A pilot study of diabetes effects on radiation attenuation characteristics of tibia and femur of rats.

This study aimed to investigate the effect of diabetes on radiation attenuation parameters of the femur and tibia of rats using Monte Carlo Simulations. First, control and diabetic rats were identified and tibias and femurs were removed. Then, the elemental ratios of the bones obtained were calculated using EDS (Energy Dissipative X-ray Spectroscopy). Therefore, radiation permeability properties of control and diabetic bones were simulated by using the content ratios in the bones in MCNP6 (Monte Carlo N-Particle) and PHITS (Particle and Heavy Ion Transport code System) 3.22 and Stopping and Range of Ions in Matter (SRIM) simulation codes. Attenuation coefficient results were compared with the NIST database via XCOM. Although differences in absorption coefficients are observed at low energies, these differences disappear as the energy increases.

The Effect of Dose Enhancement in Tumor With Silver Nanoparticles on Surrounding Healthy Tissues: A Monte Carlo Study.

Objectives: Cancer-related death rates account for approximately one-third of all deaths, and this rate is increasing remarkably every year. In this study, we examined the dose enhancement factor (DEF) in the tumor and surrounding tissues by adding different concentrations of silver nanoparticles (AgNPs) to the brain tumor using the Monte Carlo (MC) technique. Methods: This study used MCNP6.2 simulation software. A Planning Target Volume (PTV) of 1 × 1 × 1 cm3 was placed in the center of a cubic cranial model with dimensions of 5 × 5 × 5 cm3. Five different simulations were initially generated using the simple method. These simulations included pure PTV and PTV consisting of 4 different silver concentrations (5, 10, 20, and 30 mg/g). Additionally, a model was created using the nanolattice method, considering the size, position, and distribution of the AgNPs. Irradiation was performed using a source with a 6 MV linac photon spectrum. Measurements were performed using the *f8 tally, and DEF values were calculated. Results: In the simulation study using the simple method, the DEF value of PTV increased linearly with concentration, whereas the DEF values were lower than the simulation results with the nanolattice model (1.9 vs 1.4 for 30 mg/g NP concentration). Performing the simple method, we observed no remarkable dose increase in lateral OARs surrounding PTV. While a remarkable dose decrease was observed in distal OARs, a dose increase in the proximal OAR was observed, which was consistent with that of PTV. However, according to the results obtained by performing the nanolattice method, the dose increase was observed in both the proximal OAR and the distal OAR and was similar to that of PTV. Conclusion: While enhancing the dose in the tumor by adding NPs into the tumor, it is essential to consider whether it also increases the OAR dose. In addition, simulation studies on NPs showed that the dose increase varied significantly with particle size, position, and distribution. Hence, these factors should be considered carefully.

A benchmark for Monte Carlo simulations in gamma-ray spectrometry Part II: True coincidence summing correction factors.

The goal of this study is to provide a benchmark for the use of Monte Carlo simulation when applied to coincidence summing corrections. The examples are based on simple geometries: two types of germanium detectors and four kinds of sources, to mimic eight typical measurement conditions. The coincidence corrective factors are computed for four radionuclides. The exercise input files and calculation results with practical recommendations are made available for new users on a dedicated webpage.

Responses of a 6LiI Bonner Sphere Spectrometer using the Monte Carlo codes PHITS and MCNPX.

The response functions (RFs) of a Bonner Sphere Spectrometer (BSS) with a 6LiI thermal neutron detector were calculated using the Monte Carlo codes PHITS (version 3.26) and MCNPX (version 2.7.0), with their own default nuclear data libraries, and physics models. RFs were compared with other published data, obtained for the same spectrometer using the MCNP6.1 code with its own physics models. A discussion on the influence of using different nuclear data libraries and physics models using these codes/versions is analyzed.

Monte Carlo simulation of micro-x-ray fluorescence (µ-XRF) based detection of immunomagnetically labeled tumor cells.

