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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.

IntroductionA lead-acrylic protective screen is suggested to reduce radiation exposure to the unexposed breast during mammography. The presence of toxic lead in its structure may harm the tissues with which it comes in contact. This study aimed to design a CdO-rich quaternary tellurite glass screen (C40) and evaluate its efficiency compared to the Lead-Acrylic protective screen.MethodsA three-layer advanced heterogeneous breast phantom designed in MCNPX (version 2.7.0) general-purpose Monte Carlo code. Lead acrylic and C40 shielding screens were modeled in the MCNPX and installed between the right and left breast. The reliability of the absorption differences between the lead acrylic and C40 glass were assessed.Results and discussionThe results showed that C40 protective glass screen has much superior protection properties compared to the lead acrylic protective screen. The amount of total dose absorbed in the unexposed breast for C40 was found to be much less than that for lead-based acrylic. The protection provided by the C40 glass screen is 35–38% superior to that of the Lead-Acrylic screen. The C40 offer the opportunity to avoid the toxic Pb in the structure of Lead-Acrylic material and may be utilized for mammography to offer superior radioprotection to Lead-Acrylic and significantly lower the dose amount in the unexposed breast. It can be concluded that transparent glass screens may be utilized for radiation protection purposes in critical diagnostic radiology applications through mammography.

In radiology, knowing the X-ray spectrum characteristics makes it possible to estimate the absorbed dose in the patient and to improve image quality. In this study, an X-ray generator was proposed using the MCNPX code and to validate it, the simulated spectrum was compared to the data provided from AAPM Task Group 195, which resulted in a percentage difference of 8.7%. Furthermore, several X-ray spectra were generated and compared to the spectra obtained from commercially available softwares as xpecgen and SpekCalc. The percentage differences were of the order of 13% in comparison with SpekCalc and 8% with xpecgen. The major differences obtained between those spectra were concentrated in the region of characteristic peaks, independently if variations in electron beam energy, target angle or filtration thickness were performed.

IntroductionWe report the breast and chest radiation dose assessment for mammographic examinations using a three-layer heterogeneous breast phantom through the MCNPX Monte Carlo code.MethodsA three-layer heterogeneous phantom along with compression plates and X-ray source are modeled. The validation of the simulation code is obtained using the data of AAPM TG-195 report. Deposited energy amount as a function of increasing source energy is calculated over a wide energy range. The behavioral changes in X-ray absorption as well as transmission are examined using the F6 Tally Mesh extension of MCNPX code. Moreover, deposited energy amount is calculated for modeled body phantom in the same energy range.Results and discussionsThe diverse distribution of glands has a significant impact on the quantity of energy received by the various breast layers. In layers with a low glandular ratio, low-energy primary X-ray penetrability is highest. In response to an increase in energy, the absorption in layers with a low glandular ratio decreased. This results in the X-rays releasing their energy in the bottom layers. Additionally, the increase in energy increases the quantity of energy absorbed by the tissues around the breast.

In this study, three different ilmenite-limonite (ILC) concretes were modeled in terms of determining the effect of filler type and filler size on the enhancement of radiation shielding properties in the 0.142–1.33 MeV gamma-ray energy range. Accordingly, ILC concrete, micro (50 µm) and nano (50 nm) WO3, and micro (50 µm) and nano (50 nm) PbO filler added ILC samples were modeled. MCNPX simulations of micro- and nano-sized particles inside of the ILC sample were performed using lattice (LAT) and universe (U) features of MCNPX (version 2.7.0) code. The greatest increase in mass attenuation coefficient was reported from micro-PbO to nano-PbO filler as 7.88% at 0.142 MeV energy value. The study's findings indicated that decreasing the particle size of the additive material decreases the mean free path and so increases the quantity of gamma radiation interacting per unit distance. Although we validated the consistency of our input against the standard NIST database, several points, such as experimental investigations of nanoparticles in ILC concrete, nanoparticle distributions, and the effect of this distribution on overall shielding enhancement, should be investigated further. [ABSTRACT FROM AUTHOR]

Today, nuclear reactors, especially small reactors, have been significantly welcomed in different countries and have various applications, including use as desalination plants, naval propulsions, etc. During the operation of the reactor, different types of radioisotopes are produced, some of which have very high activity and half-life. The type and amount of radioisotopes produced depends on the type of fuel used, the level of enrichment and the operating time of the reactor. The atomic density and activity of the most important radioisotopes for a typical reactor with two types of high and low enrichment fuel have been calculated in terms of time using MCNPX and CINDER codes. The activity of some of these radioisotopes is high and it is necessary to pay attention to possible environmental hazards. In this research, using the MCNPX Monte Carlo code, the core of a small nuclear reactor proposed as a Virginia class submarine propulsion with a power of 150 MWth was modeled. Using the BURN and KCODE cards in this code as well as the CINDER code, the fuel burnup calculations of the core of this naval reactor have been performed. For this reactor, there are two types of metal alloy fuels, including the high-enrichment UO2-Zr fuel and the low-enrichment U-10Mo fuel, and fuel burnup calculations have been performed for both fuels. After calculations, different types of radioisotopes were identified and changes in atomic density and activity quantities were calculated for the most important of these radioisotopes with high half-life during cycle and compared. [ABSTRACT FROM AUTHOR]

We present elastic moduli, gamma radiation attenuation characteristics, and transmission factor of barium–tungstate–phosphate (BTP) glasses with the chemical formula (60-y)BaO-yWO3-40P2O5, where y = 10 (S1)–40 (S4) in steps of 10 mole%. Different types of mathematical and simulation approaches, such as the Makishima-Mackenzie model, the Monte Carlo method, and the online Phy-X/PSD software, are utilized in terms of determining these parameters. The total packing density (Vt) is enriched from 0.607 to 0.627, while the total energy dissociation (Gt) is enriched by increasing the WO3 content (from 52.2 (kJ/cm3). In the investigated glasses, increasing tungstate trioxide (WO3) contribution enhanced Young's, shear, bulk, and longitudinal moduli. Moreover, Poisson's ratio is improved by increasing the WO3 content in the BTP glasses. The 20BaO-40WO3-40P2O5 sample possessed the highest values of both linear (µ) and mass attenuation (µm) coefficients, i.e., (µ, µm)S4 > (µ, µm)S3 > (µ, µm)S2 > (µ, µm)S1. Moreover, the 20BaO-40WO3-40P2O5 sample had the lowest values of half (HVL) and tenth (TVL) layers, i.e., (half, tenth)S4 < (half, tenth)S3 < (half, tenth)S2 < (half, tenth)S1. The effective atomic number (Zeff) of the studied glasses has the same behavior as µ and µm. Finally, the 20BaO-40WO3-40P2O5 is reported with the minimum values of transmission factor (TF) for all the BTP investigated at a thickness of 3 cm. In conclusion, the sample with composition 20BaO-40WO3-40P2O5 which has the maximum WO3 reinforcement may be a beneficial glass sample, along with its advanced mechanical and gamma ray shielding properties. [ABSTRACT FROM AUTHOR]

