Macroscopic Cross Section Research Articles

A Comparative Study of Neutron Shielding Performance in Al-Based Composites Reinforced with Various Boron-Containing Particles for Radiotherapy: A Monte Carlo Simulation

This study aimed to assess and compare the shielding performance of boron-containing materials for neutrons generated in proton therapy and used in boron neutron capture therapy (BNCT). Five composites, including AlB2, Al-B4C, Al-TiB2, Al-BN, and Al-TiB2-BN, were selected as shielding materials, with concrete used as a benchmark. The mass fraction of boron compounds in these materials ranged from 10% to 50%. The Monte Carlo toolkit Geant4 was employed to calculate shielding parameters, including neutron ambient dose equivalent, dose values, and macroscopic cross-section. Results indicated that, compared to concrete, these boron-containing materials more effectively absorb thermal neutrons. When the boron compound exceeds 30 wt.%, these materials exhibit better shielding performance than concrete of the same thickness for neutrons generated by protons. For a given material, its shielding capability increases with boron content. Among the five materials when the material thickness and boron compound content are the same, the shielding performance for neutrons generated by protons, from best to worst, is as follows: Al-TiB2, Al-B4C, AlB2, Al-TiB2-BN, and Al-BN. For BNCT, the shielding performance from best to worst is in the following order: Al-B4C, AlB2, Al-TiB2, Al-TiB2-BN, and Al-BN. The results of this study provide references and guidelines for the selection and optimization of neutron shielding materials in proton therapy and BNCT facilities.

A study on the interaction parameters of charged and uncharged radiation types with some indoor plants

This study investigates how gamma rays, neutrons, and electrons interact with five commonly found indoor plants: Spathiphyllum wallisii (SW), Ficus elastica (FE), Dieffenbachia camilla (DC), Schefflera arboricola (SA), and Ficus benjamina (FB). Utilizing experimental measurements (with HPGe detector), Monte Carlo simulations (GEANT4 and FLUKA), and theoretical calculations (ESTAR and WinXCOM), some radiation interaction parameters for gamma rays, fast neutrons, thermal neutrons, and electrons were determined. Secondary particle generation was also analyzed to provide a comprehensive assessment. The determined linear attenuation coefficients with the help of the WinXCOM are 0.1376, 0.1662, 0.1385, 0.1651 and 0.1698 cm-1 for SW, FE, DC, SA and FB, respectively. The calculated total macroscopic cross sections for indoor plants in the same sample order are 2.0290, 2.0350, 2.0285, 2.0363 and 2.0362 cm-1. Among the investigated plants, FB exhibited the highest gamma ray interaction, while SA and FB showed superior interaction against fast neutrons compared to SW and DC. The findings reveal significant variations in interaction effectiveness and secondary radiation production across these plants, offering valuable insights for radiation safety and environmental health evaluations.

Research on the high-performance computing method for the neutron diffusion spatiotemporal kinetics equation based on the convolutional neural network

Due to the uncertainty of computational results and the lack of interpretability of models in solving physical field equations in current deep learning, this paper designs a convolutional neural network that can be used to solve the neutron diffusion spatiotemporal kinetics equation in polar and cylindrical coordinate systems. This algorithm directly utilizes the macroscopic cross-section of the material without using the lattice homogenization method, replaces the finite volume method with the extended matrices, and solves the extended matrices using the convolutional kernels instead of the iterative algorithms. Taking the simplified Tsinghua High Flux Reactor (THFR) as an example, the feasibility of the algorithm is verified on the PyTorch platform and compared with the calculation results of the source iteration method running on the GPU. The calculation results show that when the number of grids in the radial and axial sections of the simplified THFR model is 804,600 and 3,576,000, respectively, and the algorithm is iterated 3000 times, the normalized power of the convolutional neural network and the source iteration method converges to 10−10, and the maximum point by point error of the neutron flux density of the above two algorithms converges to 10−5. The computational time consumed by the convolutional neural network is approximately 880.64 s and 3729.62 s, which reduces the computational time by 4.66% and 5.05% compared to the GPU parallel accelerated source iteration method, and the former consumes 43.75% less memory compared to the latter. The convolutional neural network is mainly used as the virtual physics engine for the THFR digital twin system, in addition to solving the neutron diffusion spatiotemporal kinetics equation and further improving computational speed. The algorithm directly utilizes the neutron macroscopic cross-section of the material to calculate the neutron flux density distribution without using the lattice homogenization, providing theoretical guidance and algorithm support for developing the high-precision multi-physical field coupling model.

