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This paper examines the dosimetric uncertainty arising from the use of thermoplastic masks in the treatment of head and neck cancer through radiotherapy. This study was conducted through Monte Carlo simulations using the Monte Carlo N-Particle eXtended (MCNPX code), and the theoretical results are compared with radiochromic films. Using material characterization techniques, the compounds of the thermoplastic mask were identified, confirming that most of the material corresponds to the polymer C10H16O4. The theoretical results show increases ranging from 42% to 57.4% in the surface absorbed dose for 6 and 15 MV photon beams, respectively, compared to the absorbed dose without the mask. The experimental data corroborate these findings, showing dose increases ranging from 18.4% to 52.1% compared to the expected surface absorbed dose without the mask. These results highlight the need to consider the bolus effect induced by thermoplastic masks during the precise and safe planning and application of radiotherapy treatment in order to ensure its therapeutic efficacy and minimize the associated risks to patients. [ABSTRACT FROM AUTHOR]

Since the beginning of research into radiation and protection against it, the importance of searching for proper materials against radiation hazards has been recognized. Gamma radiation protection materials usually deal with heavy elements with high prices, which are hard to maintain. Polyurethane-based (PU) materials are popular in sound and thermal insulation due to, their low-weight properties and, most importantly, fast and convenient construction ingredients. PU foams (PUF) can be used as radiation shield and noise and heat resistance due to their approachability, light-weight, high resistance, and comfortable construction. This study involved simulation and an experiment to construct and investigate the properties of Polyurethane material doped with lead oxide as a gamma shield. The shield was considered in several weight fractions of lead, yielding several samples. The MCNPX 2.6 Monte Carlo code has been utilized for simulation procedure, and 137Cs was employed as the gamma source in both simulation and experiment. The results offer a promising response against the gamma radiation and are suitable for attenuating gamma rays. [ABSTRACT FROM AUTHOR]

Abstract Since the beginning of research into radiation and protection against it, the importance of searching for proper materials against radiation hazards has been recognized. Gamma radiation protection materials usually deal with heavy elements with high prices, which are hard to maintain. Polyurethane-based (PU) materials are popular in sound and thermal insulation due to, their low-weight properties and, most importantly, fast and convenient construction ingredients. PU foams (PUF) can be used as radiation shield and noise and heat resistance due to their approachability, light-weight, high resistance, and comfortable construction. This study involved simulation and an experiment to construct and investigate the properties of Polyurethane material doped with lead oxide as a gamma shield. The shield was considered in several weight fractions of lead, yielding several samples. The MCNPX 2.6 Monte Carlo code has been utilized for simulation procedure, and 137Cs was employed as the gamma source in both simulation and experiment. The results offer a promising response against the gamma radiation and are suitable for attenuating gamma rays.

This study examines the feasibility of utilizing mixed-oxide fuel [(U0.9, rgPu0.1) O2] instead of traditional UO2 in nuclear reactors. The utilization of (U0.9, rgPu0.1) O2 is particularly significant as it represents an effective approach to nuclear fuel recycling by combining reactor-grade plutonium extracted from partially used nuclear fuel and depleted uranium obtained through the enrichment process. The fundamental neutronic characteristics, such as the radial power distribution, were investigated using the MCNPX 2.7 algorithm to identify the specific channel for subsequent thermal-hydraulic (TH) analysis. The TH analysis was conducted using COMSOL-Multiphysics, allowing for the estimation of the fuel rod's axial and radial temperature profiles, as well as the determination of the departure from the nucleate boiling ratio. Furthermore, the coupling between heat transfer and solid structure (SS) was achieved using the Multiphysics tool in COMSOL-Multiphysics. This coupling facilitated the simulation of key SS parameters, including von Mises stress, volumetric strain, and displacement, while considering the influence of heat transfer. The results demonstrate significant improvements and enhanced safety margins when utilizing (U0.9, rgPu0.1) O2. [ABSTRACT FROM AUTHOR]

Abstract Regarding to their unique physical and mechanical features, glasses and glass–ceramics are suitable materials for shielding purposes. The present study evaluates the shielding properties of the CaF2–CaO–B2O3–P2O5–SrO–Ta2O5 glass system using Monte Carlo GEANT4 and MCNPX codes for X-ray radiations with an energy range of 20 to 100 keV. MAC values of the Ta0, Ta1, Ta2, Ta2.5, and Ta3 samples of the CaF2–CaO–B2O3–P2O5–SrO–Ta2O5 glass were computed using Phy-X/PSD, GEANT4, and MCNPX codes and compared. According to the results, the programs have good compatibility with each other. For instance, in the energy of 40 keV and for the Ta2 sample, GEANT4 and MCNP codes are 1.445765406 and 1.517801204 cm2/g, respectively, indicating 7.419529525 and 2.829628418% differences with 1.562 cm2/g obtained using the Phy-X/PSD software. According to recent estimations, the Ta3 sample of the CaF2–CaO–B2O3–P2O5–SrO–Ta2O5 glass system can be selected as the best shield compared with the other samples.

