Monte Carlo N-Particle eXtended Research Articles

Lead-free transparent shields for diagnostic X-rays: Monte Carlo simulation and measurements.

In recent years, the preference for using lead-free radiation protection shields has increased because of concerns regarding lead poisoning and leakage. In medical and research laboratories, glass shields are preferred because of their transparency. In this study, various glass shields were examined and compared based on the international standards. One commercially available lead-based shield, four recently studied shields, and three new lead-free shields were considered, and their shielding factors were calculated. We presented three glasses based on borate, phosphate, and silicate compounds, which were named Ir1, Ir2, and Ir3, respectively. Based on the International Electrotechnical Commission standard (IEC 61331), the air-kerma ratios (attenuation ratios) and lead equivalent values were derived using Monte Carlo N-Particle eXtended (MCNPX) calculations, and mass attenuation coefficients and effective atomic numbers (Zeff) of all the shields were obtained from XCOM database, in the diagnostic X-ray energy range of 40-120keV. In addition, some measurements were performed for the reference (lead-based) glass to validate the simulations. The above-mentioned factors for silicate-bismuth-based (Ir3) and borosilicate-barium-based (Tu) glasses were found to be higher than the others and comparable to those of commercially available lead-based glass. In conclusion, Ir3 and Tu glasses were found to be the preferred lead-free transparent shields in the diagnostic X-ray energy range.

The impact of gadolinium on the reactor production of 153Sm

Radioisotopes represent major sources of ionizing radiation, not least for use in medical applications, brachytherapy and nuclear medicine included. In this, the nuclear reactor is the main source of β- - γ emitting isotopes, an example product being 153Sm used in the treatment of pain arising from bone metastases. Present analysis relates to the potential of gadolinium neutron capture reaction, its impact on reactor production of radioisotopes and the proliferation resistant potential of thorium fuel cycle. A comparative analysis has been made of the impact of gadolinium on the production of 153Sm by UO2 and (Th, U) O2 fuels in a Westinghouse small modular reactor. Five fuel assemblies were investigated: one containing no gadolinium, the other four containing 16, 24, 34 or 44 gadolinium fuel rods. The code Monte Carlo N-Particle eXtended (MCNPX) integrated with the CINDER90 burn-up code was used for calculations. In the production of 153Sm the same trend is followed for the fuels containing gadolinium, increasing significantly with the number of gadolinium fuel rods. Zero production results from fuel assemblies without gadolinium. The concentration of 153Sm increases significantly with burn-up, indicating that gadolinium has a positive impact on the production of 153Sm.

Comparison of MCNPX and MIRDcell in assessing self-dose and cross-dose delivered to cell nuclei and the development of a realistic geometric model

Purpose: This study aims to provide a comparison between MCNPX and MIRDcell calculations for self-dose and cross-dose for three therapeutic isotopes used in internal radiotherapy (Lu-177, I-131 and Y-90) and to develop a multi-cellular geometric model to simulate an in vitro scenario.Materials and Methods: The self- and cross-dose to individual cell nuclei were assessed by Monte Carlo N-Particle eXtended (MCNPX). A close-packed cubic cell arrangement was assumed with the same amount of radioactivity per cell. Various cell sizes and subcellular distributions of radioactivity (nucleus, cytoplasm and cell membrane) were simulated. S values were obtained by MIRDcell for comparison. A Python 3.4 program was used to generate random cell coordinates in order to build a complex model that takes certain real conditions (cell size and cluster size) into account.Results: The relative differences of MCNPX versus MIRD S values (Sself) ranged from 2.88 to 10.10% for Lu-177; from 0 to 8.41% for I-131 and from 2.80 to 9.58% for Y-90. The relative differences of MCNPX versus MIRDcell cross-dose S values were 3.6%−15.7% for a sphere. The ratio of Scross max to Sself decreased for Lu-177 and I-131 with increasing cell size. The source localization within the cells had no significant impact on the cross-dosing. For single cells, the subcellular location of the source had an effect on Sself.Conclusions: MCNPX and MIRD cell-calculated S values showed good agreement. The model provided could be used to predict the biological effect caused by emitted radiation from therapeutic radionuclides at the cellular level.