Fast and Thermal Neutron Removal Cross-Section for Ceramic Glass Aluminum Oxynitride

This study investigates the effectiveness of transparent aluminum oxynitride (AlON) in neutron shielding, focusing on both fast and thermal neutrons. Using conventional radiation attenuation parameters, the macroscopic neutron removal cross-sections of AlON were calculated for varying neutron energies and material thicknesses. The Geant4 simulation toolkit was employed to model and analyze the neutron interactions with AlON. The results indicate that AlON exhibits a high neutron shielding capacity for fast neutrons, performing better than traditional shielding materials such as concrete and lead. However, for thermal neutrons, AlON's performance was less effective, only surpassing lead but not concrete or water. The findings suggest that while AlON is highly effective for fast neutron shielding, it may require complementary materials to adequately shield thermal neutrons. This could involve using AlON in combination with other materials to create a more comprehensive neutron shielding solution. AlON shows significant potential as a neutron shielding material, particularly for fast neutrons. Its integration with additional shielding materials could enhance its overall effectiveness, making it suitable for various nuclear and radiation protection applications.

An improved method for evaluating fracture density using pulsed neutron capture logging

Fracture density is a critical fracture evaluation parameter for the optimization and production prediction of hydraulic fracturing models. Nonradioactive tracer (NRT) techniques have successfully used a quantitative fracture density method based on neutron-induced gamma rays from formation elements. Furthermore, fracture density can be calculated using the formation capture cross section before and after hydraulic fracturing based on pulsed neutron capture logging. However, the response sensitivity of the macroscopic scattering cross section to the fractures decreases when the fracture density is high due to the neutron self-shielding phenomenon. Therefore, an improved fracture density evaluation method is applied, which combines the macroscopic capture cross section and the neutron self-shielding correction factor. In the new method, the peak area of titanium from the captured gamma spectrum is used to obtain a neutron self-shielding correction factor to improve the sensitivity of fracture density determination. Furthermore, the response of capture cross-section variation to fracture density at various tagged proppant concentrations and formation backgrounds is investigated. The findings indicate that the tagged proppant concentration influences the detection limit of fracture density and the sensitivity of fracture density identification. The accurate calculation range of fracture density using the new method has been extended from 5% to 10% under the condition that the tagged proppant concentration is 0.2%. Meanwhile, whereas water salinity significantly impacts capture cross-section variation, the effects of porosity, lithology, and fluid type on capture cross-section variation are negligible. A simulated fracturing example demonstrates the method’s applicability in various measurement environments. The results show that fracture density and height are consistent with the model settings after correcting for water salinity, and the fracture density calculation error is less than 3%. Therefore, our evaluation method for fracture density corrected for the neutron self-shielding effect improves response sensitivity and fracture density calculation accuracy.

Monte Carlo coupled Multi-Physics with spatially continuous material properties

In this study, we performed Monte Carlo coupled multi-physics simulations with spatially continuous material properties via Functional Expansion Tally combined with delta-tracking. The proposed multi-physics coupling framework significantly reduces the need for spatial discretization in the problem geometry, potentially decreasing simulation time as macroscopic cross section reconstructions are performed less frequently. The proposed method has been tested on pin-cell and assembly levels at hot full power to evaluate its applicability. At both pin and assembly levels, numerical experiments have demonstrated that the proposed framework can produce solutions that asymptotically approach those from conventional cell-based discretized simulations with infinitesimal cells. Furthermore, the proposed method accelerates the simulation time by over four times compared to the cell-based discretized approach with very small cells for cases without neutron absorbers. Therefore, the proposed method has the potential to be integrated into future Monte Carlo production codes, to meet the demanding requirements for improved reactor safety.

