Review on the addition of boron compounds to radiation shielding concrete

Highly Efficient and Eco-Friendly Thermal-Neutron-Shielding Materials Based on Recycled High-Density Polyethylene and Gadolinium Oxide Composites

Due to the increasing demands for improved radiation safety and the growing concerns regarding the excessive use of plastics, this work aimed to develop effective and eco-friendly thermal-neutron-shielding materials based on recycled high-density polyethylene (r-HDPE) composites containing varying surface-treated gadolinium oxide (Gd2O3) contents (0, 5, 10, 15, and 20 wt%). The results indicate that the overall thermal-neutron-shielding properties of the r-HDPE composites were enhanced with the addition of Gd2O3, as evidenced by large reductions in I/I0, HVL, and TVL, as well as the substantial increases in ∑t and ∑t/ρ of the composites. Furthermore, the results reveal that the values for tensile properties initially increased up to 5-15 wt% of Gd2O3 and then gradually decreased at higher contents. In addition, the results show that the addition of Gd2O3 particles generally increased the density (ρ), the remaining ash at 600 °C, and the degree of crystallinity (%XC) of the composites. This work also determined the effects of gamma irradiation on relevant properties of the composites. The findings indicate that following gamma aging, the tensile modulus slightly increased, while the tensile strength, elongation at break, and hardness (Shore D) showed no significant (p < 0.05) differences, except for the sample containing 5 wt% of Gd2O3, which exhibited a noticeable reduction in elongation at break. Furthermore, by comparing the neutron-shielding and mechanical properties of the developed r-HDPE composites with common borated polyethylene (PE) containing 5 wt% and 15 wt% of boron, the results clearly indicate the superior shielding and tensile properties in the r-HDPE composites, implying the great potential of r-HDPE composites to replace virgin plastics as effective and more eco-friendly shielding materials.

Liquid-liquid extraction of boron from continental brines by 2-butyl-1-octanol diluted in kerosene

Lithium production from brines generates significant quantities of salts, including boron, that are not effectively utilized and end up being stored in landfills. This study delves into a novel approach for directly extracting boron from native brines without performing solar evaporation as an alternative to traditional methods based on boron extraction from ores, offering a sustainable route to producing boric acid or borax. By exploring factors such as 2-butyl-1-octanol concentration, phase volume ratio, temperature, and pH, the research scrutinizes boron extraction efficiency from two native brines sourced from the salar de Hombre Muerto in Argentina, alongside a synthetic brine simulating these native compositions. Notably, the extractant demonstrates exceptional promise due to its limited solubility in the brine, measuring at just 18 mg L-1. Optimal conditions-2 mol L-1 2-butyl-1-octanol, O/A ratio of 4, 25 °C temperature, and pH of 5.5-resulted in a remarkable 98.2% and 94.2% recovery of boron from synthetic and native brines, respectively. Importantly, this extraction process showcased minimal co-extraction of lithium, calcium, magnesium, potassium, and sodium. Leveraging these findings, a proposed flowsheet outlines a highly selective method for extracting boron from brines, presenting an alternative avenue to conventional borax production from boron ores.

Gamma radiation shielding properties for polymer composites reinforced with bismuth tungstate in different proportions

In this study, inorganic compound (Bi2(WO4)3) was added into the composite to improve the radiation shielding properties of polymer composite. A polymer matrix was prepared by combining unsaturated polyester resin with methyl ethyl ketone peroxide and cobalt octoate (6%), and Bi2(WO4)3 was added to this polymer matrix at different ratios as filling material. In order to investigate the gamma radiation attenuation properties of the obtained polymer composites, mass attenuation coefficients, radiation shielding efficiencies, radiation transmission factors, linear attenuation coefficients, half values layer, tenth values layer, mean free path values, effective atomic numbers and effective electron densities parameters were obtained. Experimental studies were carried out with the help of HPGe detector at 22 different energies emitted from 22Na, 54Mn, 57Co, 60Co, 133Ba, 137Cs, 152Eu and 241Am radioactive sources in the photon energy range of 59.5-1408.0 keV. Each obtained experimental result was compared with the theoretical results. It was observed that the sample encoded with BiWO20 is the best radiation shielding material among all studied composites (except 59.5 keV).