The transfer of nuclear spent fuel from the reactor storage pool to dry storage or for reprocessing or final disposition requires information about its isotopic composition, decay heat, and other thermomechanical properties. The spent nuclear fuel assembly of a typical advanced pressurized water reactor, AP-1000, was characterized using the Monte Carlo MCNPX code and SCALE/ORIGEN code. The simulation of operational history started from the operation of the first fresh core for an average fuel assembly with certain physical isotopic parameters until 25 GWd/tonne U discharge burnup. The analysis considered the calculations of the radionuclide inventories, activity, neutron emission spectrum, gamma-ray emission spectrum, and decay power after 700 effective full power days and for post different time ranges until a 1 million–year cooling period. The comparison of some results of the two codes showed small differences due to the consideration of the continuous-energy variation for neutrons in the MCNPX code and the discrete energy assumption in the SCALE/ORIGEN code. [ABSTRACT FROM AUTHOR]

The plasma focus device can be utilized as an X-ray generator in radiographic applications. A few tens of nanoseconds after the pinch moment, the hard X-ray of this device is released in a time interval of 100-150 ns. The pulsed form of X-ray emitted from this device makes the common method of detection and spectrometry useless for characterizing them. In this article, passive spectrometry, using radiographic film with Al attenuation filters, was used to determine the pulsed X-ray spectrum emitted from the UIPF-1. Neural network technology was used to determine the spectrum according to recorded doses. The neural network was trained using MCNPX simulation results. The results showed that the X-ray spectrum extends from 6 keV to 50 keV with a maximum value of 8 keV when the UIPF-1 device is operated with the copper anode, the copper insert, the working voltage of 21 kV, and 0.9 mbar air gas injection. The average spectrum energy was also obtained at 17.5 keV.

The safety of electronic components used in aerospace systems against cosmic rays is one of the most important requirements in their design and construction (especially satellites). In this work, by calculating the dose caused by proton beams in geostationary Earth orbit (GEO) orbit using the MCNPX Monte Carlo code and the MULLASSIS code, the effect of different structures in the protection of cosmic rays has been evaluated. A multi-layer radiation shield composed of aluminum, water and polyethylene was designed and its performance was compared with shielding made of aluminum alone. The results show that the absorbed dose by the simulated protective layers has increased by 35.3% and 44.1% for two-layer (aluminum, polyethylene) and three-layer (aluminum, water, polyethylene) protection respectively, and it is effective in the protection of electronic components. In addition to that, by replacing the multi-layer shield instead of the conventional aluminum shield, the mass reduction percentage will be 38.88 and 39.69, respectively, for the two-layer and three-layer shield compared to the aluminum shield.

Structural properties, such as mechanical and electrical properties of materials irradiated with γ-rays are affected by displacement damage. Successive processes with different behaviours ultimately lead to the formation of sets of ‘defects’ within the material, and as an example, it may cause the substance to become brittle. In this study, Monte Carlo codes provide information using a code-based simulation method on primary knock-on atoms, or PKAs, that cause damage. Also, the rate of damage caused by the gamma radiation from the cobalt-60 source on the structural properties of iron is calculated. To access PKA information, a program called GAMMATRACK has been developed. This software provides information about the rejected atoms' properties and the interaction kinematics. The theoretical calculation method has also been used to confirm the results of the Monte Carlo method. Using the generated secondary electrons, the stopping and range of ions in matter (SRIM) code can calculate the damage caused by gamma radiation. PKA data were extracted by the GAMMATRACK program, which can be used as input for the SRIM code for the systematic analysis of gamma damage. The obtained PKA spectrum for iron agrees with the previous works. It is realised that only monovacancies are produced, and under cobalt-60 irradiation, the possibility of the atom–atom collision is negligible. Therefore, creating PKA cascades will be ruled out. Also, the damage rate for an iron target is calculated at ~10−7, 10−8 (displacement per atom (DPA) / year) for both theoretical and Monte Carlo method (MCNPX + SRIM code) computations. [ABSTRACT FROM AUTHOR]

Cerium-144 (144Ce), as a long-lived β emitter radioisotope (Eβavg = 82.25 keV, T1/2 = 284.9 d), is a potentially promising radionuclide to be used in various medical and industrial applications. In this paper, an optimized process for the separation of 144Ce from fission products is reported. Following the MCNPX calculation results, an equivalent solution was experimentally prepared that is similar to the waste solution of local 99Mo production via the fission method. For analyzing the steps of the radiochemical process, the suitable radioisotope tracers created from the activation method and were added to the simulated waste solution. The oxidation/reduction potential of Ce3+/Ce4+ together with extraction chromatography was exploited to achieve high yields of more than 70% and a reasonable radionuclidic purity of 97% and radiochemical purity of more 99% of radiocerium. [ABSTRACT FROM AUTHOR]

Due to the harmful effects of ionizing radiation, shielding has become a crucial topic for radiation protection. Finding effective, non-toxic and low-cost shielding materials is imperative in ensuring the safety of individuals exposed to ionizing radiation. Whether a material is effective in shielding against radiation depends on the linear attenuation coefficient. In this study, linear attenuation coefficients were calculated using the MCNPX code for energy values of 81 keV (Ba-133), 140 keV (Tc-99m), 662 keV (Cs-137), 1173 keV, and 1332 keV (Co-60) by incorporating Bentonite Clay (BC) nanoparticles and micro-sized particles as additives into a Polyvinyl Alcohol (PVA) matrix. BC particles with a density of 50% were added to the PVA matrix using LAT and U cards. Simulations were performed with a monoenergetic source emitting 107 particles and a narrow beam geometry, and the counts of particles with diameters of 50 nanometers and 50 micrometers were calculated using the F4 tally. When the results obtained from the simulation were compared, it was observed that as the diameters of the added particles decreased, their effectiveness in radiation shielding increased for each energy value. Among them, the 50 nm BC particles added at a rate of 50% in PVA showed the highest effect at 1332 keV, with a 9.5% increase compared to 50 µm BC particles. [ABSTRACT FROM AUTHOR]