Using advanced Monte Carlo simulation techniques and theoretical methodologies, a thorough investigation on the gamma-ray shielding properties of several heavy metal oxide glasses were performed. The general-purpose Monte Carlo code MCNPX (version 2.7.0) was used to simulate gamma-ray transmission to determine fundamental attenuation coefficients. The acquired findings were compared to Phy-X/PSD to confirm that the outputs were consistent. Additionally, other gamma-ray shielding parameters were computed and studied throughout a broad photon energy range of 0.015 MeV–15 MeV. From A to F glass samples, a sharp density increase from 5.99 g/cm3 to 8.9 g/cm3 was found. As a result, the F sample was found to have the highest linear attenuation coefficients. Our results indicate that increasing the amount of Bi reinforcement improved the material's overall gamma-ray attenuation properties. The F sample with the highest Bi reinforcement in its glass structure was subsequently shown to have superior gamma-ray shielding characteristics. Finally, we compared the F sample's half-value layer values to those of other commercial glass shields, various concretes, and other glass shields investigated in the literature. As a consequence of the benchmarking procedure, it has been determined that the F sample has better shielding capabilities than other shielding materials. It can be concluded that heavy metal oxide glasses offer apparent benefits in terms of more efficiently attenuating incoming gamma-rays. Additionally, it can be concluded that applying high Bi to heavy metal oxide glasses is a beneficial strategy for improving the gamma-ray attenuation capabilities of heavy metal oxide glasses.

Mechanical properties, elastic moduli, transmission factors (TFs), and gamma-ray shielding performance of quaternary glass systems with chemical composition (0.25−x)Bi2O3–xB2O3−0.75(50%P2O5−50%V2O5), where x = 0.05 (S1), 0.10 (S2), 0.15 (S3), and 0.20 (S4) mol%, were comprehensively studied. The MCNPX code, Phy-X/PSD software, and the Makishima–Mackenzie model were utilized to achieve the mentioned purposes. The values of the packing density (V t) decreased from 0.634432 to 0.600611, while those of the dissociation energy (G t) increased from 51.6125 kJ/cm3 for the S1 glass sample (with Bi2O3 = 5 mol%) to 56.7525 kJ/cm3 for the S4 glass sample (with Bi2O3 = 20 mol%). This means that the mechanical properties were enhanced by increasing the Bi2O3 content in glasses. Linear (µ) and mass attenuation (µ m) coefficients for the S4 glass sample were the greatest compared to those for glass materials investigated, i.e., (µ, µ m)S1 < (µ, µ m)S2 < (µ, µ m)S3 < (µ, µ m)S4. Half- and tenth-value layers (HVL and TVL, respectively) follow the trend: (HVL, TVL)S1 > (HVL, TVL)S2 > (HVL, TVL)S3 > (HVL, TVL)S4. The effective atomic number (Z eff) of investigated glasses has the same trend as of linear and mass attenuation coefficients. Our findings indicate that increasing the amount of Bi2O3 reinforcement decreased the exposure buildup factor and energy absorption buildup factor values for all mean free path values (0.5–40 mfp). All glasses recorded the minimum TF values at a thickness of 3 cm. The findings would benefit the scientific community in determining the most appropriate additive bismuth(iii) oxide/diboron trioxide type and related glass composition to provide the shielding properties previously mentioned in terms of needs and utilization requirements, as well as the most suitable glass composition.

Proton beam therapy (PBT) is a modern radiotherapy technique characterized by superior target coverage compared to conventional modalities. In this work, a comprehensive GATE Monte Carlo model was developed and then validated for a double scattering proton treatment nozzle. To this aim, a double scattering treatment nozzle was modeled in the GATE toolkit. Proton beam flatness and its symmetry, secondary neutron effective dose, and dosimetric performance were characterized. A proton beam flatness of 98.6% was observed downstream of the aperture for a 7×7 cm2 field size. The beam flatness deteriorates at the edge of the treatment field for the single scattering model while it remains approximately constant for the double scattering one. Compared to the single scattering delivery, the second scattering model results in a 1.3 times increase in neutron dose for the nickel as the optimal collimator/aperture material. Furthermore, a flat beam modulation width of 3.50 cm is formed with a distal edge at 7.86 cm in water using the GATE and MCNPX codes. The GATE model agreed with the MCNPX results. The results show that the constructed GATE model results in a fast and accurate simulation of passive scattering PBT.

This study comprehensively analyzes the gamma radiation shielding, mechanical, and acoustic properties of novel glass composites formulated from B 2 O 3 , SiO 2 , and Gd 2 O 3. Utilizing the MCNPX simulation code and Phy-X: PSD software, key parameters such as Linear Attenuation Coefficient, Half Value Layer (HVL), Mean Free Path (MFP), and Effective Atomic Number (Z eff) were meticulously evaluated for three distinct glass compositions, denoted as GL-1, GL-2, and GL-3. The investigation revealed that incorporating Gd 2 O 3 and SiO 2 notably enhances the radiation shielding efficiency of these glasses, which is evident from the decreasing trends in HVL and MFP values across the series. Employing the Makishima and Mackenzie (MM) model, this study further delved into the mechanical and acoustic characteristics of the glass samples. An increase in Gd 2 O 3 and SiO 2 content, substituting B 2 O 3 , was observed to augment the bond dissociation energy and adjust the packing density, consequently improving the elastic moduli. These mechanical enhancements were quantified through measurements of Young's modulus, bulk modulus, shear modulus, and longitudinal modulus. Acoustic properties were also calculated, including Longitudinal and Transverse velocities, mean velocity, and acoustic impedance. These parameters demonstrated a consistent improvement correlating with the increasing rigidity and mechanical strength of the glass samples, particularly for the GL-3 composition. The study concludes that the GL-3 glass sample exhibits superior performance regarding gamma photon attenuation, mechanical robustness, and acoustic properties. This underscores its potential as an effective material for radiation shielding in various industrial applications, benefiting from its enhanced mechanical and acoustic features. • GL-3 glass excels in gamma shielding, with HVL at 3.706 cm. • Increased Gd2O3 boosts GL-3's mechanical strength notably. • GL-3's acoustic properties outshine GL-1 and GL-2 series. • Study finds GL-3 ideal for industrial radiation safety. • GL-3 shows superior performance in shielding and strength. [ABSTRACT FROM AUTHOR]