Transition Metal Borides for All-in-One Radiation Shielding

All-in-one radiation shielding is an emerging concept in developing new-generation radiation protection materials since various forms of ionizing radiation, such as neutrons and gamma rays, can occur simultaneously. In this study, we examine the ability of transition metal borides to attenuate both photon and particle radiation. Specifically, fourteen different transition metal borides (including inner transition metal borides) are selected for examination based on their thermodynamic stabilities, molecular weights, and neutron capture cross-sections of the elements they contain. Radiation shielding characteristics of the transition metal borides are computationally investigated using Phy-X/PSD, EpiXS and NGCal software. The gamma-ray shielding capabilities of the transition metal borides are evaluated in terms of the mass attenuation coefficient (μm), the linear attenuation coefficient (µ), the effective atomic number (Zeff), the half-value layer (HVL), the tenth-value layer (TVL), and the mean free path (MFP). The mass and linear attenuation factors are identified for thermal and fast neutrons at energies of 0.025 eV and 4 MeV, respectively. Moreover, the fast neutron removal cross-sections (∑R) of the transition metal borides are calculated to assess their neutron shielding abilities. The results revealed that borides of transition metals with a high atomic number, such as Re, W, and Ta, possess outstanding gamma shielding performance. At 4 MeV photon energy, the half-value layers of ReB2 and WB2 compounds were found as 1.38 cm and 1.43 cm, respectively. Most notably, these HVL values are lower than the HVL value of toxic Pb (1.45 cm at 4 MeV), which is one of the conventional radiation shielding materials. On the other hand, SmB6 and DyB6 demonstrated exceptional neutron attenuation for thermal and fast neutrons due to the high neutron capture cross-sections of Sm, Dy, and B. The outcomes of this study reveal that transition metal borides can be suitable candidates for shielding against mixed neutron and gamma radiation.

A comparative neutron and gamma-ray radiation shielding investigation of molybdenum and boron filled polymer composites

This work presents a detailed radiation shielding study for polymer composites filled with Boron and Molybdenum additives. The chosen novel polymer composites were produced at different percentages of the additive materials to provide a proper evaluation of their neutron and gamma-ray attenuation abilities. The effect of additive particle size on the shielding characteristics was further investigated. On the gamma-ray side, simulation, theoretical and experimental evaluations were performed in a wide range of photon energies varying from 59.5 keV to 1332.5 keV with help of MC simulations (GEANT4 and FLUKA), WinXCOM code, a High Purity Germanium Detector, respectively. A remarkable consistency was reported between them. On the neutron shielding side, the prepared samples produced with nano and micron particle size additives were additionally examined by providing fast neutron removal cross-section (Σ_R) and the simulated neutron transmissions through the prepared samples. The samples filled with nano sized particles show better shielding capability than the one filled with micron sized particles. In other words, a new polymer shielding material that does not contain toxic content is introduced: the sample codded N-B0Mo50 exhibits superior radiation attenuation.

A detailed investigation of gamma and neutron shielding capabilities of ternary composites doped with polyacrylonitrile and gadolinium (III) sulfate

The shielding efficiencies of gamma and neutron radiations for ternary composites containing polyester resin, polyacrylonitrile and gadolinium (III) sulfate at different ratios were investigated in the present study. In order to investigate the gamma radiation shielding capacity of the produced ternary composites, linear and mass attenuation coefficients, half value layer, effective atomic number and radiation protection efficiency parameters were determined experimentally, theoretically and using the GEANT4 simulation code. The gamma shielding capabilities of the composites were studied in the photon energy range of 59.5-1332.5 keV. In order to investigate the neutron shielding abilities of composites, inelastic, elastic, capture and transport numbers, total macroscopic cross section and mean free path parameters were determined with the help of GEANT4 simulation code. In addition, the number of transmitted neutrons at different sample thicknesses and neutron energies were also determined. It was observed that gamma radiation shielding properties were improved due to the increasing amount of gadolinium (III) sulfate and neutron shielding properties were improved due to the increasing amount of polyacrylonitrile. While the composite coded P0Gd50 exhibits a better gamma radiation shielding ability than the others, the neutron shielding of the sample coded P50Gd0 is also more favorable than the others.

Recent Trends in Advanced Radiation Shielding Concrete for Construction of Facilities: Materials and Properties

Nuclear energy offers a wide range of applications, which include power generation, X-ray imaging, and non-destructive tests, in many economic sectors. However, such applications come with the risk of harmful radiation, thereby requiring shielding to prevent harmful effects on the surrounding environment and users. Concrete has long been used as part of structures in nuclear power plants, X-ray imaging rooms, and radioactive storage. The direction of recent research is headed toward concrete’s ability in attenuating harmful energy radiated from nuclear sources through various alterations to its composition. Radiation shielding concrete (RSC) is a composite-based concrete that was developed in the last few years with heavy natural aggregates such as magnetite or barites. RSC is deemed a superior alternative to many types of traditional normal concrete in terms of shielding against the harmful radiation, and being economical and moldable. Given the merits of RSCs, this article presents a comprehensive review on the subject, considering the classifications, alternative materials, design additives, and type of heavy aggregates used. This literature review also provides critical reviews on RSC performance in terms of radiation shielding characteristics, mechanical strength, and durability. In addition, this work extensively reviews the trends of development research toward a broad understanding of the application possibilities of RSC as an advanced concrete product for producing a robust and green concrete composite for the construction of radiation shielding facilities as a better solution for protection from sources of radiation. Furthermore, this critical review provides a view of the progress made on RSCs and proposes avenues for future research on this hotspot research topic.

