An investigation on protection properties of Tantalum (V) oxide reinforced glass screens on unexposed breast tissue for mammography examinations

INTRODUCTION

The utilization of radiation shielding material positioned between the both breasts are crucial for the reduction of glandular dose and the safeguarding of the contralateral breast during mammographic procedures. This study proposes an alternative substance for shielding the contralateral breast from radiation exposure during mammography screening.

METHODS

In this study, we present an analysis of the shielding effectiveness of transparent glass that has been doped with Tantalum (V) oxide encoded as BTZT6. The evaluation of this shielding material was conducted using the MCNPX code, specifically for the ipsilateral and contralateral breasts. The design of the left and right breast phantoms involved the creation of three-layer heterogeneous breast phantoms, consisting of varying proportions of glandular tissue (25%, 50%, and 75%). The design of BTZT6 and lead-acrylic shielding screens is implemented using the MCNPX code. The comparative analysis of dose outcomes is conducted to assess the protective efficacy of BTZT6 and lead-acrylic shielding screens.

RESULTS

The utilization of BTZT6 shielding material resulted in a reduction in both breast dose and skin dose exposure when compared to the lead-acrylic shield.

CONCLUSION

Based on the findings acquired, the utilization of BTZT6 shielding material screens during mammography procedures involving X-rays with energy levels ranging from 26 to 30 keV is associated with a decrease in radiation dose.

IMPLICATIONS FOR PRACTICE

It can be inferred that the utilization of BTZT6 demonstrates potential efficacy in mitigating excessive radiation exposure to the breasts and facilitating the quantification of glandular doses in mammography procedures.

Tungsten (VI) oxide reinforced antimony glasses for radiation safety applications: A throughout investigation for determination of radiation shielding properties and transmission factors

We report the functional assessment of tungsten (VI) oxide on gamma-ray attenuation properties of 60Sb₂O₃-(40-x)NaPO₃-xWO₃ antimony glasses. The elemental mass-fractions and glass-densities of each glass sample are specified separately for the MCNPX Monte Carlo code. In addition to fundamental gamma absorption properties, Transmission Factors throughout a broad radioisotope energy range were measured. According to findings, holmium (Ho) incorporation into the glass structure resulted in a net increase of 0.3406 g/cm3, whereas cerium (Ce) addition resulted in a net increase of 0.2047 g/cm3. The 40% WO₃ reinforced S7 sample was found to have the greatest LAC value, even though seven glass samples exhibited identical behavior. The S2 sample had the lowest HVL values among the glass groups evaluated in this work, computed in the energy range of 0.015-15 MeV. The lowest EBF and EABF values were reported for 40% WO₃ reinforced S7 sample with the highest LAC and density values. According to the findings of this research, WO₃ will likely make a significant contribution to the gamma ray absorption properties of antimony glasses, which are employed for optical and structural modification. Therefore, it can be concluded that WO₃ may be treated monotonically and can be employed successfully in circumstances where gamma-ray absorption characteristics, optical properties, and structural qualities need to be enhanced.

An alternative method for calculation of half-value layers without the knowledge of attenuation coefficient

Radiation protection is crucial for the safe utilization of ionizing radiation and minimizing the harmful effect upon exposure, hence some standards have been defined by some relevant organizations for the safe uses of radiation. One of the parameters relevant to the calculation of gamma ray shielding is the half-value layer (HVL), which is normally calculated using the knowledge of linear attenuation coefficient (μ). In this research, an attempt has been made to directly calculate HVL without the knowledge of μ via Monte Carlo simulation technique. For this purpose, in the Monte Carlo N-Particle eXtended (MCNPX) code, F1, F5 and Mesh Popul sequences tallies were defined and the optimal structure for the least measurement error was introduced. The MCNPX calculated values showed reasonable agreement with the experimental findings. According to the obtained results, it is suggested that in order to reduce the error of HVL calculations, in exchange for the MCNPX code, the values of the R parameter and the radiation angle of the source should be considered according to the calculations introduced in this plan. Because the results show that by considering the measurement error between 6 and 20%, the code output can be cited in different energy ranges.

Comparative analysis on application conditions of indium (III) oxide-reinforced glasses in nuclear waste management and source transportation: A Monte Carlo simulation study

This study's primary objective is to provide the preliminary findings of novel research on the design of Indium (III) oxide-reinforced glass container that were thoroughly developed for the purpose of a nuclear material container for transportation and waste management applications. The shielding characteristics of an Indium (III) oxide-reinforced glass container with a certain elemental composition against the ⁶⁰Co radioisotope was thoroughly evaluated. The energy deposition in the air surrounding the designed portable glass containers is measured using MCNPX general-purpose Monte Carlo code. Simulation studies were carried out using Lenovo-P620 workstation and the number of tracks was defined as 10⁸ in each simulation phase. According to results, the indium oxide-doped C6 (TZI8) container exhibits superior protective properties compared to other conventional container materials such as 0.5Bitumen-0.5 Cement, Pb Glass composite, Steel-Magnetite concrete. In addition to its superiority in terms of nuclear safety, it is proposed that the source's simultaneous observation and monitoring, as well as the C6 (TZI8) glass structure's transparency, be underlined as significant advantages. High-density glasses, which may replace undesirable materials such as concrete and lead, provide several advantages in terms of production ease, non-toxic properties, and resource monitoring. In conclusion, the use of Indium (III) oxide-reinforced glass with its high transparency and outstanding protection properties may be a substantial choice in places where concrete is required to ensure the safety of nuclear materials.