We investigated the shielding properties of four selected glass samples among the 66 samples. These glass samples had the following compositions and densities: S1 (55TeO 2 +45Ag 2 O, density 6.28 g/cm3), S2 (55TeO 2 +5Tl 2 O+40Ag 2 O, density 6.37 g/cm3), S3 (50TeO 2 +15Tl 2 O+35Ag 2 O, density 6.49 g/cm3), and S4 (40TeO 2 +55Tl 2 O+35Ag 2 O, density 7.20 g/cm3). These glasses exhibit high density. We compute the mass attenuation and linear attenuation coefficients for these samples using MCNPX simulations and Phy-X software between 0.001 MeV and 15 MeV energy range. The derived mass attenuation coefficient (MAC) enables the calculation of key radiation shielding metrics. The tenth value layer, the half value layer (HVL), the mean free path (MFP), and the effective atomic number (Z eff) were also determined. Values of LAC values for S1, S2, S3 and S4 glasses span 244.91 to 0.24932 cm⁻1, 319.52 to 0.2625 cm⁻1, 447.24 to 0.28491 cm⁻1 and 379.480 to 0.273 cm−1, respectively. Correspondingly, HVL values range from 0.0028 to 2.78005 cm, 0.0021–2.6403 cm, and 0.00154–2.4327 cm and 0.002135–2.917403 cm for S1, S2, S3 and S4 glasses, respectively. Our findings reveal that results from both MCNPX and Phy-X exhibit consistency, supported by computed relative differences. Notably, the glass sample with a relatively higher TlO 2 concentration demonstrates superior shielding properties across the entire energy spectrum. Importantly, these glass samples can significantly lower the strength of the x-rays and γ-rays. • The radiation shielding properties of Te glasses based on Tl and Ag were studied. • All shielding parameters were measured within the energy range of 0.001–15 MeV. • All samples exhibited an increase in Z eff. • The sample with a higher concentration of Tl showed superior shielding properties. • These glasses are promising candidates for radiation shielding materials. [ABSTRACT FROM AUTHOR]

One of the important factors of a nuclear reactor core is the fast fission factor. This paper calculates this parameter based on space and energy-dependent method using the PTRAC card of MCNPX code. Tehran research reactor (TRR) is taken as a case study, and the parameter analyses are performed on the reactor core. Fast fission factor in TRR is evaluated regarding temperature effect, control rod positions, and fuel assembly positions. Using the PTRAC card, helpful information on fast fission factors is achieved throughout the reactor core. One MCNPX runs to return a data file about neutron interaction that can be analyzed many times in different manners to reveal this useful information. The method is simple and can be applied to any nuclear reactor core. The results obtained by this method can help nuclear reactor designers and nuclear reactor fuel managers to have a precise evaluation of the parameter. The method proposed in this paper for fast fission factor calculation is compared with the results previously published in the literature.

Structural properties of materials irradiated with gamma-rays, such as mechanical and physical properties, may be modified or reduced depending on the displacement damages. Apart from increasing or decreasing that property, damage may be desirable or not. This study investigates the effects of sterilization by gamma radiation on polycaprolactone (PCL) scaffolds. Using a code-based simulation method, MCNPX provides information on primary knock-on atoms, or PKAs, that cause damage. A program has been developed called GAMMATRACK to access PKA information. These PKA data can be used as input for the SRIM code to analyze gamma damage systematically. The rate of damage caused by gamma radiation is calculated on the PCL target. The theoretical calculation method also has been used to confirm the results of the Monte Carlo method (MCNPX + SRIM code). Due to the low-energy PKAs and the thin target possibility of the displacement cascade can be ignored. It is realized that all displacements are due to single vacancies. The total number of hydrogens, carbons, and oxygen PKAs were obtained. It was found that the number of carbons PKA is more than the others, which causes three-dimensional polymer networks to be created. In the experimental analysis, it is necessary to know the appropriate depth of the sample for damage investigation. GAMMATRACK gives the gamma displacement damage graph along the length of the PCL. It shows a uniform distribution of displacement damage. The damage rate for the PCL target is calculated, and the results between the theoretical calculations and the Monte Carlo method (MCNPX + SRIM code) differ by about 17%.

We examine the nuclear safety properties of a newly designed cadmium oxide-rich glass container for nuclear material to a bitumen-reinforced concrete container. Individual transmission factors, detector modelling, and energy deposition (MeV/g) in the air are calculated using MCNPX (version 2.7.0) general purpose Monte Carlo code. Two container configurations are designed with the material properties of cadmium dioxide-rich glass and Concrete + Bitument in consideration. First, individual transmission factors for 60Co and 137Cs radioisotopes are calculated. To evaluate potential environmental consequences, energy deposition amounts in the air for 60Co and 137Cs are also determined. The minimum gamma-ray transmission rates for two container types are reported for a cadmium dioxide-rich glass container. In addition, the quantity of energy deposition is varied depending on the container type, with a lower value for cadmium dioxide-rich glass container. The 40% cadmium dioxide-doped glass container provides more effective safety than the Cement + Bitumen container, according to the overall findings. In conclusion, the utilization of cadmium dioxide-doped glass material along with its high transparency and advanced material properties may be a significant and effective option in areas where concrete is required to assure the safety of nuclear materials.

With the advancement of radiotherapy techniques, we find some challenges in small field dosimetry that are widely used in head and neck treatments, so computer simulations with the Monte Carlo method, already well established in medical physics, are a great tool for studying small field dosimetry. The present work aims to report the validate of the geometry model used for the simulations of a 6 MV LINAC beam, in addition to estimating the TPS, PDD and PDD 20,10 curves for several regular and small fields, allowing comparisons with experimental data .

In this paper, graphene/h-BN metamaterial was investigated as a new neutron radiation shielding (NRS) material by Monte Carlo N-Particle X version (MCNPX) Transport Code. The graphene/h-BN metamaterial are capable of both thermal and fast neutron moderator and neutron absorber process. The constituent phases in graphene/h-BN metamaterial are chosen to be hexagonal boron nitride (h-BN) and graphene. The introduced target was irradiated by an Am–Be neutron source with an energy spectrum of 100 keV to 15 MeV in a Monte Carlo simulation input file. The resulting current transmission rate (CTR) was investigated by the MCNPX code. Due to concrete's widespread use as a radiation shielding material, the results of this design were also compared with concrete targets. The results show a significant increase in NRS compared to concrete. Therefore, metamaterial with constituent phase's graphene/h-BN can be a suitable alternative to concrete for NRS. [ABSTRACT FROM AUTHOR]

This research aims to investigate the structural and gamma shielding properties of different resin composite dental materials for medical applications. For this study, morphological structures and elemental compositions of the investigated composite samples were obtained by SEM/EDS analysis. The FTIR spectroscopy was employed to understand the bond structures of the resin composites. In addition to structural analyses, some of vital gamma radiation attenuation parameters were determined using MCNPX general purpose Monte Carlo code (version 2–6-0). The results showed that 3 M, GC, and Tokuyama composites indicated more regular and homogeneous structure. Zn, Zr, Ba, Sr, and Yb heavy elements were added in certain proportions in order to increase the radiopacity of the restorative material in the studied samples. As FTIR results were interpreted, C–O, O–H, and Si–O bonds were observed in all of the composites. Moreover, Bisco was found to be the most successful in reducing radiation intensity among the others, while Tokuyama was the least effective material to attenuate photon intensity. The elements Zn and Yb in Bisco have increased radiopacity and radiation absorption capabilities. It can be concluded that outcomes of the present investigation would be useful for ionizing radiation application in medicine especially in diagnostic imaging and radiation treatment. [ABSTRACT FROM AUTHOR]