A miniature thermal neutron source using high power lasers

The continuous improvement of high power laser technologies is recasting the prospects of small-scale neutron sources to enable scientific communities and industries performing experiments that are currently offered at extensive accelerator-driven facilities. This paper reports moderation of laser-driven fast neutrons to thermal energies using a compact, modular, moderator assembly. A significant thermal (∼25 meV) flux of ∼106 n/sr/pulse was measured from water and plastic moderators in a proof-of-principle experiment employing a relatively moderate power laser delivering 200 J on the target in 10 ps. Using Monte Carlo N-Particle eXtended simulations, the experimental results are reproduced and discussed.

Burnup computation for HTR-10 using MCNPX as the function of radius and fuel enrichment

The HTR-10 is a gas-cooled high-temperature reactor with spherical fuel and moderator called pebble bed and operated on 10 MW thermal power. Until now, a lot of research and development on the HTR-10 in terms of neutronic computational modeling has been carried out, one of which is burnup analysis. Fuel depletion or burnup analysis is an analysis related to fuel control and reactor output power. This analysis is needed because changes in fuel composition will affect the neutronic parameter values of a reactor. The 10 MW HTR fuel depletion study has been conducted by modeling the reactor using MCNPX (Monte Carlo N-Particle e-Xtended) to analyze the effect of radius and enrichment of reactor fuel to the left value and the burnup rate. The study is conducted by modeling a 10 MW HTR with three sorts of fuel kernel radiuses using MCNPX. First, 250 μm radius, second, 275 μm radius, third, 300 μm radius and all radiuses are variated with 10-15% fuel enrichment. TRISO particles are dispersed using simple cubic (SC) lattice in the fuel zone. The fuel balls are allocated in the reactor core along with moderator balls using a body-centered cubic. To perform all the calculations, MCNPX utilizes continuous energy data library ENDF/B-VI. From the calculation result using MCNPX can be concluded that the kernels with the radius of 275 μm and 300 μm at 14% and 13% fuel enrichment respectively were more able to maintain a critical condition within a year of operation than the kernels a 250 μm radius and the fuel with 250 μm radius has the highest burnup rate that is 52,07 GWd/MTU.

Comparative evaluation of nuclear radiation shielding properties of xTeO2 + (100–x)Li2O glass system

In the present investigation, nuclear radiation shielding parameters of xTeO2 + (100–x)Li2O (where x = 95, 90, 85, 80, 75, and 70 mol%) glass system have been examined. Gamma shielding parameters such as mass attenuation coefficients (MAC), half-value layer (HVL), tenth-value layer (TVL), mean free path (MFP), effective atomic number (Zeff), effective electron density (Neff) were calculated. Moreover, neutron effective cross sections (∑R) are determined. The calculations for present materials have been performed in different photon energy ranges (0.01–20 MeV) and using Monte Carlo N-Particle eXtended (MCNPX) simulation code and theoretical results were also obtained with WinXcom program. The results obtained from the MCNPX and WinXcom program were found to be in well harmony. Moreover, for the assessment of radiation shielding success of tellurite glasses, the mass stopping power (MSP) and projected range (PR) were computed for proton and alpha particles using stopping and range of ions in matter (SRIM) code. When the results obtained from the study are examined, it is seen that 95TeLi glass has the lowest HVL, TVL, MFP, TF and the highest (∑R) values. Therefore, the 95TeLi glass has the most perfect radiation shielding achievement than other investigated glasses.

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.

Characteristics in Water Phantom of Epithermal Neutron Beam Produced by Double Layer Beam Shaping Assembly

A Double Layer Beam Shaping Assembly (DLBSA) was designed to produce epithermal neutrons for BNCT purposes. The Monte Carlo N-Particle eXtended program was used as the software to design the DLBSA and phantom. Distribution of epithermal neutron and gamma flux in the DLBSA and phantom and absorbed dose in the phantom were computed using the Particle and Heavy Ion Transport code System program. Testing results of epithermal neutron beam irradiation of the water phantom showed that epithermal neutrons were thermalized and penetrated the phantom up to a depth of 12 cm. The maximum value of the absorbed dose was 2 × 10-3 Gy at a depth of 2 cm in the phantom.

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%.

Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute.

Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN’s Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X).

