Calculation of the gamma radiation shielding efficiency of cement-bitumen portable container using MCNPX code

Incorporation of spent ion exchange resin simulate into cement composites

Solidification/Stabilization (S/S) of spent radioactive ion exchange resins (IER) is one of many critical problems facing developing nuclear industries all over the world. Immobilization technology using Ordinary Portland Cement (OPC) as an inert matrix, has been widely applied for the solidification/stabilization of spent ion exchange resin. In this study incorporation of simulated IER into cement matrix and characterization of the final solid waste form (FWF) had been searched practically in laboratory scale experiments. Factors that can affect the properties of the FWF including the water/cement (w/c) ratios and the resin concentration were studied systematically. Mechanical integrities, thermal analysis, and mass loss during hardening and curing for 28 days were evaluated for FWF hard blocks. Scanning Electron Microscope (SEM) and X-Ray Diffraction (X-RD) examinations were performed to investigate the internal architecture of the FWF. Moreover, the heat of cement hydration reactions was recorded during the IER solidification process. Based on the experimental results obtained, it is worth to state that according to its acceptable characteristics and advantages of cement as an inert matrix, it can be suggested safely for the immobilization of untreated spent radioactive ion exchange resins.

Impact of micro/nano cadmium oxide on shielding properties of cement-ball clay matrix

This study investigates the gamma radiation shielding properties of cement-ball clay matrix composites doped with micro- and nano-sized cadmium oxide (CdO) particles. The linear attenuation coefficient (LAC) was determined using a sodium iodide (NaI) detector and five radioactive point sources with energies ranging from 59.5 to 1408 keV. The LAC values obtained were compared to the XCOM database and found to be in good agreement. The composites' half-value layer (HVL), tenth value layer (TVL), mean free path (MFP), effective atomic number (Zeff), equivalent atomic number (Zeq), and absorption buildup factor (EABF) were determined. The results showed that the addition of CdO particles improved the radiation-shielding behavior of the composites and increasing the weight fraction of CdO particles increased the shielding effectiveness. The results also illustrated that when nano-sized CdO particles were compared to their micro-sized counterparts, there was a significant enhancement in radiation shielding effectiveness. For instance, a composite material composed of 50% cement, 41.7% ball clay, and 3.8% nano CdO at an energy level of 0.0595 MeV exhibited a remarkable 12.2% increase in attenuation, surpassing the performance of the micro-sized sample with an equivalent concentration. Similarly, another composite consisting of 50% cement, 33.3% ball clay, and 16.7% nano CdO demonstrated a significant 15.4% increase in attenuation at the same energy level, when compared to the micro-sized sample. The study demonstrates the potential of CdO-doped cement-ball clay matrix composites for gamma radiation shielding applications.

Comparative analysis on application conditions of indium (III) oxide-reinforced glasses in nuclear waste management and source transportation: A Monte Carlo simulation study

This study's primary objective is to provide the preliminary findings of novel research on the design of Indium (III) oxide-reinforced glass container that were thoroughly developed for the purpose of a nuclear material container for transportation and waste management applications. The shielding characteristics of an Indium (III) oxide-reinforced glass container with a certain elemental composition against the ⁶⁰Co radioisotope was thoroughly evaluated. The energy deposition in the air surrounding the designed portable glass containers is measured using MCNPX general-purpose Monte Carlo code. Simulation studies were carried out using Lenovo-P620 workstation and the number of tracks was defined as 10⁸ in each simulation phase. According to results, the indium oxide-doped C6 (TZI8) container exhibits superior protective properties compared to other conventional container materials such as 0.5Bitumen-0.5 Cement, Pb Glass composite, Steel-Magnetite concrete. In addition to its superiority in terms of nuclear safety, it is proposed that the source's simultaneous observation and monitoring, as well as the C6 (TZI8) glass structure's transparency, be underlined as significant advantages. High-density glasses, which may replace undesirable materials such as concrete and lead, provide several advantages in terms of production ease, non-toxic properties, and resource monitoring. In conclusion, the use of Indium (III) oxide-reinforced glass with its high transparency and outstanding protection properties may be a substantial choice in places where concrete is required to ensure the safety of nuclear materials.

Morphological and Gamma-Ray Attenuation Properties of High-Density Polyethylene Containing Bismuth Oxide

For extensive radiation exposure, inventing a novel radiation shielding material is a burning issue at present for the purpose of life saving. Considering this thought, in this study, by adding sundry amounts of Bi₂O₃ into pure high-density polyethylene (HDPE), six HDPE systems were prepared to evaluate the radiation shielding efficiency. These HDPE systems were HDPEBi-0 (pure HDPE), HDPEBi-10 (10 wt% Bi₂O₃), HDPEBi-20 (20 wt% Bi₂O_3-), HDPEBi-30 (30 wt% Bi₂O₃), HDPEBi-40 (40 wt% Bi₂O₃), and HDPEBi-50 (50 wt% Bi₂O₃). The values of the linear attenuation coefficients of the experimental results (calculated in the lab using HPGe) were compared with the theoretical results (obtained using Phy-X software) at 0.060, 0.662, 1.173, and 1.333 MeV energies. To ensure the accurateness of the experimental results, this comparison was made. It was crystal clear that for energy values from 0.06 MeV to 1.333 MeV, all the experimental values were in line with Phy-X software data, which demonstrated the research setup's reliability. Here, the linear attenuation coefficient (LAC), and mean free path (MFP) shielding parameters were assessed. At the energy of 1.333 MeV, sample HDPEBi-0 showed an HVL value 1.7 times greater than that of HDPEBi-50, yet it was 23 times greater at 0.0595 MeV. That means that for proper radiation protection, very-low-energy HDPE systems containing 10-50% Bi₂O₃ could be used; however, the thickness of the HDPE system must be increased according to the energy of incident radiation.