In the ongoing research, the Monte Carlo simulation of three different Barite Concrete samples coded as C-B0, C-B50 and C-B100 was estimated via MCNPX Monte Carlo Code to investigate the neutron-gamma photon shielding characteristics. Simulations were performed for 252Cf source which is surrounded by the polyethylene cylinder. Using Watt Fission (WF) distribution and Doppler Effect (DE), the neutron-gamma photon flux was obtained and was compared to various types of Barite Concrete samples. The total macroscopic cross section, linear attenuation coefficient (µ), Half Value Thickness (HVT) and Mean Free Path (MFP) were calculated for gamma rays and neutrons. The equivalent absorbed dose rate was estimated for neutrons and showed that an increment of the Barite rate causes a reduction in the equivalent absorbed dose rate from 0.0003572 to 0.0003542 (µSV/h). Moreover, total macroscopic cross section which shows total macroscopic cross section which shows the neutron attenuation coefficient (∑t) undergoes descending trend of (∑t)C-B0 < (∑t)C-B50 < (∑t)C-B100 which confirms the disability of the Barite Concrete sample against neutrons. The evaluated gamma photon attenuation parameters for energy ranging from 0.05 to 11 MeV are shown graphically. Findings reveal that by increasing the rate of the Barite, the samples' shielding ability will reduce against neutron whereas it will enhance the gamma photons shielding capability. [ABSTRACT FROM AUTHOR]

Purpose: In the current study, we aimed to look into the macroscopic and microscopic dose enhancement effect of metallic nanoparticles in interstitial brachytherapy of gastric adenocarcinoma by Iodine-125 source using a nano-lattice model in MCNPX (2.7) and MCNP6.1 codes. Materials and methods: Based on a nano-lattice simulation model containing a radiation source and a tumor tissue with cellular compartments loaded with 7 mg/g spherical nanoparticles, the microscopic and macroscopic levels of energy deposition by the secondary electrons was estimated. Results: The results show that the values of macroscopic DEF are higher than microscopic DEF values and the macroscopic DEF values decrease by increasing the distance from the surface of brachytherapy source. Accordingly, it could be noted that gold nanoparticles have the highest radiosensitization effect among the other nanoparticles and the related DEF value is close to the resultant DEF values for bismuth nanoparticles. Moreover, the results revealed a remarkable discrepancy between the DEF and secondary electron spectra calculated by MCNPX (2.7) and MCNP6.1 codes, which could be justified by the difference in energy cut-off and electron transport algorithms of two codes. Conclusions: According to the both MCNPX (2.7) and MCNP6.1 outputs, it could be concluded that the presence of metallic nanoparticles in the tumor tissue of gastric adenocarcinoma increases the physical effectiveness of brachytherapy by I-125 source. This study aims to provide recommendations for future preclinical studies. Actually, the results presented herein give a physical view of radiosensitization potential of different metallic nanoparticles and could be considered in design of analytical and experimental radiosensitization studies in tumor regions using various radiotherapy modalities in the presence of heavy nanomaterials. [ABSTRACT FROM AUTHOR]

The Holmium-166 radionuclide is one of the most effective radionuclides used to treat bone marrow cancer and rheumatoid arthritis. Among the recommended radionuclides used in radiation synovectomy, 166Ho has got much attention due to suitable decay properties such as short half-life, its high beta energy, gamma-ray emission with suitable energy for nuclear imaging, and the possibility of large-scale production in medium flux reactor. One method to deliver 166Ho to the target tissue is via the 166Dy/166Ho-Chitosan in vivo generator. Compared with other similar radiopharmaceuticals, using the in vivo generator to deliver 166Ho, causes minimal non-target tissue exposure and increased absorbed dose in the target tissue. In this work, the absorbed dose of 166Dy/166Ho-Chitosan radio-complex for radio-synovectomy purposes was calculated by GEANT4 and MCNPX. The obtained results were compared with each other. In addition, the dosimetry results of the mentioned radio-complex have been compared with the common radio complexes used for radio-synovectomy.

In this article, a criticality and depletion analysis of the European Lead-Cooled Training Reactor (ELECTRA), a low-power, compact, fast lead-cooled reactor, was performed. The active core model and simulations were made by using the MCNPX code. First, the model is compared with the reference and the influence of the reflector radius on the system is analyzed. Secondly, it was simulated the fuel combustion after 30 years of continuous operation of the ELECTRA, managing to evaluate the transmutation of a reprocessed fuel in a compact lead-cooled reactor.

The nuclear spectroscopy method has long been used for advanced studies on nuclear physics. In order to decrease costs and increase the efficiency of nuclear radiation investigations, quick and efficient solutions are required. The purpose of this research was to calculate the whole energy peak efficiency values for a range of gamma-ray energies, from 30.973 keV to 1408 keV, at various source-detector distances using the MCNPX Monte Carlo code, which is extensively used in nuclear medicine, industry, and scientific research. As a result, the modeled detectors' full-energy peak efficiencies were calculated and compared to both experimental data and Monte Carlo simulations. Experiment results and prior studies using Monte Carlo simulations were found to be very consistent with these results. The counting efficiency against source-detector distance is then calculated using the modeled detectors. The data we have show that LaBr3(Ce) has outstanding detection properties. This study’s findings might be used to improve the design of detectors for use in wide range of high-tech gamma spectroscopy and nuclear research applications.

We report the gamma-ray shielding properties of five different lithium silicate glasses based on the (40 − x) Li2O–10Pb3O4–50SiO2 nominal composition. Transmission factor values and some basic shielding parameters such as linear (µ) and mass attenuation coefficients (µ/ρ), half-value layer, tenth value layer, and mean free path (MFP) values of the investigated glass samples are determined in a large photon energy range. Using the G–P fitting method at various MFP values, the exposure buildup factor and energy absorption buildup factor values of the examined glasses are also calculated. Based on the findings, it can be concluded that the S5 glass specimen, which exhibits the greatest Ag2O additive and density among the various glass samples, represents a favorable choice for the purpose of shielding against gamma radiation.