Monte Carlo simulation of X-ray spectra produced by Linac

This study was conducted to determine the intensity distribution of X-ray energy spectrum of that produced by Linear Accelerator (Linac) with tungsten target (W). The x-ray spectra on the space behind of target is calculated through the simulation method using Monte Carlo N-Particle eXtended version software (MCNPX). Linac is simulated with acceleration voltage 6 MV up to 15 MV using tungsten metal target (W). Based on the simulation of electron source accelerated, X-rays generated mostly (40-75) % are X-rays with energy of 1 MeV to 2 MeV, whereas the intensity of X-rays with energies fit the energies of the shooter’s electrons is very small (0.03 -0.340)%. It was indicated that the energy spectra dominated in the 1 MeV to 2 MeV than maximum energy in the x-ray spectrum range along the energy variation of Linac radiotherapy simulation. In addition, also found electrons about 2.24% to 5.67% compared to the intensity of X-rays, so as to increase the equivalent dose in the patient’s tissue later. This information can be used for the manufacture of low energy range filter design that dominate the spectrum and consideration of radiation protection efforts.

3D GaN-based betavoltaic device design with high energy transfer efficiency

A combined GaN 3D core-shell and planar pin structure is being developed and demonstrated to achieve the highest potential to increase energy transfer efficiency from the source (ηsrc) and power generated per cm2 (PGaN/cm2) in a betavoltaic (BV) device configuration. Physics-based Sentaurus TCAD and Monte Carlo N-Particle extended (MCNPX) software are employed to obtain the maximum ηsrc and PGaN/cm2 by a parametric study of device dimensions coupled with a 63NiCl2 source. Idealized structure dimensions are determined to be 2 µm wide, 4 µm tall GaN pin core-shell mesas, with 63Ni source conformally surrounding the structure with a 2 µm gap for maximum efficiency of energy transfer. For maximizing power deposited (10 µm mesa separation) a 3.75x increase in PGaN/cm2 at approximately half the activity density compared to a planar device is achieved for 4 µm mesa height, with 5.82x improvement in ηsrc.

Beam Shaping Assembly Optimization for Boron Neutron Capture Therapy Facility Based on Cyclotron 30 MeV as Neutron Source

A design of beam shaping assembly (BSA) installed on cyclotron 30 MeV model neutron source for boron neutron capture therapy (BNCT) has been optimized using simulator software of Monte Carlo N-Particle Extended (MCNPX). The Beryllium target with thickness of 0.55 cm is simulated to be bombarded with 30 MeV of proton beam. In this design, the parameter regarding beam characteristics for BNCT treatment has been improved, which is ratio of fast neutron dose and epithermal neutron flux. TiF3 is replaced to 30 cm of 27Al as moderator, and 1.5 cm of 32S is combined with 28 cm of 60Ni as neutron filter. Eventually, this design produces epithermal neutron flux of 2.33 × 109, ratio between fast neutron dose and epithermal neutron flux of 2.12 × 10-13,ratio between gamma dose and epithermal neutron flux of 1.00 × 10-13, ratio between thermal neutron flux and epithermal neutron flux is 0.047, and ration between particle current and total neutron flux is 0.56.

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.

OA167] Development of a realistic geometry model for cellular dosimetry of different therapeutic isotopes used in internal radiotherapy

Purpose This study is conducted to improve the multi-cellular model which would simulate an in-vitro scenario and to estimate the absorbed dose at the cellular and subcellular level for different therapeutic isotopes used in internal radiotherapy such as I-131, Y-90 and Lu-177. The objective of this study is to develop a complex model of a multicellular cell-cluster. Methods S-factors (absorbed dose per unit cumulated activity) calculations using Monte Carlo simulations were performed by using Monte Carlo N-Particle eXtended (MCNPX) code. S-cross (cross-dose) minimum, S-cross maximum and S-self (self-dose) were calculated. S-cross mean values were provided by using the Medical Internal Radiation Dose (MIRD) cell2 software. Nucleus (N), cytoplasm (Cy), cell surface (CS) and three isotopes, Y-90, I-131 and Lu-177 used for internal radiotherapy were simulated. Python software was used to generate random coordinates of cells in order to build a complex model taking into account realistic conditions (cell size, cluster size). The calculated coordinates of the cells are used as MCNPX input parameters. Results The coordinates of the cells are calculated and a multi-cellular model is developed. The comparison between MCNPX results and the MIRD cell2 values agreed well since the relative difference for S-self is less than 12% for the three radionuclides studied, Lu-177, I-131 and Y-90. The S-cross mean results provided by using MIRDcell2 software are between S-cross max and min data calculated for Lu-177, I-131 and Y-90 and for 3 configurations N to N; Cy to N and CS to N. The calculated ratios of S-cross vs S-self for three configurations (N to N, Cy to N, CS to N) are ranging from 10 to 50 for Lu-177, 20 to 70 for I-131 and from 40 to 140 for Y-90. It was showed that this ratio decreased for Lu-177 and I-131 with increasing the size of the cell and consequently the radius of the cell cluster. Conclusions The realistic geometry model provided by this study is designed to achieve as accurate results of estimating absorbed dose to the target cell and to predict the biological effect caused by the emitted radiations of therapeutic radionuclides.

