Bolus Effect Caused by Use of Thermoplastic Masks in Head and Neck Radiotherapy Treatments.

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.

Investigating polyurethane foam loaded with high-z nanoparticles for gamma radiation shielding compared to Monte Carlo simulations.

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.

Investigation of nano MgO loaded polyvinyl chloride polymer in protective clothing as a nonlead materials.

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.

A comparative Monte Carlo simulation study on shielding features of the CaF2-CaO-B2O3-P2O5-SrO-Ta2O5 glass system against X-ray by GEANT4 and MCNPX codes.

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.

Potential production of Ac-225 at Egyptian second research reactor (ETRR-2) through neutron induced transmutation of Ra-226.

A considerable focus has been paid to the production of 225Ac due to its effective therapeutic action in alpha-targeted radiotherapy. Considering the future global clinical demand, it is necessary to increase the production capacity of 225Ac. A feasibility study was conducted to investigate the production of 225Ac through neutron induced transmutation of 226Ra at the Egyptian Second Research Reactor (ETRR-2) using the MCNPX code. The calculations were carried out for 1 g of 226Ra target exposed to the highest neutron flux in the irradiation grid surrounding the reactor core. The 227Ra, 225Ra, 227Ac, and 225Ac generated activities as a function of irradiation and decay times were estimated. Our study revealed that in this non-linear production process, 39.22 MBq of pure 225Ac could be obtained after three days of irradiation, while 148.74 MBq could be obtained after fifteen days of continuous irradiation.

An investigation on protection properties of Tantalum (V) oxide reinforced glass screens on unexposed breast tissue for mammography examinations.

INTRODUCTION: 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. METHODS: 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. RESULTS: 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. CONCLUSION: 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. IMPLICATIONS FOR PRACTICE: 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.

A comparative study of a traditional localization algorithm and a deep learning model for radioactive particle tracking application.

Radioactive particle tracking is a nuclear technique that tracks a sealed radioactive particle inside a volume through a mathematical location algorithm, which is widely applied in many fields such as chemical and civil engineering in hydrodynamics flows. It is possible to reconstruct the trajectory of the radioactive particle using a traditional mathematical algorithm or artificial intelligence methods. In this paper, the traditional algorithm is based on solving a minimization problem between the simulated events and a calibration dataset, and it was written using C++ language. The artificial intelligence method is represented by a deep neural network, in which hyperparameters were defined using a Python optimization library called Optuna. This paper aims to compare the potentiality of both methods to evaluate the accuracy of the radioactive particle tracking technique. This study proposes a simplified model of a concrete mixer, six NaI(Tl) detectors, and a137Cs sealed radioactive particle. The simulated measurement geometry and the dataset (3615 patterns) were developed using the MCNPX code, which is a mathematical code based on the Monte Carlo Method. The results show a mean absolute percentage error (MAPE) of 20.81%, 10.33%, and 16.84% for x, y and z coordinates, respectively, for the traditional algorithm. For the deep neural network, MAPE is 6.87%, 2.70%, and 22.79% respectively for x, y and z coordinates. In addition, an investigation is carried out to analyze whether the size of the calibration dataset influences the performance of both methods.

Tungsten (VI) oxide reinforced antimony glasses for radiation safety applications: A throughout investigation for determination of radiation shielding properties and transmission factors.

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.

An alternative method for calculation of half-value layers without the knowledge of attenuation coefficient.

Radiation protection is crucial for the safe utilization of ionizing radiation and minimizing the harmful effect upon exposure, hence some standards have been defined by some relevant organizations for the safe uses of radiation. One of the parameters relevant to the calculation of gamma ray shielding is the half-value layer (HVL), which is normally calculated using the knowledge of linear attenuation coefficient (µ). In this research, an attempt has been made to directly calculate HVL without the knowledge of µ via Monte Carlo simulation technique. For this purpose, in the Monte Carlo N-Particle eXtended (MCNPX) code, F1, F5 and Mesh Popul sequences tallies were defined and the optimal structure for the least measurement error was introduced. The MCNPX calculated values showed reasonable agreement with the experimental findings. According to the obtained results, it is suggested that in order to reduce the error of HVL calculations, in exchange for the MCNPX code, the values of the R parameter and the radiation angle of the source should be considered according to the calculations introduced in this plan. Because the results show that by considering the measurement error between 6 and 20%, the code output can be cited in different energy ranges.

