Monte Carlo N-Particle eXtended Code Research Articles

Designing Flexible Shield From Lead With High Protection Against X-rays in Operating Room Using Monte Carlo Simulation

Background: The use of non-flexible lead shields in X-ray detection systems requires a high cost, and it is heavy and subject to frequent cracks and fractures. Therefore, the shield should be designed in a way that has less weight, appropriate efficiency in reducing ionizing radiation, and a more reasonable price. Methods: In the operating room using the MCNPX (Monte Carlo N-Particle extended) code, the shield compounds were modeled to determine the optimal composition (with less weight, higher efficiency in removing ionizing radiation, and in more appropriate price), and then the amount of dose was calculated around the lead chamber with 10 cm from the walls of a detection system. Results: Using a weight combination of 80% lead and 20% polyvinyl chloride was a suitable option in the diagnostic radiology domain which has less weight, lower cost, and good efficiency in reducing X-rays compared to lead shields. Conclusion: The amount of dosage rate on all sides outside the chamber became less than 1 µSv/h, indicating that the combination of lead and polyvinyl chloride is suitable for the body of the chamber.

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.

Monte Carlo simulation of the mass attenuation coefficient and effective atomic number of the Eremurus-Rhizophora ssp. particleboard phantom at the mammography energy range

The earlier investigations proved the potential of the Rhizophora spp. as a phantom material. In this study, the mass attenuation coefficients and empirical and theoretical effective atomic numbers of Eremurus-Rhizophora spp. particleboards in three particle sizes and two adhesive treatment levels were evaluated in 16.63–60 keV photon energy range, using theoretical (XCOM: NIST Photon Cross Sections Database Code), simulation (MCNPX: Monte Carlo N-Particle eXtended Code), and experimental methods. The findings of this study were discussed in terms of the various calculating approaches, and also comparing them with raw wood, binderless particleboard, water, and different breast tissues. The mass attenuation coefficients of Eremurus-Rhizophora spp. samples were seriously agreed with young age breast tissue in three determination methods. Although the empirical effective atomic number (Zeff) is independent of photon energy and the theoretical Zeff is affected by it, the empirical and theoretical effective atomic numbers of the Eremurus-Rhizophora spp. particleboards are in good agreement in the mammography photon energy range. Moreover, the results of this study showed that the radiation features of Rhizophora spp. particleboard can be significantly improved by adding the Eremurus spp. root. The results prove the validity of the MCNP method as an alternative to the XCOM and experimental methods for evaluating the attenuation features and effective atomic numbers of the bio-composites.

COMPARISON OF SPECTRA AND MEAN GLANDULAR DOSE WITH TUBE VOLTAGES USED IN DIGITAL BREAST TOMOSYNTHESIS FROM SIMULATED, METROLOGICAL AND CLINICAL CASES.

Digital breast tomosynthesis (DBT) is a screening and diagnostic modality that acquires images of a breast at multiple angles during a short scan. The Selenia Dimensions (Hologic, Bedford, Mass) DBT system can perform both full-field digital mammography and DBT. The system acquires 25 projections over a 15° angular range (from -7.5° to +7.5°). X-ray spectroscopy is generally linked with a high-resolution semiconductor detector through a correction to its energy response function. The energy spectrum describes the radiation field, in which several quality parameters can be extracted, such as the effective energy, half-value layer and exposure. X-ray spectroscopy is usually performed with solid-state semiconductor detectors. Radiation dose is a concern in mammography, as the current protocols recommend that medical physicians evaluate mean glandular dose (MGD) as a part of service quality control. Studies are needed for radiation dose optimization from tomosynthesis patients. The COMET metrological X-ray tube, considered as with a constant potential and cooled, has proved to be a crucial tool in order to obtain the high energy resolution for low-energy radiographs in mammography. The Monte Carlo method, through Monte Carlo N-Particle eXtended (MCNPX), was proven to be an essential tool for image formation and posterior analysis of the deposited dose from breast simulators and radiographic contrast evaluation, for several anode/filter combinations. The purpose of this work was to assess the MGD and spectra in slabs of polymethyl methacrylate (PMMA) and breast equivalent thicknesses using four experiments with a Hologic Selenia Dimensions mammography X-ray tube with multimeter, a spectrometer (only for spectra, in this case), a metrological X-ray tube with a multimeter, and the MCNPX code. References indicate that the real conditions for a mammography X-ray tube that conducts tomosynthesis include tube voltages of 26, 29, 30 and 33kVp. Taking into account several thicknesses of PMMA, both the MGD and spectral results were in accordance with the references. Most of the spectra were in accordance with the references, showing that the resources used in the experiments can evaluate the energy level received by a patient. The MGD values were lower than those in the references from 30 to 50mm PMMA, and the data can be used for improvements in the detectors used in the Laboratory of Metrology in the State of Rio de Janeiro University, Brazil. Additionally, in the future, optimization of image quality can be performed for both semiconductors and mammography X-ray equipment.