Fixed Source Sensitivity Calculations for Inertial Confinement Fusion Applications

A numerical code library was developed for the radiation transport code MCNP6.3 to calculate generalized response sensitivity coefficients for fixed source neutron transport problems with applications to inertial confinement fusion (ICF) experiments. The new MCNP6.3 dependency is used to generate a novel time convolution response that represents a neutron time-of-flight (nToF) signal. The traditional suite of macroscopic cross-section sensitivities and constrained fixed source probability distribution sensitivities are available for both the standard and the new response tallies in this library. However, novel sensitivity coefficients for the constrained hyperparameters of analytic fixed source probability distributions are emphasized in this work for their connection to ICF neutron transport models. Particularly, advanced Monte Carlo methods are developed for calculating the sensitivity of a nToF signal to perturbations in an ICF plasma’s ion temperature and burn history as well as perturbations in the target liner mass density and the shape parameters of the nToF detector’s impulse response function. Together, these capabilities form an advanced suite of computational tools that can be used to analyze and extract information from any ICF experimental platform.

Physical, chemical and radiation shielding properties of metakaolin-based geopolymers containing borosilicate waste glass

Upcycling waste borosilicate glass (BS) is a viable and sustainable way of preserving the natural resources required for the production of new materials and ensuring environmental sanitation and safety. This study was aimed at investigating how the addition of BS influences the physical, chemical, and mechanical properties and shielding performance against neutrons and charged particles on metakaolin-based geopolymer G. Four batches of G infused with 10%, 20%, and 30% wt BS samples and coded as G, G-10BS, G-20BS, and G-30BS, respectively, were fabricated based on the solid-state reaction method. The samples were characterized accordingly using experimental and theoretical procedures. X-ray diffraction analysis of the materials showed that the addition of BS suppressed crystal peaks in the geopolymers. The Vickers hardness values were calculated as 554, 503, 517, and 528 HV under a 0.1 kg load. The values of fast neutron macroscopic cross-section increased from 0.0601 cm−1 to 0.0706 cm−1 as BS content increased from 0 to 30 wt%. The coherent, incoherent, absorption, and total interaction probabilities of thermal neutrons increased in the geopolymer samples. For electrons, the stopping powers were in the range of 1.575–13.17 cm2/g, 1.571–13.15 cm2/g, 1.571–13.16 cm2/g, and 1.571–13.18 cm2/g for G, G-10BS, G-20BS, and G-30BS, respectively. The projected ranges of carbon ions in G, G-10BS, G-20BS, and G-30BS are within the limits 0.069–17.87 μm, 0.064–16.99 μm, 0.064–16.99 μm, and 0.057–15.24 μm. The trend of the projected ranges of electrons, protons, α-particles, and C ions with kinetic energies within 15 keV and 15 MeV was mostly in the sequence G > G-10BS > G-20BS > G-30BS. This order was consistent with the density and BS content of the geopolymer samples. The addition of BS to the metakaolin-based geopolymer improved the fast neutron moderating capacity and thermal neutron cross-sections and reduced the range of charged particle radiation in the geopolymers. The neutron shielding ability of G-30BS was found to be better than that of some existing shielding materials. The BS-infused concrete can be useful for the production of concrete used for radiation control.

Multiple semi-coherent particles strengthened ultra-fine-grained Al composites for neutron shielding materials