Simulation study of a simultaneous beta-gamma-ray detection using a 3-layer phoswich detector and Monte Carlo methods

Combination of two or three dissimilar scintillator materials as a radiation detector has found major role in environmental radiation monitoring. In this paper, a three-layer Phoswich detector including BC-400, YAG, and CsI was designed to efficiently discriminate gamma-ray in the beta events up to 3.2 MeV using a simple rise-time discrimination method. MCNPX Monte Carlo code was used to obtain interaction probability of beta and gamma-rays as well as optimum thicknesses of the layers in the designing process. The optical transport of the system was simulated by GEANT4. In this regard, the pulses from simultaneous beta-gamma emitter sources were detected and discriminated based on pulse's rise-time so that the minimum number of gamma-ray contaminating events was observed in the beta spectrum. The results showed that using the proposed configuration and the method, output pulses with a rise-time shorter than 9 ns have been successfully detected as a beta particle while those with rising time longer than 15 ns have been identified as gamma-ray events. Overall results revealed that using the proposed system, an individual spectrum of beta particles or gamma-rays can be recorded from a simultaneous beta-gamma emitter source that minimizes contribution of the other radiation.

Utilization of three-layers heterogeneous mammographic phantom through MCNPX code for breast and chest radiation dose levels at different diagnostic X-ray energies: A Monte Carlo simulation study

Introduction

We report the breast and chest radiation dose assessment for mammographic examinations using a three-layer heterogeneous breast phantom through the MCNPX Monte Carlo code.

Methods

A three-layer heterogeneous phantom along with compression plates and X-ray source are modeled. The validation of the simulation code is obtained using the data of AAPM TG-195 report. Deposited energy amount as a function of increasing source energy is calculated over a wide energy range. The behavioral changes in X-ray absorption as well as transmission are examined using the F6 Tally Mesh extension of MCNPX code. Moreover, deposited energy amount is calculated for modeled body phantom in the same energy range.

Results and discussions

The diverse distribution of glands has a significant impact on the quantity of energy received by the various breast layers. In layers with a low glandular ratio, low-energy primary X-ray penetrability is highest. In response to an increase in energy, the absorption in layers with a low glandular ratio decreased. This results in the X-rays releasing their energy in the bottom layers. Additionally, the increase in energy increases the quantity of energy absorbed by the tissues around the breast.

Introduction of Graphene/h-BN Metamaterial as Neutron Radiation Shielding by Implementing Monte Carlo Simulation

In this paper, graphene/h-BN metamaterial was investigated as a new neutron radiation shielding (NRS) material by Monte Carlo N-Particle X version (MCNPX) Transport Code. The graphene/h-BN metamaterial are capable of both thermal and fast neutron moderator and neutron absorber process. The constituent phases in graphene/h-BN metamaterial are chosen to be hexagonal boron nitride (h-BN) and graphene. The introduced target was irradiated by an Am-Be neutron source with an energy spectrum of 100 keV to 15 MeV in a Monte Carlo simulation input file. The resulting current transmission rate (CTR) was investigated by the MCNPX code. Due to concrete's widespread use as a radiation shielding material, the results of this design were also compared with concrete targets. The results show a significant increase in NRS compared to concrete. Therefore, metamaterial with constituent phase's graphene/h-BN can be a suitable alternative to concrete for NRS.

Influence of radiotherapy room shielding on ambient dose equivalent due to photons H*(10)p and neutrons H*(10)n in the patient's plane

This study discusses a computer simulation for the equivalent ambient dose due to photons, H*(10)p, and neutrons, H*(10)n, in the patient's plane undergoing radiation therapy. A standard radiotherapy room with an additional shielding made by one lead or steel tenth-value layer was considered. A Varian 2100/2300 C/D linear accelerator head operating at 18 MV was modeled. Jaw openings of 5 cm × 5 cm, 10 cm × 10 cm, 20 cm × 20 cm, and 30 cm × 30 cm, as well as the multileaf collimator under eight different angles of gantry inclination, were also modeled. The use of steel in the shield generates a slightly raised average value of H*(10)p (0.527%) compared to when using lead. This finding can be interpreted as that the use of lead or steel coating makes no difference to the additional shield calculations when only photons are considered. When considering the contribution to H*(10)n, there is a significant difference (11.7% increase) for using lead compared to steel shielding.

