Monte Carlo N-Particle eXtended Research Articles

Assessment of PCXMC for patients with different body size in chest and abdominal x ray examinations: a Monte Carlo simulation study

A PC Program for x ray Monte Carlo (PCXMC) has been used to calculate organ doses in patient dosimetry and for the exposure assessment in epidemiological studies of radiogenic health related risks. This study compared the dosimetry from using the built-in stylized phantoms in the PCXMC to that of a newer hybrid phantom library with improved anatomical realism. We simulated chest and abdominal x ray projections for 146 unique body size computational phantoms, 77 males and 69 females, with different combinations of height (125–180 cm) and weight (20–140 kg) using the built-in stylized phantoms in the PCXMC version 2.0.1.4 and the hybrid phantom library using the Monte Carlo N-particle eXtended transport code 2.7 (MCNPX). Unfortunately, it was not possible to incorporate the hybrid phantom library into the PCXMC. We compared 14 organ doses, including dose to the active bone marrow, to evaluate differences between the built-in stylized phantoms in the PCXMC and the hybrid phantoms (Cristy and Eckerman 1987 Technical Report ORNL/TM-8381/V1, Oak Ridge National Laboratory, Eckerman and Ryman 1993 Technical Report 12 Oak Ridge, TN, Geyer et al 2014 Phys. Med. Biol. 59 5225–42). On average, organ doses calculated using the built-in stylized phantoms in the PCXMC were greater when compared to the hybrid phantoms. This is most prominent in AP abdominal exams by an average factor of 2.4-, 2.8-, and 2.8-fold for the 10-year-old, 15-year-old, and adult phantoms, respectively. For chest exams, organ doses are greater by an average factor of 1.1-, 1.4-, and 1.2-fold for the 10-year-old, 15-year-old, and adult phantoms, respectively. The PCXMX, due to its ease of use, is often selected to support dosimetry in epidemiological studies; however, it uses simplified models of the human anatomy that fail to account for variations in body morphometry for increasing weight. For epidemiological studies that use PCXMC dosimetry, associations between radiation-related disease risks and organ doses may be underestimated, and to a greater degree in pediatric, especially obese pediatric, compared to adult patients.

Monte Carlo simulations used to calculate the energy deposited in the coronary artery lumen as a function of iodine concentration and photon energy

Purpose: To better understand the risks of cumulative medical X-ray investigations and the possible causal role of contrast agent on the coronary artery wall, the correlation between iodinated contrast media and the increase of energy deposited in the coronary artery lumen as a function of iodine concentration and photon energy is investigated.Materials and methods: The calculations of energy deposition have been performed using Monte Carlo (MC) simulation codes, namely PENetration and Energy LOss of Positrons and Electrons (PENELOPE) and Monte Carlo N-Particle eXtended (MCNPX). Exposure of a cylinder phantom, artery and a metal stent (AISI 316L) to several X-ray photon beams were simulated.Results and discussion: For the energies used in cardiac imaging the energy deposited in the coronary artery lumen increases with the quantity of iodine. Monte Carlo calculations indicate a strong dependence of the energy enhancement factor (EEF) on photon energy and iodine concentration. The maximum value of EEF is equal to 25; this factor is showed for 83 keV and for 400 mg Iodine/mL. No significant impact of the stent is observed on the absorbed dose in the artery for incident X-ray beams with mean energies of 44, 48, 52 and 55 keV.Conclusion: A strong correlation was shown between the increase in the concentration of iodine and the energy deposited in the coronary artery lumen for the energies used in cardiac imaging and over the energy range between 44 and 55 keV. The data provided by this study could be useful for creating new medical imaging protocols to obtain better diagnostic information with a lower level of radiation exposure.

Patient-Specific Computational Model and Dosimetry Calculations for PET/CT of a Patient Pregnant with Twins.