Neutron shielding materials face imbalanced behaviors among shielding, strength, and ductility properties. Based on the requirement of the high property shielding particles, a superior semi-coherent τ(Al4MgGd) phase was designed and predicted by cluster expansion (CE) method using density functional theory calculations. To realize its shielding property, the Powder Metallurgy-based routines (i.e., powder fabrication, spark plasma sintering, and hot extrusion techniques) are used to fabricate 6TiB2/Al-6Mg-5Gd (wt.%) composite with dispersed refined τ phases and homogenized TiB2 distribution. The atomic structure of ternary phase τ is examined by aberration-corrected high-angle annual dark-field (HAADF) scanning transmission electron microscope (STEM) and energy dispersive X-ray spectroscopy (EDXS) STEM experiments, which is well complied with the calculated compound (Al4MgGd). In detail, the τ(Al4MgGd) phase has a semi-coherent interface both with α-Al and TiB2, which is consistent with the prediction of interface relationships. With the optimized interfaces, the TiB2 and τ phases can effectively promote recrystallization and suppress grain growth, leading to the formation of ultra-fine grain structure. Then, the composite exhibits advanced shielding properties (Macroscopic transmission cross section ∼24.1 /cm, higher than 30 %B4C/Al) and optimized synergic mechanical properties (Ultimate tensile strength ∼506 MPa, elongation ∼12.9 %), which are far higher than available Al-based neutron shielding materials. Finally, the underlying strength-ductility mechanisms are discussed. Critically, the design and optimization of shielding particle interfaces are reliable strategies for developing novel structural-functional integrated materials.

Quantum motion of oxygen and hydrogen in water: Atomic and total kinetic energy across melting from neutron scattering measurements.

We provide a concurrent measurement of the hydrogen and oxygen nuclear kinetic energies in the water molecule across melting at 270K in the solid phase and 276K in the liquid phase. Experimental values are obtained by analyzing the neutron Compton profiles of each atomic species in a deep inelastic neutron scattering experiment. The concurrent measurement of the atom kinetic energy of both hydrogen and oxygen allows the estimate of the total kinetic energy per molecule due to the motion of nuclei, specifically 35.3 ± 0.8 and 34.8 ± 0.8kJ/mol for the solid and liquid phases, respectively. Such a small difference supports results from abinitio simulations and phenomenological models from the literature on the mechanism of competing quantum effects across the phase change. Despite the experimental uncertainties, the results are consistent with the trend from state-of-the-art computer simulations, whereby the atom and molecule kinetic energies in the liquid phase would be slightly lower than in the solid phase. Moreover, the small change of nuclear kinetic energy across melting can be used to simplify the calculation of neutron-related environmental dose in complex locations, such as high altitude or polar neutron radiation research stations where liquid water and ice are both present: for neutron energies between hundreds of meV and tens of keV, the total scattering cross section per molecule in the two phases can be considered the same, with the macroscopic cross section only depending upon the density changes of water near the melting point.

The role of MoO3 on the physical, elasto-mechanical and nuclear shielding efficiency of barium-boro-bismuthate glass system: Comparative investigation

Investigations were conducted on the effect of molybdenum addition on the physical, elasto-mechanical and nuclear shielding characteristics of five different glasses, designated S1 through S5, with varying amounts of MoO3. The mechanical parameters of the examined glasses, including their Young (Y), bulk (B), shear (S) and longitudinal (L) moduli, microhardness (H), and poisons ratio, were calculated utilizing the Makishima-Mackenzie technique. The physical and mechanical characteristics of the examined glasses are improved by the addition of MoO3. The Y, B, S, and L moduli rose with increasing MoO3 concentrations from 34.6680 to 35.9989 GPa, 14.13578–14.664 GPa, 21.10714–22.60943 GPa, and 31.70897–33.59673 GPa, respectively. The H values of the glass network also dropped from 0.016431 to 0.016885 GPa. Moreover, as the concentration of MoO3 rose, the Poisson ratio value changed from 0.427021 to 0.429533. The MCNP5 code and WinXCOM program were used to compute various shielding parameters for X/gamma-rays in the energy range between 0.015 and 15 MeV. A good agreement was found between the estimated values using both MCNP5 code and WINXCOM program. The S5 glass sample exhibited the greatest gamma-ray attenuation characteristics among the examined glasses. When compared to other radiation shielding materials including recently studied glasses and polymers as well as other typical gamma ray shielding materials, the current glass composition offers complete (100 %) x-ray protection and effective efficacy for gamma-ray shielding at energies up to 15 MeV. Calculations were made of the studied glasses' macroscopic removal cross sections (∑R), which were then compared to those of commonly used neutron shielding materials. S1 sample performed the best ∑R value while S5 displays a value (0.10516 cm−1) that is comparable to those of graphite, concrete, basalt magnetite, and a few other recently studied glasses and alloys. The heavy ions (proton and alpha particle) attenuation parameters have been evaluated in energy range between 0.01 and 15 MeV. Regarding the studied glasses, the S5 sample was shown to have the best proton and alpha shielding properties. These results imply that the glass samples under study make good nuclear shielding materials for a range of applications.