Radiations of atomic/nuclear particles during earthquakes have been detected and verified by many experiments. Some tests also verify some radiations during the fracture of small-size piezoelectric rocks in the laboratories. Some researchers studied the mechanism of such reactions and considered the photonuclear reaction as a possible mechanism for such phenomena. In this study, to understand the dominant mechanisms of radiation of particles before or during the earthquakes and also to find some relationships between applying the mechanical energy and the mentioned radiations in small and giant scales of piezoelectric rocks, we first analyzed the piezoelectricity equations in these rocks. Then, we performed an energy analysis of the earthquakes, considering the amount of energy stored in a block of rock during applying the uniaxial stress. Thereafter, MCNPX simulation code is employed to simulate the tracks and energies of the created particles in a small piece of quartz and in three giant granite blocks. We have found that in the small piece of quartz, the atomic interactions cause low-energy electrons' and photons' radiations, and no nuclear interactions have been observed. However, in giant granite blocks, in addition to the atomic interactions, the photonuclear reactions play big roles, resulting in the creation of high-energy photons, neutrons and protons. Furthermore, some equations have been extracted to relate the mechanical energy stored in a piezoelectric rock and the number and energy of the created atomic/nuclear particles. The outcome of this research would be very beneficial in understanding the earthquake mechanism and its prediction. [ABSTRACT FROM AUTHOR]

Introduction: In this research, a new material (Ti2V0.7Cu97.3) was proposed for the target of medical linear accelerators (linacs) to reduce the production of unwanted photoneutrons in the radiotherapy. So, the fluence, dose equivalent and kerma of the photoneutrons were calculated in a soft tissue phantom. Materials and Methods: The medical linac was the Varian 2100 C/D 18 MV, which its tungsten target was replaced with a new multi-metal target (Ti2V0.7Cu97.3). Desired quantities were computed in a ICRU soft tissue phantom, using the Monte Carlo code MCNPX (v. 2.6). Results: The ratio of the maximums of fluence, kerma, and dose equivalent of photoneutrons along the central axis of the ICRU phantom with new target rather than tungsten target were 72 %, 59 % and 61 %, respectively. Average of the Ratio of fluence, kerma, and dose equivalent in inner area (distances less than 5 cm from central axis) at different depths of the phantom with new target rather than tungsten target were 78 %, 70 % and 75 %, respectively. Uncertainties at all points were less than 5 % (except for a few points which were less than 10 %). Conclusion: This work showed that applying Ti2V0.7Cu97.3 alloy for the target of linac, can reduce the produced photoneutrons up to 38 % by an applicable and inexpensive way. [ABSTRACT FROM AUTHOR]

The direct effects of tungsten trioxide (WO3) on gamma radiation attenuation properties of tungsten/barium/phosphate glasses with chemical form xWO3–(50-x/2)BaO–(50-x/2)P2O3, x = 0 (S1)–60 (S7) in steps of 10 mol% has been investigated. To this end, Phy-X/PSD software and Monte Carlo code were applied. The increased amount of WO3 in the glass composition, on the other hand, contributed positively to the increase in density and radiation sensing properties. At the lowest and highest WO3 contributions, a nearly twofold increase in HVL value was seen for the S1 and S7 glass samples, respectively, with S7 having the lowest HVL values. The HVL has obeyed the trend as: (HVL)S1 > (HVL)S2 > (HVL)S3 > (HVL)S4 > (HVL)S5 > (HVL)S6 > (HVL)S7. The variation of the tenth value layer (TVL) for all studied glasses has the same trend of HVL. In terms of mean free path (MFP), there was negative effect of WO3 concentration on the trend of MFP. Consequently, S7 glass sample has the lowest values of MFP, while the S1 glass sample has the highest values. Therefore, (MFP)S1 > (MFP)S2 > (MFP)S3 > (MFP)S4 > (MFP)S5 > (MFP)S6 > (MFP)S7. A comparison of the HVL of S7 glass sample with some commercial radiation shielding materials such as different types of concrete and RS-253-G18 glasses has been performed and concluded that the S7 glass sample is superior as radiation shielding material than several commercial materials. It can be concluded that WO3 reinforcement serves a multipurpose of increasing the density and hence the gamma-ray-shielding characteristics of comparable glass compositions. [ABSTRACT FROM AUTHOR]

In this work, Esfahan light water subcritical reactor (ELWSCR) is analysed using experimental and theoretical diagnostic methods. Important neutronic parameters of the system such as prompt neutron lifetime, delayed neutron fraction, prompt neutron decay constant, negative reactivity of the core, fuel and moderator temperature coefficient of reactivity, and overall and local void coefficient of reactivity are estimated. Also, neutron flux distribution, reflector saving, water level effect, and lattice pitch of the core including operating point of the facility are studied in details. Theoretical results are calculated by MCNPX and measurements are performed utilizing zero power reactor noise method. Detailed descriptions of the results are explained in the text.

In this study, a promising relationship between antibacterial zone and radiation attenuation properties was investigated in Ag2O doped bioactive glasses with a chemical composition of xAg2O–20Li2O–25TeO2-(55-x)B2O3 (where x = 0, 0.5, 1.0, 1.5, and 2.0 mol%). For this aim, a wide-ranging radiation attenuation characterization procedure was performed on Ag2O substituted bioactive glasses. The general-purpose Monte Carlo code MCNPX (v.2.7.0) was used to model bioactive glasses. The mass attenuation coefficients were calculated using a gamma-ray transmission setup. The coefficients obtained were used to determine other important attenuation properties. Finally, for particular behaviors, exposure (EBF) and energy absorption (EABF) build-up factors were calculated for specific attitudes of Ag2O substitutions in bioactive glasses during the interaction process. The results showed that there is direct relationship between Ag2O substitution amount and radiation attenuation properties. In addition to its well-behaviors on inhibition zone against bacterial occurrences, it can be concluded that increasing Ag2O would increase the gamma-ray attenuation properties of studied bioactive glass system.

This study aimed to determine the contribution of B2O3/CdO substitution on gamma-ray attenuation behaviors of 46V2O5–46P2O5–(8−x)B2O3–xCdO (x = 0–8 mol%) glass system. Accordingly, attenuation coefficients along with half and tenth value layers of five different samples were determined in 0.015 MeV–15 MeV photon energy range. Moreover, effective atomic numbers and effective atomic weight along with exposure and energy absorption buildup factors were determined in same energy range. The result showed that B2O3/CdO substitution has a direct effect on behaviors of studied semiconducting oxide glasses against ionizing gamma-rays. Our findings showed that increasing CdO reinforcement has an obvious impact on gamma-ray attenuation properties especially in the low energy range, where photoelectric effect dominates the photon–matter interaction. Moreover, half-value layer, mean-free path and tenth value layer also decrease with an increase in the content of CdO in the composition. Consequently, VPBCd8 sample with 8% mole CdO additive was reported with the minimum half-value layer, the mean-free path, tenth value layer exposure build-up factor and energy absorption build-up factors. The outcomes would be useful for scientific community to observe the most suitable substitution type along with related semiconducting oxide glass composition to provide the aforementioned shielding properties in terms of needs and utilization requirements.

