Search Results (72)

This study investigates neutron radiation sources in medical cyclotrons used for PET isotope production, focusing on differences between ¹⁸F and ¹¹C. Neutron and gamma dose rates were measured in the bunker and operator control room during routine production with an 11 MeV Eclipse cyclotron. ¹⁸F production generated approximately 2.5 times higher neutron levels in the bunker than ¹¹C. Shielding performance also varied: the same wall reduced neutron fluxes by factors of k_F = 14,000 for ¹⁸F and k_C = 86,000 for ¹¹C, while gamma shielding was similar for both isotopes (k_γ ≈ 28,000). However, the neutron shielding factor calculated from the data for ¹⁸F should be taken as k_F ≥ 1.4 × 10⁴, because several neutron readings reached the upper limit of the detector range, which indicates a partial underestimation of the dose in the bunker. Consequently, neutron levels in the control room during ¹⁸F production were about 15-fold higher than during ¹¹C production. These differences result from distinct neutron generation mechanisms. The ¹⁸O(p,n)¹⁸F reaction produces primary neutrons with a Maxwellian spectrum (~2.5 MeV), while ¹¹C neutrons arise solely from secondary interactions in structural materials. The findings emphasize the need for composite shielding adapted to isotope-specific spectra. Annual dose estimates (260 ¹⁸F and 52 ¹¹C productions) showed neutron exposure (3.78 mSv/year, 57%) exceeded gamma exposure (2.82 mSv/year, 43%). The total dose of 6.6 mSv/year is ~33% of regulatory limits, supporting compliance but underscoring the need for dedicated neutron dosimetry. Full article

Compact fusion reactors are receiving increasing interest as a promising route for accelerating the path toward commercial fusion, thanks to their reduced size and cost. However, this compactness introduces new technological challenges, including higher radiation loads on critical functional components, such as the magnet system. Neutron shielding is therefore of utmost importance to guarantee the expected lifetime of the device, and its selection must account for the harsh environment imposed by the high radiation flux. Shielding materials should be structurally stable, not melt within the operational temperature windows, and be relatively low-cost. For nuclear reactor applications, binary compounds are typically the preferred choice as they often meet these requirements, particularly in terms of availability and cost. In this work, we present a systematic Monte Carlo analysis of more than 700 binary compounds, exposed to the neutron spectrum at the most loaded position of the vacuum vessel in a simplified model of a compact fusion reactor. Shielding performances were evaluated in a toroidal geometry in terms of neutron attenuation, power deposition, and activation, leading to the identification of several promising compositions for effective neutron shielding in future fusion applications. Full article

Addressing the challenge of synergistically optimizing shielding performance and mechanical properties in nuclear radiation shielding materials, this study designed and prepared as-cast Mg-12Dy-xSm-0.4Zr (x = 1, 2, 3) alloys by incorporating rare earth elements Dy and Sm, which possess high thermal neutron absorption cross-sections. The co-addition of Sm and Dy significantly refined the grains and promoted the precipitation of bone-like Mg₅(Sm,Dy) and Mg₄₁Sm₅ phases along grain boundaries. The alloys exhibited favorable mechanical properties, with ultimate tensile strength (UTS) reaching up to 194.6 MPa and elongation (EL) up to 10.9%. However, higher Sm content led to an increased amount of secondary phases at grain boundaries, resulting in stress concentration and a subsequent decline in both yield strength and elongation. Moreover, the combined addition of Dy and Sm markedly enhanced the thermal neutron shielding performance. Experimental results agreed well with Geant4 simulations, showing that both the neutron shielding rate and linear attenuation coefficient improved with increasing Sm content, demonstrating the positive role of Dy and Sm in neutron absorption. The developed alloy achieves simultaneous improvement in mechanical properties and neutron shielding capacity, providing valuable insights for the development of lightweight “function–structure integrated” radiation shielding materials for applications such as nuclear medicine and aerospace. Full article