The utilization of radiation shielding material positioned between the both breasts are crucial for the reduction of glandular dose and the safeguarding of the contralateral breast during mammographic procedures. This study proposes an alternative substance for shielding the contralateral breast from radiation exposure during mammography screening. In this study, we present an analysis of the shielding effectiveness of transparent glass that has been doped with Tantalum (V) oxide encoded as BTZT6. The evaluation of this shielding material was conducted using the MCNPX code, specifically for the ipsilateral and contralateral breasts. The design of the left and right breast phantoms involved the creation of three-layer heterogeneous breast phantoms, consisting of varying proportions of glandular tissue (25%, 50%, and 75%). The design of BTZT6 and lead-acrylic shielding screens is implemented using the MCNPX code. The comparative analysis of dose outcomes is conducted to assess the protective efficacy of BTZT6 and lead-acrylic shielding screens. The utilization of BTZT6 shielding material resulted in a reduction in both breast dose and skin dose exposure when compared to the lead-acrylic shield. Based on the findings acquired, the utilization of BTZT6 shielding material screens during mammography procedures involving X-rays with energy levels ranging from 26 to 30 keV is associated with a decrease in radiation dose. It can be inferred that the utilization of BTZT6 demonstrates potential efficacy in mitigating excessive radiation exposure to the breasts and facilitating the quantification of glandular doses in mammography procedures. [ABSTRACT FROM AUTHOR]

Using the experimental data in nuclear computing to verify the calculation methods and tools based on numerical and statistical methods has many benefits such as illustrating the quality, ensuring the capabilities, and validating computer codes. Simulation by computer tools is also applicable in the safety analysis of research reactors. In this research, the computer tool (MCNPX 2.7.0: 2011) was verified against the experimental data of neutron flux and spectrum on the sample position of the Tehran Research Reactor (TRR) neutron imaging system by the neutron activation method. To determine the benchmark specifications, the simulation of the system was done at the first step by considering a well-defined facility geometric, material specification and reactor core configuration, fuel elements, and radiation facility (beam tubes and collimator, reactor core, and neutron imaging components). Then the flux and neutron spectrum at the sample position were calculated. In the second step, a set of In (bare and covered by cd) and Au foils and a set of Au, Ni, Ti, and Zr, were placed and exposed almost in front of the reactor E beam tube. The neutron energy spectrum was unfolded by calculating the saturation activity of each foil by SAND-II code, and the neutron flux was calculated. A comparison of the results obtained in two steps shows a relatively good and acceptable agreement (Max. 30% deviation) between the flux and the shape of the flux profile obtained from calculations and experimental data. [ABSTRACT FROM AUTHOR]

The LEU miniature plate irradiated in the Tehran Research Reactor, which produces the radioisotope molybdenum-99, is a radioactive source consisting of various radioisotopes. Before any practical application, it is necessary that the dose rate of gamma rays emitted from this source be calculated. This was done for comparison with the recommended dose rates for radioactive materials transport and radiation protection of employees during hot tests. In this paper, the validity of a proposed method for calculating the dose rate of gamma rays emitted from hot targets was investigated, experimentally. The Monte Carlo code MCNPX and multi-step algorithm written in MATLAB were used in the proposed method. The results showed that the calculated dose rate was always lower than the measured dose rate. Therefore, based on a conservative view, it is better to multiply the calculated dose rate values by 2 and then compare them with the allowable limits. This will enable you to determine the appropriate shield thickness.

Germanium crystals have a dead layer that causes a decrease in efficiency since the layer is not useful for detection but strongly attenuates photons. The thickness of this inactive layer is not well known due to the existence of a transition zone where photons are increasingly absorbed. Therefore, using data provided by manufacturers in the detector simulation model, some strong discrepancies appear by changing the dead layer. Investigations into the Ge detector response functions for gamma rays have been conducted using straightforward physical mechanisms implemented by Monte Carlo simulations. The detector response function feature's most probable interaction mechanisms are described. The Monte Carlo method is applied to simulate the calibration of a HPGe detector in order to determine the total inactive germanium layer thickness and the active volume that is needed in order to study the response function for both types of detectors. Results indicated a strong impact of dead layer variations on the response function of the simulated detectors.

This paper examines the dosimetric uncertainty arising from the use of thermoplastic masks in the treatment of head and neck cancer through radiotherapy. This study was conducted through Monte Carlo simulations using the Monte Carlo N-Particle eXtended (MCNPX code), and the theoretical results are compared with radiochromic films. Using material characterization techniques, the compounds of the thermoplastic mask were identified, confirming that most of the material corresponds to the polymer C10H16O4. The theoretical results show increases ranging from 42% to 57.4% in the surface absorbed dose for 6 and 15 MV photon beams, respectively, compared to the absorbed dose without the mask. The experimental data corroborate these findings, showing dose increases ranging from 18.4% to 52.1% compared to the expected surface absorbed dose without the mask. These results highlight the need to consider the bolus effect induced by thermoplastic masks during the precise and safe planning and application of radiotherapy treatment in order to ensure its therapeutic efficacy and minimize the associated risks to patients.

We report the gamma-ray shielding properties, transmission factors and, effective removal cross section values of several aluminoborosilicate glasses that have been synthesized through various glass-forming materials such as Al 2 O 3 , B 2 O 3 , SiO 2 , and ZrO 2. The study utilized the elemental compositions and densities of eight different glass samples as input variables in theoretical calculations and Monte Carlo simulations. According to the results obtained, it was seen that the network-forming type used in aluminoborosilicate glasses had a direct effect on the radiation absorption properties of the glasses. The utilization of ZrO 2 and Cs 2 O at the highest concentration as the network former and glass network modifier in the NCBZ-6 sample yielded the most advantageous results in relation to its gamma-ray absorption capabilities. The benefit is intrinsically related to the heightened density of glass and the incorporation of compounds with increased atomic numbers, both of which are fundamental characteristics desired in materials designed for the purpose of gamma-ray absorption. However, the enhanced capability of ZrO 2 to absorb gamma-rays excludes the absorption of high-energy neutrons. The absence of boron trioxide (B 2 O 3) in the NCBZ-6 sample can be ascribed to its restricted availability against fast neutrons. The continued existence of ZrO 2 as a network forming in the investigated ABS glasses is likely to result in improved material homogeneity and progressive enhancement of gamma-ray absorption characteristics. It can be concluded that the incorporation of ZrO 2 as a network-former component may be an appropriate strategy to enhance the gamma-ray shielding capabilities of aluminoborosilicate glasses. [ABSTRACT FROM AUTHOR]