Medium-thickness-dependent proton dosimetry for radiobiological experiments

A calibration method was proposed in the present work to determine the medium-thickness-dependent proton doses absorbed in cellular components (i.e., cellular cytoplasm and nucleus) in radiobiological experiments. Consideration of the dependency on medium thickness was crucial as the linear energy transfer (LET) of protons could rise to a sharp peak (known as the Bragg peak) towards the end of their ranges. Relationships between the calibration coefficient R vs medium-layer thickness were obtained for incident proton energies of 10, 15, 20, 25, 30 and 35 MeV, and for various medium thicknesses up to 5000 μm, where R was defined as the ratio DA/DE, DA was the absorbed proton dose in cellular components, and DE was the absorbed proton dose in a separate radiation detector. In the present work, DA and DE were determined using the MCNPX (Monte Carlo N-Particle eXtended) code version 2.4.0. For lower incident proton energies (i.e., 10, 15 and 20 MeV), formation of Bragg-peak-like features were noticed in their R-vs-medium-layer-thickness relationships, and large R values of >7 and >6 were obtained for cytoplasm and nucleus of cells, respectively, which highlighted the importance of careful consideration of the medium thickness in radiobiological experiments.

Optimization of a Neutron Long Counter Design by Monte Carlo Simulation.

In a search to optimize neutron long counter design for overall efficiency and flat energy response, Monte Carlo simulations were carried out for a variety of detector design parameters using the Monte Carlo N-Particle Extended code. Based on the standard long counter design by McTaggart, moderator diameter, moderator back length, and longitudinal hole diameter were sequentially varied, and the sensitivity of each parameter to the long counter response was systematically analyzed. For each design, simulations were done in the neutron energy range of 1 keV to 10 MeV. From the simulation results, it turned out that out of the three moderator parameters, the moderator diameter is most sensitive for optimizing the long counter response. As the last design parameter, the effect of the central slow-neutron counter was investigated, which showed a significant difference in the response. The investigation of each design parameter gave clear insight on its effect on the long counter response and enabled one to determine the optimum condition.

Calculation of air kerma inside the radiation field of X-ray tube

The air kinetic energy released per unit of mass (air kerma) quantity is of importance in both medicine and industry for evaluating the radiation field of X-ray tubes and also for calibrating the reference instruments. The air kerma spatial distribution inside the X-ray tube's radiation field is not uniform because of the heel effect phenomenon. In this paper, a combination of Monte Carlo N Particle eXtended (MCNPX) simulation code and artificial neural network (ANN) has been utilized to estimate the air kerma inside the radiation field of X-ray tube for a wide range of tube voltages (50–600 kV) which covers both medical and industrial applications. At first, an X-ray tube with an anode angle of 20° was simulated using MCNPX code. Next, in order to provide the required data for training the network, 1375 point detectors were placed in different positions inside the conical radiation field of the simulated X-ray tube and then the air kerma was calculated for tube voltages in the range of 50–600 kV with a step of 50 kV. The tube voltage and point detector's spherical coordinates including distance from target, tangential angle and polar angle were utilized as the 4 inputs and the air kerma was used as the output of the ANN. After training the ANN, the proposed ANN model could estimate the air kerma in every position inside the radiation field of X-ray tube for a wide range of voltages (50–600 kV) with a mean relative error (MRE) of less than 0.67%.

Investigation of different sources in order to optimize the nuclear metering system of gas–oil–water annular flows

The used metering technique in this paper is based on the multienergy (at least dual) gamma-ray attenuation. The aim of the current study is investigation of different combinations of sources in order to find the best combination for precise metering gas, oil and water percentages in annular three-phase flows. The required data were generated numerically using Monte Carlo N Particle extended (MCNPX) code. As a matter of fact, the current investigation devotes to predict the volume fractions in the annular three-phase flow, on the basis of a multienergy metering system including different radiation sources and one sodium iodide detector, using the hybrid model. Since the summation of volume fractions is constant, a constraint modeling problem exists, meaning that the hybrid model must predict only two volume fractions. Six hybrid models associated with the number of applied radiation sources are employed. The models are applied to predict the oil and gas volume fractions. For the next step, the hybrid models are trained based on numerically obtained data from the MCNPX code. The results show that the best prediction results are obtained for the oil and gas volume fractions of a system with the (241Am & 137Cs) radiation sources.