With the rapid growth of nuclear energy, effective shielding of radioactive nuclear by-products is critical for safety and environmental protection. Gadolinium (Gd) is ideal for neutron shielding due to its exceptionally high thermal neutron capture cross-section. Despite significant progress in developing various Gd-based shielding materials, poor interfacial compatibility between Gd₂O₃ and polymer matrices remains a significant limitation. In this study, we addressed this challenge by successfully modifying Gd₂O₃ nanoparticles (Gd₂O₃@SIT-M) through the construction of a dual-layer molecular coating using electrostatic interactions. Initially, Gd₂O₃ was functionalized with the silane coupling agent 3-(trihydroxysilyl) propyl-1-propane-sulfonic acid (SIT), followed by subsequent assembly of polyether amine M2070 onto this modified surface. The combined presence of hydrophilic sulfonic acid groups from SIT and amine-ether groups from M2070 endowed Gd₂O₃@SIT-M nanoparticles with excellent hydrophilicity, significantly reducing their aqueous contact angle to 14.34°. Consequently, this modification strategy notably enhanced the dispersion stability of Gd₂O₃ nanoparticles in aqueous solutions and polymer matrices. The developed approach thus provides an effective pathway for fabricating advanced polymer-based neutron shielding materials with improved dispersibility, stability, and overall performance. Full article

A novel high-temperature-resistant W-B₄C-poly(4-methyl-1-pentene) (PMP) composite shielding material against neutron and gamma rays was developed and fabricated. Firstly, utilizing the ²³⁵U-induced fission spectrum as the source term, the compositional ratio of the W-B₄C-PMP ternary composite was optimized using the genetic algorithm-based GENOCOPIII program coupled with MCNP simulations. Then, the composite was fabricated through coupling agent modification, melt mixing, and hot pressing. Finally, the effects of coupling modification and tungsten content on the thermomechanical properties of the composite were investigated. Results demonstrated that functional groups from the silane coupling agent KH550 were successfully grafted onto the filler surfaces. For composites containing 30 wt% modified B₄C and 40 wt% modified W in the PMP matrix, the heat deflection temperature (HDT) increased by 18.5% and 19.1%, respectively, compared to their unmodified counterparts. The impact strength also improved by 31.6% and 5.0%, respectively. The variation trend of the composite’s modulus approximately followed the classical Einstein model, while its tensile strength and flexural strength conformed precisely to the model: ${\displaystyle \frac{{\mathsf{\sigma }}_{\mathrm{c}}}{{\mathsf{\sigma }}_{\mathrm{m}}}}=0.88{\mathrm{V}}_{\mathrm{f}}^{-0.02}$. Thermal analysis indicated that the composites possessed a melting point exceeding 230 °C, and their thermal stability improved with increasing filler content. Full article

In this study, novel Al₆₀₆₁-(30-x)B₄C-xSm₂O₃ (x = 0, 1, 3, 5, 7, and 9 wt%) composites were fabricated using high-energy ball milling followed by cold pressing and sintering. The aim was to improve both the mechanical performance and radiation-shielding capabilities by integrating Sm₂O₃ as a reinforcement phase. Microstructural analyses via XRD and SEM-EDX revealed that the addition of Sm₂O₃ significantly enhanced phase uniformity, reduced porosity, and improved interfacial bonding, especially by mitigating the inherent poor wettability between Al₆₀₆₁ and B₄C. As a result, the relative density, hardness, and wear resistance were considerably improved with an increasing Sm₂O₃ content. Monte Carlo simulations (MCNP6.2) demonstrated that while thermal neutron shielding showed a slight decline due to the reduced boron content, fast neutron and gamma-ray attenuation were substantially enhanced owing to the high atomic number and density of Sm₂O_3. The results demonstrate that the mechanical performance and superior neutron-shielding properties contribute to new visions in material design and applications and have the potential to provide safer and more effective radiation-protection solutions that are environmentally sustainable. Full article

The capability to discriminate among nuclear fuel properties is essential for a successful nuclear safeguard and security program. Accurate nuclear material identification is hindered due to challenges such as differing levels of enrichments, weak radiation signals in the case of fresh nuclear fuel, and complex self-shielding effects. This study explores the application of supervised machine learning algorithms to digitized radiation detector data for classifying signatures of special nuclear materials. Three scintillation detectors, an EJ-309 liquid scintillator, a CLYC crystal scintillator, and an EJ-276 plastic scintillator, were used to measure gamma-ray and neutron data from special nuclear material at the National Criticality Experiments Research Center (NCERC) at the National Nuclear Security Site (NNSS), at Nevada, USA. Radiation detector pulse data was extracted from the collected digitized data and applied to three separate supervised learning models: Random Forest, XGBoost, and a feedforward Deep Neural Network, chosen for their wide-spread use and distinct data ingest and processing analytics. Through model refinement, such as adding an additional parameter feature, an accuracy of greater than 95% was achieved. Analysis on model parameter feature importance revealed Countrate, which is the overall gamma-ray and neutron incidents for each detector, was the most influential parameter and essential to include for improved classification. Initial model versions not including the Countrate parameter feature failed to classify. Supervised learning models allow for measured gamma-ray and neutron pulse data to be used to develop effective identification and discrimination between material compositions of different fuel assemblies. The study demonstrated that traditional pulse shape parameters alone were insufficient for discriminating between special nuclear materials; the addition of Countrate markedly improved model accuracy but all models were heavily dependent on this specific feature, thus illustrating the need for alternative, more distinct parameter features. The machine learning development framework captured in this study will be beneficial for future applications in discriminating between different fuel enrichments and additives such as burnable poisons. Full article