A Monte Carlo study to investigate the feasibility to use the Moroccan panoramic irradiator in sterile insect technique programs

Mediterranean fly pest (Ceratitis) is one of the most destructive pests of fruit species in Morocco. The sterile insect technique (SIT) is an environmentally friendly strategy that uses ionizing radiation to sterilize adult insects. Morocco has a panoramic gamma irradiator used to irradiate agri-food products. This irradiator is not dedicated to SIT programs due to its geometry that does not allow to obtain a dose uniformity ratio (DUR) recommended for such applications. This article presents a Monte Carlo study to investigate the feasibility of using the panoramic gamma irradiator at the National Institute for Agronomic Research (NIAR) of Tangier, Morocco, to setting up SIT methods and contributing to Ceratitis control programs. The Monte Carlo method was used to simulate the concrete bunker in which the panoramic gamma irradiator is installed. To obtain a recommended DUR required for SIT programs, two cells similar of the Gammacell-220 irradiator, which is mainly used in the SIT programs around the world, were simulated inside the concrete bunker. The simulation and calculations were performed using the MCNPX-2.7e Monte Carlo simulation code. It is demonstrated that at both investigated positions, the spatial distribution of dose rates in the two modeled irradiation cells, which were similar to a gammacell-220 irradiator cell, are uniform enough that the cells can be used for SIT programs. It is concluded that the panoramic irradiator at NIAR can be used to contribute to the control of Mediterranean fly pest and other insect pests in Morocco.

Impact of Eye and Breast Shielding on Organ Doses During Cervical Spine Radiography: Design and Validation of MIRD Computational Phantom

Purpose: The study aimed to design and validate computational phantoms (MIRD) using the MCNPX code to assess the impact of shielding on organ doses. Method: To validate the optimized phantom, the obtained results were compared with experimental results. The validation of the optimized MIRD phantom was provided by using the results of a previous anthropomorphic phantom study. MIRD phantom was designed by considering the parameters used in the anthropomorphic phantom study. A test simulation was performed to compare the dose reduction percentages (%) between the experimental anthropomorphic phantom study and the MCNPX-MIRD phantom. The simulation was performed twice, with and without shielding materials, using the same number and locations of the detector. Results: The absorbed dose amounts were directly extracted from the required organ and tissue cell parts of output files. Dose reduction percentages between the simulation with shielding and simulation without shielding were compared. The highest dose reduction was noted in the thymus (95%) and breasts (88%). The obtained dose reduction percentages between the anthropomorphic phantom study and the MCNPX-MIRD phantom were highly consistent and correlated values with experimental anthropomorphic data. Both methods showed Relative Difference (%) ranges between 0.88 and 2.22. Moreover, the MCNPX-MIRD optimized phantom provides detailed dose analysis for target and non-target organs and can be used to assess the efficiency of shielding in radiological examination. Conclusion: Shielding breasts and eyes during cervical radiography reduced the radiation dose to many organs. The decision to not shield patients should be based on research evidence as this approach does not apply to all cases.

MCNPX computational modeling applied to the potential dose rates calculation of cargo scanning

This study focusses on the risk of potential exposure to radiation for personnel driving a truck as well as illegal individuals being transported in cargo containers. Inspection facilities usually use a high energy linear accelerator (linac) in order to inspect the cargo. Since this type of equipment has associated health risks due to potential unwanted exposure, the occupational and public dose limits should be calculated in order to develop safer work conditions. This work used a computation model running the code MCNPX to simulate a typical cargo inspection facility which used a linac operating at 4.5 MeV. Two scenarios were considered: (1) exposure of the driver to the primary beam due to a potential failure of the safety sensors; and (2) dose received by an illegal individual being transported inside the cargo container. The results show a dose of 0.8514 mSv per scan for the driver exposed to the primary X-ray beam, and 0.1997 mSv per scan for an individual transported in the cargo box. In conclusion, both the individual and the driver received a dose below the acceptable limit considered safe for an individual (1 mSv/year). However, that was the value of one scan; in a case in which multiple scans would be performed, the dose limit can be quickly exceeded. In that case, the limit would be exceeded by the driver faster than by the individual in the cargo.

Calculation of Photoneutron Contamination of Varian Linac in ICRU Soft-Tissue Phantom Using MCNPX Code

Purpose

The aim of this research was to calculate the fluence, dose equivalent (DE), and kerma of thermal, epithermal and fast photoneutrons separately, within ICRU soft-tissue-equivalent phantom in the radiotherapy treatment room, using MCNPX Monte Carlo code.

Materials and Methods

For this purpose, 18 MV Varian Linac 2100 C/D machine was simulated and desired quantities were calculated on the central axis and transverse directions at different depths.

Results

Maximum fluence, DE and kerma of total photoneutrons on central axis of the phantom were 43.8 n.cm^-2.Gy^-1, 0.26, and 3.62 mGy.Gy^-1, at depths 2, 0.1, 0.1 cm, respectively. At any depth, average of fluence, DE and kerma in the outer area of the field were less than the inner area and in general were about 72%, 52%, and 45%, respectively.

Conclusion

According to this research, within the phantom; variation of fluence, DE and kerma in transverse direction were mild, and along the central axis at shallow area were sharp. DE of fast photoneutrons at shallow and deep areas were one order of magnitude greater than thermal photoneutrons.

