Monte Carlo simulation of X-ray spectra and evaluation of filter effect using MCNP4C and FLUKA code

The general-purpose MCNP4C and FLUKA codes were used for simulating X-ray spectra. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic X-ray production were considered in this work. Tungsten/aluminum combination was used as target/filter in the simulation. The results of two codes were generated in 80, 100, 120 and 140 kV and compared with each other. In order to survey filter effect on X-ray spectra, the attenuation coefficient of filter was calculated in 120 kV. More details of filter effect have been investigated. The results of MCNP4C and FLUKA are comparable in the range of bremsstrahlung spectra, but there are some differences between them especially in specific X-ray peaks. Since the specific peaks have not significant role on image quality, both FLUKA and MCNP4C codes are fairly appropriate for production of X-ray spectra and evaluating image quality, absorbed dose and improvement in filter design.     

Borehole parametric study for neutron induced capture gamma-ray spectrometry using the MCNP code.

The MCNP Monte Carlo code has been used to simulate neutron transport from an Am-Be source into a granite formation surrounding a borehole. The effects of the moisture and the neutron poison on the thermal neutron flux distribution and the capture by the absorbing elements has been calculated. Thermal and nonthermal captures for certain absorbers having resonance structures in the epithermal and fast energy regions such as W and Si were performed. It is shown that for those absorbers having large resonances in the epithermal regions when they are present in dry formation or when accompanied by neutron poisons the resonance captures may be significant compared to the thermal captures.     

Computed radiography simulation using the Monte Carlo code MCNPX

Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology’s development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.     

Optimization of the target of an accelerator-driven neutron source through Monte Carlo numerical simulation of neutron and gamma transport by the PRIZMA code

At Budker Institute of Nuclear Physics, epithermal neutron source for neutron-capture therapy was built and neutron generation was realized. Source is based on tandem accelerator and uses near-threshold neutron generation from the reaction ⁷Li(p,n)⁷Be. The paper describes target optimization through the numerical simulation of proton, neutron and gamma transport by Monte Carlo method (PRIZMA code). It is shown that the near-threshold mode attractive due low activation provides high efficiency of the dose and acceptable therapeutic ratio and advantage depth.     

Image enhancement using MCNP5 code and MATLAB in neutron radiography

This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality.This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work.     

Monte Carlo studies of the Tunisian gamma irradiation facility using GEANT4 code.

This paper describes a complete Monte Carlo study of the Tunisian gamma irradiation facility (CNSTN) using the GEANT4 CERN's code. The work focused on the optimization of the absorbed dose distribution inside the irradiation cell, with and without product. For this purpose, 32 different points at the middle plane of the source rack, 29 positions along Z axis and 7 critical points, were carried out using PMMA dosimeters. Then, to achieve a given bulk density, boxes loaded with "dummy" product were used. Simulated and experimental results are compared and good agreement is observed. It is shown that Monte Carlo simulation improves process understanding, predicts absorbed dose distributions and calculates dose uniformity within product.     

Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry.

The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-dose-ratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer's data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators.     

Methodology for digital radiography simulation using the Monte Carlo code MCNPX for industrial applications.

This work presents a methodology for digital radiography simulation for industrial applications using the MCNPX radiography tally. In order to perform the simulation, the energy-dependent response of a BaFBr imaging plate detector was modeled and introduced in the MCNPX radiography tally input. In addition, a post-processing program was used to convert the MCNPX radiography tally output into 16-bit digital images. Simulated and experimental images of a steel pipe containing corrosion alveoli and stress corrosion cracking were compared, and the results showed good agreement between both images.     

Monte Carlo simulation of neutron irradiation facility developed for accelerator based in vivo neutron activation measurements in human hand bones.

The neutron irradiation facility developed at the McMaster University 3 MV Van de Graaff accelerator was employed to assess in vivo elemental content of aluminum and manganese in human hands. These measurements were carried out to monitor the long-term exposure of these potentially toxic trace elements through hand bone levels. The dose equivalent delivered to a patient during irradiation procedure is the limiting factor for IVNAA measurements. This article describes a method to estimate the average radiation dose equivalent delivered to the patient's hand during irradiation. The computational method described in this work augments the dose measurements carried out earlier [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724]. This method employs the Monte Carlo simulation of hand irradiation facility using MCNP4B. Based on the estimated dose equivalents received by the patient hand, the proposed irradiation procedure for the IVNAA measurement of manganese in human hands [Arnold et al., 2002. Med. Phys. 29(11), 2718-2724] with normal (1 ppm) and elevated manganese content can be carried out with a reasonably low dose of 31 mSv to the hand. Sixty-three percent of the total dose equivalent is delivered by non-useful fast group (> 10 keV); the filtration of this neutron group from the beam will further decrease the dose equivalent to the patient's hand.     

In vivo prompt gamma neutron activation analysis of cadmium in the kidney and liver.