Investigating the effects of gold and titania nanoparticles in the treatment of breast cancer using a compressed breast phantom in the presence of high intensity low mono-energetic x-ray radiation.

Today, with the advancement of nanotechnology, nanomaterials with high atomic numbers such as gold and titania nanoparticles can be specifically concentrated in tumor cells in various ways and benefited from the advantage of increasing the dose due to the proximity of these elements next to cancer cells as a radiation sensitizer. In this research, parameters such as radiation energy (in the range of kilo voltage), nanoparticles concentration, and tumor depth were investigated in the compressed breast phantom by using the MCNPX code to investigate the effect of radio sensitivity. The tumor in the form of a cube with sides of 0.5 cm was labeled with spherical nanoparticles of titania and gold with a radius of 70 nm with different concentrations. The results showed that exposure to the range of kilo voltage causes photoelectric absorption to occur with a high probability and a relatively large dose is delivered to the tumor. The energy that can cause the most damage to the tumor was obtained at 65 keV in the presence of gold nanoparticles and in the range of 40 to 45 keV in the presence of titania nanoparticles. Also, with increasing concentration the dose enhancement factor increases, but with increasing depth, for dose enhancement factor does not change.

Toward the strengthening of radioprotection during mammography examinations through transparent glass screens: A benchmarking between experimental and Monte Carlo simulation studies.

Introduction: A 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. Methods: A 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 discussion: The 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.

Utilization of three-layers heterogeneous mammographic phantom through MCNPX code for breast and chest radiation dose levels at different diagnostic X-ray energies: A Monte Carlo simulation study.

Introduction: We report the breast and chest radiation dose assessment for mammographic examinations using a three-layer heterogeneous breast phantom through the MCNPX Monte Carlo code. Methods: A 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 discussions: The 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.

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

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

Simulation study of a simultaneous beta-gamma-ray detection using a 3-layer phoswich detector and Monte Carlo methods.

Combination of two or three dissimilar scintillator materials as a radiation detector has found major role in environmental radiation monitoring. In this paper, a three-layer Phoswich detector including BC-400, YAG, and CsI was designed to efficiently discriminate gamma-ray in the beta events up to 3.2 MeV using a simple rise-time discrimination method. MCNPX Monte Carlo code was used to obtain interaction probability of beta and gamma-rays as well as optimum thicknesses of the layers in the designing process. The optical transport of the system was simulated by GEANT4. In this regard, the pulses from simultaneous beta-gamma emitter sources were detected and discriminated based on pulse's rise-time so that the minimum number of gamma-ray contaminating events was observed in the beta spectrum. The results showed that using the proposed configuration and the method, output pulses with a rise-time shorter than 9 ns have been successfully detected as a beta particle while those with rising time longer than 15 ns have been identified as gamma-ray events. Overall results revealed that using the proposed system, an individual spectrum of beta particles or gamma-rays can be recorded from a simultaneous beta-gamma emitter source that minimizes contribution of the other radiation.

An extensive study on the neutron-gamma shielding and mass stopping power of (70-x) CRT-30K2O-xBaO glass system for 252Cf neutron source.

The present study focuses on the charged-uncharged particles shielding performance of the addition of a mixed type of cathode ray tube (CRT) in a glass system that is irradiated by the 252Cf neutron source via the MCNPX simulation and analytical calculations, as well as Phy-X: PSD and SRIM software. The CRT waste glass is inserted into the glass system with (70-x) CRT-30K2O-xBaO general formula for x = 0, 10, 20 mol% that produces CG1, CG2, and CG3 glass shielding materials. Using Watt Fission Distribution (WFD) and Doppler Effect (DE) the neutron-gamma photon spectra were extracted for shielded (in the presence of the glass materials) and unshielded (in air) cases. Some calculated attenuation parameters related to the neutron deduced that CG1 is the best neutron attenuator among the selected glass samples. Moreover, by increasing the density of the glass from CG1 to CG3, the ascending trend is observed for the linear attenuation coefficient (LAC, cm-1) of the studied glass, and the best shielding competence is monitored for CG3. Furthermore, two sharp peaks are found in Zeff graphs which may be due to K-edge absorption of Ba and Pb elements and by decreasing the Pb element from CG1 to CG3 the second peak gradually becomes smooth. In addition, Mass Stopping Power/ Projected Ranges of the proton (H1) and alpha particles (He+2) are also estimated by SRIM code and findings show that CG1 can better stop proton and alpha particles in comparison with the other chosen glass structures.