In this study, the neutron and gamma radiation-shielding properties of serpentinites from the Guleman ophiolite complex were investigated, and results were evaluated in comparison with rare earth element (REE) content. The linear and mass attenuation coefficients (LAC and MAC), half-value layer (HVL), mean free path (MFP), and effective atomic numbers (Z_eff) of serpentinite samples were experimentally measured in the energy range of 80.99–383.85 keV. Theoretical MAC values were calculated. Additionally, fast neutron removal cross-sections, as well as thermal and fast neutron macroscopic cross-sections, were theoretically determined. The absorbed equivalent dose rates of serpentinite samples were also measured. The radiation protection efficiency (RPE) for gamma rays and neutrons were determined. It was observed that the presence of rare earth elements within serpentinite structure has a significant impact on thermal neutron cross-sections, while crystalline water content (LOI) plays an influential role in fast neutron cross-sections. Moreover, it has been observed that the concentration of gadolinium exerts a more substantial influence on the macroscopic cross-sections of thermal neutrons than on those of fast neutrons. The research results reveal the mineralogical, geochemical, morphological and radiation-shielding properties of serpentinite rocks contribute significantly to new visions for the use of this naturally occurring rock as a geological repository for nuclear waste or as a wall-covering material in radiotherapy centers and nuclear facilities instead of concrete. Full article

The Volumetric Neutron Source (VNS) is a pivotal facility proposed for advancing fusion nuclear technology, particularly for the qualification of breeding blanket systems, a key component of DEMO and future fusion reactors. This study focuses on the design and optimisation of the VNS Thermal Shield, adopting a multidisciplinary approach to address its thermal and structural behaviours. The Thermal Shield plays a crucial role in protecting superconducting magnets and other cryogenic components by limiting heat transfer from higher-temperature regions of the tokamak to the cryostat, which operates at temperatures between 4 K and 20 K. To ensure both thermal insulation and structural integrity, multiple design iterations were conducted. These iterations aimed to reduce electromagnetic (EM) forces induced during magnet charge and discharge cycles by introducing strategic cuts and reinforcements in the shield design. The optimisation process included the evaluation of various aluminium alloys and composite materials to achieve a balance between rigidity and weight while maintaining structural integrity under EM and mechanical loads. Additionally, an integrated thermal study was performed to ensure effective temperature management, maintaining the shield at an operational temperature of around 80 K. Cooling channels were incorporated to homogenise temperature distribution, improving thermal stability and reducing thermal gradients. This comprehensive approach demonstrates the viability of advanced material solutions and design strategies for thermal and structural optimisation. The findings reinforce the importance of the VNS as a dedicated platform for testing and validating critical fusion technologies under operationally relevant conditions. Full article

This study presents a hybrid computational investigation into the radiation shielding behavior of Fe-enriched Al-based alloys (Al-Fe-Mo-Si-Zr) for potential use in nuclear applications. Four alloy compositions with varying Fe contents (7.21, 6.35, 5.47, and 4.58 wt%) were analyzed using a combination of Monte Carlo simulations, machine learning (ML) predictions based on multilayer perceptrons (MLPs), EpiXS, and SRIM-based charged particle transport modeling. Key photon interaction parameters—including mass attenuation coefficient (MAC), half-value layer (HVL), buildup factors, and effective atomic number (Z_eff)—were calculated across a wide energy range (0.015–15 MeV). Results showed that the 7.21Fe alloy exhibited a maximum MAC of 12 cm²/g at low energies and an HVL of 0.19 cm at 0.02 MeV, indicating improved gamma attenuation with increasing Fe content. The ML model accurately predicted MAC values in agreement with Monte Carlo and XCOM data, validating the applicability of AI-assisted modeling in material evaluation. SRIM calculations demonstrated enhanced charged particle shielding: the projected range of 10 MeV protons decreased from ~55 µm (low Fe) to ~50 µm (high Fe), while alpha particle penetration reduced accordingly. In terms of fast neutron attenuation, the 7.21Fe alloy reached a maximum removal cross-section (Σ_R) of 0.08164 cm⁻¹, showing performance comparable to conventional materials like concrete. Overall, the results confirm that Fe-rich Al-based alloys offer a desirable balance of lightweight design, structural stability, and dual-function radiation shielding, making them strong candidates for next-generation protective systems in high-radiation environments. Full article