Neutron and photon out-of-field doses at cardiac implantable electronic device (CIED) depths

The accuracy of an out-of-field dose from an Elekta Synergy accelerator calculated using the X-ray Voxel Monte Carlo (XVMC) dose algorithm in the Monaco treatment planning system (TPS) for both low-energy (6 MV) and high-energy (15 MV) photons at cardiac implantable electronic device (CIED) depths was investigated through a comparison between MCNPX simulated out-of-field doses and measured out-of-field doses using three high spatial and sensitive active detectors. In addition, total neutron equivalent dose and fluence at CIED depths of a 15-MV dose from an Elekta Synergy accelerator were calculated, and the corresponding CIED relative neutron damage was quantified. The results showed that for 6-MV photons, the XVMC dose algorithm in Monaco underestimated out-of-field doses in all off-axis distances (average errors: -17% at distances X < 10 cm from the field edge and -31% at distances between 10 < X ≤ 16 cm from the field edge), with an increasing magnitude of underestimation for high-energy (15 MV) photons (up to 11%). According to the results, an out-of-field photon dose at a shallower CIED depth of 1 cm was associated with greater statistical uncertainty in the dose estimate compared to a CIED depth of 2 cm and clinical depth of 10 cm. Our results showed that the relative neutron damage at a CIED depth range for 15 MV photon is 36% less than that reported for 18 MV photon in the literature.

Evaluation of gamma ray and neutron attenuation capability of thermoplastic polymers

The fast neutron and gamma ray attenuation capability of the most common thermoplastic polymers used in nuclear applications has been evaluated theoretically. Monte Carlo simulation has been used to compute the gamma-ray energy absorption buildup factor in the energy range 0.015-15 MeV at penetration depths up to 40 MFP. The results of MCNPX calculations have been validated against the results derived from the Geometric Progression fitting method. To evaluate neutron attenuation performance of the polymers, the fast neutron removal cross-section has been determined using theoretical database. Despite the superior ability of polysulfone and poly (ether sulfone) in gamma ray attenuation, high-density polyethylene has been found to have the best fast neutron removal ability among all. The detailed insights into the fast neutron and gamma ray shielding properties of selected polymers in the present work might have great potential applications in nuclear systems.

TENIS - ThErmal Neutron Imaging System for use in BNCT

A thermal neutron flux measurement tool with two perpendicular sets of plastic scintillator arrays was designed and simulated (Ghal-Eh and Green, 2016) with the MCNPX code (Version 2.6.0, with ENDF/B-VII cross section library (ENDF, 2011)). The proposed system aimed to provide a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from ¹H(n_th, γ)²D reactions. In the present work, using Monte Carlo code FLUKA and its scintillation light transport capability, several important upgrades were carried out to include the light transport modeling in the response of plastic scintillators, analyze the cross-talk phenomenon, optimize the system geometry, and also provide a new approach in thermal neutron image reconstruction. The results showed that the last two cases played a significant role in improving the longitudinal profile of thermal neutron flux.

Modelling of reusable target materials for the production of fission produced 99Mo using MCNP6.2 and CINDER90

Current fission-based methods of ⁹⁹Mo production require single use uranium targets which leads to spent uranium waste. This waste could be reduced if a target is developed that does not require dissolution so that it can be reused for multiple production runs. MCNP6.2 was used to model reusable targets of 20% and 1% enrichment for activity produced, target efficiency and burnup. The 1% enriched target was found to be much more efficient but had a lower activity produced compared to the 20% enriched target. The ideal target design for ⁹⁹Mo production that optimises efficiency and reusability and reduces the self-shielding effect of UO₂ was found to be a target that is made from 1% enriched UO₂ with density as high as allowable for sufficient yields, efficient ⁹⁹Mo extraction and having an irradiation time of 5 days, with the target able to be re-irradiated and re-processed 2-4 times.

Development of a non-invasive method for monitoring variations in salt concentrations of seawater using nuclear technique and Monte Carlo simulation

In the oil production industry, water is used as a fluid injected into the well to raise the oil when the well is depressurized. Water thus produced presents variations in the concentrations of dissolved salts, as there is a mixture of different types of water, related to its origin (such as connate water, sea water). Because it is reused in oil production, water needs to be monitored to maintain the standard suitable for its use as it can be hypersaline, contributing to the encrustation of pipes and contamination of underground water reservoirs. In this study, a noninvasive method was developed to determine the salt concentration in seawater. The method uses a detection system that contains a NaI(Tl) detector, a²⁴¹Am source, and a sample holder to measure the mass attenuation coefficient of saltwater samples. For validation, the same setup was also simulated using the MCNPX code. Saltwater samples with different concentrations of NaCl and KBr were used as a proxy for seawater. The mass attenuation coefficients for the simulation exhibited the smallest relative errors (up to 6.2%), and the experimental ones exhibited the highest relative errors (up to 25%) when compared with theoretical values.

Effect of human body position on gamma radiation dose rate from granite stones

The use of granite stones as building materials in homes or offices can result in the residents' long-term whole-body exposure to gamma radiation. Although the whole-body annual dose has been investigated in the literature, it is obvious that different human organs receive different equivalent dose due to different position respect to the walls and floor covered by granite stones. In this paper, the effect of distance from the walls and floor of a room on the equivalent dose is investigated by using MCNPX code. An "ORNL" phantom is simulated in three situations; standing (P1), sleeping one meter above the floor (P2) and sleeping on the floor (P3) and the equivalent dose in different organs is calculated. Excess lifetime cancer risk (ELCR) is calculated in the whole of the body for these three positions. By the results, the value of ELCR in the third position is more than the average world value (2.9 × 10-4). The results show the importance of considering body position in dose determination, especially in some organs such as the brain and eyes which are close to the granite stones in certain positions such as sleeping.