The radiation dose delivered to pregnant patients during radiologic imaging procedures raises health concerns because the developing embryo and fetus are considered to be highly radiosensitive. To appropriately weigh the diagnostic benefits against the radiation risks, the radiologist needs reasonably accurate and detailed estimates of the fetal dose. Expanding our previously developed series of computational phantoms for pregnant women, we here describe a personalized model for twin pregnancy, based on an actual clinical scan. Methods: The model is based on a standardized hybrid pregnant female and fetus phantom and on a clinical case of a patient who underwent an 18F-FDG PET/CT scan while expecting twins at 25 weeks' gestation. This model enabled us to produce a realistic physical representation of the pregnant patient and to estimate the maternal and fetal organ doses from the 18F-FDG and CT components. The Monte Carlo N-Particle Extended general-purpose code was used for radiation transport simulation. Results: The 18F-FDG doses for the 2 fetuses were 3.78 and 3.99 mGy, and the CT doses were 0.76 and 0.70 mGy, respectively. Therefore, the relative contribution of 18F-FDG and CT to the total dose to the fetuses was about 84% and 16%, respectively. Meanwhile, for 18F-FDG, the calculated personalized absorbed dose was about 40%-50% higher than the doses reported by other dosimetry computer software tools. Conclusion: Our approach to constructing personalized computational models allows estimation of a patient-specific radiation dose, even in cases with unusual anatomic features such as a twin pregnancy. Our results also show that, even in twins, the fetal organ doses from both 18F-FDG and CT present a certain variability linked to the anatomic characteristics. The CT fetal dose is smaller than the 18F-FDG PET dose.

Reconstruction Algorithm of Calibration Map for RPT Techniques in Quadrilateral Bubble Column Reactor Using MCNPX Code

Radioactive Particle Tracking (RPT) is non-invasive evaluation technique capable of visualising and tracking the motion of the identified phase in bubble column reactor. The single radioactive particle emits γ-ray, and its movement in the column is tracked with the aid of arrays radiation detectors. In this study, Monte Carlo approach was programmed to reconstruct the particle tracer position so-called calibration map. The iterative reconstruction algorithms is used to generate and calculate the 2600 coordinates of calibration map from the number of photon counts from the ten NaI scintillation detectors. To validate the simulation precision, a spiral trajectory of radioactive particle inside the bubble column region consist of 84 grid point locations of the particle were applied. Calibration algorithm was developed for radioactive isotopes Au-198 and Sc-46 particle position verification and determination of statistical uncertainty from the introduction of a various number of primary photon emission. The result of the studies proved that higher number of particles used in the algorithm for position reconstruction gives more accuracy and we found that the Sc-46 obtain more accurate calculation and shows low statistical level than Au-198 particle. The outcome of this simulation based on random sampling from Monte Carlo N-Particle Extended (MCNPX) demonstrated that the calibration map could successfully be implemented in RPT technique to observe the dynamic movement of radioactive particle which represents the tracked media in the quadrilateral bubble column.

Reconstruction Algorithm of Calibration Map for RPT Techniques in Quadrilateral Bubble Column Reactor Using MCNPX Code

Radioactive Particle Tracking (RPT) is non-invasive evaluation technique capable of visualising and tracking the motion of the identified phase in bubble column reactor. The single radioactive particle emits ?-ray, and its movement in the column is tracked with the aid of arrays radiation detectors. In this study, Monte Carlo approach was programmed to reconstruct the particle tracer position so-called calibration map. The iterative reconstruction algorithms is used to generate and calculate the 2600 coordinates of calibration map from the number of photon counts from the ten NaI scintillation detectors. To validate the simulation precision, a spiral trajectory of radioactive particle inside the bubble column region consist of 84 grid point locations of the particle were applied. Calibration algorithm was developed for radioactive isotopes Au-198 and Sc-46 particle position verification and determination of statistical uncertainty from the introduction of a various number of primary photon emission. The result of the studies proved that higher number of particles used in the algorithm for position reconstruction gives more accuracy and we found that the Sc-46 obtain more accurate calculation and shows low statistical level than Au-198 particle. The outcome of this simulation based on random sampling from Monte Carlo N-Particle Extended (MCNPX) demonstrated that the calibration map could successfully be implemented in RPT technique to observe the dynamic movement of radioactive particle which represents the tracked media in the quadrilateral bubble column.

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.