Germanium crystals have a dead layer that causes a decrease in efficiency since the layer is not useful for detection but strongly attenuates photons. The thickness of this inactive layer is not well known due to the existence of a transition zone where photons are increasingly absorbed. Therefore, using data provided by manufacturers in the detector simulation model, some strong discrepancies appear by changing the dead layer. Investigations into the Ge detector response functions for gamma rays have been conducted using straightforward physical mechanisms implemented by Monte Carlo simulations. The detector response function feature's most probable interaction mechanisms are described. The Monte Carlo method is applied to simulate the calibration of a HPGe detector in order to determine the total inactive germanium layer thickness and the active volume that is needed in order to study the response function for both types of detectors. Results indicated a strong impact of dead layer variations on the response function of the simulated detectors. [ABSTRACT FROM AUTHOR]

A novel simulation to calculate the neutronic parameters of the TRIGA 2000 reactor using plate-type fuel has been performed. The plate fuel used was produced by the Indonesian Nuclear Industry (PT INUKI) with U3Si2-Al material. Neutronic parameters based on INUKI’s plate-type fuel dimension and the current TRIGA’s configuration were simulated using MCNPX. The simulation was performed by modeling the complete reactor’s configuration on a fresh fuel core state. We obtained the kinetic parameter values from the simulation, i.e., delayed neutron fraction of 8.11×10‑3, a prompt neutron lifetime of 2.0551×10‑4 s, and an average neutron generation time of 1.87×10‑4 s. The excess reactivity of the reactor was 9.02 %Δk/k, while reactivity in the one-stuck-rod state was below ‑0.5 $ with an average value of ‑3.40 %Δk/k (‑4.19 $). The average thermal neutron flux peak occurred at the central irradiation position with the value of 3.0×1013 to 3.1×1013 n/(cm2 s). The reactor has a power peaking factor of 1.379 in the control rod position of 0 % on D3 fuel. The reactor had a negative feedback reactivity coefficient, except for the moderator coefficient. These results suggest that the current configuration of plate-type fuel met the nuclear reactor neutronic safety standards.

Protection of astronauts and electronic components in satellites and spacecraft against space rays is one of the most important primary requirements in space missions. In this work, the effect of three materials, aluminum, as the most common material, polyethylene and a graded-z structure, in the protection of space radiations has been evaluated. The calculations of the dose caused by these radiations on the human body and a silicon piece have been carried out by MCNPX Monte Carlo code,. The dose caused by cosmic rays has been calculated after applying shields of aluminum, graded-z structure and polyethylene. The results showed that by using polyethylene and about 4.4% increase in weight compared to the aluminum shield, it is possible to reduce the dose caused by photons by more than 50% in the human body and 30% in silicon parts, and the dose caused by protons by about 30%. It cut both for astronauts and electronic components. Graded-z shielding performed very well in the dose attenuation caused by photons, but appeared ineffective in the dose attenuation caused by protons.

Moduli of elasticity and the ability of gamma radiation shielding as well as transmission factor (TF) of four samples of tungsten bario-phosphate glasses with chemical formula xBaO-40WO3-(60-x)P2O5: (x = 10, 20, 30, and 40 mol% glasses) have been examined. The glass samples have been coded as S1: 10BaO-40WO3-50P2O5, S2: 20BaO-40WO3-40P2O5, S3: 30BaO-40WO3-30P2O5, and S4: 40BaO-40WO3-20P2O5. Makishima-Mackenzie model, the newly online Phy-X/PSD software, and Monte Carlo code were applied to achieve the mentioned purposes. The total dissociation energy (Gt) values were decreased from 62.58 (kJ/cm3) for S1 glasses to 55.92 (kJ/cm3) for S4 glasses. The total packing density (Vt) was 0.613, 0.626, 0.612, and 0.643, for S1, S2, S3, and S4, respectively. Values of Young's (EMM) modulus were 76.759 GPa, 75.631 GPa, 71.167 GPa, and 71.966 GPa and Poisson's ratio (σMM) possessed the values 0.273 GPa, 0.278 GPa, 0.273 GPa, and 0.284 GPa for S1, S2, S3, and S4, respectively. The S4 sample possessed the highest values of both linear (µ) and mass (µm) attenuation coefficients. However, (µ, µm)S1 < (µ, µm)S2 < (µ, µm)S3 < (µ, µm)S4. Also, The S4 sample possessed the lowest values of half value layer (HVL), tenth (TVL) value layers, and mean free path (MFP) among all studied glasses. Therefore, (HVL, TVL, MFP)S4 < (HVL, TVL, MFP)S3 < (HVL, TVL, MFP)S2 < (HVL, TVL, MFP)S1. Both effective atomic number (Zeff) and electron density (Neff) of the investigated glasses have as similar trend. In terms of transmission factor (TF), the sample S4 achieved the minimum values at a thickness of 3 cm. However, the S4 sample exhibited superior radiation shielding characteristics. [ABSTRACT FROM AUTHOR]

The Heavy Water Zero Power Reactor (HWZPR) is a 100‑W research reactor that uses natural uranium metallic fuel and heavy water as moderator. The HWZPR is designed for training and research in reactor physics fields. The main aim of this paper is to evaluate the variation of neutronic parameters of a HWZPR resulting from changes in the fuel lattice pitch using MCNPX code. In such a manner, neutronic parameters including effective multiplication factor (keff) for critical water level, effective delayed neutrons fraction (βeff), prompt neutrons lifetime (lp), and neutron flux distribution are investigated and benchmarked. The results illustrated that, as the lattice pitch is increased, more heavy water is needed for criticality condition according to the decrease in the number of fuel rods and fewer available fissile materials. Comparison of the results with the reactor Safety Analysis Report shows reasonable agreement. [ABSTRACT FROM AUTHOR]