Objective. Circulating tumor cells (CTCs) carry crucial information related to the spreading and proliferation of tumors, especially at early stages of the disease. Despite the huge clinical potential held by CTCs in cancer therapy, capture and detection of these cells from the patient's peripheral blood system is rather challenging since CTCs are extremely rare cells. The objective of this paper is, based on Monte Carlo simulations, to propose the detection of immunomagnetically labelled tumor cells by micro-x-ray fluorescence (µ-XRF).Approach. The simulations were carried out with the Monte Carlo N-Particle, version 6.2, (MCNP6.2) code. The model simulates 20µm cancer cell lines and 10µm CTCs tagged with Fe3O4@SiO2spherical nanoparticles of diameters 25 nm, 60 nm and 110 nm. A 17.5 keV monochromatic, micro-focused x-ray beam of diameter 15µm, impinges on cancer cells immersed in a phosphate-buffered saline solution. The simulations also include a polymeric sample holder and a silicon drift detector with a beryllium window and silver collimator.Main results. The results show the dependence of the signal intensity (Fe Kαline) on cell and nanoparticle sizes. Samples containing two and three CTCs were also simulated in particular geometrical configurations. It is presented how the inter-cell distances and cell positions relative to the incident x-ray beam affect the signal. In addition, within the parameters used in the simulations,µ-XRF method provides a minimum detection limit of 9.4 pg of Fe, which corresponds to detecting a single 10µm CTC labeled with 110 nm Fe3O4@SiO2nanoparticles at 6.3% binding.Significance. Theµ-XRF based method proposed in this paper for detecting CTCs, combined with immunomagnetic nanoparticles (NPs), has the potential to be innovative in the field of liquid biopsy.

Fast Neutron Measurement System Using Prompt Gamma Neutron Activation Solid Converter: Monte Carlo Study.

Measuring fast neutron emission around accelerators is important for purposes of environmental monitoring and radiation safety. It is necessary to detect two types of neutrons: thermal and fast. Fast neutron spectroscopy is commonly employed using a hydrogen-recoil proportional-counter; however, its threshold is 2 MeV. The aim of this study was to expand PGNA converters based on KCl to fulfil the need to detect neutron energies ranging from 0.02 MeV to 3 MeV. In our previous research, we established a counting system comprised of a large converter of KCl with a NaI(Tl) gamma radiation spectrometer. The KCl converter is efficient for fast neutron prompt gamma emission. The potassium naturally includes a radioisotope that emits 1.460 MeV gamma rays. The presence of the constant level of 1.460 MeV gamma ray counts offers an advantage, providing a stable background for the detector. The study was carried out using MCNP simulations of the counting system with a variety of PGNA converters based on KCl. We concluded that KCl mixtures combined with other elements, such as PGNA converters, demonstrated improved detection performance for fast neutron emissions. Furthermore, an explication of how to add materials to KCl to provide a proper converter for fast neutrons was introduced.

Internal dosimetry of 64Cu-DOX-loaded microcapsules by Monte Carlo method.

Some of the issues regarding introducing new radiocompounds in nuclear medicine are the distribution patterns, delivered dose to different organs, diagnostic abilities and side effects. In this study, in order to assess the biodistribution of 64Cu-DOX-loaded microcapsules, rats were IV-injected with the microcapsules, and 1, 4, 14, and 24 h later, the activities of the targeted organs were measured (%ID/g). The accumulated activities were achieved by %ID/g curves, and S-factors were obtained by MCNP outputs. The MIRD formulation and Monte Carlo method were used to determine the absorbed dose in the target organs. The biodistribution data and PET-CT images showed that the lungs were where the majority of activity was seen. According to MIRD and MCNP, the maximum dose delivered in the lungs was 5.79E+01 mGy/MBq and 4.70E+01 mGy/MBq, respectively. Also, the effective dose was 1.2E+01 for MIRD and 8.31E+00 mSv/MBq for MCNP. These results indicate that 64Cu-DOX microcapsules can be considered a new radiocompound in pulmonary imaging, and MCNP simulation can be a reliable method for internal dosimetry.

Monte Carlo calculation of whole body counter efficiency factors for different computational phantoms.