Mass Attenuation Coefficients of Human Body Organs using MCNPX Monte Carlo Code

Introduction: Investigation of radiation interaction with living organs has always been a thrust area in medical and radiation physics. The investigated results are being used in medical physics for developing improved and sensitive techniques and minimizing radiation exposure. In this study, mass attenuation coefficients of different human organs and biological materials such as adipose, blood, bone, brain, eye lens, lung, muscle, skin, and tissue have been calculated. Materials and Methods: In the present study, Monte Carlo N-Particle eXtended (MCNP-X) version 2.4.0 was used for determining mass attenuation coefficients, and the obtained results were compared with earlier investigations (using GEometry ANd Tracking [GEANT4] and FLUKA computer simulation packages) for blood, bone, lung, eye lens, adipose, tissue, muscle, brain, and skin materials at different energies. Results: The results of this study showed that the obtained results from MCNP-X were in high accordance with the National Institute of Standards and Technology data. Conclusion: Our findings would be beneficial for use of present simulation technique and mass attenuation coefficients for medical and radiation physics applications.

Effects of body habitus on internal radiation dose calculations using the 5-year-old anthropomorphic male models

Computational phantoms are commonly used in internal radiation dosimetry to assess the amount and distribution pattern of energy deposited in various parts of the human body from different internal radiation sources. Radiation dose assessments are commonly performed on predetermined reference computational phantoms while the argument for individualized patient-specific radiation dosimetry exists. This study aims to evaluate the influence of body habitus on internal dosimetry and to quantify the uncertainties in dose estimation correlated with the use of fixed reference models. The 5-year-old IT’IS male phantom was modified to match target anthropometric parameters, including body weight, body height and sitting height/stature ratio (SSR), determined from reference databases, thus enabling the creation of 125 5-year-old habitus-dependent male phantoms with 10th, 25th, 50th, 75th and 90th percentile body morphometries. We evaluated the absorbed fractions and the mean absorbed dose to the target region per unit cumulative activity in the source region (S-values) of F-18 in 46 source regions for the generated 125 anthropomorphic 5-year-old hybrid male phantoms using the Monte Carlo N-Particle eXtended general purpose Monte Carlo transport code and calculated the absorbed dose and effective dose of five 18F-labelled radiotracers for children of various habitus. For most organs, the S-value of F-18 presents stronger statistical correlations with body weight, standing height and sitting height than BMI and SSR. The self-absorbed fraction and self-absorbed S-values of F-18 and the absorbed dose and effective dose of 18F-labelled radiotracers present with the strongest statistical correlations with body weight. For 18F-Amino acids, 18F-Brain receptor substances, 18F-FDG, 18F-L-DOPA and 18F-FBPA, the mean absolute effective dose differences between phantoms of different habitus and fixed reference models are 11.4%, 11.3%, 10.8%, 13.3% and 11.4%, respectively. Total body weight, standing height and sitting height have considerable effects on human internal dosimetry. Radiation dose calculations for individual subjects using the most closely matched habitus-dependent computational phantom should be considered as an alternative to improve the accuracy of the estimates.

In vitro Dosimetric Study of Biliary Stent Loaded with Radioactive 125I Seeds.

Background:A novel radioactive 125I seed-loaded biliary stent has been used for patients with malignant biliary obstruction. However, the dosimetric characteristics of the stents remain unclear. Therefore, we aimed to describe the dosimetry of the stents of different lengths — with different number as well as activities of 125I seeds.Methods:The radiation dosimetry of three representative radioactive stent models was evaluated using a treatment planning system (TPS), thermoluminescent dosimeter (TLD) measurements, and Monte Carlo (MC) simulations. In the process of TPS calculation and TLD measurement, two different water-equivalent phantoms were designed to obtain cumulative radial dose distribution. Calibration procedures using TLD in the designed phantom were also conducted. MC simulations were performed using the Monte Carlo N-Particle eXtended version 2.5 general purpose code to calculate the radioactive stent's three-dimensional dose rate distribution in liquid water. Analysis of covariance was used to examine the factors influencing radial dose distribution of the radioactive stent.Results:The maximum reduction in cumulative radial dose was 26% when the seed activity changed from 0.5 mCi to 0.4 mCi for the same length of radioactive stents. The TLD's dose response in the range of 0–10 mGy irradiation by 137Cs γ-ray was linear: y = 182225x − 6651.9 (R2= 0.99152; y is the irradiation dose in mGy, x is the TLDs’ reading in nC). When TLDs were irradiated by different energy radiation sources to a dose of 1 mGy, reading of TLDs was different. Doses at a distance of 0.1 cm from the three stents’ surface simulated by MC were 79, 93, and 97 Gy.Conclusions:TPS calculation, TLD measurement, and MC simulation were performed and were found to be in good agreement. Although the whole experiment was conducted in water-equivalent phantom, data in our evaluation may provide a theoretical basis for dosimetry for the clinical application.