Platinum compounds are the main chemotherapeutic agents used in cancer treatment. 195mPt is a promising theranostic radionuclide for targeted radionuclide therapy. This study presents 195mPt production through neutron activation of natural Platinum target in Tehran Research Reactor. Theoretical production yield of 195mPt was calculated through simultaneous solution of differential equations of decay chains, and also simulation with MCNPX. Theoretical results were compared with experimental values. Moreover, liquid–liquid extraction using ethyl acetate was used for chemical separation. The results showed a good agreement between the experimental and theoretical values. In addition, high separation efficiency (> 99%) was achieved via solvent extraction separation. [ABSTRACT FROM AUTHOR]

A Lead-Cooled Fast Reactor (LFR) is one of the six advanced fourth-generation reactors that many countries have been interested in designing, constructing, or develop them in the world. ALFRED reactor design investigated in this paper has 125 (MW) electric power and is being cooled by lead. The main purpose of the ALFRED project is to design and construct a reactor with high safety performance and reliability in all operation modes. In the present study, Monte Carlo codes MCNPX and SuperMC are used to calculate the neutronic parameters of the ALFRED reactor core including effective multiplication factor, kinetic parameters, neutron flux, and power distribution based on the latest design data. According to the results, the insertion of the control rods into the core can move the maximum flux height of about 40% to 50% to the end. Also, comparing codes' results show an appropriate consistency. [ABSTRACT FROM AUTHOR]

In this paper the goal is to optimize a shield for the americium-beryllium radioactive source, which generates a mixed neutron and gamma-ray radiation field, using two different materials designed with simultaneous moderation and absorption of neutrons and gamma-ray shielding in mind: a new type of stainless steel and epoxy resin (C21H25ClO5) composite with a 40% NiO additive. For this purpose, MCNPX was used and the equivalent dose rate for both types of radiation was calculated. After the optimization process, the shield volume experienced a 54% reduction while its weight, due to the use of a steel alloy, was increased by 3.56 %. At a distance of 1 m from the center of the source, the equivalent dose rate was reduced by 10.98% in comparison to that of the original system.

The present work concerns the development of a technique for activity determination of 68Ge/68Ga and 57Co disused radioactive sources. This technique aims to determine the activity of these sources by measuring and Monte Carlo simulation using the MCNPX code. Therefore, efficiency calibrations of the 3 x 3 NaI(Tl) detector for specific source geometries were carried out. Spectrums for two types of disused radioactive sources were collected for different measurement times. The characteristic gamma rays of a flood source containing 57Co and a line source containing 68Ge/68Ga, were used. In case of 68Ge/68Ga, the annihilation peak of 511 keV was also used considering the disused radioactive sources as a positron emitter. Sources of the mentioned types with certified nominal activities were used to validate MCNPX models. For the 57Co source, 15 minutes measuring time was adequate for activity determination, and 2 hours measuring time provided adequate sensitivity, at the level of general clearance. For 68Ge/68Ga line source, 15 minutes measurement was adequate for activity determination by using the 511 keV from the annihilation of 1899.1 keV positrons.

We report the functional assessment of tungsten (VI) oxide on gamma-ray attenuation properties of 60Sb2O3-(40-x)NaPO3-xWO3 antimony glasses. The elemental mass-fractions and glass-densities of each glass sample are specified separately for the MCNPX Monte Carlo code. In addition to fundamental gamma absorption properties, Transmission Factors throughout a broad radioisotope energy range were measured. According to findings, holmium (Ho) incorporation into the glass structure resulted in a net increase of 0.3406 g/cm3, whereas cerium (Ce) addition resulted in a net increase of 0.2047 g/cm3. The 40% WO3 reinforced S7 sample was found to have the greatest LAC value, even though seven glass samples exhibited identical behavior. The S2 sample had the lowest HVL values among the glass groups evaluated in this work, computed in the energy range of 0.015–15 MeV. The lowest EBF and EABF values were reported for 40% WO3 reinforced S7 sample with the highest LAC and density values. According to the findings of this research, WO3 will likely make a significant contribution to the gamma ray absorption properties of antimony glasses, which are employed for optical and structural modification. Therefore, it can be concluded that WO3 may be treated monotonically and can be employed successfully in circumstances where gamma-ray absorption characteristics, optical properties, and structural qualities need to be enhanced.

Neutron detection techniques are widely studied in many articles. Most of this research requires a lot of electronic equipment. In this study, using the Thick Gas electron multiplier (THGEM) detector, a new method for neutron detection is proposed to reduce electronic equipment. In the neutron detection system, the converter material is used for converting neutrons to protons that are directed to the THGEM detector. By filling the detector space with noble gas and applying special voltage, THGEM enters to Self-Quenched Streamer (SQS) mode for protons detection. All these steps are examined by simulation, then the detection system is made and is examined in the laboratory. Finally, the simulation results and laboratory results are compared. The results show that the 1 mm Plexiglas layer is suitable for converting neutrons to protons. The suitable distance between the converter layer and the THGEM detector is 3 cm. Also, the SQS mode happens in the most number of THGEM holes when the THGEM voltage is 980 volt. Investigating an approach to neutron detection by placing THGEM in SQS mode can be useful because, firstly, placing the THGEM detector in SQS mode simplifies electrical circuits and secondly, with this proposed detection system; it is possible to design detectors with different dimensions for neutrons. [ABSTRACT FROM AUTHOR]