In this paper, multi layer radiation shield is designed for electrons and protons space environments to protect in all satellite orbits. The designed shield layer material and thickness is analyzed and optimized by MCNPX linked with Matlab. In Matlab, genetic algorithm is employed for optimization where the cost function is obtained from MCNPX. The designed shield achieves more than 53% TID reduction for proton environment and more 72% TID reduction for electron environment compared with aluminum by similar thickness. This good specifications prove the capability and ability of the deigned multilayer shields to protect the electronic parts in the satellite applications.

To investigate the 106Ru dose distribution in an intraocular pseudotumor by micro silica glass beads thermo-luminescence dosimeters (TLD). Dosimetry was conducted using Monte Carlo (MC) simulation as well as an experimental setup by employing an in-house eye equivalent phantom. A BEBIG CCA model of 106Ru plaque was used. Bead TLDs were modeled using MCNPX code, choosing water and plexiglass as phantom mediums along with CCA plaque. The deposited energy and electron energy spectra were calculated using *f8 and f4 tallies, respectively. A dedicated eye-equivalent phantom was designed and fabricated by 3D-printing techniques to place glass beads. Absorbed dose at the central axis of the CCA plaque by beads TLDs was compared to manufacturer-reported data and simulation results. Maximum, minimum, and mean relative differences of 23%, 7%, and 15% with the manufacturer and 14%, 0%, and 1.6% with the simulation were obtained, respectively. Replacing glass bead material with water resulted in an average relative difference of 33.9% in absorbed energy per mass unit, corresponding to 21% and 44.8% at the first (0.5 mm depth) and last dosimetry points (9.5 mm depth), respectively. It has also been shown that glass beads can shadow each other, leading to a dose reduction of 15%. A specific eye equivalent phantom with a pseudotumor was constructed, and silica glass beads were used for dosimetry. The results of the experiment and the MC simulation are in good accord. In the case of 106Ru ophthalmic brachytherapy plaque dosimetry, replacing glass beads material with water led to a mean difference of about 34% as well as shadowing effects. This emphasizes how crucial it is to employ a reliable correction factor in clinical applications. • Creating an eye equivalent phantom capable of introducing different materials as tumour foundation. • Using TLD glass beads as a dosimeter in ocular plaque dosimetry. • Monte Carlo analysis of glass beads' material as a dosimeter. • Glass beads vs. water show 33.95% dose difference in CCA plaque dosimetry, highlighting correction factor importance. [ABSTRACT FROM AUTHOR]

This study investigates the mechanical as well as shielding properties of 70P 2 O 3 –5Al 2 O 3 -(25-x)Na 2 O-xPbO, while using advanced MCNPX simulations code and Phy-X software. Advanced MCNPX coding simulation and online Phy-X software are used to compare the properties of three aluminum lead phosphate glasses named S1, S2 and S3 with PbO concentrations 0, 10, and 20 mol%, respectively. Various shielding parameters were calculated in the photon energy ranging from 0 to 15 MeV. These parameters include mass attenuation coefficient, effective atomic number, half-value layer, tenth-value layer, mean free path and linear attenuation coefficient. Our comparative analysis of the results from MCNPX and Phy-X in the case of mass attenuation coefficient indicates a high level of agreement, differing by less than 2% from each other. The half-value layer for S3 was found at 0.003 cm, 1.709 cm, and 5.984 cm at energies of 0.01 MeV, 0.5 MeV, and 15 MeV, respectively, while for S1 these values were found at 0.010 cm, 3.35 cm, and 13.96 cm in the same energy spectra. The results show that S3, with 20% PbO, is more likely to attenuate the gamma and X-ray photons; due to their minimum values of tenth value layer, half value layer, mean free path, and greater mass attenuation coefficient, effective atomic number as compared to S1 (undoped) and S2 (10% doped with PbO). The mass attenuation coefficient for S3 was found between 67.95 cm2/g to 0.030 cm2/g, while for undoped S1, these values varied between 27.64 cm2/g to 0.021 cm2/g in the energy range of 0–15 MeV. Furthermore, the Makishima-Mackenzie model is utilized to evaluate the mechanical behavior of our glass samples, influenced by factors such as energy of dissociation and packing density related to components of oxide. The mechanical features encompass various moduli, Poisson's coefficient and microhardness. The introduction of PbO in glass samples S2 and S3 increased all these properties, excluding Poisson's coefficient which decreases with the addition of PbO. Our findings provide valuable insights into the shielding and mechanical parameters of these glasses, suggesting probable applications in fields such as optics and nuclear physics. • The MCNPX simulations code and Phy-X program gave the same result. • The Shielding properties of the studied samples improved with the increasing PbO content. • The increasing PbO content resulted in enhanced mechanical properties of the investigated glass systems. [ABSTRACT FROM AUTHOR]

In this study, three heavy metal oxide glasses (A :46.0PbO-42.0Bi 2 O 3 -12.0Ga 2 O 3, B: 45.94PbO-42.66Bi 2 O 3 -10.0Ga 2 O 3 -1.4BaO, C: 72.8PbO-17.0GeO 2 -10.2Ga 2 O 3) were synthesized to determine their optical and gamma-ray shielding properties in terms of assessing their potential applications in medical and industrial radiation facilities. Glasses were synthesized using melt quenching method. The optical band gap energy is calculated by the absorption spectrum measured at room temperature. We found a large band at 500 nm that refers to Bi+ions and appears to samples A and B that contain Bi 2 O 3 in their compositions. Optical band gap energies were reported as 2.014 ev, 2.055 eV and 2.430 eV for A, B and C samples, respectively. Next, fundamental gamma-ray parameters were also determined using MCNPX general Monte Carlo code and Phy-X/PSD in 0.15–15 MeV photon energy. Our findings clearly showed that the B sample, which includes the highest concentration of Bi 2 O 3 , has a considerable advantage in terms of gamma-ray attenuation. Moreover, the results also showed that sample B has significantly higher attenuation properties than shielding concretes and several glass shields. It can be concluded that Bi 2 O 3 is a useful component for heavy metal oxide glasses in terms of improving gamma-ray shielding capabilities for radiation shielding applications. [ABSTRACT FROM AUTHOR]

Purpose: The aim of this research was to calculate the fluence, dose equivalent (DE), and kerma of thermal, epithermal and fast photoneutrons separately, within ICRU soft-tissue-equivalent phantom in the radiotherapy treatment room, using MCNPX Monte Carlo code. Materials and Methods: For this purpose, 18 MV Varian Linac 2100 C/D machine was simulated and desired quantities were calculated on the central axis and transverse directions at different depths. Results: Maximum fluence, DE and kerma of total photoneutrons on central axis of the phantom were 43.8 n.cm-2.Gy-1, 0.26, and 3.62 mGy.Gy-1, at depths 2, 0.1, 0.1 cm, respectively. At any depth, average of fluence, DE and kerma in the outer area of the field were less than the inner area and in general were about 72%, 52%, and 45%, respectively. Conclusion: According to this research, within the phantom; variation of fluence, DE and kerma in transverse direction were mild, and along the central axis at shallow area were sharp. DE of fast photoneutrons at shallow and deep areas were one order of magnitude greater than thermal photoneutrons.