This study evaluated the impact of mill scale (MS), a steel manufacturing waste product, as a replacement for natural fine aggregate (up to 100% by volume) in high-calcium fly ash-based geopolymer concrete (GC) and ordinary Portland cement concrete (CC). We compared the workability, compressive strength, splitting tensile strength, modulus of elasticity, density, water absorption, porosity, ultrasonic pulse velocity, thermal conductivity, acid resistance, chloride penetration, and radiation attenuation (gamma rays and fast neutrons) of the resulting materials. Results showed that GC and CC with 100% MS achieved 28-day compressive strengths of 23.6 MPa and 35.2 MPa, respectively, representing 58% and 90% of the strengths of plain GC and CC. MS-modified GC exhibited superior acid and chloride resistance compared to CC. Importantly, MS enhanced radiation shielding, with GC and CC containing 100% MS, demonstrates the best performance, suggesting its potential use in radiation-shielding construction materials. Full article

An important aspect of pressurized water reactor (PWR) lifetime monitoring is supporting radiation shielding analyses which can quantify various in-core and out-core effects induced in reactor materials by varying neutron–gamma fields. A good understanding of such a radiation environment during normal and accidental operating conditions is required by plant regulators to ensure proper shielding of equipment and working personnel. The complex design of a typical PWR is posing a deep penetration shielding problem for which an elaborate simulation model is needed, not only in geometrical aspects but also in efficient computational algorithms for solving particle transport. This paper presents such a hybrid shielding approach of FW-CADIS for characterization of the reactor pressure vessel (RPV) irradiation using SCALE6.2.4 code package. A fairly detailed Monte Carlo model (MC) of typical reactor internals was developed to capture all important streaming paths of fast neutrons which will backscatter the biological shield and thus enhance RPV irradiation through the cavity region. Several spatial differencing and angular segmentation options of the discrete ordinates S_N flux solution were compared in connection to a S_N mesh size and were inspected by VisIt code. To optimize MC neutron transport toward the upper RPV head, which is a particularly problematic region for particle transport, a deterministic solution of discrete ordinates in forward/adjoint mode was convoluted in a so-called contributon flux, which proved to be useful for subsequent S_N mesh refinement and variance reduction (VR) parameters preparation. The pseudo-particle flux of contributons comes from spatial channel theory which can locate spatial regions important for contributing to a shielding response. Full article

Dense lithium hydride (LiH) is widely used in neutron shielding applications for thermonuclear reactors and space systems due to its unique properties. However, traditional sintering methods often lead to cracking in LiH products. This study investigates the densification sintering of LiH using microwave technology. A multiphysics model was established based on the measured dielectric properties of LiH at different temperatures, allowing for a detailed analysis of the electromagnetic and thermal field distributions during the microwave heating of cylindrical LiH samples. The results indicate that the electric field distribution within the LiH is relatively uniform, with resistive losses concentrated primarily in the LiH region of the microwave cavity. LiH rapidly absorbs microwave energy, reaching the sintering temperature of 520 °C in just 415 s. Additionally, the temperature difference between the low- and high-temperature regions during the sintering process remains below 5 °C, demonstrating excellent uniform heating characteristics. The microwave sintering process enhances interface migration within the LiH samples, resulting in dense metallurgical bonding between grains. In summary, this research provides valuable insights and theoretical support for the rapid densification of LiH materials, highlighting the potential of microwave technology in improving material properties. Full article

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. Full article

There are various types of charged particles in the space environment, which can cause different types of radiation damage to materials and devices, leading to on-orbit failures and even accidents for spacecraft. Developing lightweight and efficient radiation-shielding materials is an effective approach to improving the inherent protection of spacecraft. The protective performance of different materials against proton and electron spectra in the Earth’s radiation belts is evaluated using a Geant4 simulation. Based on the simulation results, suitable hardening components were selected to design composite materials, and B₄C/PEEK composites with different B₄C contents were successfully prepared. The experimental results demonstrate that the simulated and experimental results for the electron, proton and neutron shielding performance of the B₄C/PEEK composites are consistent. These composites exhibit excellent radiation shielding capabilities against electrons, protons and neutrons, and the radiation protection performance improves with increasing B₄C content in the B₄C/PEEK composite materials. Full article