Voxel model of a rabbit: assessment of absorbed doses in organs after CT examination performed by two different protocols.

The objective of this work was to assess absorbed doses in organs and tissues of a rabbit, following computed tomography (CT) examinations, using a dedicated 3D voxel model. Absorbed doses in relevant organs were calculated using the MCNP5 Monte Carlo software. Calculations were perfomed for two standard CT protocols, using tube voltages of 110 kVp and 130 kVp. Absorbed doses were calculated in 11 organs and tissues, i.e., skin, bones, brain, muscles, heart, lungs, liver, spleen, kidney, testicles, and fat tissue. The doses ranged from 15.3 to 28.3 mGy, and from 40.2 to 74.3 mGy, in the two investigated protocols. The organs that received the highest dose were bones and kidneys. In contrast, brain and spleen were organs that received the smallest doses. Doses in organs which are stretched along the body did not change significantly with distance. On the other hand, doses in organs which are localized in the body showed maximums and minimums. Using the voxel model, it is possible to calculate the dose distribution in the rabbit's body after CT scans, and study the potential biological effects of CT doses in certain organs. The voxel model presented in this work can be used to calculated doses in all radiation experiments in which rabbits are used as experimental animals.

Monte Carlo investigation of electron specific energy distribution in a single cell model.

Knowledge of microdosimetric quantities of certain radionuclides is important in radio immune cancer therapies. Specific energy distribution of radionuclides, which are bound to the cell, is the microdosimetric quantity essential in the process of radionuclide selection for patient tumour treatment. The aim of this paper is to establish an applicable method to determine microdosimetric quantities for various radionuclides. The established method is based on knowledge of microdosimetric quantities of monoenergetic electrons. In this paper these quantities are determined for the single-cell model for a range of electron energies up to [Formula: see text], using the Monte Carlo transport code PENELOPE. The results show that using monoenergetic specific energies, reconstruction of the specific energy of beta-emitting radionuclides can be successfully done with very high accuracy. Microdosimetric quantities share information about the physical processes involved and give insight about energy depositions, which is of use in the procedure of radionuclide selection for a given type of therapy.

Simulation of H p (10) and effective dose received by the medical staff in interventional radiology procedures.

Interventional radiology and cardiology are widespread employed techniques for diagnosis and treatment of several pathologies because they avoid the majority of the side-effects associated with surgical treatments, but are known to increase the radiation exposure to patient and operators. In recent years many studies treated the exposure of the operators performing cardiological procedures. The aim of this work is to study the exposure condition of the medical staff in some selected interventional radiology procedures. The Monte Carlo simulations have been employed with anthropomorphic mathematical phantoms reproducing the irradiation scenario of the medical staff with two operators and the patient. A personal dosemeter, put on apron, was modelled for comparison with measurements performed in hospitals, done with electronic dosemeters, in a reduced number of interventional radiology practices. Within the limits associated to the use of numerical anthropomorphic models to mimic a complex interventional procedure, the personal dose equivalent, H p (10), was evaluated and normalised to the simulated Kerma-Area Product, KAP, value, indeed the effective dose has been calculated. The H p (10)/KAPvalue of the first operator is about 10 µSv/Gy.cm2, when ceiling shielding is not used. This value is calculated on the trunk and it varies of +/-30% moving the dosemeter to the waist or to the neck. The effective dose, normalised to the KAP value, varies between 0.03 and 0.4 µSv/Gy.cm2. Considering all the unavoidable approximation of this kind of investigations, the comparisons with hospital measurement and literature data showed a good agreement allowing to use of the present results for dosimetric characterisation of interventional radiology procedures.

Monte Carlo studies on photon interactions in radiobiological experiments.