Conceptual Design of Collimator at Boron Neutron Capture Therapy Facility with 30 MeV Cyclotron and Target 9Be as Neutron Generator Using Monte Carlo N-Particle Extended Simulator

The optimization of collimator has been studied which resulted epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) using Monte Carlo N Particle Extended (MCNPX). Cyclotron 30 MeV and <sup>9</sup>Be target is used as a neutron generator. The design criteria were based on recommendation from IAEA. Mcnpx calculations indicated by using 25 cm and 40 cm thickness of PbF<sub>2</sub> as reflector and back reflector, 15 cm thickness of TiF<sub>3</sub> as first moderator, 35 cm thickness of AlF<sub>3</sub> as second moderator, 25 cm thickness of <sup>60</sup>Ni as neutron filter, 2 cm thickness of Bi as gamma filter, and aperture with 20 cm of diameter size, an epithermal neutron beam with an intensity 1.21 × 10<sup>9</sup> n.cm<sup>-2</sup>.s<sup>-1</sup>, fast neutron and gamma doses per epithermal neutron of 7.04 × 10<sup>-13</sup> Gy.cm<sup>2</sup>.n<sup>-1</sup> and 1.61 × 10<sup>-13</sup> Gy.cm<sup>2</sup>.n<sup>-1</sup>, minimum thermal neutron per epithermal neutron ratio of 0.043, and maximum directionality of 0.58, respectively could be produced. The results have not passed all the IAEA’s criteria in fast neutron component and directionality.

Electrical Properties of MWCNT/HDPE Composite-Based MSM Structure Under Neutron Irradiation

Because of their low cost, low energy consumption, high performance, and exceptional electrical properties, nanocomposites containing carbon nanotubes are suitable for use in many applications such as sensing systems. In this research work, a metal–semiconductor–metal (MSM) structure based on a multiwall carbon nanotube/high-density polyethylene (MWCNT/HDPE) nanocomposite is introduced as a neutron sensor. Scanning electron microscopy, Fourier-transform infrared, and infrared spectroscopy techniques were used to characterize the morphology and structure of the fabricated device. Current–voltage (I–V) characteristic modeling showed that the device can be assumed to be a reversed-biased Schottky diode, if the voltage is high enough. To estimate the depletion layer length of the Schottky contact, impedance spectroscopy was employed. Therefore, the real and imaginary parts of the impedance of the MSM system were used to obtain electrical parameters such as the carrier mobility and dielectric constant. Experimental observations of the MSM structure under irradiation from an americium–beryllium (Am–Be) neutron source showed that the current level in the device decreased significantly. Subsequently, current pulses appeared in situ I–V and current–time (I–t) curve measurements when increasing voltage was applied to the MSM system. The experimentally determined depletion region length as well as the space-charge-limited current mechanism for carrier transport were compared with the range for protons calculated using Monte Carlo n-particle extended (MCNPX) code, yielding the maximum energy of recoiled protons detectable by the device.

Theoretical estimation of 64Cu production with neutrons emitted during 18F production with a 30 MeV medical cyclotron