An intelligent gamma-ray technique for determining wax thickness in pipelines

Measuring the wax deposition inside pipelines is one of the critical parameters in the oil, gas and petrochemical industries to control the flow through the pipelines. This paper presents a novel method using artificial neural networks to measure the thickness of the wax. This method was based on counting the backscattered gamma-ray from different thicknesses of the wax inside the pipes with different diameters. For this purpose, the system was simulated by MCNPX code and the designed setup was optimized. The main analyses were based on the simulation results but the verification was performed using a real experimental setup. The results showed a good agreement between the simulation results and the experimental data with a root mean square error less than 1%. Response of the detector was simulated for a standard industrial nominal pipe ranged from 2 to 4.5 inches and for radiation sources ¹³⁷Cs and ⁶⁰Co. Using these data, a multilayer perceptron for different energy sources was trained. The best prediction of the wax thickness was obtained for the case of using two radiation sources, simultaneously. The output of the trained neural network showed that the proposed method is capable of measuring the wax thickness inside the pipe with a good accuracy.

Dose reduction in HDR brachytherapy of esophageal cancer using gold and gold alloy plaques: a Monte Carlo study

In this work, the use of gold and gold alloy plaques is proposed for the first time, to reduce the dose to healthy organs in brachytherapy with Ir-192 sources. For dose simulations in tumour and healthy tissue, the MCNPX Monte Carlo code was used. The radiation source implemented in those simulations was benchmarked with well-known TG-43 criteria of dose rate constant, air-kerma strength, radial dose function, and 2D anisotropy function. For various arrangements of iridium sources and plaques of gold and gold alloy of various thicknesses, the dose distributions in an esophagus tumour and in surrounding healthy organs were simulated. The results showed that while the dose to the tumour is not much affected by the presence of gold plaques with a thickness of 3.5 mm in an optimized ¹⁹²Ir sources' configuration, a relative reduction in average organ dose of 64%, 65%, 73%, 67%, and 35% was observed, for esophagus, thyroid, heart, stomach, and liver, respectively. Moreover, it was found that a gold plaque leads to smaller doses to healthy organs than a gold alloy plaque. It is concluded that gold plaques can be used to improve the treatment of esophageal cancer by HDR brachytherapy and to protect surrounding non-target organs.

Evaluation of effectiveness of equivalent dose during proton boron fusion therapy (PBFT) for brain cancer: A Monte Carlo study

Recently it has been suggested that the presence of boron-11 during proton therapy leads to a significant dose increasement in the BUR. Three high-LET alpha particles with an average energy of 4 MeV are generated at the point of interaction between proton and boron-11. Nevertheless, the cross-section of p+B11→3α interaction is negligible and dose increasement is unlikely. The purpose of this study is dose evaluation of the proton therapy with and without the boron-11. All simulations were performed using MCNPX 2.6.0 code at the Snyder head phantom. At the elderly stage, the range of Bragg-peaks was adapted to the tumor volume, with and without boron-11. Then, the different concentrations of boron-11 were assumed including 65,500,10³,10⁵,2.5×10⁵ and 5×10⁵ppm in the tumor region. To investigate the maximum effectiveness of PBFT (proton boron fusion therapy), the entire tumor was assumed full of boron-11, and the dose components were calculated. Consequently, In the best case, the maximum dose amplification was less than 5%, in which the entire tumor was assumed full boron-11. The total number of alpha particles generated from p+B11→3α interaction is negligible. As well as the presence of boron-11 during the proton therapy makes that the Bragg-peaks happen in greater depth. Hence, from the Monte Carlo standpoint, the effectiveness of the proton boron fusion therapy is not related to the alpha particles because the dose component of alpha particles is negligible.

Comparison of MCNPX and FLUKA Monte Carlo codes in the simulating a nuclear gauge

In this paper, a nuclear gauge was simulated using MCNPX and FLUKA codes. This device consists of a fluid-containing vessel, three BC400 rod plastic scintillator and a⁶⁰Co gamma source. Simulation studies show that the changes in the count of each of the three detectors and the sum of their counts decrease with increasing vessel water height. The simulation results were compared with the measurement results. Comparison results show that the mean count difference between MCNPX code and experimental results is about 3% lower than FLUKA code, but the computation time using FLUKA code is approximately 2.8 times less than the MCNPX code.

Shielding features, to non-ionizing and ionizing photons, of FeCr-based composites

This paper has been focused on the a detail study on non-ionizing and ionizing electromagnetic (EM) shielding features and build-up factors of reinforced with ferrochrome (FeCr) composites. The non-inozing electromagnetic shielding performance quantities of composites have been determined in the frequency range between 12.4 and 18.0 GHz. Also, the experimental mass attenuation coefficients (MAC) have been estimated using gamma spectrometer and various radioactive point, and compared to those of theoretical and simulation (MCNPX) results. With help of the obtained linear attenuation coefficients, several attenuation quantities, i.e., effective atomic number (Z_eff), half value layer (HVL), and mean free path (MFP) have been discussed. In addition, buildup factors (EBF and EABF) values have been estimated utilizing the G-P fitting method. The results showed that composite encoded FeCr(15%) is superior shielding attenuation properties among the investigated samples.