In this study, glass composition based on tungsten oxide (WO3) doped tellurium, titanium glasses: (100 − x − y) TeO2–xTiO2–yWO3: where (x = 5) and (y = 5, 10, 15, 20, 25) coded as TT5W5, TT5W10, TT5W15, TT5W20, and TT5W25 were investigated for shielding properties against ionizing radiation. Gamma radiation shielding parameters such as mass attenuation coefficients (MAC) are calculated through MCNPx code and Phy-X/PSD software in the energy range of 0.015–15 MeV. Obtained MAC values are then used to calculate other gamma radiation shielding parameters such as mean free path (MFP) and effective atomic number (Zeff). Besides this, the exposure buildup factor (EBF) was also calculated by using EXABCal software at different penetration depths (PDs) in the energy range of 0.015–15 MeV. Sample TT5W25, which has a larger WO3 content of 25 mol% shows higher values of MAC and lower values of MFP among all the examined glass samples. Our investigated TeO2–TiO2–WO3 glass samples possess the lowest MFP values in comparison with the different types of concretes and commercially available shielding glasses. In addition, fast neutron shielding characteristics in light of fast neutron removal cross-section have also been computed. Glass sample TT5W25 possesses the higher values of fast neutron removal cross-section as compared to the other glass samples. The results indicate that the adding up of WO3 improves shielding against the fast neutron and gamma radiation. [ABSTRACT FROM AUTHOR]

In radiotherapy, ionizing radiation is used to damage cells. Monte Carlo simulations of radiation interactions with DNA give us good information about the type of damage and their repair process, which can be very useful in treating cancer and protecting against radiation. The ionizing radiation damage includes single- and double-strand breaks as well as base lesions (SSB, DSB, BL). The DNA damage can be repaired through some processes within the cell. DNA damage, especially of DSBs that are mis-repaired or unrepaired, can result in cell death. This process plays a crucial role in killing cancer cells and treating cancer. In this research, first, the dose was calculated in the cell nucleus with the MCNPX code. Then, the possibility of different damage types in DNA was investigated by simulating the physical and chemical processes of low and high-energy electrons with the Geant4-DNA code. Then with the help of the Matlab software and mathematical modeling, we investigated the repair of DSBs for the latter energies. The results show that at ranging energies of 100 to 300 eV, the number of breaks increases, and for higher energies, it decreases due to the electron range in the cell nucleus. The repair reaction rate is also calculated from the NHEJ presynaptic process for ranging energies 100 eV to 1 MeV. In the repair section, the repair time at low energies is longer than the higher energies due to more DSBs.

The aim of this study is to assess the individual gamma-ray transmission factors (TFs) and some fundamental gamma-ray attenuation properties of several types of glasses based on WO3–TeO2–B2O3 glasses system. MCNPX (version 2.7.0) is used for the calculation of TFs. Other critical parameters are determined using the Phy-X/PSD program. To determine the TFs of studied glasses, several medical radioisotopes are determined along with their characteristic gamma-ray energies. The superior values for the investigated parameters are found in glass sample S6. Furthermore, the exposure build-up factor and energy absorption build-up factor values for glass sample S6 were the lowest. S6 glass sample with the chemical composition 0.03833B + 0.26075O + 0.11591Zn + 0.52783Te + 0.05718W and a density of 3.3579 g/cm3 is found to have exceptional gamma-ray attenuation qualities, according to our findings. It can be concluded that the prospective attributes of WO3-doped glass systems and associated glass compositions would be beneficial for scientific community in terms of providing a clearer view for some advanced applications of these glass types.

We report some fundamental gamma-ray shielding properties and individual transmission factors (TFs) of five distinct glass samples with a nominal composition of xSb2O3·(40 − x)PbO·60B2O3·0.5CuO and (where; 0 ≤ x ≤ 40 mol%). Phy-X/PSD and MCNPX (version 2.7.0) Monte Carlo code are utilized to determine several critical parameters, such as cross-sections, attenuation coefficients, half and tenth value layers, build-up factors, and TFs. A general transmission setup is designed using basic requirements. Accordingly, TFs are evaluated for several medical radioisotopes. Next, the gamma-ray shielding parameters and TFs are assessed together in terms of providing the validity of the findings. Our results showed that there is a positive contribution of increasing Sb2O3 amount in the glass matrix owing its direct effect to the density increment as well. This positive effect on gamma-ray shielding properties is also observed for decreasing mean free path values from S1 to S5 samples. The exposure build-up factor (EBF) and energy absorption build-up factor (EABF) values, increasing the quantity of Sb2O3 supplementation, resulted in a general reduction in EBF and EABF values (i.e., from 0.5 to 40 mfp). When the quantity of Sb2O3 rises from S1 to S5, the collision rate of incoming gamma rays in glass samples increases significantly. The TF figures reveal that S5 showed the least transmission behavior across all the above-mentioned studied glass thicknesses. It can be concluded that increasing the Sb2O3 additive is a beneficial and monotonic technique, when the gamma-ray shielding qualities or TF values must be further enhanced.

Nowadays, the irradiation of blood and blood products, as well as sterilization of pharmaceuticals and biological materials used in the medical industry, are being carried out in developed countries and some developing countries. Usually for this purpose, self-contained dry source storage irradiators containing cesium-137 or cobalt-60 gamma ray sources with activity up to 3 kCi are used as blood irradiator facilities. In this research, an attempt was made to use the MCNPX code to perform the calculations related to the nuclear design of a blood irradiator facility using cobalt-60 line sources in two parts of iso-dose curves and shielding according to internationally accepted standards. The dose uniformity ratio inside the irradiation chamber was about 2 and the external dose rate at a distance of 5 cm from the surface of the gamma cell device was about 36 µSv/h and below the allowable limit. The results showed that the iso-dose curves were completely symmetric and the delivered dose increased near the cobalt-60line sources while decreasing as the distance away from the line sources and closer to the cylindrical axis of the irradiation chamber. In addition, the results showed that the designed lead shield with an average thickness of 30 cm easily blocks the gamma rays and in terms of health physics, working with such a blood irradiator will not pose any risk of radiation to employees.