PurposeThis work presents the theoretical estimation of a combined production of 18F and 64Cu isotopes for PET applications. 64Cu production is induced in a secondary target by neutrons emitted during a routine 18F production with a 30MeV cyclotron: protons are used to produce 18F by means of the 18O(p,n)18F reaction on a [18O]-H2O target (primary target) and the emitted neutrons are used to produce 64Cu by means of the 64Zn(n,p)64Cu reaction on enriched zinc target (secondary target). MethodsMonte Carlo simulations were carried out using Monte Carlo N Particle eXtended (MCNPX) code to evaluate flux and energy spectra of neutrons produced in the primary (Be+[18O]-H2O) target by protons and the attenuation of neutron flux in the secondary target. 64Cu yield was estimated using an analytical approach based on both TENDL-2015 data library and experimental data selected from EXFOR database. ResultsTheoretical evaluations indicate that about 3.8 MBq/μA of 64Cu can be obtained as a secondary, ‘side’ production with a 30MeV cyclotron, for 2h of irradiation of a proper designed zinc target. Irradiating for 2h with a proton current of 120 μA, a yield of about 457 MBq is expected. Moreover, the most relevant contaminants result to be 63,65Zn, which can be chemically separated from 64Cu contrarily to what happens with proton irradiation of an enriched 64Ni target, which provides 64Cu mixed to other copper isotopes as contaminants. ConclusionsThe theoretical study discussed in this paper evaluates the potential of the combined production of 18F and 64Cu for medical purposes, irradiating a properly designed target with 30MeV protons. Interesting yields of 64Cu are obtainable and the estimation of contaminants in the irradiated zinc target is discussed.

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.

Evaluation of equivalent and effective dose by KAP for patient and orthopedic surgeon in vertebral compression fracture surgery

Clinical scenarios were virtually modeled to estimate both the equivalent and effective doses normalized by KAP (Kerma Area Product) to vertebra compression fracture surgery in patient and surgeon. This surgery is known as kyphoplasty and involves the use of X-ray equipment, the C-arm, which provides real-time images to assist the surgeon in conducting instruments inserted into the patient and in the delivery of surgical cement into the fractured vertebra. The radiation transport code used was MCNPX (Monte Carlo N-Particle eXtended) and a pair of UFHADM (University of Florida Hybrid ADult Male) virtual phantoms. The developed scenarios allowed us to calculate a set of equivalent dose (HT) and effective dose (E) for patients and surgeons. In additional, the same scenario was calculated KAP in the tube output and was used for calculating conversion coefficients (E/KAP and HT/KAP). From the knowledge of the experimental values of KAP and the results presented in this study, it is possible to estimate absolute values of effective doses for different exposure conditions. In this work, we developed scenarios with and without the surgical table with the purpose of comparison with the existing data in the literature. The absence of the bed in the scenario promoted a percentage absolute difference of 56% in the patient effective doses in relation to scenarios calculated with a bed. Regarding the surgeon, the use of the personal protective equipment (PPE) reduces between 75% and 79% the effective dose and the use of the under table shield (UTS) reduces the effective dose of between 3% and 7%. All these variations emphasize the importance of the elaboration of virtual scenarios that approach the actual clinical conditions generating E/KAP and HT/KAP closer to the actual values.

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.

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.

A novel conformal superficial high-dose-rate brachytherapy device for the treatment of nonmelanoma skin cancer and keloids

PurposeTo develop a novel conformal superficial brachytherapy (CSBT) device as a treatment option for the patient-specific radiation therapy of conditions including superficial lesions, postsurgical positive margins, Dupuytren's contractures, keloid scars, and complex anatomic sites (eyelids, nose, ears, etc.). Methods and MaterialsA preliminary CSBT device prototype was designed, built, and tested using readily available radioactive seeds. Iodine-125 (125I) seeds were independently guided to the treatment surface to conform to the target. Treatment planning was performed via BrachyVision Planning System (BPS) and dose distributions measured with Gafchromic EBT3 film. Percent depth dose curves and profiles for Praseodymium-142 (142Pr), and Strontium-90/Yttrium-90 (90Sr–90Y) were also investigated as potential sources. Results achieved with 90Sr–90Y and electron external beam radiation therapy were compared and Monte Carlo N-Particle eXtended 2.6 simulations of 142Pr seeds were validated. ResultsBPS was able to predict clinical dose distributions for a multiple seeds matrix. Calculated and measured doses for the 125I seed matrix were 500 cGy and 473.5 cGy at 5 mm depth, and 171.0 cGy and 201.0 cGy at 10 mm depth, respectively. Results of 90Sr–90Y tests demonstrate a more conformal dose than electron EBRT (1.6 mm compared to 4.3 mm penumbra). Measured 142Pr doses were 500 cGy at surface and 17.4 cGy at 5 mm depth. ConclusionsThe CSBT device provides a highly conformal dose to small surface areas. Commercially available BPS can be used for treatment planning, and Monte Carlo simulation can be used for plans using beta-emitting sources and complex anatomies. Various radionuclides may be used in this device to suit prescription depths and treatment areas.