Radiation damage is one of the aging important causes in nuclear reactors. Radiation damage causes changes in material properties. In this study, this effect has been evaluated and analyzed on the clad of the Tehran research reactor (TRR). A grade 6061 aluminum is used as a clad in the TRR. The MCNPX code is used to designate the most sensitive location of the reactor and calculate neutron flux distribution. Then, a software using FORTRAN language programming is developed to process the particle track (PTRAC) output file of the MCNPX code. The SRIM code is used here to calculate the rate of displacement per atom. Moreover, the SPECOMP and SPECTER codes are also applied to estimate the displacement rate and compared with the results attained using the SRIM code. The rate of displacement per atom by the SPECTER and SRIM codes have been obtained 2.54×10−7 dpa/s and 2.44×10−7dpa/s (QD method), respectively. Also, the mechanical properties have been evaluated using the RCC-MRx code and have been compared with experimental results. Finally, the change in the matter specification has been analyzed as a function of time.

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

Nowadays, multiphase fluid flowmetry is of great importance in the industry. The most important parameter required for multi-phase fluid flowmeter is the volume fraction of each phase. In this study, three flow regimes (annular, stratified and homogeneous) for two-phase (liquid–gas) as well as three-phase (liquid–liquid–gas) to determine the volume fraction of phases with use gamma-ray attenuation were simulated by MCNPX code. The obtained mean relative error is about 0.02 for determination of volume fraction in best set-up. It is important to note that in this study, the small error value was obtained only by reducing the effect of the buildup factor on the Lambert–Beer relation without using artificial intelligence. It is also worth noting that the simulations and experiments of this study were performed using only a transmission detector (NaI) and a Gamma source Cs-137 (0.662 keV), for two-phases, and two Gamma source Cs-137(0.662 MeV) and Am-241(0.0595 MeV), for three phases. [ABSTRACT FROM AUTHOR]

The influence of the casing and liquid inside the borehole in neutron porosity tool response is analyzed. For this, MCNPX simulation of neutron tool inside of a slim borehole was performed. The results showed that the presence of PVC casing and liquid inside the borehole reduces the tool response in about 73% for low porosities, and 90% for high porosities. [ABSTRACT FROM AUTHOR]

The accuracy of each simulating beta-voltaic battery parameter is very important, especially in micro-batteries. The aim of this study is to improve the calculation of beta-battery parameters’ accuracy. For this purpose, at first, using the Monte Carlo N-Particle code (MCNPX), the energy accumulation distribution of the 63Ni beta particle spectrum inside a silicon semiconductor has been simulated. Then, the ATLAS C-Interpreter function in C ++ was defined, using one of the SILVACO code abilities (the parameter F.RADIATE BEAM statement). Finally, the device electric parameters have been obtained using ATLAS-SILVACO based on the location-dependent of MCNPX results. For validation, the calculations were performed for a battery sample made of 16 mm2 cross-section and 1 mCi activity of radioisotope 63Ni as a source, and finally, the results were compared with one experimental result and two analytical methods. The calculations repeated for the other sample with 100 mCi activity and 1 cm2 of geometry, and compared its results with an analytical method results. The results showed that the simulation of micro-battery characteristics by the MCNPX-SILVACO hybrid code using three-dimensional electron-hole pairs’ distribution in semiconductor and the full spectrum of beta particles creates a significant increase in the accuracy of the computation, and provides a good capability to optimize the design of the battery.

محصورسازی چشمه‌های تابش یکی از مهمترین روش‌های کنترل خطرات بهداشت محیط برای جلوگیری از آسیب‌‌‌رسانی به انسآن‌هاست. معمولاً از اتاقک سربی با دیوارهایی از آجر سربی برای محصورسازی چشمه‌های دارای خطر پرتوزایی متوسط و کم استفاده می‌شود. در این پژوهش، بیشترین میزان فعالیت‌ مجاز چشمۀ میله‌ای کبالت‌ـ60 به دو روش تحلیلی و مونت‌کارلویی با استفاده از کد شبیه‌ساز MCNPX، طوری محاسبه شد که کار با این میزان از فعالیت در داخل یک اتاقک سربی با دیوار‌هایی از آجرهای سربی استاندارد به ضخامت 5 سانتی‌متر، توانست استانداردهای پذیرفته‌شده بین‌المللی برای کار با اتاقک سربی در نواحی محصورشده را برآورده کند. اختلاف بین نتایج روش تحلیلی و مونت‌کارلو به طور میانگین کمتر از 16 درصد به دست آمد. نتایج نشان داد که برای کار در وجه‌های کاربردی پرتودهی/ذخیره و نمونه‌گذاری/نمونه‌برداری و همینطور انتقال، باید میزان فعالیت چشمه در روش تحلیلی به ترتیب از 5.61 و 48.25 میلی‌کوری و در روش مونت‌کارلو از 6.67 و 56.36 میلی‌کوری کمتر باشد. این روش تحلیلی با تقریب نسبتاً خوبی داده‌های حاصل از کد قدرتمند MCNPX را تولید کرد.

Prompt Gamma Ray Neutron Activation Analysis (PGNAA) is a useful nondestructive method with numerous applications. In this study neutron yield from 7Li(p,n) reaction with proton energies of 2.5, 3, 4, 4.5 and 5 MeV was used in order to provide the necessary thermal beam to the investigated sample. The facility was designed and simulated using the MCNPX Monte Carlo code. The primary goal was the enhancement of the signal to noise ratio and the improvement of the detection limit of the system. An extensive series of simulations were performed for each source in order to examine the consequence of the solid angle which formed between the neutron beam and the collimator detector. Boron was used as an element which emits the prompt gamma rays and according to the simulations, 2.5 and 3 MeV protons beams offer the best performance, while for proton beam with energy in the range between 4-5 MeV the results indicate slightly poorer performance. Simulations are presented for the in-vivo PGNAA in the human liver.

Recently, investigation of advanced shielding materials to be used as an alternative to lead apron has become important. In the current study, MgO loaded into PVC matrix as a non-lead modern shielding composite was modeled to evaluate its performance on radiation protective clothing (RPC). Parameters such as mass attenuation coefficient (MAC), mean free path (MFP), flux buildup factor (FBF), transmission factor (TF) and lead equivalent value (LEV) of samples were calculated using MCNPX Code. The simulation of the MCNP code was validated, by comparing the mass attenuation of concrete sample, with standard XCOM data and very good agreement was attended between XCOM and MC Code results. The MAC of nano and micro-sized samples were also compared with pure PVC and it was found that the nano MgO particle exhibits higher attenuation compared to micro MgO particle and pure PVC. The results show that, the MAC of samples increased to 63.13 % in 1.332 MeV with increasing filler concentration of nano MgO to 50 wt% relative to pure PVC. Investigation of LEV shows that nano MgO sample has more effective than Pb in 1.173 and 1.332 MeV gamma ray energy so that it provides 36.46 % and 11.13 % lighter RPC than Pb ones., Competing Interests: The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (© 2024 The Authors. Published by Elsevier Ltd.)

