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 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>

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>

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.

Effect of Boron Carbide Content on Properties of Thermoplastic Natural Rubber Composite as Thermal Neutron Shielding

2017 ◽  
Vol 888 ◽  
pp. 179-183
Author(s):  
Nurazila Mat Zali ◽  
Hafizal Yazid ◽  
Megat Harun Al Rashid Megat Ahmad ◽  
Irman Abdul Rahman ◽  
Yusof Abdullah
Keyword(s):  
Tensile Strength ◽  
Natural Rubber ◽  
Boron Carbide ◽  
Thermal Neutron ◽  
Cross Section ◽  
Weight Percent ◽  
Neutron Shielding ◽  
Blending Method ◽  
Macroscopic Cross Section ◽  
Thermoplastic Natural Rubber

In this work, thermoplastic natural rubber (TPNR) composites were produced through melt blending method. Boron carbide (B4C) as filler was added into the polymer blend (TPNR) with different weight percent from 0% to 30% and the effect of different B4C contents on mechanical and thermal neutron attenuation properties of TPNR composites has been studied. The phase formation in composites was analyzed using XRD technique. From the results, it showed that the incorporation of B4C fillers into TPNR matrix has enhanced the macroscopic cross section of the composites, however it lessens the tensile strength. Macroscopic cross section of the composites were increased from 3.34 cm-1 to 14.8 cm-1, while the tensile strength of the composites decreased from 3.79 MPa to 1.06 MPa with increasing B4C from 0 wt% to 30 wt%. B4C diffraction peaks were also increased in intensity with increasing B4C content.

γ-RAYS PRODUCED BY SLOW NEUTRON CAPTURE IN BERYLLIUM, CARBON, AND NITROGEN

1951 ◽  
Vol 29 (1) ◽  
pp. 1-13 ◽  
Author(s):  
B. B. Kinsey ◽  
G. A. Bartholomew ◽  
W. H. Walker
Keyword(s):  
Excited States ◽  
Cross Section ◽  
Neutron Capture ◽  
Radiative Capture ◽  
Slow Neutron ◽  
Binding Energies ◽  
Carbon And Nitrogen ◽  
Radiative Capture Cross Section ◽  
Γ Rays ◽  
Γ Ray

The γ-rays produced by slow neutron capture in beryllium, carbon, and nitrogen have been investigated with a pair spectrometer. Only one γ-ray was observed to follow capture in beryllium and in carbon; nine were observed from nitrogen. The neutron binding energies computed from the energies of these radiations are: Be10, 6.797 ± 0.008 Mev.; C13, 4.948 ± 0.008 Mev.; N15, 10.823 ± 0.012 Mev. The nitrogen γ-rays can be fitted into a term scheme containing the previously known excited states of N15, obtained from the study of (d, p) reactions, together with one excited state at 9.16 Mev., hitherto unreported. A comparison of the intensities of the nitrogen and beryllium radiations shows that the radiative capture cross section of nitrogen is at most 160 millibarns.

CLEANING UP EMERGENCY OIL SPILLS FROM THE WATER SURFACE WITH MAGNETIC ADSORBENTS

2017 ◽  
Author(s):  
Yury Rubanov ◽  
Yury Rubanov ◽  
Yulia Tokach ◽  
Yulia Tokach ◽  
Marina Vasilenko ◽  
...  
Keyword(s):  
Oil Recovery ◽  
Water Surface ◽  
Iron Ore ◽  
Oil Spills ◽  
Air Cooling ◽  
Steelmaking Slag ◽  
Magnetic Adsorbents ◽  
Ratio Of Components ◽  
Iron Ore Concentrate ◽  
Electric Furnace Steelmaking

There was suggested a method of obtaining a complex adsorbent with magnetic properties for the oil spill clean-up from the water surface by means of controlled magnetic field. As magnetic filler a finely-dispersed iron-ore concentrate in the form of magnetite, obtained by wet magnetic separation of crushed iron ore, was suggested. As an adsorbing component the disintegrating electric-furnace steelmaking slag, obtained by dry air-cooling method, was selected. The mass ratio of components slag:magnetite is 1(1,5÷2,0). For cleaning up emergency oil spills with the suggested magnetic adsorbent a facility, which is installed on a twin-hulled oil recovery vessel, was designed. The vessel contains a rectangular case between the vessel hulls with inlet and outlet for the treated water, the bottom of which is a permanently moving belt. Above the belt, at the end point of it there is an oil-gathering drum with magnetic system. The adsorbent is poured to oil-products layer from a hopper, provided with drum feeder. Due to the increased bulk weight the adsorbent sinks rapidly into the oil layer on the water surface. If the large non-floating flocculi are formed, they sink and sedimentate on the moving belt and are moved to the oil-gathering drum. The saturated adsorbent is removed from the drum surface with a scraper, connected with a gutter, with contains a rotating auger.

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.

Mechanism on reduction swelling of pellets prepared from Bayan Obo iron ore concentrate

2021 ◽  
pp. 1-11
Author(s):  
Yingjie Fan ◽  
Yunhao Zhang ◽  
Zhichao Li ◽  
Yifan Chai ◽  
Yici Wang ◽  
...  
Keyword(s):  
Iron Ore ◽  
Reduction Swelling ◽  
Bayan Obo ◽  
Iron Ore Concentrate
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