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Atomic Number ◽  
Absorbed Dose ◽  
Neutron Shielding ◽  
Factors Affecting ◽  
Average Atomic Number ◽  
Shielding Properties ◽  
Relationship Of ◽  
The Relationship

Four novel boron containing ores/epoxy composites were prepared by using nature ludwigite green ore (NLGO), artificial ores including boron containing iron ore concentrate (BCIOC), boron rich slag (BRS) and boron mud (BM) as neutron absorbers, which were obtained by dressing from NLGO, blast furnace separation from BCIOC and borax production, respectively. The microstructure of the composites was investigated by X-ray diffraction (XRD), scanning electron microscope (SEM) and Fourier transform infrared spectroscopy (FT-IR). Shielding properties of the composites against thermal and Cf-252 fast neutron were measured. Factors affecting the shielding properties were investigated. Energy deposition, absorbed dose and half value layers of the composites against neutron penetration were simulated by Monte Carlo method, and a simulated function was obtained. The results showed that the composites were compact and the particles of ores were homogeneous. Ascending boron mole numbers per unit volume (nB) in the composites can obviously enhance the macroscopic absorbing cross section (Σthermal) and improve the shielding properties for thermal neutron. The relationship of Σthermal and nB follows the equation of Σthermal=0.218+450.490nB. And the relationship of average atomic number (Z) of boron containing ores composites and the macroscopic removal cross section for fast neutron follows the equation of Σremoval=0.042exp (-Z/5.70)+0.032. Enhancement of the shielding abilities against Cf-252 fast neutron can be achieved by using the composites with low average atomic number.

Benzoxazine resin as an interesting building block for advanced neutrons shields

2021 ◽  
pp. 095400832110218
Author(s):  
Oussama Mehelli ◽  
Mehdi Derradji ◽  
Abdelmalek Habes ◽  
Nour Elislem Leblalta ◽  
Raouf Belgacemi ◽  
...  
Keyword(s):  
Cross Section ◽  
Research Reactor ◽  
Building Blocks ◽  
Mean Free Path ◽  
Sample Thickness ◽  
Experimental Setup ◽  
Reinforced Composites ◽  
Neutron Shielding ◽  
Recent Developments ◽  
Macroscopic Cross Section

The design of lightweight neutrons shields has been restricted for quite some time to the use of the epoxy thermosets as the main building blocks. Meanwhile, the recent developments in the field of polymers suggest otherwise. Indeed, benzoxazine resins have taken the lead over the traditional thermosets in many exigent applications. Therefore, in a vision to introduce newer matrices with better performances and to further expand the applications of the benzoxazine resins into the nuclear field, the neutron shielding efficiency along with the thermal and thermomechanical performances of the neat benzoxazine polymer and its subsequent B4C-reinforced composites were investigated. The neutron shielding measurements were performed using an optimized experimental setup at NUR research reactor, Algiers. The neat benzoxazine polymer displayed almost similar thermal neutrons screening performances than the epoxy with a macroscopic cross-section (Σ) of a 0.724 cm− 1 equivalent to a mean free path (λ) of 0.957 cm. The effect of the particle amount was also studied to maximize the shielding ability of the developed materials. For instance, the benzoxazine composite containing 20 wt.% of B4C displayed the outstanding screening ratio of about 96% for a sample thickness of 13 mm. Finally, the remarkable findings were put into context by providing multifaceted comparisons with the available shielding materials.

scholarly journals Mechanical and Gamma Ray Absorption Behavior of PbO-WO3-Na2O-MgO-B2O3 Glasses in the Low Energy Range

Materials ◽  
2021 ◽  
Vol 14 (13) ◽  
pp. 3466
Author(s):  
Aljawhara H. Almuqrin ◽  
Badriah Albarzan ◽  
O. I. Olarinoye ◽  
Ashok Kumar ◽  
Norah Alwadai ◽  
...  
Keyword(s):  
Fast Neutron ◽  
Cross Section ◽  
Unit Volume ◽  
Gamma Ray ◽  
Glass System ◽  
Gamma Photon ◽  
Neutron Shielding ◽  
Removal Cross Section ◽  
Good Potential ◽  
B2o3 Glass

The Makishima and Mackenzie model has been used to determine the mechanical properties of the PbO-WO3-Na2O-MgO-B2O3 glass system. The number of bonds per unit volume of the glasses (nb) increases from 9.40 × 1022 to 10.09 × 1022 cm−3 as the PbO content increases from 30 to 50 mol%. The Poisson’s ratio (σ) for the examined glasses falls between 0.174 and 0.210. The value of the fractal bond connectivity (d) for the present glasses ranges from 3.08 to 3.59. Gamma photon and fast neutron shielding parameters were evaluated via Phy-X/PSD, while that of electrons were calculated via the ESTAR platform. Analysis of the parameters showed that both photon and electron attenuation ability improve with the PbO content. The fast neutron removal cross section of the glasses varies from 0.094–0.102 cm−1 as PbO molar content reduced from 50–30 mol%. Further analysis of shielding parameters of the investigated glass system showed that they possess good potential to function in radiation protection applications.

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