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

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

Applications of MCNP simulation in treatment planning: a comparative study

Monte Carlo codes have been used for approximately 80 years to solve various problems in medical physics. In this paper, the importance of the MCNPX code in treatment planning is highlighted. As illustrative examples of the role of MCNPX in this field, some dosimetric parameters, isodose distribution curves, and figures of merit (FOMs) were considered for photon beams of various energies. To the best of the authors' knowledge, such a systematic study has not been done before. Tissue-air ratio values were obtained as a function of depth in tissue as well as field size. The results of the simulations were in agreement within 3.5% with experimental results reported in the literature. Backscatter factor values were calculated as a function of beam energy, and found to be in agreement with published experimental values within 5.9%. The isodose curves for different conditions and beam arrangements were also simulated. Besides, FOMs were calculated for different radiation energies. All the results were in agreement with related data in the literature. It is concluded that the MCNPX code and the models developed in the present study can be used in different conditions where these parameters are involved, improving individualized treatment planning for individual patients.

A commentary on the impact of modelling results to inform mission planning and shield design

A correspondence has been received in reference to a recently published article titled "On the decision making criteria for cis-lunar reference scenarios". The intent of the paper was to demonstrate: (i) a novel methodology for calculating the dose from solar particle events (SPEs), and (ii) the impact of the SPE parametric model, shield thickness, dose metric, and radiation transport code on choosing a worst-case scenario. This effort assumed a spherical, aluminum spacecraft with an internal diameter of 3.8 m and with varying wall thickness ranging from 2 to 10 cm. A brief component of this article compared the dose from several solar particle events (SPEs) inside the spherical spacecraft geometry as calculated with Monte Carlo radiation transport code MCNPX and the on-line tool OLTARIS. In this comparison, the MCNPX simulation parameters assumed a volume-averaged dose while OLTARIS calculations assumed a point-dose estimate at the center of the spherical geometry. These modeling assumptions were detailed in the initial publication. The differences in the neutron, proton, and light-ion fluences and doses obtained between both codes were generally attributed to differences transport methodologies, nuclear physics models, boundary condition setup and detector regions. The commentary received demonstrated when both codes used a point-detector geometry and/or volume-averaged geometries, the two would yield similar proton fluences. This is a worthwhile observation that further emphasizes the impact of modeling assumption. The commentary further suggested however that the volume-averaged dose results "artificially reduced" estimates and that it was both "misleading" and "not-applicable" for use in storm shelter design. The response presented here will reiterate the context of the initial assumptions made, demonstrate the variability in point-dose estimates relative to a volume-averaged dose estimate, state why a volume-averaged estimate is equally applicable in this context, and lastly reference other factors that can give rise to increased uncertainty.

Dosimetric investigation of a new high dose rate 192Ir brachytherapy source, IRAsource, by Monte Carlo method

Purpose

The purpose of the present study was to perform an independent calculation of dosimetric parameters associated with a new ¹⁹²Ir brachytherapy source model, IRAsource.

Materials and methods

The parameters of air kerma strength (AKS), dose rate constant (DRC), geometry function (GF), radial dose function (RDF), as well as two-dimensional (2D) anisotropy function (AF) of IRAsource ¹⁹²Ir source model were calculated in this study. The MC n-particle extended (MCNPX) code was also employed for simulating high dose rate (HDR), IRAsource and ¹⁹²Ir source; and formalism was used for calculating dosimetry parameters based on task group number 43 updated report (TG-43 U1).

Results

The results of this study were consistent with the ones reported about the IRAsource source by Sarabiasl et al. The AKS per 1 mCi activity and the DRC values were also equal to 3.65 cGycm² h^-1 mCi^-1 and 1.094 cGyh^-1U^-1; respectively. The comparison of the results of the DRC and the RDF reported by Sarabiasl et al. also validated the ¹⁹²Ir IRAsource simulation in this study. Moreover, the AFs of IRAsource source model were in a good agreement with those of Sarabiasl et al. at different distances, which could be attributed to identical geometries.

Conclusion

In line with those reported by Sarabiasl et al., the results of this study confirmed the IRAsource ¹⁹²Ir source for clinical uses. The calculated dosimetric parameters of the IRAsource source could be utilized in clinical practices as input data sets or for validation of treatment planning system calculations.

Production assessment of non-carrier-added 199Au by (n,γ) reaction

Gold-199 is a promising theranostic radionuclide for targeted radioimmunotherapy as well as for scintigraphy and dosimetry. ¹⁹⁹Au can be produced in two methods in the direct and indirect routes of the reactor production via ¹⁹⁷Au(n,γ)¹⁹⁸Au(n,γ)¹⁹⁹Au as the direct or ¹⁹⁸Pt(n,γ)¹⁹⁹Pt→¹⁹⁹Au as the indirect method. This investigation described the development of a method for the reactor production of no-carrier-added (NCA) ¹⁹⁹Au through neutron activation of natural Pt in Tehran Research Reactor (TRR) at a thermal neutron flux of 3.5 × 10¹³ n cm^-2 s^-1. Also, in this paper, the activity of ¹⁹⁹Au has been estimated using the MCNPX code. In this case, first, the reactor core is simulated. Then the calculated results are compared with the corresponding experimental values. Moreover, two different chemical separation methods are investigated experimentally in details.