Application of artificial neural network for assembly homogenized equivalence parameter generation

The development of a new reactor core and optimizations for fuel assembly (FA) structures necessitates the generation of homogenized equivalence parameters, including macroscopic cross-sections (XS), for numerous FA configurations. A feed-forward convolutional neural network, XSNET, has been created to streamline this process. Currently, XSNET can produce these parameters for fresh and burned UO2 fuel, both with and without Gadolinia (Gd) burnable absorber rods. In this study, we integrated XSNET with our in-house nodal diffusion code, RAST-K, to create a conventional two-step approximation code system. This system was tested on various tasks, including analyzing a 3-dimensional pressurized water reactor employing 16 × 16 FA types with Gd rods. Over 1,000 unique loading patterns were considered for nodal calculation, which involved steady-state depletion calculations and pin power reconstruction. The results obtained using XSNET/RAST-K demonstrated good agreement with the reference STREAM/RAST-K code system.

Enhancing gamma and neutron radiation shielding efficiency of LDPE/PVC polymers using cobalt, aluminum, and magnesium oxide fillers

A procedure for creating and assessing LDPE/PVC polymers using Co, Al, and Mg oxides as fillers for use as radiation shielding material was conducted. The SEM/EDX equipment was utilized for elemental analysis and mapping in order to visually represent the morphology of the produced polymers. The characterizations indicate that the MgCo2O4 particles possess an irregular and tiny morphology, displaying a spherical form with an average diameter of less than 100 nm. In addition, the Al2CoO4 particles are characterized as being aggregated, diminutive, and exhibiting a spherical morphology, with an average diameter of less than 150 nm. The current study involved experimental investigation, simulation using FLUKA-MC code, and theoretical calculations using Phy-X-PSD code to examine the radiological properties of the shielding material. The results depend on several factors, including the half-value layer (HVL), the effective atomic number (Zeff), the effective electron number (Neff), the effective conductivity (Ceff), the exposure buildup factor (EBF), the energy-absorption buildup factor (EABF), and the fast neutron macroscopic removal cross-section (FNRC). The SR-NIEL-7 platform has been utilized to compute the mass-stopping power for protons (H+) and alpha particles (He++) based on their kinetic energy. The experimental findings indicate that the density of LDPE/PVC filled with MgCo2O4 is greater than that of Al2CoO4. Furthermore, polymer filled with MgCo2O4 exhibits higher values for MAC, LAC, and FNRCs than Al2CoO4 filler. The measured LAC and MAC conform to the simulated values.

VINUS: A neutron transport solver based on the variational nodal method for reactor core analysis

Compared to traditional transverse integration methods, the variational nodal method, with its unique advantages, is more suitable for high-fidelity calculations of reactor physics in reactors with complex geometries and finer detail descriptions. In this study, the basic theory of the variational nodal method was derived and the VINUS code is developed. The neutron solver based on this method is adaptable to various geometric models, and showcased the code's fundamental framework. On this basis, a set of self-designed macroscopic cross-section benchmarks, actual macroscopic cross-section benchmark VVER-440, and few-group microscopic cross-section benchmark RBEC-M for fast spectrum reactors were used to verify different functionalities of VINUS. The results were shown that VINUS code maintains good computational accuracy and convergence trends. For the VVER-440 benchmarks, the deviation of keff of VINUS from reference is less than 100 pcm, and the maximum power deviation is less than 4 %. For the RBEC-M, the deviation of keff is 125 pcm, and the maximum power deviation is less than 5 %. These outcomes collectively demonstrate the solver's potential for engineering applications in future advanced reactor designs.