A closer look at the efficiency calibration of LaBr3(Ce) and NaI(Tl) scintillation detectors using MCNPX for various types of nuclear investigations.

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.

Introduction of Graphene/h-BN Metamaterial as Neutron Radiation Shielding by Implementing Monte Carlo Simulation.

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.

Investigative study of radiotoxicity of spent nuclear fuel assembly of some commercial nuclear power plants.

The European Pressurized Water Reactor (EPR) and Hualong One Pressurized Water Reactor (HPR) are two of the reactors under consideration by the Ghana Nuclear Power Programme. Radiotoxicity analysis of Spent Nuclear Fuel (SNF) assembly was carried out with these commercial Pressurized Water Reactor (PWR) nuclear power technology as case study. This will help determine which one is less radiotoxic on the environment between the two reactor technologies, in the long run. Burnup depletion calculation for the Uranium Oxide (UOX) fuel of these reactor technologies was simulated, using Monte Carlo Neutron Particle Extended (MCNPX), a code used in nuclear fuel management analysis, being a well validated code and also due to its versatile nuclei reactions cross section library. Determination of radiotoxicity for EPR and HPR SNF is the main objective of this study. The radiotoxicity was achieved taking into consideration the radioactive decay rate of the radionuclides and the Dose Factor of each radionuclide present in the SNF using the International Commission on Radiological Protection (ICRP) compendium of Dose Factors due to ingestion. The radiotoxicity for the two reactor's SNF were compared. The initial radiotoxicity for HPR SNF was higher in the duration below one hundred years but at about a hundred years and above, the radiotoxicity was higher for EPR SNF. The radiotoxicity was tremendously reduced for the reprocessed spent UOX fuel (with the Pu and U extracted) to be used as mixed oxide (MOX) fuel. The main finding is that Pu isotopes are the major contributors to the radiotoxicity of the SNF for the two reactors systems due to their very high radioactivity, long half-lifes and high dose factors as compared to other actinides and fission products present in the SNF.

Study of alpha spectrometry for detection of radon and progeny using gas micro-strip detector.

In this paper, simulation of a gas micro-strip detector by using the MCNPX code, the feasibility of alpha spectroscopy for radon and its progeny has been investigated. Initially, for the verification of the code, the range of alpha particles released from 222Rn gas in the air has been obtained in the standard condition, which is consistent with the experimental results. Subsequently, the energy loss per unit path length, range of alpha particles and radon progeny was measured, then the relationship between the range, the energy and the air pressure has been achieved. Finally, the simulation results have been compared with the results of the particle range relationship, pressure and energy, and the radon spectroscopy has been performed with the studied detector. The comparison of the spectrum obtained with the simulated micro-strip detector and the experimental results shows that the introduced micro-strip detector, in addition to the ability to measure radon and daughters, also has the ability to extract the spectrum from it.

An investigation of corium in Fukushima Daiichi Unit-1 accident.

The core melt composition resulting from the Fukushima Daiichi Unit1 Nuclear Power Plant (FD-U1) accident is essential for the corium characterization phase before decommissioning or handling of radioactive waste. Several models were applied by different research groups for the estimation of corium composition. In this paper, the investigation of the isotopic composition, and radioactivity of the radio-nuclides in the corium10 and 50 years post-accident were calculated using Monte Carlo code, MCNPX 2.7. The results showed that the estimated core materials inventory at reactor scram before core melt was about 123.97 ton, and after the formation of the corium melt was about 140.702 ton, which agrees with the predictions calculated using other models. Also, the total corium activity was about 6.046E+17 Bq and 1.89E+17 Bq 10 and 50 years post- accident, respectively. The radionuclide compositions in the corium are necessary for decommissioning plan of F-D-U1. Furthermore, RELAP/SCDAPSIM MOD3.4 code was used for the analysis of thermal performance of the FD-U1 reactor core starting from the time of the accident up to corium formation and slumping to the lower head of the reactor pressure vessel (RPV). Our analysis indicates that the hydrogen generation started on March11th, around 18:39. The results were compared with MELCOR code and OECD/NEA BSAF Phase I results, which were found in good agreement.