Individual monitoring can provide an estimate of the radioactivity present in the body of the exposed individuals. Periodic monitoring of occupationally exposed individuals is of great importance in case of accidental incorporation. Computational phantoms and Monte Carlo codes are often used to complement the calibration method of counting systems in internal dosimetry. Here, counting efficiency (CE) factors for a WBC system were calculated using MC simulations. The WBC system with a NaI(Tl) detector and the BOMAB phantom was modeled using three MC codes. After validation, the models were used to obtain CE values for a wide range of energies, and a CE curve was generated for the WBC system. To estimate the effects of anatomical differences on the measurement process, two anthropomorphic voxel phantoms were modeled using the VMC code. For the detector position with the highest CE value, the differences when comparing BOMAB results with the MaMP and Yale results were (-1 ± 6)% and (-1 ± 3)%, respectively. The results confirm that the use of the BOMAB phantom is a good approach for the calibration of the whole-body counter system. Measurements should be made at detector position with the highest CE values, and it is recommended to use the mean Monte Carlo CE values calculated in this work.

Spectra, fluence and absorbed doses in sensitive organs due to scattered X-rays during a chest CT.

The use of ionizing radiation for the treatment and diagnosis of diseases is becoming more frequent. The technologies associated with diagnostic imaging are constantly evolving, allowing faster and cheaper diagnoses to benefit the patient. However, this has caused an increase in the exposure to ionizing radiation of patients and health professionals. One of the diagnostic techniques for obtaining high-resolution anatomical images of patients is computed tomography (CT). Due to the detail and quality of the images obtained with CT, its use is becoming more frequent. The information provided by these images allows the specialist to make better diagnoses; however, exposure to X-rays deposits a dose in the patient. CT represents approximately 20% of all X-ray examinations but it is responsible for 70% of the medical dose accumulated by the patient. During the acquisition of the images, the highest dose is deposited in the area of the body whose image is to be obtained. During the incidence of X-rays, there is dispersion of these that reach sensitive organs whose dose is not evaluated. The objective of this work was to estimate, using Monte Carlo methods, the fluence and X-ray spectra and to obtain a factor that allows knowing the absorbed dose in sensitive organs due to scattered radiation during a chest CT. With the MCNP5 code, the CT equipment and a hybrid anthropomorphic phantom, type BOMAB it was found that the absorbed dose in these organs depends on the size of the organ and the distance between the organ and the surface of the slice on the thorax where the X-rays are incident.

MCNPX simulation and experimental validation of an unmanned aerial radiological system (UARS) for rapid qualitative identification of weak hotspots.

Nuclear threats such as dirty bombs and illicit trafficking of radioactive sources are major concerns of humanity. Fast detection and accurate localization of radioactive material out of regulatory control (MORC) by autonomous and semi-autonomous monitoring systems like robots can help to reduce radiation exposure to the public and workers, and it will improve security and peace in the world. This study proposes an autonomous radiological monitoring system consisting of a 2-inch NaI detector coupled to a PM tube and mounted on a multi-rotor UAV to detect radioactive sources. First, an experimental scenario was modeled using the MCNPX Monte Carlo (MC) code. In this modeling, the gamma spectra in 15 detectors were recorded from the rays emitted simultaneously from the areas' sources. The total count under the spectrum was measured for each of the detectors at different heights. The experimental tests were also performed to detect the simultaneous effect of five low-level Co-60 and Cs-137 point sources on a soccer field. Next, the modeling results were compared with the experimental ones, which showed good agreement and the capability to use MC modeling to simulate different radiological scenarios. The experimental results also showed that at 50 cm, all radioactive sources were successfully detected in their actual location. By decreasing the flight height, the ability of the monitoring unmanned aerial to detect radioactive sources was increased significantly.

Benchmarking of stray neutron fields produced by synchrocyclotrons and synchrotrons in compact protontherapy centers (CPTC) using MCNP6 Monte Carlo code.