Considering the importance of deuterium in nuclear science including medical and industrial researches such as (BNCT) and nuclear reactors respectively, it is important to study various possible ways in addition to common methods for measuring its concentration. This study is an effort to measure deuterium concentration using PGNAA. The main idea is to calculate the area under 2.23 MeV gamma-rays photo peak resulting from neutron collision with Hydrogen atoms which are in mix with deuterium in samples. The study carried out by both simulation and experiment. Monte Carlo MCNPX2.6 code has been used for simulation and based on its acceptable results an experimental setup has been arranged. The coordination of results was in the range of R = 0.99 and R = 0.98 in simulation and experiment respectively. The accuracy of the study has been investigated by measuring the concentration of an unknown sample by both PGNAA and Fourier transform infrared spectroscopy (FT-IR) methods in which there were acceptable correlation between these two methods.

This paper aims to design a suitable irradiation box of topaz stones and improve the neutronic parameters by maintaining the necessary conditions for their cooling in Tehran Research Reactor (TRR). In this way, by defining an appropriate neutron filter and simulating the irradiation box in the TRR core with MCNPX2.7 code, the neutronic parameters of the core are calculated, and the best configuration is selected. Then, various components of the irradiation channel are simulated using the Solidworks software, and thermal calculations are performed using the Flow Simulation plug-in. The results showed that although the presence of water in the irradiation channel and topaz box results in good cooling for topaz stones, the slowdown of fast neutrons causes intense activation of the stones. However, by using a closed-loop independent of the reactor pool water and a gaseous fluid such as air, the thermal neutron flux and consequently the activity of the stones is reduced. Moreover, the results demonstrated that, in this case, acceptable conditions could be provided for long-term and safe irradiation of stones by selecting the appropriate fluid velocity and adjusting the core power.

Half Value Layer calculations theoretically need prior specification of linear attenuation calculations, since the HVL value is derived by dividing ln(2) by the linear attenuation coefficient. The purpose of this study was to establish a direct computational model for determining HVL, a vital parameter in nuclear radiation safety studies and shielding material design. Accordingly, a typical gamma-ray transmission setup has been modeled using MCNPX (version 2.4.0) general-purpose Monte Carlo code. The MCNPX code's INPUT file was designed with two detection locations for primary and secondary gamma-rays, as well as attenuator material between those detectors. Next, Half Value Layer values of some well-known gamma-ray shielding materials such as lead and ordinary concrete have been calculated throughout a broad gamma-ray energy range. The outcomes were then compared to data from the National Institute of Standards and Technology. The Half Value Layer values obtained from MCNPX were reported to be highly compatible with the HVL values obtained from the NIST standard database. Our results indicate that the developed INPUT file may be utilized for direct computations of Half Value Layer values for nuclear safety assessments as well as medical radiation applications. In conclusion, advanced simulation methods such as the Monte Carlo code are very powerful and useful instruments that should be considered for daily radiation safety measures. The modeled MCNPX input file will be provided to the scientific community upon reasonable request.

Boron Neutron Capture Therapy (BNCT) is a relatively safer technology for killing cancer cells, one of which is the Glioblastoma multiforme. One of the main components of the BNCT equipment is the collimator which functions as an exit point for epithermal neutron particles that hit cancer cells. In addition to the experimental method, BNCT research can be carried out by modeling, including using the MCNPX software based on the Monte Carlo Method. This research aimed to determine the flux distribution of fast and epithermal neutrons and the dose rate of fast neutrons and gamma that hit the target cancer cells in the phantom head of ORNL MIRD. Modeling using the MCNPX software has three main parts: cell cards, surface cards, and data cards. A tally is used on the data card to calculate the neutron flux. Based on the calculation of the modeling results, the flux of epithermal neutron is 2.87 x 109 n/cm2.s. The dose ratio of the epithermal to the fast neutron flux is 2.29 x 10-14 Gy.cm2/n. Then, the balance of the dose rate of the epithermal to the gamma is 1.64 x 10-14 Gy.cm2/n, and the ratio of epithermal to thermal neutron flux is 0.004. In this study, the epithermal neutron flux hitting the target cancer cells in cell target was moderated at 4 cm so that at a depth of 8 cm, the energy was converted into thermal neutrons. Based on the analysis of the results, it can be concluded that the neutron flux that will interact with cancer tissue is thermal neutrons, not epithermal neutron flux.

The role of dysprosium oxide (Dy2O3)-doped PbO–B2O3–SiO2 glasses on gamma and neutron radiation shielding properties was investigated and evaluated in terms of potential use in radiation protection. The glass composition was arranged according to the formula (55–x)B2O3–25PbO–20SiO2–xDy2O3 (0 ≤ x ≤ 5) (mol%). The Monte Carlo simulation (MCNPX) code was utilised to simulate the mass attenuation coefficient (MAC) of the glasses. When photon energies were in the range of 0.015–15 MeV, the resulting simulated mass attenuation (µm) values were evaluated with conventional Phy-X/PSD database findings. The Phy-X/PSD collected data and the predicted µm findings were in good agreement. The half-value layer (HVL), effective atomic number (Zeff), exposure and energy absorption build-up factors (EBF and EABF) of the glass samples were evaluated. 5 mol% Dy2O3-doped sample has significant γ-ray shielding efficacy in the complete energy range due to lower values of HVL, higher values of MAC and effective atomic number. The build-up factors follow a similar pattern with incident photon energy and penetration depth, growing on a steeply increasing trend up to a maximum at 0.5–1.5 MeV energy, then gradually reducing as photon energy increases up to 15 MeV. The build-up factors show the same routine with penetration depth and incident photon energy. They tend to steeply rise to a peak level at 0.5–1.5 MeV energy, then gradually decrease as the photon energy increases to 15 MeV. Also, the neutron shielding ability was detected and the effective removal cross-section Σ R (cm−1) was estimated theoretically. Due to the greatest value of effective removal cross-section, 5 mol% Dy2O3-doped glass exhibits stronger neutron shielding characteristics than the other glass samples. [ABSTRACT FROM AUTHOR]