Tuning the 3D Printability and Thermomechanical Properties of Radiation Shields

Additive manufacturing, with its rapid advances in materials science, allows for researchers and companies to have the ability to create novel formulations and final parts that would have been difficult or near impossible to fabricate with traditional manufacturing methods. One such 3D printing technology, direct ink writing, is especially advantageous in fields requiring customizable parts with high amounts of functional fillers. Nuclear technology is a prime example of a field that necessitates new material design with regard to unique parts that also provide radiation shielding. Indeed, much effort has been focused on developing new rigid radiation shielding components, but DIW remains a less explored technology with a lot of potential for nuclear applications. In this study, DIW formulations that can behave as radiation shields were developed and were printed with varying amounts of porosity to tune the thermomechanical performance.

The fabrication and neutron shielding property of polyphenylene sulfide containing salicylic acid/Gd2O3 composites

A series of neutron absorbing materials with good neutron absorbing capacity, high strength and good thermal property were designed and prepared in this work. First of all, polyphenylene sulfide containing different mole content of salicylic acid (SAPPS) in the main chain was synthesized by nucleophilic substitution polymerization under high pressure. Then the composites with different content of nano Gd2O3 and modified PPS were prepared by melt blending method. The testing results indicated that the copolymers SAPPS were synthesized successfully, there was an interface interaction between nano Gd2O3 and the matrix without the need for surfactants or coupling agents. Additionally, the content of nano Gd2O3 had no obvious influence on the thermal property of the composites. While following with the increase of the content of nano Gd2O3, the tensile strength of the composites increased firstly and then decreased, when the content of nano Gd2O3 was 10 wt%, the tensile strength of 10%Gd2O3/5%SAPPS reached the maximum value of 74.9 MPa. The results of neutron shielding testing indicated that the content of nano-particles had a large effect on the neutron shielding rate of composites. The neutron shielding rate of 50%Gd2O3/5%SAPPS composite was up to 83%. All of these results indicated that the Gd2O3/5%SAPPS had potential to be applied to the high-temperature resistance and thermal shielding materials in nuclear and aerospace applications.

Neutron shielding calculations for neutron imaging facility at the Maâmora Triga reactor

The build of neutron imaging facility is planned in the tangential beam around the TRIGA 2 MW, located in the Mâamora Nuclear Research Centre of Morocco. The design of this equipment and the plans for all operations should be evaluated in advance for the adequacy of radiation protection. Cement and concrete has been widely used as shielding material in a nuclear reactor in order to minimize exposure to individuals. In this paper GEANT4 and FLUKA simulation toolkit were used to study neutron attenuation, and optimizing the layers of shielding material to minimize thickness. Materials being studied include different types of concrete such as Hematite and Hormirad mixed borax with different boron and iron compositions. Initial studies indicate that optimized shielding material is Hematite. A total thickness of 700 mm produces the required attenuation factor. Further studies may allow optimizing the shielding enclosures. Because of the low floor load (5000 kg/m²), one solution that could be adopted is the substitution one bloc shielding by two blocks (with a half thickness) with air gap between them. The model allows studying the performance of the shielding materials in this particular space arrangement. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation.

Effect of Borosilicate Glass Wastes and Synthetic Silica on Cement Products Properties

This paper presents the findings of research study on the effect of borosilicate glass wastes on properties of cement paste and mortar. The borosilicate glass contains three times less alkali than soda-lime glass and about 12 % of boron oxide, so pozzolanic activity of borosilicate glass is three times higher compared to soda-lime glass. In order to increase the pozzolanic activity of glass precipitated synthetic silica was used. Mathematical models were used in order to test the effect of synthetic silica on pozzolanic activity of borosilicate glass. Test results indicate that replacement of 1 % of borosilicate glass by synthetic silica increased the pozzolanic activity up to 9.4 mg CaO/g of additive. By adding 5 % of borosilicate glass instead of cement, the standard compressive strength is reduced by about 20 %. However, complex additive of borosilicate glass (2–5 %) and synthetic silica (0.5 %) increased initial compressive strength of cement mortars to 10 %.

Basic Mechanical and Neutron Shielding Performance of Mortar Mixed with Boron Compounds with Various Alkalinity

This study conducted fundamental tests on mortars using the boron compounds recycled industrial wastes to replace uneconomic boron products. The boron compounds were three types according to the pH and the physical and neutron shielding performance of mortar mixed with boron compounds was examined. The adopted boron compounds classified as acid, slightly alkaline, and strongly alkaline with respect to the pH are acidic boric acid, alkali borax, and high alkali borax, respectively. The physical properties were evaluated by measuring the compressive strength and setting time as well as the thermal neutron shielding performance. The measured compressive strength revealed that the strengths of the specimens mixed with boron compounds were generally lower than that of the basic specimen made of control specimen. In addition, the initial and final setting times were longer than those of the control specimen. The thermal neutron shielding performances of the specimens mixed with boron compounds were higher than that of the control specimen. Consequently, the differences of the type and chemical composition of boron compounds influenced the physical properties and thermal neutron shielding performance of mortar, including its compressive strength, setting time, and neutron shielding performance. Therefore, it is important to determine the optimal amount of boron compounds in the fabrication of mortar.