Proton therapy is an external radiotherapy using proton beams with energies between 70 and 230 MeV to treat some type of tumours with outstanding benefits, due to its energy transfer plot. There is a growing demand of facilities taking up small spaces and Compact Proton Therapy Centers (CPTC), with one or two treatment rooms, supposing the technical response of manufacturers to this request. A large amount of stray radiation is yielded in the interaction of proton beam used in therapy, neutrons mainly, hence, optimal design of shielding and verifications must be carried out in commissioning phases. Currently, almost 50 proton centers are under construction and start up in several countries, including ten in Spain. In the present work the effectiveness of shielding in two CPTC was verified with the Monte Carlo code MCNP6 by calculating the ambient dose equivalent, H*(10) due to secondary neutrons, outside the enclosures and walls of the center. The facilities modelled were the two centers currently operating in Spain, the first, since December 2019, with a superconductor synchrocyclotron, and the second, since March 2020, with a compact synchrotron. The geometry and materials are based on dimensions proposed a priori by the vendors, therefore, the paper is focused on check the suitability of the materials and thickness of the walls of the centers. Several models of the radiation sources were simulated, starting from a conservative assumptions, followed by more realistic scenarios. In all cases, the results reached for the ambient dose equivalent, H*(10), were below 1 mSv/year, which is the legal limit considered for the public in international references. Finally, considering that the recent ICRU Report 95 proposes changes in the operational quantities, the dose outside shieldingt has been evaluated in terms of the new next area surveillance quantity, H*, known as ambient dose, in the process of implementation.

Calculation of absorbed dose in paediatric phantoms using Monte Carlo techniques for 18F-FDG and 99mTc-DMSA and the new TIAC.

The radiopharmaceuticals most commonly used in nuclear medicine are 18F-FDG and 99mTc-DMSA, both of which are administered to paediatric and adult patients using the same time activity coefficient. However, the IAEA recommends specific paediatric dosimetry. The aim of this work (TW) was to estimate the absorbed dose for 18F-FDG and 99mTc-DMSA using two paediatric voxel phantoms (Baby and Child) by Monte Carlo techniques. Biokinetic data for both radiopharmaceuticals were obtained from ICRP 128. In addition, the new time-integrated activity coefficient (TIAC) values from a recent publication were examined for the following organs: Brain, urinary bladder wall, liver, heart wall, and lung. The absorbed dose per injected activity (AD/IA) and effective dose per injected activity (E/IA) values were calculated for both phantoms and the results were compared with simulated data of paediatric phantoms from ICRP 128, MIRDcalc software and available literature. Regarding AD/IA in organs, differences of up to 61% and 115% were found for the Baby phantom and 120% and 167% for the Child phantom using 18F-FDG and 99mTc-DMSA, respectively. For FDG using the new TIAC, a maximum difference of 244% was observed. For E/IA, the maximum differences were 27% and 31% for the Baby and Child phantoms, respectively, for FDG and DSMA. In this study, new dosimetric data were calculated using Baby and Child phantoms and the newly recommended TIAC.

Characterization of photoneutron fluxes emitted by electron accelerators in the 4-20 MeV range using Monte Carlo codes: A critical review.

Applications of electron accelerators range from nuclear waste package assay and security-related tasks to radiation therapy. Studies aiming at characterizing photoneutron fluxes generated by electron accelerators are usually based on Monte Carlo simulation. In this paper, we critically review the performance of Monte Carlo transport codes to simulate photoneutron fluxes emitted by electron accelerators operating between 4 and 20 MeV, typically the energy range of interest for the aforementioned applications. First, we go through the state of the art and lay the foundations of current theoretical knowledge on photoneutrons. By carrying out additional investigations, we show that contamination of photoneutron fluxes by electroneutrons is likely to lie between 0 and 2%. Second, we assess the characteristics of photoneutron fluxes emitted by tungsten or tantalum conversion targets and by heavy water or beryllium secondary targets. This characterization step is conducted with MCNP6.2, which is often considered as one of the reference Monte Carlo codes, and built around three parameters, i.e., photoneutron yield cross-sections, energy spectrum and angular distribution. In particular, we demonstrate that erroneous parameters in the nuclear data of MCNP6.2 lead to (γ, xn) cross-section threshold errors for two tungsten isotopes, i.e., 182W and 186W, inducing in turn a global underestimation of photoneutron production in tungsten. Furthermore, by taking an in-depth look at nuclear data libraries, we show that photoneutron yield cross-sections are sometimes poorly evaluated below 20 MeV, e.g., 2H, 9Be and 184W. Third, thanks to vanadium and aluminium foils, we benchmark MCNP against photoneutron activation measurements conducted in the vicinity of three different electron accelerators, including a medical one. Simulation of these measurements denote a systematic underestimation trend extending from a few percent to a factor ten. Recent findings reported in the literature proved that photoneutron kinematics is implemented in MCNP with erroneous equations related to neutron inelastic scattering, causing hardening of photoneutron energy spectrum and may explain in part the discrepancies encountered in this MCNP benchmark study. Finally, in light of the three main sources of errors that potentially lead to unreliable results when simulating photoneutron fluxes with Monte Carlo codes - implementation of nuclear data, modelling of photonuclear physics and fundamental knowledge of photoneutron yield cross-sections - we issue recommendations for the code developers and users. Until further progress is made in the field of photoneutron simulation, mastering the current limitations of Monte Carlo codes could be the first milestone for their users.