Optimisation of a hybrid photoneutron source in a linear accelerator using GEANT4 and MCNPX Monte Carlo codes

Radiotherapy is important for treating oral cancer in advanced stages. Boron neutron capture therapy represents a unique modality in which neutron beams penetrate into the tissue within an epithermal range. Different methods are available for neutron production, and in this study, we used an electron accelerator. A photoneutron source based on three different energy values (10, 25, and 50 MeV) of a linear accelerator electron beam was designed using GEANT4 (GEometry ANd Tracking) and MCNPX (Monte Carlo N-Particle eXtended) simulation codes. The results indicate that a hybrid photoneutron source introduced tungsten and uranium with BeO. The statistical uncertainties in all simulations were less than 0.3 and 0.07% for MCNPX and the standard electromagnetic physics packages of GEANT4, respectively. Various cross-sectional and stopping-power data and different physics simulations produced different distributions.

Capabilities of the Monte Carlo Simulation Codes for Modeling of a Small Animal SPECT Camera.

This study aims to compare Monte Carlo-based codes' characteristics in the determination of the basic parameters of a high-resolution single photon emission computed tomography (HiReSPECT) scanner. The geometry of this dual-head gamma camera equipped with a pixelated CsI(Na) scintillator and lead hexagonal hole collimator were accurately described in the GEANT4 Application for the Tomographic Emission (GATE), Monte Carlo N-particle extended (MCNP-X), and simulation of imaging nuclear detectors (SIMIND) codes. We implemented simulation procedures similar to the experimental test for calculation of the energy spectra, spatial resolution, and sensitivity of HiReSPECT by using 99mTc sources. The energy resolutions simulated by SIMIND, MCNP-X, and GATE were 17.53, 19.24, and 18.26%, respectively, while it was calculated at 19.15% in experimental test. The average spatial resolutions of the HiReSPECT camera at 2.5cm from the collimator surface simulated by SIMIND, MCNP-X, and GATE were 3.18, 2.9, and 2.62mm, respectively, while this parameter was reported at 2.82mm in the experiment test. The sensitivities simulated by SIMIND, MCNP-X, and GATE were 1.44, 1.27, and 1.38cps/μCi, respectively, on the collimator surface. Comparison between simulation and experimental results showed that among these MC codes, GATE enabled to accurately model realistic SPECT system and electromagnetic physical processes, but it required more time and hardware facilities to run simulations. SIMIND was the most flexible and user-friendly code to simulate a SPECT camera, but it had limitations in defining the non-conventional imaging device. The most important characteristics like time and speed of simulation, preciseness of results, and user-friendliness should be considered during simulations.

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.

The Optimization of Collimator Material and In Vivo Testing Dosimetry of Boron Neutron Capture Therapy (BNCT) on Radial Piercing Beam Port Kartini Nuclear Reactor by Monte Carlo N-Particle Extended (MCNPX) Simulation Method

Boron Neutron Capture Therapy (BNCT) on radial piercing beam port Kartini nuclear reactor by MCNPX simulation method has been done in the National Nuclear Energy Agency Yogyakarta. BNCT is a type of therapy alternative that uses nuclear reaction 10B (n, α) 7Li to produce 2.79 MeV total kinetic energy. To be eligible IAEA conducted a study of design improvements and variations on some parameters to optimum condition which are Ni-nat thickness of 1.75 cm as collimator wall, Al2S3 as thick as 29 cm as moderator, Al2O3 0.5 cm thick as filter, Pb and Bi thickness of 4 cm as the end of the gamma shield collimators and Bi thickness of 1.5 cm as the base gamma shield collimators. The total dose was accepted in the tumor tissue 900 × 10-4 Gy/s. Radiation dose on the tumor tissue is 50±3 Gy with time irradiation of 9 minutes and 10 seconds. That dose was given into skin tissue and healthy liver tissue consecutively (6.00±0.05) × 10-2 Gy and (10.00±0.05) × 10-2 Gy. It shows the dose received by healthy tissue is still within safe limits.