Influence of Gd2O3 on the structure, physical, neutrons and gamma-photon protection features of newly prepared (50-x)B2O3 + 25Bi2O3 + 25Na2O + xGd2O3 glasses with different Gd2O3 concentrations was investigated. Glasses have been prepared via solid-state conventional method and encoded as Gd0, Gd1, Gd2, Gd3, and Gd4, respectively, according to xGd2O3 values. X-ray diffraction measurements proved that all Gd-glasses were in an amorphous nature. The density of the samples was increased from 5.1021 to 6.3210 g.cm−3. The mass attenuation coefficient (MAC) values were simulated via MCNPX in photon energy range of 15 keV–15 MeV and verified via XCOM program. The MACs were ranged from 28.69 to 32.52 cm2.g−1, while linear attenuation coefficient values were found to be ranged from 142 to 201 cm−1 for Gd0 and Gd4, respectively. The Gd4 glass has the lowest half value layer (HVL, highest radiation shielding capacity) among all the prepared Gd-glasses, several concretes, and conventional glasses. The atomic cross section (ACS) and electronic cross section (ECS) values were enhanced with the increment of Gd ratio in the sample compositions, and the measures of ACS were much greater than the ECS values. The effective atomic number (Zeff) values were ranged from 62.7 to 63.4 for Gd0 to Gd4, respectively. The effective conductivity (Ceff) is directly proportional to the density and effective electron density of the glass samples. By utilizing the Phy-X/PSD program, the maximum values of exposure buildup factor were slowly decreased with the addition of Gd2O3 contents and decreased from 55.8 to 55.3, while the highest energy absorption buildup factor increases slowly from 70.5 to 71.1 for glasses Gd0 and Gd4, respectively. The lowest values of fast neutron removal cross section (∑R, cm−1) were obtained for glasses Gd0 (∑R = 0.179 cm−1), while the highest value was obtained for glasses Gd4 (∑R = 0.179 cm−1). [ABSTRACT FROM AUTHOR]

Characteristics of tellurite-tungstate-antimonate glasses containing heavy metal oxide were investigated in detail using two methods: the MCNPX Monte Carlo code and the Phy-X/PSD platform. The influence of Sm2O3, translocating with TeO2 at ratios of 0.2, 0.5, 0.8, 1, and 1.5 mol% on radiation shielding properties of glasses, was set forth with five glass structures determined according to the (75-x)TeO2-15Sb2O3-10WO3-xSm2O3 glass composition. Densities of the glasses were prepared by doping a low ratio of Sm2O3 that varied between 5.834 and 5.898 g/cm3. Sample densities, which have an important role in determining radiation shielding character, increased depending on the increase in Sm2O3 concentration. Effective removal cross-section (∑R) values against fast neutrons, as well as linear and mass attenuation coefficients, half-value layer, mean free path, variation of effective atomic number against photon energy, exposure, and energy built-up factors, were simulated with the help of these two methods. As a result of these estimates, it can be concluded that values obtained using both methods are consistent with each other. From the obtained values, it can be concluded that the SM1.5 sample containing 1.5 mol% would have the most efficient role in radiation shielding. An increase of Sm2O3 resulted in a significant increase in linear and mass attenuation coefficients and effective removal cross-section values belonging to fast neutrons and, in addition, resulted in a decrease in the half value layer. Doping HMO glasses with Sm2O3 was observed to contribute directly to the development of radiation shielding properties of the glass. [ABSTRACT FROM AUTHOR]

Mediterranean fly pest (Ceratitis) is one of the most destructive pests of fruit species in Morocco. The sterile insect technique (SIT) is an environmentally friendly strategy that uses ionizing radiation to sterilize adult insects. Morocco has a panoramic gamma irradiator used to irradiate agri-food products. This irradiator is not dedicated to SIT programs due to its geometry that does not allow to obtain a dose uniformity ratio (DUR) recommended for such applications. This article presents a Monte Carlo study to investigate the feasibility of using the panoramic gamma irradiator at the National Institute for Agronomic Research (NIAR) of Tangier, Morocco, to setting up SIT methods and contributing to Ceratitis control programs. The Monte Carlo method was used to simulate the concrete bunker in which the panoramic gamma irradiator is installed. To obtain a recommended DUR required for SIT programs, two cells similar of the Gammacell-220 irradiator, which is mainly used in the SIT programs around the world, were simulated inside the concrete bunker. The simulation and calculations were performed using the MCNPX-2.7e Monte Carlo simulation code. It is demonstrated that at both investigated positions, the spatial distribution of dose rates in the two modeled irradiation cells, which were similar to a gammacell-220 irradiator cell, are uniform enough that the cells can be used for SIT programs. It is concluded that the panoramic irradiator at NIAR can be used to contribute to the control of Mediterranean fly pest and other insect pests in Morocco. [ABSTRACT FROM AUTHOR]

The present study aims to investigate the gamma photon-neutron attenuation parameters of the marble concrete such as the linear attenuation coefficient ( μ), Mass Attenuation Coefficient (MAC), Half Value Layer (HVL), neutron removal cross-section (∑R), and Exposure Buildup Factor (EBF) by MCNPX Monte Carlo code to examine the possibility of utilizing selected samples against gamma photon-neutron. The six types of marble concrete were exposed to different energies ranging from 511 to 1332 keV. The shielding effectiveness of the selected marble concrete was compared to each other and the obtained results show that the increment of the penetration depth causes an increase in the value of EBF and the sharp peak is monitored at 200 keV. The ∑R (neutron removal cross-section) values were also reported for all types of marble concrete. The outcomes depict that ∑R is strongly relevant to the weight fractions of light elements of the selected shielding materials. Furthermore, the variations of σ a (total atomic cross-section) were reviewed and discussed versus wide energy ranges. Findings in all sections confirmed that NM-15 has the best gamma photon shielding quality in comparison with the other types of marble concrete, while NM-0 and NM-15 show the highest neutron proficiency. [ABSTRACT FROM AUTHOR]

The time distributions of neutron and gamma events for the plutonium metal plates of different isotopic compositions in a singles and coincidence acquisition mode were measured with the Fast Neutron Multiplicity Counter (FNMC) consisting of 8 EJ-309 liquid scintillators and 8 stilbene detectors. Since time intervals between detected events depend on the source activity, estimation of 240Pu effective mass in an investigated sample was based on the calibration curve obtained by using a subset of the remaining Pu metal plates. It was demonstrated that the neutron time interval distributions can provide the evaluation of 240Pu effective mass with high accuracy. The Kolmogorov complexity (KC) and Lyapunov exponent were calculated to reveal the presence of a chaotic and non-random time series in all the measured ones. [ABSTRACT FROM AUTHOR]