Monte Carlo N-Particle eXtended Simulation Research Articles

Monte Carlo Methods to Simulate the Propagation of the Created Atomic/ Nuclear Particles from Underground Piezoelectric Rocks through the Fractures Before the Earthquakes

Until now, many studies have been performed on particle radiations before or during earthquakes (EQs). Neutron, gamma, electron, proton, and ultra-low frequency (ULF) photons are among the particles, detected during EQs. In our previous study, with the help of piezoelectricity relationships and the elastic energy formula, the Monte Carlo N‐Particle eXtended (MCNPX) simulation code was applied to find the amount of created atomic/nuclear particles, the dominant interactions; and the energy of the particles for various sizes of quartz and granite blocks. In this study, using the MCNPX simulation code, we have estimated the flux of the particles (created from under-stressed granitic rocks) at different distances from the EQ hypocenter inside the fractures, filled with air, water, and CO2. It was found that inside a water-filled fracture, the particles do not show the flux far from the EQ hypocenter. However, inside the gases like air and CO2 with the normal condition density, different types of particles can have a flux far from the source (more than a kilometer) and they might reach themselves to the surface in the case that the EQ hypocenter is very shallow (0­-5 km). However, for deep EQs, it seems that the most detected nuclear particles on the surface should pass via the vacuum-filled fractures and reach the surface. Moreover, it was concluded that the higher the density of the fracture’s filling fluid, the less distance that the particles can have a flux.

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.

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.

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.

Light output response of EJ-309 liquid organic scintillator to 2.86–3.95 MeV carbon recoil ions due to neutron elastic and inelastic scatter

We present the first measurements of energy-dependent light output from carbon recoils in the liquid organic scintillator EJ-309. For this measurement, neutrons were produced by an associated particle deuterium–tritium generator and scattered by a volume of EJ-309 scintillator into stop detectors positioned at four fixed angles. Carbon recoils in the scintillator were isolated using triple coincidence among the associated particle detector, scatter detector, and stop detectors. The kinematics of elastic and inelastic scatter allowed data collection at eight specific carbon recoil energies between 2.86 and 3.95 MeV. We found the light output caused by carbon recoils in this energy range to be approximately 1.14% of that caused by electrons of the same energy, which is comparable to the values reported for other liquid organic scintillators. A comparison of the number of scattered neutrons at each angle to a Monte Carlo N-Particle eXtended simulation indicates that the ENDF/B-VII.1 evaluation of differential cross sections for 14.1 MeV neutrons on carbon has discrepancies with the experiment as large as 55%, whereas those reported in the JENDL-4.0u evaluation agree with experiment.

The Voxel Mice Model of MCNPX for Simulation In Vivo Test BNCT

A study of voxel mice model of MCNPX has been done for in vivo test Boron Neutron Capture Therapy (BNCT). Mathematical and parameters were used to construct the stylized Mice model phantom. The geometry was modified into simulation software MCNPX (Monte Carlo N-Particle eXtended) simulation input. The result of mice stylized model phantom has been showed Figure 3.

Estimation of photoneutron yield in linear accelerator with different collimation systems by Geant4 and MCNPX simulation codes

At present, the bremsstrahlung photon beams produced by linear accelerators are the most commonly employed method of radiotherapy for tumor treatments. A photoneutron source based on three different energies (6, 10 and 15 MeV) of a linac electron beam was designed by means of Geant4 and Monte Carlo N-Particle eXtended (MCNPX) simulation codes. To obtain maximum neutron yield, two arrangements for the photo neutron convertor were studied: (a) without a collimator, and (b) placement of the convertor after the collimator. The maximum photon intensities in tungsten were 0.73, 1.24 and 2.07 photon/e at 6, 10 and 15 MeV, respectively. There was no considerable increase in the photon fluence spectra from 6 to 15 MeV at the optimum thickness between 0.8 mm and 2 mm of tungsten. The optimum dimensions of the collimator were determined to be a length of 140 mm with an aperture of 5 mm × 70 mm for iron in a slit shape. According to the neutron yield, the best thickness obtained for the studied materials was 30 mm. The number of neutrons generated in BeO achieved the maximum value at 6 MeV, unlike that in Be, where the highest number of neutrons was observed at 15 MeV. Statistical uncertainty in all simulations was less than 0.3% and 0.05% for MCNPX and the standard electromagnetic (EM) physics packages of Geant4, respectively. Differences among spectra in various regions are due to various cross-section and stopping power data and different simulations of the physics processes.

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.

HDR Brachytherapy Dose Distribution is Influenced by the Metal Material of the Applicator.

Applicators containing metal have been widely used in recent years when applying brachytherapy to patients with cervical cancer. However, the high dose rate (HDR) treatment-planning system (TPS) that is currently used in brachytherapy still assumes that the treatment environment constitutes a homogeneous water medium and does not include a dose correction for the metal material of the applicator. The primary purpose of this study was to evaluate the HDR 192Ir dose distribution in cervical cancer patients when performing brachytherapy using a metal-containing applicator. Thermoluminescent dosimeter (TLD) measurements and Monte Carlo N-Particle eXtended (MCNPX) code were used to explore the doses to the rectum and bladder when using a Henschke applicator containing metal during brachytherapy. When the applicator was assumed to be present, the absolute dose difference between the TLD measurement and MCNPX simulation values was within approximately 5%. A comparison of the MCNPX simulation and TPS calculation values revealed that the TPS overestimated the International Commission of Radiation Units and Measurement (ICRU) rectum and bladder reference doses by 57.78% and 49.59%, respectively. We therefore suggest that the TPS should be modified to account for the shielding effects of the applicator to ensure the accuracy of the delivered doses.