Radiation Shielding Concrete with alternate constituents: An approach to address multiple hazards

Radiation Shielding Concrete (RSC) is a superior alternative to many conventional and modern shields against gamma and neutron radiation hazards. The present work is the first comprehensive review on utilization of alternate materials, emphasizing hazardous industrial byproducts, as constituents of RSC. Such usage enhances the performance, sustainability, and affordability of RSC. Added advantages are the immobilization of wastes and the conservation of natural resources for RSC. The review analyses incorporation of ferrous and non-ferrous slags, mines wastes, plastics, red mud, cathode ray tube's glass, metallic wastes, fly ash, silica fume, and miscellaneous residues. Besides, utilization of fibers, nanoparticles, and calcined clay is investigated. The influence on shielding efficiency is adjudged by scrutinizing changes in parameters such as half-value layer and linear attenuation coefficients. Similarly, variations in mechanical and durability properties are investigated and compared. The underlying responsible factors related to the physical, chemical and morphological characteristics of materials and their consequences on RSC's behavior are correlated. In association with alternatives, the advantages, disadvantages, and possible treatment methods are discussed. The country-wise, material-specific, and progressive research trends are revealed to facilitate future work in this upcoming field. Finally, conclusions are drawn with exposition of current bottlenecks and scope of future research.

Total Ambient Dose Equivalent Buildup Factors for Portland Concrete

In this work, total ambient dose equivalent buildup factors for Portland concrete slabs are calculated using Monte Carlo n-particle software MCNP6™. Buildup factor calculations could approach intractable solutions in general as they depend on a large number of variables. These include geometry, source energy, and the composition of the shield (which itself can be heterogeneous). In this work, Cf and americium-beryllium sources are considered, as well as monoenergetic incident neutrons in the energy range from 0.025 eV to 14 MeV at multiple incident angles. The shielding material of interest was taken to be standard Portland concrete. The transmitted neutron and gamma-ray ambient dose rate was calculated first and then used for total buildup factor calculations. Perhaps more telling than the calculated theoretical buildup factor, the credible dispersion in expected resultant buildup factors was also calculated by conducting a very rudimentary sensitivity analysis, varying the water content in the first case and then varying the amount of aggregate. An additional aim of this work is to provide a model based on the machine-learning technique called the support vector regression method in the calculation of concrete buildup factors.

Enhancement of thermal neutron shielding of cement mortar by using borosilicate glass powder

Concrete has been used as a traditional biological shielding material. High hydrogen content in concrete also effectively attenuates high-energy fast neutrons. However, concrete does not have strong protection against thermal neutrons because of the lack of boron compound. In this research, boron was added in the form of borosilicate glass powder to increase the neutron shielding property of cement mortar. Borosilicate glass powder was chosen in order to have beneficial pozzolanic activity and to avoid deleterious expansion caused by an alkali-silica reaction. According to the experimental results, borosilicate glass powder with an average particle size of 13µm showed pozzolanic activity. The replacement of borosilicate glass powder with cement caused a slight increase in the 28-day compressive strength. However, the incorporation of borosilicate glass powder resulted in higher thermal neutron shielding capability. Thus, borosilicate glass powder can be used as a good mineral additive for various radiation shielding purposes.

Hybrid nanocomposites based on poly aryl ether ketone, boron carbide and multi walled carbon nanotubes: evaluation of tensile, dynamic mechanical and thermal degradation properties

In this study an attempt has been made to incorporate a radiation resistant filler like boron carbide (B4C) nanopowder along with multi walled carbon nanotubes (MWCNT) in a high performance polymer namely poly aryl ether ketone (PAEK) for potential applications in the nuclear industry. The dispersion of nanofillers in PAEK was established by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Infra red (IR) spectroscopy indicated the interaction between functionalized MWCNT (F-MWCNT) and PAEK. The optimum combination of B4C and F-MWCNT was obtained from the tensile property analysis. It was found from the dynamic mechanical analysis that the storage modulus of the composite at elevated temperature was enhanced by B4C inclusions. Mechanical damping factor spectra showed the shift of PAEK glass transition temperature to higher values due to the presence of B4C and F-MWCNT. Thermogravimetric analysis (TGA) presented the resistance offered by B4C to the degradation of PAEK especially at elevated temperatures.