X-ray and γ-ray photons have been widely used for studying radiobiological effects of ionizing radiations. Photons are indirectly ionizing radiations so they need to set in motion electrons (which are a directly ionizing radiation) to perform the ionizations. When the photon dose decreases to below a certain limit, the number of electrons set in motion will become so small that not all cells in an "exposed" cell population can get at least one electron hit. When some cells in a cell population are not hit by a directly ionizing radiation (in other words not irradiated), there will be rescue effect between the irradiated cells and non-irradiated cells, and the resultant radiobiological effect observed for the "exposed" cell population will be different. In the present paper, the mechanisms underlying photon interactions in radiobiological experiments were studied using our developed NRUphoton computer code, which was benchmarked against the MCNP5 code by comparing the photon dose delivered to the cell layer underneath the water medium. The following conclusions were reached: (1) The interaction fractions decreased in the following order: 16O > 12C > 14N > 1H. Bulges in the interaction fractions (versus water medium thickness) were observed, which reflected changes in the energies of the propagating photons due to traversals of different amount of water medium as well as changes in the energy-dependent photon interaction cross-sections. (2) Photoelectric interaction and incoherent scattering dominated for lower-energy (10 keV) and high-energy (100 keV and 1 MeV) incident photons. (3) The fractions of electron ejection from different nuclei were mainly governed by the photoelectric effect cross-sections, and the fractions from the 1s subshell were the largest. (4) The penetration fractions in general decreased with increasing medium thickness, and increased with increasing incident photon energy, the latter being explained by the corresponding reduction in interaction cross-sections. (5) The areas under the angular distribution curves of photons exiting the medium layer and subsequently undergoing interactions within the cell layer became smaller for larger incident photon energies. (6) The number of cells suffering at least one electron hit increased with the administered dose. For larger incident photon energies, the numbers of cells suffering at least one electron hit became smaller, which was attributed to the reduction in the photon interaction cross-section. These results highlighted the importance of the administered dose in radiobiological experiments. In particular, the threshold administered doses at which all cells in the exposed cell array suffered at least one electron hit might provide hints on explaining the intriguing observation that radiation-induced cancers can be statistically detected only above the threshold value of ~100 mSv, and thus on reconciling controversies over the linear no-threshold model.

MCNPX CALCULATIONS OF SPECIFIC ABSORBED FRACTIONS IN SOME ORGANS OF THE HUMAN BODY DUE TO APPLICATION OF 133Xe, 99mTc and 81mKr RADIONUCLIDES.

Monte Carlo simulations were performed to evaluate treatment doses with wide spread used radionuclides 133Xe, 99mTc and 81mKr. These different radionuclides are used in perfusion or ventilation examinations in nuclear medicine and as indicators for cardiovascular and pulmonary diseases. The objective of this work was to estimate the specific absorbed fractions in surrounding organs and tissues, when these radionuclides are incorporated in the lungs. For this purpose a voxel thorax model has been developed and compared with the ORNL phantom. All calculations and simulations were performed by means of the MCNP5/X code.

Monte Carlo studies on neutron interactions in radiobiological experiments.

Monte Carlo method was used to study the characteristics of neutron interactions with cells underneath a water medium layer with varying thickness. The following results were obtained. (1) The fractions of neutron interaction with 1H, 12C, 14N and 16O nuclei in the cell layer were studied. The fraction with 1H increased with increasing medium thickness, while decreased for 12C, 14N and 16O nuclei. The bulges in the interaction fractions with 12C, 14N and 16O nuclei were explained by the resonance spikes in the interaction cross-section data. The interaction fraction decreased in the order: 1H > 16O > 12C > 14N. (2) In general, as the medium thickness increased, the number of "interacting neutrons" which exited the medium and then further interacted with the cell layer increased. (3) The area under the angular distributions for "interacting neutrons" decreased with increasing incident neutron energy. Such results would be useful for deciphering the reasons behind discrepancies among existing results in the literature.

Conversion coefficients for determination of dispersed photon dose during radiotherapy: NRUrad input code for MCNP.

Radiotherapy is a common cancer treatment module, where a certain amount of dose will be delivered to the targeted organ. This is achieved usually by photons generated by linear accelerator units. However, radiation scattering within the patient's body and the surrounding environment will lead to dose dispersion to healthy tissues which are not targets of the primary radiation. Determination of the dispersed dose would be important for assessing the risk and biological consequences in different organs or tissues. In the present work, the concept of conversion coefficient (F) of the dispersed dose was developed, in which F = (Dd/Dt), where Dd was the dispersed dose in a non-targeted tissue and Dt is the absorbed dose in the targeted tissue. To quantify Dd and Dt, a comprehensive model was developed using the Monte Carlo N-Particle (MCNP) package to simulate the linear accelerator head, the human phantom, the treatment couch and the radiotherapy treatment room. The present work also demonstrated the feasibility and power of parallel computing through the use of the Message Passing Interface (MPI) version of MCNP5.

Characteristics of Protons Exiting from a Polyethylene Converter Irradiated by Neutrons with Energies between 1 keV and 10 MeV.