Rhyolite as a Naturally Sustainable Thermoluminescence Material for Dose Assessment Applications

Thermoluminescence characteristics of natural rhyolite have been studied. Dose response at a wide dose range of 0.5–2000 Gy has been determined. Minimum detectable dose and thermal fading rate are evaluated. Glow curve deconvolution is conducted after determining the best read-out conditions. The repeated initial rise (RIR) method is used to detect the overlapping peaks, and a glow curve deconvolution procedure is used to extract the thermoluminescence parameters of rhyolite. According to the findings, rhyolite glow curves show five interfering peaks corresponding to five electron trap levels at 142, 176, 221, 298, and 355 °C, respectively, at a heating rate of 3 °C/s. The obtained kinetic order for the deconvoluted peaks showed mixed-order kinetic. The reported results might be useful to introduce rhyolite as a natural sustainable material for radiation dosimetry applications.

Asphaltene or Polyvinylchloride Waste Blended with Cement to Produce a Sustainable Material Used in Nuclear Safety

The current research uses sustainable methods to preserve the environment, such as exploiting municipal or industrial waste that may harm the environment. The wreckage of polyvinyl chloride (PVC) pipes and asphaltene are used as additives to cement to improve its mechanical properties, while stabilizing the radioactive waste resulting from the peaceful uses of nuclear materials, or enhancing its radiation shielding efficiency. New composites of Portland cement with ground PVC or asphaltene up to 50% are investigated. Fast neutron removal cross-section (ƩR) and gamma shielding parameters, such as mass attenuation coefficient (MAC), half-value layer (HVL), effective atomic number (Zeff), and exposure build-up factor (EBF) at wide energy range and thickness, are determined. The compressive strength and apparent porosity of the examined composites are examined to test the durability of the prepared composites as stabilizers for radioactive waste. The obtained results show that the bulk density of hardened cementitious composites was slightly increased by increasing the additive amount of PVC or asphaltene. The compressive strength of cement composites reached more than 4.5 MP at 50 wt.% PVC and 8.8 MPa at 50 wt.% asphaltene. These values are significantly higher than those recommended by the US Nuclear Regulatory Commission (3.4 MPa). Additionally, the obtained results demonstrate that although the gamma MAC is slightly decreased by adding asphaltene or PVC, the neutron removal cross-section was highly increased, reaching 171% in the case of 50 wt.% asphaltene and 304% in the case of 50 wt.% PVC. We can conclude that cement composites with PVC or asphaltene have optimized radiation shielding properties and can stabilize radioactive waste.

Novel Shielding Mortars for Radiation Source Transportation and Storage

New types of mortar, M1 (60% sand, 25% cement, 10% ball clay, and 15% WO3), M2 (50% sand, 25% cement, 10% ball clay, and 25% WO3), M3 (60% sand, 25% cement, 10% Barite, and 15% WO3), and M4 (50% sand, 25% cement, 10% Barite, and 25% WO3), were prepared and the impact of WO3 and barite on their radiation shielding performance and mechanical properties was evaluated. The radiation attenuation factors were evaluated using five radioactive point sources, and a sodium iodide (NaI) scintillation detector (3″ × 3″) was used to detect the attenuation of gamma ray photons emitted from radioactive sources. The density values of the mortar samples lie within the range of 2.358 and 2.602 g/cm3. The compressive strength and the tensile strength of the prepared mortars increased with the increasing percentage of WO3. The M4 mortar had the highest linear attenuation coefficient (LAC) value. The LAC results demonstrated that adding barite and a high percentage of WO3 into the mortars notably enhanced the radiation shielding performance of the prepared mortar. The relationship between the half value layer (HVL) and the energy is direct, and so was used to calculate the thickness of mortar needed to absorb or scatter half the number of low-energy photons falling on the samples. At 0.06 MeV, the HVL values of the samples were 0.412, 0.280, 0.242, and 0.184 cm for samples M1–M4, respectively. The highest HVL values, obtained at 1.408 MeV, were 5.516, 5.202, 5.358, and 5.041 cm. Thus, a thinner layer of the M4 sample provided comparable attenuation of photons and radiation protection to the thicker M1–M3 samples. The new material is promising as an effective shield of radiation-emitting sources during transportation and long-term storage.

Enhancement of Ceramics Based Red-Clay by Bulk and Nano Metal Oxides for Photon Shielding Features

We prepared red clays by introducing different percentages of PbO, Bi₂O₃, and CdO. In order to understand how the introduction of these oxides into red clay influences its attenuation ability, the mass attenuation coefficient of the clays was experimentally measured in a lab using an HPGe detector. The theoretical shielding capability of the material present was obtained using XCOM to verify the accuracy of the experimental results. We found that the experimental and theoretical values agree to a very high degree of precision. The effective atomic number (Z_eff) of pure red clay, and red clay with the three metal oxides was determined. The pure red clay had the lowest Z_eff of the tested samples, which means that introducing any of these three oxides into the clay will greatly enhance its Z_eff, and consequently its attenuation capability. Additionally, the Z_eff for red clay with 10 wt% CdO is lower than the Z_eff of red clay with 10 wt% Bi₂O₃ and PbO. We also prepared red clay using 10 wt% CdO nanoparticles and compared its attenuation ability with the red clay prepared with 10 wt% PbO, Bi₂O₃, and CdO microparticles. We found that the MAC of the red clay with 10 wt% nano-CdO was higher than the MAC of the clay with microparticle samples. Accordingly, nanoparticles could be a useful way to enhance the shielding ability of current radiation shielding materials.