2D inversion of in situ gamma-ray spectrometric measurements of 137Cs for site characterization.

Environmental contamination by radioactive materials can be characterized by in situ gamma surface measurements. During such measurements, the field of view of a gamma detector can be tens of meters wide, resulting in a count rate that integrates the signal over a large measurement support volume/area. The contribution of a specific point to the signal depends on various parameters, such as the height of the detector above the ground surface, the gamma energy and the detector properties, etc. To improve the spatial resolution of the activity concentration, contributions of a radionuclide from nearby areas to the count rate of a single measurement should be disentangled. The experiments described in this paper, deployed 2D inversion of in situ gamma spectrometric measurements using a non-negative least squares-based Tikhonov regularization method. Data were acquired using a portable LaBr3 gamma detector. The detector response as a function of the distance of the radioactive source, required for the inversion process, was simulated using the Monte Carlo N-Particle (MCNP) transport code. The uncertainty on activity concentration was calculated using the Monte Carlo error propagation method. The 2D inversion methodology was first satisfactorily assessed for 133Ba and 137Cs source activity distributions using reference pads. Secondly, this method was applied on a 137Cs contaminated site, making use of above-ground in-situ gamma spectrometry measurements, conducted on a regular grid. The inversion process results were compared with the results from in-situ borehole measurements and laboratory analyses of soil samples. The calculated 137Cs activity concentration levels were compared against the activity concentration value for exemption or clearance of materials which can be applied by default to any amount and any type of solid material. Using the 2D inversion and the Monte Carlo error propagation method, a high spatial resolution classification of the site, in terms of exceeding the exemption limit, could be made. The 137Cs activity concentrations obtained using the inversion process agreed well with the results from the in-situ borehole measurements and those from the soil samples, showing that the 2D inversion is a convenient approach to deconvolute the contribution of radioactive sources from nearby areas within a detector's field of view, and increases the resolution of spatial contamination mapping.

Comparison between MCNP and planning system in brachytherapy of cervical cancer.

Absorbed doses in uterus during brachytherapy were calculated with MCNP in relevant points and compared with planning system for one patients. MCNP was applied with two different humanoid phantoms in input, ORNL and voxel models, which represent human body in mathematical way. Good agreement between both phantoms, as well as, between MCNP and planning system were found. In addition the doses in critical organs (bladder and colon in this kind of therapy), were calculated and compared with maximal doses in these organs obtained from planning system for 15 other patients. MCNP doses agree well with planning system in points of uterus for those 15 patients, where radioactive source is used to apply. However, there are systematical discrepancies between doses in colon and bladder obtained by MCNP and planning system.

Toolkit implementation to exchange phase-space files between IAEA and MCNP6 monte Carlo code format.

PURPOSE: Some Monte Carlo simulation codes can read and write phase space files in IAEA format, which are used to characterize accelerators, brachytherapy seeds and other radiation sources. Moreover, as the format has been standardized, these files can be used with different simulation codes. However, MCNP6 has not still implemented this capability, which complicate the studies involving this kind of sources and the reproducibility of results among independent researchers. Therefore, the purpose of this work is to develop a tool to perform conversions between IAEA and MCNP6 phase space files formats, to be used for Monte Carlo simulations. MATERIALS AND METHODS: This paper presents a toolkit written in C language that uses the IAEA libraries to convert phase space files between IAEA and MCNP6 format and vice versa. To test the functionality of the provided tool, a set of verification tests has been carried out. In addition, a linear accelerator treatment has been simulated with the PENELOPE library using the PenEasy framework, which is already capable to read and write IAEA phase space files, and MCNP6 using the developed tools. RESULTS: Both codes show compatible depth dose curves and profiles in a water tank, demonstrating that the conversion tools work properly. Moreover, the phase space file formats have been converted from IAEA to MCNP6 format and back again to IAEA format, reproducing the very same results. CONCLUSION: The toolkit developed in this work offers MCNP6 scientific community an external and validated program able to convert phase space files in IAEA format to MCNP6 internal format and use them for Monte Carlo applications. Furthermore, the developed tools provide also the reverse conversion, which allow sharing MCNP6 results with users of other Monte Carlo codes. This capability in the MCNP6 ecosystem provides to the scientific community the ability not only to share radiation sources, but also to facilitate the reproducibility among different groups using different codes via the standard format specified by the IAEA.