DEVELOPMENT OF SIMULATED WORKPLACE FIELDS AT KRISS FOR PERFORMANCE TEST OF NEUTRON PERSONAL DOSEMETERS.

Simulated workplace neutron fields have been developed at the Korea Research Institute of Standards and Science (KRISS). An (241)Am-Be neutron source and a cylindrical moderator composed of stainless steel and heavy water were installed in a 10-cm-thick concrete block with dimensions of 150 × 120 × 120 cm(3) The neutron energy spectrum at a distance of 66.5 cm was measured using a Bonner sphere spectrometer and was found to agree with the spectrum obtained from the Monte Carlo N-Particle Extended simulation to within 5 %. The neutron fluence-to-personal dose equivalent conversion coefficients were (20.8-43.6) pSv·cm(2) and were thus in good agreement with those of reactor fields. The results showed that the KRISS-simulated workplace neutron fields can be used for performance tests and the calibration service of neutron personal dosemeters.

APPLICATION OF THE S3M AND MCNPX CODES IN PARTICLE DETECTOR DEVELOPMENT

Semiconductor detectors can be used to detect neutrons if they are covered by a conversion layer. Some neutrons transfer their kinetic energy to hydrogen via elastic nuclear scattering in the conversion layer, and protons are produced as recoils. These protons enter the sensitive volume of the detector and are detected. In the process of detector development, Monte Carlo computer codes are necessary to simulate the detection process. This paper presents the main features of the S3M code (SRIM Supporting Software Modules) and shows its application potential. Examples are given for the neutron detectors with a conversion layer and for CVD (Chemical Vapor Deposition) diamond detectors for beam-condition monitors at the LHC (Large Hadron Collider). Special attention is paid to the S3M statistical modules that can be of interest also for other application areas like beam transport, accelerators, ion therapy, etc. The results are generated by MCNPX (Monte Carlo N-Particle eXtended) simulations used to optimize the thickness of the HDPE (high density polyethylene) conversion layer.

High-fidelity passive neutron multiplicity measurements and simulations of uranium oxide

When measuring low-self-multiplication uranium samples by passive neutron multiplicity counting, a technique not commonly used on uranium, there should be a decrease in the doubles counts with increasing enrichment. The reason for this decrease is explained as being due to the reduction in 238U mass, which has a large specific spontaneous fission rate in comparison to the other naturally occurring uranium isotopes. However, this is an over simplification of the actual nuclear physics which is occurring in bulk oxide items. The neutron multiplicity production rates are affected by many factors, such as the items self-multiplication, uranium isotopic composition which controls the spontaneous fission to (α,n) neutron production ratio, and the energy distribution of the emitted neutrons which includes induced fission neutrons. This paper presents a high-fidelity analysis of the passive neutron multiplicity distributions for varying enrichments of known uranium oxide standards. This analysis was performed through physical measurements of standards using the Los Alamos Epithermal Neutron Multiplicity Counter (ENMC) and detailed Monte Carlo N Particle Extended (MCNPX) neutron transport simulations supplemented with best available source term data. The MCNPX simulations reproduced the generic trends observed experimentally but absolute agreement is modest, especially for the triples counting. This suggests improvements in basic nuclear data are needed.

Monte Carlo Simulation of Performance of a Time-Resolved Range Telescope Using Selected Image Quality Assurance Phantoms

The superior dose conformation from protons is attributed to the Bragg peak near the end of the proton range. One challenge in proton cancer treatment is to assess the proton range fluctuations due to organ motion such as respiration. A time-resolved proton range telescope that measures coordinates, direction cosines, and the residual range of each proton can be useful in detecting and quantifying variations in radiological path length during the course of proton radiotherapy. In this paper, the Monte Carlo N-Particle eXtended (MCNPX) code was used to simulate the range telescope and study the image quality. To validate the MCNPX simulations, a simulated proton radiograph was compared with an experimentally acquired film for the same phantom. In addition, four quality assurance phantoms were simulated to investigate the quality of simulated proton radiography. Finally, the methods were applied to one phase of a patient four-dimensional computed tomography (4DCT) data set for proton radiography simulations. The results indicate that Monte Carlo simulations offer data that are useful in analyzing image spatial and temporal resolutions. Simulations show that it is useful to quantify the tumor position changes due to respiration by using a proton telescope.

Proton radiography and fluoroscopy of lung tumors: a Monte Carlo study using patient-specific 4DCT phantoms.

Monte Carlo methods are used to simulate and optimize a time-resolved proton range telescope (TRRT) in localization of intrafractional and interfractional motions of lung tumor and in quantification of proton range variations. The Monte Carlo N-Particle eXtended (MCNPX) code with a particle tracking feature was employed to evaluate the TRRT performance, especially in visualizing and quantifying proton range variations during respiration. Protons of 230 MeV were tracked one by one as they pass through position detectors, patient 4DCT phantom, and finally scintillator detectors that measured residual ranges. The energy response of the scintillator telescope was investigated. Mass density and elemental composition of tissues were defined for 4DCT data. Proton water equivalent length (WEL) was deduced by a reconstruction algorithm that incorporates linear proton track and lateral spatial discrimination to improve the image quality. 4DCT data for three patients were used to visualize and measure tumor motion and WEL variations. The tumor trajectories extracted from the WEL map were found to be within 1 mm agreement with direct 4DCT measurement. Quantitative WEL variation studies showed that the proton radiograph is a good representation of WEL changes from entrance to distal of the target. MCNPX simulation results showed that TRRT can accurately track the motion of the tumor and detect the WEL variations. Image quality was optimized by choosing proton energy, testing parameters of image reconstruction algorithm, and comparing to ground truth 4DCT. The future study will demonstrate the feasibility of using the time resolved proton radiography as an imaging tool for proton treatments of lung tumors.