Simulation of a complete X-ray digital radiographic system for industrial applications

Simulating X-ray images is of great importance in industry and medicine. Using such simulation permits us to optimize parameters which affect image's quality without the limitations of an experimental procedure. This study revolves around a novel methodology to simulate a complete industrial X-ray digital radiographic system composed of an X-ray tube and a computed radiography (CR) image plate using Monte Carlo N Particle eXtended (MCNPX) code. In the process of our research, an industrial X-ray tube with maximum voltage of 300 kV and current of 5 mA was simulated. A 3-layer uniform plate including a polymer overcoat layer, a phosphor layer and a polycarbonate backing layer was also defined and simulated as the CR imaging plate. To model the image formation in the image plate, at first the absorbed dose was calculated in each pixel inside the phosphor layer of CR imaging plate using the mesh tally in MCNPX code and then was converted to gray value using a mathematical relationship determined in a separate procedure. To validate the simulation results, an experimental setup was designed and the images of two step wedges created out of aluminum and steel were captured by the experiments and compared with the simulations. The results show that the simulated images are in good agreement with the experimental ones demonstrating the ability of the proposed methodology for simulating an industrial X-ray imaging system.

Commissioning dosimetry and in situ dose mapping of a semi-industrial Cobalt-60 gamma-irradiation facility using Fricke and Ceric-cerous dosimetry system and comparison with Monte Carlo simulation data

Characterization of a 90 kCi (3330 TBq), semi-industrial, cobalt-60gamma irradiator was performed by commissioning dosimetry and in-situ dose mapping experiments with Ceric-cerous and Fricke dosimetry systems. Commissioning dosimetry was carried out to determine dose distribution pattern of absorbed dose in the irradiation cell and products. To determine maximum and minimum absorbed dose, overdose ratio and dwell time of the tote boxes, homogeneous dummy product (rice husk) with a bulk density of 0.13g/cm3 were used in the box positions of irradiation chamber. The regions of minimum absorbed dose of the tote boxes were observed in the lower zones of middle plane and maximum absorbed doses were found in the middle position of front plane. Moreover, as a part of dose mapping, dose rates in the wall positions and some selective strategic positions were also measured to carry out multiple irradiation program simultaneously, especially for low dose research irradiation program. In most of the cases, Monte Carlo simulation data, using Monte Carlo N-Particle eXtended code version MCNPX 2.7., were found to be in congruence with experimental values obtained from Ceric-cerous and Fricke dosimetry; however, in close proximity positions from the source, the dose rate variation between chemical dosimetry and MCNP was higher than distant positions.

Application of Prompt Self-Powered Neutron Detectors to the Lead-Cooled Fast Reactor Demonstrator ALFRED: Validation of the Monte Carlo Model for Selected SPNDs

The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the industrial prototype European lead fast reactor proposed in lead-cooled European advanced demonstration reactor. It has a relatively low power (125 MWe) with a compact design to reduce the cost but maintaining its representativeness and it is cooled by pure lead. One of the open issues is linked to the neutron flux in-core monitoring system because of the harshness of the environment the detectors should be installed in, due to high temperatures, and the neutron-gamma radiation field levels. Monte Carlo simulation is a possible way of facing the problem, reproducing into a virtual world the reactor core, the surrounding environment and radiation interactions. In previous works, neutron spectra and gamma doses at possible detectors' locations in ALFRED were retrieved, with consideration on the applicability of each suitable device currently available. Fission chambers (FCs) were found to be exploited at reactor start-up and intermediate power range. Prompt self-powered neutron detectors (SPNDs) seemed to be the best solution to monitor the reactor full power, becoming the main research target: their effective applicability on field has to be demonstrated. SPND applications do not include reactor control purposes usually. Moreover, their irradiation experience involved thermal and epithermal neutron spectra monitoring, mainly. The lack of data when SPNDs sense fast neutron fluxes in terms of prompt-response pushed the authors to deepen the study in such direction. The work herein shows the mathematical approach based on Monte Carlo simulation of SPNDs by the Monte Carlo N-particle eXtended code (MCNPX), so as to study the capability of the code in reproducing real devices' signals while experimented on field. Such a verification turned out to be the preliminary stage for studying new concepts for SPNDs, in terms of sensitive materials and geometries, envisaging the possibility for designing, prototyping, and testing new devices in suitable fast neutron-flux facilities.