A comparison study of GEANT4 and MCNP6 on neutron-induced gamma simulation.

Neutron-induced gamma simulation has multiple applications in various fields, such as radiation therapy, imaging, and nuclear well logging. Among them, Monte Carlo Codes, MCNP6, and GEANT4 are suitable solutions for nuclear detection. Since there are few published comparisons of GEANT4 with MCNP6, especially for the simulation of neutron-induced gamma spectra, therefore, the aim of this paper is to compare the GEANT4 and MCNP6 in the simulation of low-energy (less than 20 MeV) neutron-induced inelastic and capture gamma spectra, the feasibility of which could help to provide an alternative approach for researchers where MCNP6 is not available. Two representative models pertaining to nuclear well logging applications are designed and employed for the purpose of comparison, based on which neutron-induced inelastic and capture gamma spectra are analyzed using different methods associated with each code specifically, such as Time Window, Energy Cutoff, and Physics Tracking. Based on the cross-section library ENDF/B-VII.1, the results obtained from MCNP 6.1 and GEANT4 10.6 match well, which demonstrates the feasibility of using either code in such applications.

Comparison of Monte Carlo simulations and theoretical calculations of nuclear shielding characteristics of various borate glasses including Bi, V, Fe, and Cd.

The aim of this study is to evaluate both buildup factors and photon attenuation effectiveness of some borate glasses doped by Cd, Fe, V, and Bi. The mass attenuation coefficients (µm) of these glass systems have been calculated via MCNP6 and GEANT4 computer simulation codes over 0.02-10 MeV energy range and compared with the theoretical results of WinXCOM program. And then, half-value layer, (HVL), mean free path (MFP), and effective atomic number (Zeff) for the different content based glasses have been determined. Through G-P fitting process, exposure buildup factors (EBF) have been found in the energy range of 0.015-15 MeV up to 40 mfp. It can be concluded that SrBiO20 glass is alternative material in terms of photon attenuation, respectively.

Sensitivity study of C/O logging measurements by Monte Carlo method.

Carbon oxygen ratio (C/O) logging has great importance in the accurate determination of hydrocarbon saturation in the reservoir region. This measurement is independent of the salinity of the formation water, unlike alternative logging methods. Analysis of the measurement requires modelling of the time-dependent coupled neutron-gamma field produced by the tool, which is most efficiently done by the Monte Carlo (MC) method. MC simulation can be used to generate the gamma spectrum at the detectors of the probe for a variety of rock physics conditions and borehole environments and thus the C/O can be determined by the processing of the simulated gamma spectrum. The simulation results are used to derive the interpretation diagrams for the basic petrophysical effects and to investigate the role of the side effects. Considering the industrial practice of log evaluation, the resolution and limitations of the method is quantified in the measurement space by defining a goodness factor based on the area of the interpretation chart. The focus of this paper is on detector arrangement, but it also covers the effect of porosity, lithology, and the casing.

Comparison of PHITS2.88 and MCNP6.1 for the characterization of a LUPIN-II neutron area monitor.

The use of different nuclear data libraries and physics models can be a source of discrepancies in neutron transport simulation. Different Monte Carlo simulation toolkits can be used to characterize neutron monitors, these codes usually employ by default different nuclear data libraries and physics models. This work presents, for the first time, a comparison of MCNP and PHITS for the characterization of a LUPIN-II neutron rem-meter. The most significant discrepancies between the codes have been found around 100 MeV.