Influence of the size of nano- and microparticles and photon energy on mass attenuation coefficients of bismuth-silicon shields in diagnostic radiology

Recent studies have shown that the particle size of the shielding material and photon energy has significant effects on the efficiency of radiation-shielding materials. The purpose of the current study was to investigate the shielding properties of the bismuth-silicon (Bi-Si) composite containing varying percentages of micro- and nano-sized Bi particles for low-energy X-rays. Radiation composite shields composed of nano- and micro-sized Bi particles in Si-based matrix were constructed. The mass attenuation coefficients of the designed shields were experimentally assessed for diagnostic radiology energy range. In addition, the mass attenuation coefficients of the composite were comprehensively investigated using the MCNPX Monte Carlo (MC) code and XCOM. The X-ray attenuation for two different micro-sized Bi composites of radii of 50 µm and 0.50 µm showed enhancement in the range of 37-79% and 5-24%, respectively, for mono-energy photons (60-150 keV). Furthermore, the experimental and MC results indicated that nano-structured composites had higher photon attenuation properties (approximately 11-18%) than those of micro-sized samples for poly-energy X-ray photons. The amount of radiation attenuation for lower energies was more than that of higher energies. Thus, it was found that the shielding properties of composites were considerably strengthened by adding Bi nano-particles for lower energy photons.

Conceptual design for a new heterogeneous 241Am-9Be neutron source assembly using SOURCES4C-MCNPX hybrid simulations

In this study, an approach to simulate a novel variable-yield heterogeneous ²⁴¹Am-⁹Be was proposed with a hybrid use of SOURCES4C and MCNPX codes, and its energy spectrum and neutron emission yield were simulated. In these simulations, the energy spectra of the alpha particles emitted from the americium source and the neutrons produced within the beryllium and oxygen contents as a result of ⁹Be(α,n) and ^17,18O(α,n) reactions were calculated with SOURCES4C whilst the neutron transport from neutron production points to the space outside the source assembly were performed with the MCNPX code. The neutron energy spectrum and neutron emission yield for two different configurations of single-rod and multi-rod sources (i.e., americium or americium oxide rods in beryllium medium) were compared to a source of homogeneous americium (or its oxides) and beryllium mixture. The proposed heterogeneous geometry was aimed to provide a neutron source with a variable neutron yield, easy-to-shut down and easy-to-waste process features. The results confirmed that the homogeneous source represented the largest neutron yield compared to single- and multi-rod geometries. However, the neutron yield in heterogenous geometry could be altered by varying the number of americium (or americium oxide) rods to reach the desired neutron yield.

Processing of JEFF-3.3 and ENDF/B-VIII.0 and testing with critical benchmark experiments and TRIGA Mark II research reactor using MCNPX

A comparative study has been performed with the latest evaluated nuclear data libraries JEFF-3.3 and ENDF/B-VIII.0. The study has been conducted through the benchmark calculations for 120 criticality problems and the TRIGA Mark II research reactor with the libraries processed using NJOY21 for MCNPX Monte Carlo transport code. The criticality benchmarks assemblies, taken from the ICSBEP benchmark, cover Uranium (highly enriched uranium, intermediate enriched uranium, low enriched uranium, and²³³U) and Plutonium fuel systems in a various metal forms, and under a various spectral conditions. The Moroccan TRIGA Mark II research reactor calculation is used to look into the predictive capability of those nuclear data libraries and then to compare the accuracy of the predicted results with the experimental data published elsewhere. Actually, the purpose of this study is to investigate some neutronic and kinetic parameters of those benchmarks for both libraries. The former consist of effective multiplication factor, heat distribution, neutron flux distribution, effective delayed neutron fraction (β_eff), prompt removal lifetime (τ_r) and the mean neutron generation time (Λ). The results show that the calculated effective multiplication factor, heat distribution, neutron flux distribution, and the kinetic parameters are in good agreement with references. However, it is found that the computed values are strongly depending on the nuclear data set used in calculations.

Proposing an intelligent approach for measuring the thickness of metal sheets independent of alloy type

Radiation based gauges have been widely utilized as a nondestructive and robust tool for measuring the thickness of metal sheets in industry. The typical radiation thickness meter can just work accurately when the composition of the material is fixed during the measurement process. In conditions that material composition may differ substantially from the nominal composition, such as manufacturing rolled metals factories, the thickness measurements would be along with errors. The purpose of the present research is resolving the problem of measuring the thickness of metal sheets with various alloys. The aluminum is investigated in this work as a case study but the procedure can be applied for other types of metals. As the first step, the performance of various arrangements of two main detection techniques, named dual energy and dual modality, were investigated using MCNPX code to obtain optimum technique and arrangement. The simulation results indicated that a binary combination of ²⁴¹Am-⁶⁰Co isotopes as the source and one transmission detector in dual energy technique is the most appropriate choice. After then, an experimental setup based on the obtained optimal technique from simulation investigations was established. The aluminum sheets with 4 alloy types of 1050, 3105, 5052 and 6061 and thicknesses in the range of 0.2-4 cm with a step of 0.2 cm were tested and the obtained data were implemented for testing and training the artificial neural network (ANN). The proposed methodology could predict the thickness of aluminum sheet independent of its alloy type with an error of less than 0.04 cm in experiments.