A new approach for radiosynoviorthesis: A dose-optimized planning method based on Monte Carlo simulation and synovial measurement using 3D slicer and MRI.

Recently, there has been a growing interest in a methodology for dose planning in radiosynoviorthesis to substitute fixed activity. Clinical practice based on fixed activity frequently does not embrace radiopharmaceutical dose optimization in patients. The aim of this paper is to propose and discuss a dose planning methodology considering the radiological findings of interest obtained by three-dimensional magnetic resonance imaging combined with Monte Carlo simulation in radiosynoviorthesis treatment applied to hemophilic arthropathy. The parameters analyzed were: surface area of the synovial membrane (synovial size), synovial thickness and joint effusion obtained by 3D MRI of nine knees from nine patients on a SIEMENS AVANTO 1.5 T scanner using a knee coil. The 3D Slicer software performed both the semiautomatic segmentation and quantitation of these radiological findings. A Lucite phantom 3D MRI validated the quantitation methodology. The study used Monte Carlo N-Particle eXtended code version 2.6 for calculating the S-values required to set up the injected activity to deliver a 100 Gy absorbed dose at a determined synovial thickness. The radionuclides assessed were: 90Y, 32P, 188Re, 186Re, 153Sm, and 177Lu, and the present study shows their effective treatment ranges. The quantitation methodology was successfully tested, with an error below 5% for different materials. S-values calculated could provide data on the activity to be injected into the joint, considering no extra-articular leakage from joint cavity. Calculation of effective treatment range could assist with the therapeutic decision, with an optimized protocol for dose prescription in RSO. Using 3D Slicer software, this study focused on segmentation and quantitation of radiological features such as joint effusion, synovial size, and thickness, all obtained by 3D MRI in patients' knees with hemophilic arthropathy. The combination of synovial size and thickness with the parameters obtained by Monte Carlo simulation such as effective treatment range and S-value, from which is calculated the injected activity, could be used for treatment planning in RSO. Data from this methodology could be a potential aid to clinical decision making by selecting the most suitable radionuclide; justifying the procedure, fractioning the dose, and the calculated injected activity for children and adolescents, considering both the synovial size and thickness.

Semi-insulating GaAs based detector of fast neutrons produced by D–T nuclear reaction

We have examined semi-insulating (SI) GaAs detectors with high density polyethylene (HDPE) conversion layer by a mono-energetic neutrons with kinetic energy of 16.755 MeV generated by a deuterium—tritium nuclear reaction. First, the influence of HDPE layer thickness on the relative detection efficiency of fast neutrons was studied. The MCNPX (Monte Carlo N-particle eXtended) code has been used to support the analysis of the experiment. The theoretical optimum thickness of the conversion layer was determined to 1.9 mm using the MCNPX code. The HDPE conversion layers of various thicknesses, in the range from 50 μ m to 3200 μ m, were glued on the top Schottky contact of SI GaAs detector in the experiment. The neutron detection efficiency was evaluated from measured spectra and compared to results from simulations. The experimental data showed very good agreement with simulation results. Then the effect of active detector thickness modified by detector reverse bias on neutron detection efficiency was studied. Finally, the effect of the angle of irradiation on neutron detection efficiency was evaluated exhibiting decreasing tendency with increasing deviation from perpendicular direction of impinging neutrons.

Radiation detector based on 4H-SiC used for thermal neutron detection

In this work we have focused on detection of thermal neutrons generated by 239Pu–Be isotopic neutron source. A high quality liquid phase epitaxial layer of 4H-SiC was used as a detection region. The thickness of the layer was 70 μ m and the diameter of circular Au/Ni Schottky contact was 4.5 mm. Around the Schottky contact two guard rings were created. The detector structure was first examined as a detector of protons and alpha particles for energy calibration. Monoenergetic protons of energies from 300 keV up to 1.9 MeV were used for detector energy calibration and a good linearity was observed. The energy resolution of 35 keV was obtained for 1.9 MeV protons. The 6LiF conversion layer was applied on the detector Schottky contact. In the experiment we used different thicknesses of conversion layers from 5 μ m up to 35 μ m. Measured detected spectra show two parts corresponding to alpha particles detection in lower energy channels and 3H in higher energy channels. We have also performed simulations of thermal neutron detection using MCNPX (Monte Carlo N-particle eXtended) code. The detection efficiency and the detector response to thermal neutrons was calculated with respect to the 6LiF layer thickness. The detection efficiency calculation is found to be in good agreement with the experiment.