Monte Carlo method has been used to determine the efficiency for proton production and to study the energy and angular distributions of the generated protons. The ENDF library of cross sections is used to simulate the interactions between the neutrons and the atoms in a polyethylene (PE) layer, while the ranges of protons with different energies in PE are determined using the Stopping and Range of Ions in Matter (SRIM) computer code. The efficiency of proton production increases with the PE layer thickness. However the proton escaping from a certain polyethylene volume is highly dependent on the neutron energy and target thickness, except for a very thin PE layer. The energy and angular distributions of protons are also estimated in the present paper, showing that, for the range of energy and thickness considered, the proton flux escaping is dependent on the PE layer thickness, with the presence of an optimal thickness for a fixed primary neutron energy.

Monte Carlo calculations of lung dose in ORNL phantom for boron neutron capture therapy.

Monte Carlo simulations were performed to evaluate dose for possible treatment of cancers by boron neutron capture therapy (BNCT). The computational model of male Oak Ridge National Laboratory (ORNL) phantom was used to simulate tumours in the lung. Calculations have been performed by means of the MCNP5/X code. In this simulation, two opposite neutron beams were considered, in order to obtain uniform neutron flux distribution inside the lung. The obtained results indicate that the lung cancer could be treated by BNCT under the assumptions of calculations.

Neutron detection by a CR-39 detector and analysis of proton tracks etched in the same and opposite directions.

A program code to simulate neutron interactions with a CR-39 detector and calculate parameters describing the induced etched proton tracks in the CR-39 material was previously developed(( 1)). This code was used to understand the mechanisms involved during interactions with neutrons in the CR-39 material and the influence of the etching process, enabling an improvement in the efficiency of the CR-39 detector. Due to neutron interaction with atoms of the detector material, the created protons are emitted in different directions and their latent tracks are oriented randomly within the detector. The aim of this paper is to show differences between the number of visible tracks etched in the same and opposite directions from both sides of the detector. The efficiency of neutron detection was analysed as a function of the removed layer and neutron energy for both sides of detector.

Efficiency of whole-body counter for various body size calculated by MCNP5 software.

The efficiency of a whole-body counter for (137)Cs and (40)K was calculated using the MCNP5 code. The ORNL phantoms of a human body of different body sizes were applied in a sitting position in front of a detector. The aim was to investigate the dependence of efficiency on the body size (age) and the detector position with respect to the body and to estimate the accuracy of real measurements. The calculation work presented here is related to the NaI detector, which is available in the Serbian Whole-body Counter facility in Vinca Institute.

Hit probability of a disk shaped detector with particles with a finite range emitted by a point-like source.

An analytical analysis of the geometrical efficiency of a circular detector for particles with a finite range, emitted from a point-like source, is given. Several different cases were determined, depending on the particle range, radius of the detector and the position of the source with respect to the detector. These cases were analyzed separately and different expressions for calculating the hit probability were obtained for each of them. Results were compared with Monte Carlo calculations and good agreement was found. The problem considered here might be relevant for alpha-particle detection under specific conditions.

Calculation of the effective dose from natural radioactivity in soil using MCNP code.

Effective dose delivered by photon emitted from natural radioactivity in soil was calculated in this work. Calculations have been done for the most common natural radionuclides in soil (238)U, (232)Th series and (40)K. A ORNL human phantoms and the Monte Carlo transport code MCNP-4B were employed to calculate the energy deposited in all organs. The effective dose was calculated according to ICRP 74 recommendations. Conversion factors of effective dose per air kerma were determined. Results obtained here were compared with other authors.

Absorbed fractions for electrons and beta particles in sensitive regions of human respiratory tract.

The absorbed fractions (AF) of electrons in sensitive layers of human respiratory tract were calculated in this paper. For that purpose the source code for simulation package PENELOPE, based on Monte Carlo method, was developed. The human respiratory tract was modeled according to ICRP66 publication, where AF of electrons was calculated using EGS4 simulation software. Some approximations used in ICRP66 were corrected in this work, and new values of AF for radon progeny are given. Minimal energy (EABS) that electron can have during transport through material is 1 keV in ICRP66, while it is set as low as 100 eV in the presented work. Lowering value of EABS gives more accurate results for AF when initial energy of electrons is below 50 keV. To represent tissue, water is used in ICRP66, while in this work epithelia tissue is used.

Room model with three modal distributions of attached 220Rn progeny and dose conversion factor.

Jacobi parametric room model was extended to the three modal distribution of aerosols, and applied to (220)Rn progeny. The computer program was developed to calculate ratios of progeny activity concentrations in different modes to (220)Rn concentration. The ratios are relatively small and they are given as functions on ventilation rate. Dose conversion factor (DCF) for (220)Rn progeny was calculated as 4.5 mSv WLM(-1), which is smaller by over three times than that for (222)Rn progeny.