The present work concerns the development of a technique for activity determination of 57Co and 68Ge/68Ga disused radioactive sources. This technique aims to determine the activity of these sources by measuring and Monte Carlo simulation using the MCNPX code. Therefore, efficiency calibrations of the 3 x 3 NaI(Tl) detector for specific source geometries were carried out. Spectrums for two types of disused radioactive sources were collected for different measurement times. The characteristic gamma rays of a flood source containing 57Co and a line source containing 68Ge/68Ga, were used. In case of 68Ge/68Ga, the annihilation peak of 511 keV was also used considering the disused radioactive sources as a positron emitter. Sources of the mentioned types with certified nominal activities were used to validate MCNPX models. For the 57Co source, 15 minutes measuring time was adequate for activity determination, and 2 hours measuring time provided adequate sensitivity, at the level of general clearance. For 68Ge/68Ga line source, 15 minutes measurement was adequate for activity determination by using the 511 keV from the annihilation of 1899.1 keV positrons. [ABSTRACT FROM AUTHOR]

Neutron data and cross-sections are highly regarded and are essential for developing nuclear equipment such as advanced fission and fusion reactors, accelerators, neutron shielding, physics studies, etc. The neutron cross-section should preferably be measured using a single-energy neutron beam, although the presence of a background in research reactors can affect its accurate determination. The Neutron Powder Diffraction (NPD) facility of Tehran Research Reactor (TRR) has been taken into consideration for measuring the neutron cross-section based on its properties, including neutron monochromator and multiple collimators. In this work, radiative capture cross-sections of Au, In, and Rh materials have been calculated using TRR monochromatic beam. MCNPX is a Monte Carlo particle transport code that has been applied to simulate the measurement system of the neutron cross-section and calculate the reaction rates. The effect of the presence and absence of different sections of the background on the cross-section values was investigated and the results were compared with EXFOR data library for validation. According to the findings, neutron backgrounds can have varying impacts depending on factors such as sample material, the isotope resonance regions, neutron source spatial distribution, and neutron monochromatic energy. However, the presence of fast neutron background contributes to the most uncertainty in the cross section values while its removal produces an average discrepancy from experimental libraries of 7.16%. Also, removing the cold neutron background also causes a relative difference equal to 7.65%. [ABSTRACT FROM AUTHOR]

We study the theoretical and simulated shielding properties of 69P2O5–10Gd2O3–10BaO–10ZnO–1Er2O3 and 69P2O5–10GdF3–10BaO–10ZnO-1Er2O3 glasses. We exploit advanced simulation (MCNPX) and Phy-X, an online Photon Shielding and Dosimetry software, in the 0.001–15 MeV energy range, to calculate the linear attenuation coefficient and mass attenuation coefficients of the samples. The obtained mass attenuation coefficient is then used to calculate the critical radiation shielding characteristics, such as the half-value layer, mean free path, tenth value layer, and effective atomic number. The calculated LAC values range from 108.759 to 0.097 and from 84.124 to 0.087 cm−1 while HVL values range from 0.006 to 7.164 and from 0.008 to 8.013 cm, and MFP values vary from 0.009 to 10.335 and from 0.012 to 11.561 cm for ZBaOEr and ZBaFEr, respectively, with energy ranging from 0.015 to 15 MeV. We find that the results obtained from MCNPX and Phy-X are consistent with each other, based on the computed relative difference. The glass sample containing Gd2O3 exhibits better shielding properties than the one with GdF3, over the entire energy range. Our examined glass samples are thick enough to substantially lessen the intensity of incident X-rays and gamma rays. [ABSTRACT FROM AUTHOR]

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]

Purpose: Absorbed dose enhancement due to the presence of high atomic number nanoparticles (NP)s has been estimated and modeled by Monte Carlo (MC) simulation methods. In the current study, two MC codes of MCNPX and EGSnrc codes were compared by calculation of secondary electron energy spectra and nano-scaled dose values around four types of spherical NPs. Materials and Methods: The MC model was composed of a spherical nanoparticle with a diameter of 50 nm and mono-energetic sources of photons with energies of 30,60, and 100 keV. The secondary electrons emitted from the nanoparticle were scored on the nanoparticle surface and the delivered dose to water around the nanoparticle was tallied using concentric shells with a thickness of 25 nm. Four different elements were used as materials of NPs, including Gold, Bismuth, Gadolinium, and Hafnium. Results: Our results showed a considerable difference in the number of emitted electrons per incident photon between the two codes. There were also discrepancies between the two codes in the energy spectra of secondary electrons. Calculated radial dose values around NPs in nano-scale had a similar pattern for both codes. However, significant differences existed for some elements. Conclusion: It can be concluded that the results of nano-scaled MC modeling for nanoparticle-based radiation therapy are dependent on the code type and its algorithm for electron transport as well as exploited cross-section libraries.

The risk assessment of local nuclear fuel melting in nuclear power plants (NPPs) due to its localized nature and the difficulty in accurate monitoring is of vital importance for safe operation of nuclear reactors. Experiences have shown that flow blockage accident without safety control rod axe man (SCRAM) can lead to local nuclear fuel melting which consequently affect the safety of NPPs. The purpose of this study is to analyze the transient states leading to local fuel melting based on ATWS-related events given in FSAR of the VVER-1000/V446 nuclear reactor including pump failure, local blockage, power level increasing and their combinations. In this work, a coupling framework is first developed based on the MCNPx and the COBRA/EN codes. After validation with available data, the results showed that despite 18% deviation from the mass flow rate reduction limitation, 470 kPa from the channel pressure drop limitation, and 204K from the clad temperature limitation in the most pessimistic situation, the reactor SCRAM does not occur. However, in these conditions (where SCRAM does not occur), 70% void fraction for 12 minitues is observed in some channels. Therefore, there may be dry spots and local melting of fuel in normal operational and ATWS conditions that need to be identified. According to the results, the occurrence of the void fraction above zero is locally expectant and an appropriate monitoring system should be used to identify weakness points of the system.