SINGLE ANODE TRIPLE GEM TISSUE EQUIVALENT PROPORTIONAL COUNTER AS THE BASIS FOR A PERSONAL NEUTRON DOSIMETER.

This paper reports on a tissue-equivalent proportional counter (TEPC) based on a triple gas electron multiplier structure, with a single pad readout, as a basis for a personal neutron dosimeter. Its dosimetric response was studied using the 252Cf neutron source at the Health Physics Generator Facility of the Canadian Nuclear Laboratories. Measured lineal energy spectra were found to be in agreement with numerical simulations performed with Monte Carlo N-Particle eXtended (MCNPX). Both simulations and measurements showed that the mean pathlength of secondary charged particles in the TEPC gas was best represented by the thickness of the drift region of the device. It was determined that the Cauchy Theorem, used to calculate the mean chord length in spherical and cylindrical TEPCs, overestimated the simulated mean chord length by nearly a factor of two. Important operational characteristics of the device were investigated, including gas gain, sensitivity and dosimetric response, as functions of tissue-equivalent gas pressure. This work demonstrates that the proposed design can serve as the basis for a personal neutron dosimeter device, which would satisfy the angular dosimetric response criteria of the personal dosimeter standard IEC61526.

Assessment of Mean Glandular Dose in Mammography System with Different Anode-Filter Combinations Using MCNP Code.

BackgroundX-ray mammography is one of the general methods for early detection of breast cancer. Since glandular tissue in the breast is sensitive to radiation and it increases the risk of cancer, the given dose to the patient is very important in mammography.ObjectivesThe aim of this study was to determine the average absorbed dose of X-ray radiation in the glandular tissue of the breast during mammography examinations as well as investigating factors that influence the mean glandular dose (MGD). One of the precise methods for determination of MGD absorbed by the breast is Monte Carlo simulation method which is widely used to assess the dose.Materials and MethodsWe studied some different X-ray sources and exposure factors that affect the MGD. “Midi-future” digital mammography system with amorphous-selenium detector was simulated using the Monte Carlo N-particle extended (MCNPX) code. Different anode/filter combinations such as tungsten/silver (W/Ag), tungsten/rhodium (W/Rh), and rhodium/aluminium (Rh/Al) were simulated in this study. The voltage of X-ray tube ranged from 24 kV to 32 kV with 2 kV intervals and the breast phantom thickness ranged from 3 to 8 cm, and glandular fraction g varied from 10% to 100%.ResultsMGD was measured for different anode/filter combinations and the effects of changing tube voltage, phantom thickness, combination and glandular breast tissue on MGD were studied. As glandular g and X-ray tube voltage increased, the breast dose increased too, and the increase of breast phantom thickness led to the decrease of MGD. The obtained results for MGD were consistent with the result of Boone et al. that was previously reported.ConclusionBy comparing the results, we saw that W/Rh anode/filter combination is the best choice in breast mammography imaging because of the lowest delivered dose in comparison with W/Ag and Rh/Al. Moreover, breast thickness and g value have significant effects on MGD.

Scan MDCs for GPS-Based Gamma Radiation Surveys.

A method for estimating the minimum detectable concentration of a contaminant radionuclide in soil when scanning with gamma radiation detectors (known as the "scan MDC") is described in the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM). This paper presents an alternate method for estimating scan MDCs for GPS-based gamma surveys based on detector efficiencies modeled with the probabilistic Monte Carlo N-Particle Extended (MCNPX) Transport simulation code. Results are compared to those provided in MARSSIM. An extensive database of MCNPX-based detection efficiencies has been developed to represent a variety of gamma survey applications and potential scanning configurations (detector size, scan height, size of contaminated soil volume, etc.), and an associated web-based user interface has been developed to provide survey designers and regulators with access to a reasonably wide range of calculated scan MDC values for survey planning purposes.