Assessment of environmental radon hazard using human respiratory tract models.

Radon is a natural radioactive gas derived from geological materials. It has been estimated that about half of the total effective dose received by human beings from all sources of ionizing radiation is attributed to 222Rn and its short-lived progeny. In this paper, the use of human respiratory tract models to assess the health hazard from environmental radon is reviewed. A short history of dosimetric models for the human respiratory tract from the International Commission on Radiological Protection (ICRP) is first presented. The most important features of the newest model published by ICRP in 1994 (as ICRP Publication 66) are then described, including the morphometric model, physiological parameters, radiation biology, deposition of aerosols, clearance model and dose weighting. Comparison between different morphometric models and comparison between different deposition models are then given. Finally, the significance of various parameters in the lung model is discussed, including aerosol parameters, subject related parameters, target and cell related parameters, and parameters that define the absorption of radon from the lungs to blood. Dosimetric calculations gave a dose conversion coefficient of 15 mSv/WLM, which is higher than the value 5 mSv/WLM derived from epidemiological studies. ICRP stated that dosimetric models should only be used for comparison of doses in the human lungs resulted from different exposure conditions.

Room model with three modal distributions of attached radon progeny.

In this paper a room model with three modal distributions of attached radon progeny is developed. Recoil factors are recalculated for each of the modes, and different recoil factors than usually used are obtained. Dependence of progeny concentration in various modes on ventilation and attachment rate is presented. Unattached Pb is overestimated up to 15% if one modal distribution is used, which can lead to the overestimation of lung dose.

Influence of variability of 214Pb recoil factor on lung dose.

Different parameters enter models of the human respiratory tract. The unattached fraction of the radon progeny was identified as the most important parameter, with the strongest influence on lung dose. The unattached fraction depends on the indoor aerosol concentration and other environmental conditions. The recoil factor, p, which influences the unattached fraction of 214Pb and 214Bi, defined as the average detachment probability from the aerosol after an alpha decay of 218Po, has almost always been taken as a constant. Here the recoil factor was recalculated under different assumptions and found to be in the range between 0.1 and 0.8. A smaller recoil factor means lower unattached fractions of 214Pb and 214Bi. The influence of the recoil factor on lung dose was also estimated. The lung dose is smaller by about 10% if p = 0.1 is assumed in calculating the unattached fraction instead of p = 0.8.

Absorbed fraction of radon progeny in human bronchial airways with bifurcation geometry.

PURPOSE: The absorbed fraction, defined as the portion of the initial particle energy which is absorbed in the tissue of interest, was calculated, under bifurcation geometry of the airway tubes, for alpha-particles emitted from radon progeny in the human respiratory tract. The results are given for all branching generations and compared with the data obtained for the commonly used infinite straight cylinders adopted by the International Commission on Radiological Protection (ICRP) Report 66. MATERIALS AND METHODS: A model was created to calculate the absorbed fraction of alpha-particle energy in the human lung using bifurcation geometry. Monte Carlo simulations of alpha-particle propagation in tissue and air were performed. The stopping powers of alpha-particles were adopted from the International Commission of Radiation Units and Measurements (ICRU) Report 49. RESULTS: The absorbed fractions for the bifurcation geometry are given for the 15 generations in the tracheobronchial tree for alpha-particle energies of 6 and 7.69 MeV. The sources were assumed to be the fast and slow moving mucus. CONCLUSIONS: Comparisons with ICRP66 data reveal that the assumption of long, straight cylinders was appropriate in some cases, but not in all. Adoption of the absorbed fractions obtained from the bifurcation model instead of the ICRP66 data caused 'redistribution' of doses in the bronchial (BB) and bronchiolar (bb) regions.

Absorbed fraction of alpha-particles emitted in bifurcation regions of the human tracheo-bronchial tree.

A model for bifurcation regions of the human tracheo-bronchial tree was developed. Equations for the surfaces are given to enable calculations of doses from alpha-particles emitted in these regions. It has been found that a bifurcation region is well approximated by a quasi-ellipsoid. The absorbed fractions of alpha-particles emitted in bifurcation regions were calculated by the Monte Carlo method. The average absorbed fraction under the bifurcation geometry is close to that found under the cylindrical geometry in the bronchial region. In the bronchiolar region, the absorbed fractions under the bifurcation geometry are up to 20% larger than those under the cylindrical geometry.