An existing McMaster University in vivo prompt gamma neutron activation analysis system has been improved in order to reduce the cadmium detection limit in the kidney and liver. The detection limit for the kidney was found to be 1.7 mg, a greater than factor of 2 improvement over the previous results obtained at McMaster. The liver detection limit was determined to be 3.3 ppm. The corresponding skin dose for these measurements was only 0.5 mSv. The effect of kidney position on the detection limit also was examined. Figures of merit were calculated in order to compare the performance of the current system to others.     

Dose mapping using MCNP code and experiment for SVST-Co-60/B irradiator in Vietnam

By using MCNP code and ethanol–chlorobenzene (ECB) dosimeters the simulations and measurements of absorbed dose distribution in a tote-box of the Cobalt-60 irradiator, SVST-Co60/B at VINAGAMMA have been done. Based on the results Dose Uniformity Ratios (DUR), positions and values of minimum and maximum dose extremes in a tote-box, and efficiency of the irradiator for the different dummy densities have been gained. There is a good agreement between simulation and experimental results in comparison and they have valuable meanings for operation of the irradiator.     

Radiation shielding evaluation of the BNCT treatment room at THOR: a TORT-coupled MCNP Monte Carlo simulation study.

This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using "TORT-coupled MCNP" method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency.     

蒙特卡罗程序MCNP、EGSnrc、DPM剂量计算比较研究

目的 验证3个蒙特卡罗程序MCNP、EGSnrc、DPM计算结果的一致性问题。方法基于简单均匀及非均匀模型和复杂临床头部实例模型，对3个蒙特卡罗程序MCNP、EGSnrc、DPM的模型建模、计算结果、计算时间进行了比较研究。结果尽管3个程序在粒子输运原理、模拟参数设置、几何描述模型以及材料截面数据等方面存在不同，但是计算结果仍符合很好。结论3个蒙特卡罗程序计算复杂模型具有相当的可靠性；简单快速蒙特卡罗程序DPM具有明显的优势。     

Development of a prompt gamma neutron activation analysis facility for B concentration measurements at RA-3: Design stage

A PGNAA facility is being developed for ¹⁰B concentration measurements at RA-3 reactor. Its design targets detection limits better than tenths of a microgram and irradiation times on the order of minutes. Computational models were developed, which estimated thermal neutron fluxes in irradiation position to be larger than 10⁹ n cm⁻² s⁻¹. Calculated amounts of photons and fast neutrons make necessary for filter/moderator arrangements. An irradiation device was designed and numerically tested, which is being built and is to be used for performing characterizing measurements.     

蒙特卡罗方法在γ谱测量中效率刻度的研究

目的 建立探测器的无源效率刻度方法。方法 利用¹³⁷Cs点源,对HPGe和NaI晶体尺寸进行调整以获得正确的探测器几何参数,基于BOMAB体模的数学模型,在γ线能量126～1836 keV范围内,采用蒙特卡罗方法,结合核应急情况下体内核素探测几何模式,分别计算了HPGe探测器和NaI探测器对BOMAB体模的计数效率,根据计算结果拟合出相关的效率曲线和函数。结果 通过拟合函数计算获得的探测效率与应用蒙特卡罗算法得到的探测效率一致性较好。对于NaI探测器,残差为-19%～18%;对于HPGe探测器,残差为-11%～17%。结论 根据BOMAB体模的数学模型应用蒙特卡罗方法对便携式γ谱仪进行无源刻度省时、省力,具有较强的可操作性,是一种非常方便的谱仪校准方法。     

Grid and slot scan scatter reduction in mammography: comparison by using Monte Carlo techniques

PURPOSE: To evaluate a comprehensive array of scatter cleanup techniques in mammography by using a consistent methodology. MATERIALS AND METHODS: Monte Carlo techniques were used to evaluate the Bucky factor (BF) and the contrast improvement factor (CIF) for linear and cellular grids and for slot scan and scanning multiple-slot assembly (SMSA) systems. RESULTS: For a 28-kVp molybdenum anode-molybdenum filter spectrum with a standard detector and a 6-cm-thick 50% adipose-50% glandular breast phantom, slot scan techniques delivered an ideal BF. For slot widths greater than 4 mm, however, the CIF was lower than that achieved by the high-transmission cellular grid with a grid ratio of 3.8:1. A tungsten-septa air-interspaced cellular grid with a 4:1 grid ratio outperformed the high-transmission cellular grid in both BF and CIF. The SMSA was shown to be efficacious when 4-mm-wide slots were separated by at least 20 mm. In comparison with the literature, 3.6% agreement was achieved with other Monte Carlo studies, 3.3% with an experimental study that used a digital detector, and 13%-29% agreement was demonstrated in comparison to film-based experimental studies. CONCLUSION: With use of consistent methods for comparison, cellular grids were shown to substantially outperform linear grids but have slightly higher BFs compared with that of slot scan geometries at the same CIF.     