MCNPX simulations of the silicon carbide semiconductor detector response to fast neutrons from D–T nuclear reaction

Silicon Carbide (SiC) has been long recognized as a suitable semiconductor material for use in nuclear radiation detectors of high-energy charged particles, gamma rays, X-rays and neutrons. The nuclear interactions occurring in the semiconductor are complex and can be quantified using a Monte Carlo-based computer code. In this work, the MCNPX (Monte Carlo N-Particle eXtended) code was employed to support detector design and analysis. MCNPX is widely used to simulate interaction of radiation with matter and supports the transport of 34 particle types including heavy ions in broad energy ranges. The code also supports complex 3D geometries and both nuclear data tables and physics models. In our model, monoenergetic neutrons from D–T nuclear reaction were assumed as a source of fast neutrons. Their energy varied between 16 and 18.2 MeV, according to the accelerating voltage of the deuterons participating in D–T reaction. First, the simulations were used to calculate the optimum thickness of the reactive film composed of High Density PolyEthylene (HDPE), which converts neutral particles to charged particles and thusly enhancing detection efficiency. The dependency of the optimal thickness of the HDPE layer on the energy of the incident neutrons has been shown for the inspected energy range. Further, from the energy deposited by secondary charged particles and recoiled ions, the detector response was modeled and the effect of the conversion layer on detector response was demonstrated. The results from the simulations were compared with experimental data obtained for a detector covered by a 600 and 1300 [Formula: see text]m thick conversion layer. Some limitations of the simulations using MCNPX code are also discussed.

Medical staff extremity dosimetry in CT fluoroscopy: an anthropomorphic hand voxel phantom study

This work aims to contribute to the study of the radiation dose distribution delivered to the hands of medical staff members during a general computed tomographic (CT) fluoroscopic guided procedure. In this study, both Monte Carlo simulations and measurements were performed. For free-in-air and computed tomography dose index (CTDI) body phantom measurements, a standard pencil ionization chamber (IC) 100 mm long was used. The CT scanner model was implemented using MCNPX (Monte Carlo N-Particle eXtended) and was successfully validated by comparing the simulated results with measurements. Subsequently, CT images of a hand, together with an anthropomorphic phantom, were voxelized and used with the MCNPX code for dose calculations. The hand dose distribution study was performed both by using thermo-luminescent detector measurements and Monte Carlo simulations. The validated simulation tool provides a new perspective for detailed investigations of CT-irradiation scenarios. Simulations show that there is a strong dose gradient, namely the even zones of the hand that are in precise vicinity to the x-ray beam only receive about 4% of the maximum dose delivered to adjacent areas which are directly exposed to the primary x-ray beam. Finally, the scatter contribution of the patient was also studied through MC simulations. The results show that for directly exposed parts of the hand surface, the dose is reduced by the body of the patient (due to the shielding), whereas the dose is increased by scattered radiation from the patient for parts of the skin that receive scattered radiation only.

Alternatives for Helium-3 in Multiplicity Counters

Alternatives to helium-3 are being actively pursued due to the shortage and rising costs of helium-3. For safeguards applications, there are a number of ongoing investigations to find alternatives that provide the same capability in a cost-effective manner. One of the greatest challenges is to find a comparable alternative for multiplicity counters, since they require high efficiency and short collection or die-away times. Work has been progressing on investigating three commercially available alternatives for high efficiency multiplicity counters: boron trifluoride (BF3) filled proportional tubes, boron-lined proportional tubes, and lithium fluoride with zinc sulfide coated light guides. The baseline multiplicity counter used for the investigation is the Epithermal Neutron Multiplicity Counter with 121 helium-3 filled tubes at 10 atmosphere pressure, which is a significant capability to match. The primary tool for the investigation has been modeling and simulation using the Monte Carlo N-Particle eXtended (MCNPX) radiation transport program, with experiments to validate the models. To directly calculate the coincidence rates in boron-lined (and possibly other) detectors, the MCNPX code has been enhanced to allow the existing coincidence tally to be used with energy deposition rather than neutron capture reactions. This allows boron-lined detectors to be modeled more accurately. Variations of tube number and diameter along with variations in the amount of inter-tube moderator have been conducted for the BF3 and boron-lined cases. Tube pressure was investigated for BF3 , up to two atmospheres, as well as optimal boron thickness in the boron-lined tubes. The lithium fluoride was modeled as sheets of material with light guides in between, and the number and thickness of the sheets investigated. The amount of light guide, which in this case doubles as a moderator, was also optimized. The results of these modeling and simulation optimization investigations are described and results presented.