Multi-physics analysis of SBLOCA by coupled Neutronics/Thermal-Hydraulics full plant model

With the development of computational tools, the accident analysis would be performed by best estimate approaches. The employment of multi-physics codes would be considered as new calculation procedure to obtain high valuable results by improving calculations with specialized codes. The small break loss of coolant accident (SBLOCA) is an accident that has typically been analyzed using system codes with a limited number of Thermal-Hydraulics (TH) channels in the core region, neglecting the use of multi-physics calculations. This approach results in insufficient accuracy and resolution in both neutronics and TH calculations within the core zone. To address this issue, this study proposes the implementation of a Neutronics/Thermal-Hydraulics coupling with one-to-one mapping in the core zone, achieved through the development of a comprehensive plant TH model. This approach aims to enhance the accuracy and resolution of calculations, enabling the identification of local phenomena, parameters, and feedback during accident simulations, and the incorporation of these effects on the macroscopic cross-sections at each time step. The results of SBLOCA simulation reveals the effects of conservative/best estimate approach on the sequence of events, the initiation time of the safety systems, the critical controlled parameters of the power plant, and the precise understanding of the accident. Also, it was found that the radial power distribution becomes disrupted and loses its symmetry immediately after the SCRAM leading to different characteristics of steam generators (SGs) and natural circulation flow rate in each loop of primary side.

Effect of ZrO2 on physical and radiation shielding properties of tellurium −lead- calcium −borate (TPCB)glass system

This research investigates the impact of adding Zirconia (ZrO2) to tellurium-lead-borate glasses, focusing on how it affects their physical, optical, and radiation shielding properties, particularly against gamma-rays and neutrons. Key parameters for gamma-ray shielding (mass attenuation coefficient (µm), effective atomic numbers (Zeff), and effective electron densities (Neff)) were measured across an energy range of 0.080 to 2.614 MeV. Notably, excellent agreement (within 0.5 %) was observed between the measured and calculated µm values. For slow neutron attenuation measurements, we employed a 241Am-Be neutron source in conjunction with a BF3 detector. Fast neutron attenuation was assessed using the NXCOM software to determine the effective macroscopic removal cross-section (ΣR). The addition of Zirconia notably enhanced the glass density (∼6.8 %) and its ability to shield against gamma-rays (∼10 %) and slow neutrons (∼71 %). However, this improvement comes with a slight trade-off: a decrease of approximately 1.5 % in fast neutron attenuation was observed. This trade-off between shielding effectiveness for different radiation types warrants further investigation to optimize glass compositions, considering the desired balance between shielding performance across various radiation spectrums.

Optimizations of the Accelerated Axial Polynomial Method of Characteristics of the APOLLO3® Deterministic Neutron Transport Code

For some years now, the TDT (two- and three-dimensional transport) solver of the APOLLO3® deterministic neutron transport code has been able to perform lattice calculations on three-dimensional extruded and unstructured geometries. A polynomial expansion of the angular flux has been implemented to better describe the flux gradient axially to reduce the number of computational meshes required to reach a given accuracy. Then the polynomial approximation was extended to macroscopic cross sections to perform evolution calculations. Besides these transport schemes, synthetic acceleration has also been implemented, relying on double PN approximations of the angular flux on the boundaries of the spatial regions. The solver has already introduced several techniques to reduce the transport and memory footprint; for example, for the storage of the surfaces crossed by a trajectory or the classification of chords. In this paper, new optimizations are presented. One deals with how monomials of the polynomial basis are integrated along trajectories. Another one concerns the computation of the source term of the transmission equation in the case of polynomial cross sections. The last optimization exploits the fact that, along horizontal trajectories, the flux and the cross sections are constant to speed up the sweep algorithm. Calculations on 5 × 5 and 7 × 7 pressurized water reactor assemblies were performed to assess the gains of these recently developed strategies. The results show good improvements both in computing time and in memory footprint reductions.