A comparison of HDR near source dosimetry using a treatment planning system,Monte Carlo simulation,and radiochromic film

This study aimed to investigate the high-dose rate Iridium-192 brachytherapy, including near source dosimetry, of a catheter-based applicator from 0.5 mm to 1 cm along the transverse axis. Radiochromic film and Monte Carlo (MC) simulation were used to generate absolute dose for the catheter-based applicator. Results from radiochromic film and MC simulation were compared directly to the treatment planning system (TPS) based on the American Association of Physicists in Medicine Updated Task Group 43 (TG-43U1) dose calculation formalism. The difference between dose measured using radiochromic film along the transverse plane at 0.5 mm from the surface and the predicted dose by the TPS was 24%±13%. The dose difference between the MC simulation along the transverse plane at 0.5 mm from the surface and the predicted dose by the TPS was 22.1%±3%. For distances from 1.5 mm to 1 cm from the surface, radiochromic film and MC simulation agreed with TPS within an uncertainty of 3%. The TPS under-predicts the dose at the surface of the applicator, i.e., 0.5 mm from the catheter surface, as compared to the measured and MC simulation predicted dose. MC simulation results demonstrated that 15% of this error is due to neglecting the beta particles and discrete electrons emanating from the sources and not considered by the TPS, and 7% of the difference was due to the photon alone, potentially due to the differences in MC dose modeling, photon spectrum, scoring techniques, and effect of the presence of the catheter and the air gap. Beyond 1 mm from the surface, the TPS dose algorithm agrees with the experimental and MC data within 3%.     

Characterization of scatter and penetration using Monte Carlo simulation in 131I imaging.

In 131I SPECT, image quality and quantification accuracy are degraded by object scatter as well as scatter and penetration in the collimator. The characterization of energy and spatial distributions of scatter and penetration performed in this study by Monte Carlo simulation will be useful for the development and evaluation of techniques that compensate for such events in 131I imaging. METHODS: First, to test the accuracy of the Monte Carlo model, simulated and measured data were compared for both a point source and a phantom. Next, simulations to investigate scatter and penetration were performed for four geometries: point source in air, point source in a water-filled cylinder, hot sphere in a cylinder filled with nonradioactive water, and hot sphere in a cylinder filled with radioactive water. Energy spectra were separated according to order of scatter, type of interaction, and gamma-ray emission energy. A preliminary evaluation of the triple-energy window (TEW) scatter correction method was performed. RESULTS: The accuracy of the Monte Carlo model was verified by the good agreement between measured and simulated energy spectra and radial point spread functions. For a point source in air, simulations show that 73% of events in the photopeak window had either scattered in or penetrated the collimator, indicating the significance of collimator interactions. For a point source in a water-filled phantom, the separated energy spectra showed that a 20% photopeak window can be used to eliminate events that scatter more than two times in the phantom. For the hot sphere phantoms, it was shown that in the photopeak region the spectrum shape of penetration events is very similar to that of primary (no scatter and no penetration) events. For the hot sphere regions of interest, the percentage difference between true scatter counts and the TEW estimate of scatter counts was <12%. CONCLUSION: In 131I SPECT, object scatter as well as collimator scatter and penetration are significant. The TEW method provides a reasonable correction for scatter, but the similarity between the 364-keV primary and penetration energy spectra makes it difficult to compensate for these penetration events using techniques that are based on spectral analysis.     

Monte Carlo calculation of Varian 2300C/D Linac photon beam characteristics: a comparison between MCNP4C, GEANT3 and measurements.

Different codes are used for Monte Carlo (MC) calculations in radiation therapy. In this research, MCNP4C and GEANT3 codes have been compared in calculations of dosimetric characteristics of Varian Clinac 2300C/D. The parameters of influence in the differences seen in dosimetric features were discussed. This study emphasizes that both MCNP4C and GEANT3 MC can be used in radiation therapy computations and their differences in photon spectra calculations have a negligible effect on percentage depth dose computations in radiation therapy.     

Monte Carlo simulation of a PhosWatch detector using Geant4 for xenon isotope beta–gamma coincidence spectrum profile and detection efficiency calculations

A simulation tool has been developed using the Geant4 Toolkit to simulate a PhosWatch single channel beta–gamma coincidence detection system consisting of a CsI(Tl)/BC404 Phoswich well detector and pulse shape analysis algorithms implemented digital signal processor. The tool can be used to simulate the detector's response for all the gamma rays and beta particles emitted from ¹³⁵Xe, ^133mXe, ¹³³Xe, ^131mXe and ²¹⁴Pb. Two- and three-dimensional beta–gamma coincidence spectra from the PhosWatch detector can be produced using the simulation tool. The accurately simulated spectra could be used to calculate system coincidence detection efficiency for each xenon isotope, the corrections for the interference from the various spectral components from radon and xenon isotopes, and system gain calibration. Also, it can generate two- and three-dimensional xenon reference spectra to test beta–gamma coincidence spectral deconvolution analysis software.