Extension of RPI-adult male and female computational phantoms to obese patients and a Monte Carlo study of the effect on CT imaging dose

Although it is known that obesity has a profound effect on x-ray computed tomography (CT) image quality and patient organ dose, quantitative data describing this relationship are not currently available. This study examines the effect of obesity on the calculated radiation dose to organs and tissues from CT using newly developed phantoms representing overweight and obese patients. These phantoms were derived from the previously developed RPI-adult male and female computational phantoms. The result was a set of ten phantoms (five males, five females) with body mass indexes ranging from 23.5 (normal body weight) to 46.4 kg m−2 (morbidly obese). The phantoms were modeled using triangular mesh geometry and include specified amounts of the subcutaneous adipose tissue and visceral adipose tissue. The mesh-based phantoms were then voxelized and defined in the Monte Carlo N-Particle Extended code to calculate organ doses from CT imaging. Chest–abdomen–pelvis scanning protocols for a GE LightSpeed 16 scanner operating at 120 and 140 kVp were considered. It was found that for the same scanner operating parameters, radiation doses to organs deep in the abdomen (e.g., colon) can be up to 59% smaller for obese individuals compared to those of normal body weight. This effect was found to be less significant for shallow organs. On the other hand, increasing the tube potential from 120 to 140 kVp for the same obese individual resulted in increased organ doses by as much as 56% for organs within the scan field (e.g., stomach) and 62% for those out of the scan field (e.g., thyroid), respectively. As higher tube currents are often used for larger patients to maintain image quality, it was of interest to quantify the associated effective dose. It was found from this study that when the mAs was doubled for the obese level-I, obese level-II and morbidly-obese phantoms, the effective dose relative to that of the normal weight phantom increased by 57%, 42% and 23%, respectively. This set of new obese phantoms can be used in the future to study the optimization of image quality and radiation dose for patients of different weight classifications. Our ultimate goal is to compile all the data derived from these phantoms into a comprehensive dosimetry database defined in the VirtualDose software.

Detection of fast neutrons using detectors based on semi-insulating GaAs

Detectors with AuZn square Schottky contact of the area of 2.5 × 2.5 mm2 were fabricated. On the back side, the whole area AuGeNi eutectic ohmic contact was evaporated. The thickness of the base material (semi-insulating GaAs) was 220 μm. The connection of 4 detectors in parallel was tested to get the detection area of 25 mm2. The 239Pu-Be fast neutron source with energies between 0.5 and 12 MeV was used in experimental measurements. We have investigated the optimal thickness of HDPE (high-density polyethylene) conversion layer for fast neutron detection. The spectra of the neutrons were measured by detectors covered by HDPE converter of different thicknesses. The fast neutron detection efficiency proved experimentally was compared with results from simulations performed by MCNPX (Monte Carlo N-Particle eXtended) code.

X-ray fluorescence Kα/Kβ ratios for a layered specimen: Comparison of measurements and Monte Carlo calculations with the MCNPX code

This paper shows how the Monte Carlo N-Particle eXtended (MCNPX) computer code can easily be applied to calculate the Kα/Kβ ratios of elements distributed heterogeneously in an analyzed specimen. Experiments were performed on specimens made of thick substrates covered by one or two mono-elemental layers (Ti, Ni, and Cu) of intermediate thickness. The Kα/Kβ ratios of the elements present in a substrate and in the layers were measured and then compared with the ratios obtained with the MCNPX computer code and with simple analytical calculations. Good correspondence was achieved between the experimental and calculated values. It was demonstrated that the MCNPX code makes it possible to predict the Kα/Kβ ratios for a specimen composed of layers and a substrate. This code can be applied for recognizing sample homogeneity or the depth heterogeneity of a specimen. In addition, the depth distribution of elements can be estimated and then the coating thicknesses in layered samples can also be determined.