Developed concrete reinforced by waste of lead fishing weights and steel nails as gamma rays and neutrons shields for nuclear power reactors

Wastes of steel nails WSN and wastes of lead fishing weights WLFW were successfully recycled in concrete to enhance its ability to attenuate neutrons and gamma rays. A series of studies on the effect of partially replacing coarse aggregates of concrete with different percentages (5, 10, and 20%) of WSN or WLFW on the density, slump, compressive strength, and attenuation of neutrons and gamma rays were implemented. The compressive strength of WSN-reinforced concrete ranged between 36.2 and 45.2 MPa, while that of WLFW-reinforced concrete ranged between 28.6 and 45.1 MPa. The ability of WSN- or WLFW-reinforced concrete to attenuate neutrons and gamma rays was tested by using two types of neutron energies; slow and total slow neutrons; and three gamma rays energies 661.64, 1173.23, and 1332.51 keV. Slow and total slow neutrons macroscopic cross-section, linear and mass attenuation coefficients of gamma rays, and half value layer for both types of nuclear radiation were obtained. The results showed that there is a significant improvement in the slow and total slow neutron attenuation efficiency of the considered concrete when reinforced by WSN or WLFW. On the other hand, due to the increase in the atomic number and density of both Fe present in WSN and Pb in WLFW, the attenuation ability of the studied gamma ray energies increased. The WLFW-reinforced concrete showed better value in the gamma rays and neutrons attenuation parameters compared to the WSN-reinforced concrete. Hence, the considered concretes reinforced by WSN or WLFW are a distinguished choice as radiation shields in nuclear power reactors, in addition to their adaptability to high concentrations of waste.

Influence of aluminum and vanadium oxides on copper borate glass: A physical/radiological study

Due to the radiation released by commonly used isotopes, many nuclear, medical, and industrial facilities require proper radiation shielding. In this work, distinct copper borate glasses intercalated with varied aluminum and vanadium oxide (Al2O3 and V2O5) content have been synthesized and used against radiation (gamma rays and fast/thermal neutrons). The different percents were as follows: [60% B2O3 + 35% CuO + (5-x)% Al2O3 + xV2O5], where x = 0, 1, and 2.5 wt.%, which was coded as BCu(5-x)Al:xV. The synthesized glass samples were characterized using Fourier transforms, infrared, and X-Raydiffraction analysis. Experimentally, the radiation shielding possessions of the samples were established using an HPGe detector at the gamma energy lines 0.356 MeV, 0.661 MeV, 1.173 MeV, and 1.332 MeV. Also, the prepared glasses were investigated theoretically using the Monte Carlo code (MCNP5) at photon energies of 0.015–15 MeV. Also, the fast and thermal neutron macroscopic effective removal cross-sections were calculated using MRCsC and JANIS-4.1 software, respectively. The prepared sample BCu2.5Al:2.5V, which has a vanadium and aluminum content of 2.5%, has the highest linear attenuation coefficient as well as the highest removal cross-section for fast, and thermal neutrons.

Microstructural design to Al-6Mg-5Gd alloy for the unification of structural and neutron shielding properties

Design of composition and microstructure can effective improve the structural and neutron shielding properties of neutron shielding materials. Based on the shielding standard of 30%B4C/Al, an Al-6Mg-5Gd (wt%) alloy was designed by Shmakov-Yamamoto model and successfully fabricated by Powder Metallurgy-based routines (e.g., powder fabrication, spark plasma sintering and hot extrusion techniques). By the rapid solidification in the atomization process, the submicron cell structures were formed in powders. Due to the difference in the original powder size, the as-sintered alloys had heterogeneous microstructures, leading to the bimodal grain structures of as-extruded alloys. During these processes, the formation of nano-sized τ(C15) (Al4MgGd) phase in combination with its firstly observed transformation behavior showed the great impacts on microstructure evolutions. Therefore, the as-extruded alloy exhibited the superior yield strength (335±7 MPa), ultimate tensile strength (462±5 MPa) and elongation (10.2±0.3%). Through the neutron absorption test, the as-extruded alloy showed the comparable of macroscopic transmission cross section with 30%B4C/Al. In comparison with the available Al-based neutron shielding materials, our report exhibited the best unification of mechanical and functional properties. The underlying mechanisms for synergic strength-ductility behavior were discussed on the nano-sized phase strengthening, α-Al/τ interface modulation, and bimodal grain features. Critically, this work designed and studied a high-performance neutron shielding material, shedding light on the advanced structural-functional integrated materials.