Defining Processing Times for Accelerator Produced 225Ac and Other Isotopes from Proton Irradiated Thorium

During the purification of radioisotopes, decay periods or time dependent purification steps may be required to achieve a certain level of radiopurity in the final product. Actinum-225 (Ac-225), Silver-111 (Ag-111), Astatine-211 (At-211), Ruthenium-105 (Ru-105), and Rhodium-105 (Rh-105) are produced in a high energy proton irradiated thorium target. Experimentally measured cross sections, along with MCNP6-generated cross sections, were used to determine the quantities of Ac-225, Ag-111, At-211, Ru-105, Rh-105, and other co-produced radioactive impurities produced in a proton irradiated thorium target at Brookhaven Linac Isotope Producer (BLIP). Ac-225 and Ag-111 can be produced with high radiopurity by the proton irradiation of a thorium target at BLIP.

Developing a new target design for producing 99Mo in a MTR reactor

⁹⁹Mo is an important radioisotope and mainly produced using uranium fission reaction in a nuclear reactor. Investigation for probable improvements, especially on target geometry and in-core location of target is the main goal of this research. This is for producing more efficient of ⁹⁹Mo in a typical Material Testing Reactors (MTRs). A parametric investigation is done focused on the target characteristics such as geometry, location, material, density, heat flux (power density) and also usability in Tehran Research Reactor (TRR) as a MTR case study. Stochastic code MCNPX 2.6.0 along with CFD code ANSYS are used to perform neutronic and thermal-hydraulic analyses. A target with plate type design is specified and proposed as a final and most favorable design. Taking into account the safety criteria, the production yield, the chemical process and radioactive waste, it is demonstrated that the new target design meets the key design requirements without compromising the reactor safety. This research results indicate that the new target gives rise to a higher production of ⁹⁹Mo using less amount of initial material causing to a reduction in nuclear waste and process difficulties.

Monte Carlo simulations and analysis of transmitted gamma ray spectra through various tissue phantoms

Studies of radiation interactions with tissue equivalent material find importance in efforts that seek to avoid unjustifiable dose to patients, also in ensuring quality control of for instance nuclear medicine imaging equipment. Use of the Monte Carlo (MC) simulation tool in such characterization processes allows for the avoidance of costly experiments involving transmitted X- and γ-ray spectrometry. Present work investigates MC simulations of γ-ray transmission through tissue equivalent solid phantoms. Use has been made of a range of radionuclide gamma ray sources, ^99mTc, ¹³¹I, ¹³⁷Cs, ⁶⁰Co (offering photons in the energy range from a few keV up to low MeV), popularly applied in medicine and in some cases for gauging in industry, obtaining the transmission spectra following their interaction with various phantom materials and thicknesses. In validation of the model, the simulated values of mass attenuation coefficients (μ/ρ) for different phantom materials and thicknesses were found to be in good agreement with reference values (NIST, 2004) to within 1.1% for all material compositions. For all of the primary photon energies and medium thicknesses of interest herein, results show that multiple scattering peaks are generally located at energies lower than 100 keV, although for the larger phantom thicknesses it is more difficult to distinguish single, double and multiple scattering in the gamma spectra. Transmitted photon spectra investigated for water, soft tissue, breast, brain and lung tissue slab phantoms are demonstrated to be practically independent of the phantom material, while a significant difference is observed for the spectra transmitted through bone that was proved to be due to the density effect and not material composition.

Gamma and neutron dosimetry of Tehran Research Reactor containment during and after LOCA accident

The radiation effects of loss of coolant accident on the reactor operators are critical issues. In the present study, computational codes were used to investigate gamma and neutron dose rates inside the 5 MW pool-type Tehran Research Reactor containment. The carried out calculations showed, when the remaining coolant over the core is decreasing less than 1 m, the containment should be evacuated. In addition, 24 h after LOCA, the containment gamma dose rates drop to one tenth of their initial values immediately after LOCA.

Equivalent dose calculation in simulation of lung cancer treatment and analysis of dose distribution profile

Currently, lung cancer is one of the most lethal types of cancer (IARC, 2012), the pathology being detected in advanced stage, when the tumor has considerable volume because the disease in most cases asymptomatic in the early stages (INCA, 2016). Dosimetry analysis of healthy organs under real conditions is not feasible. Therefore, computational simulations are used to aid in dose verification in organs of patients submitted to radiotherapy. The goal of this study was to calculate the equivalent dose, due to photons, in the surrounding of healthy organs of patients submitted to radiotherapy for lung cancer, through computational modeling. The simulation was performed using the MCNPX code (MNCPX, 2006), Rex and Regina phantoms (ICRP 110, 2009), radiotherapy room, Siemens Oncor Expression accelerator operating at 6 MV and treatment protocol adopted at the INCA (National Cancer Institute - Brazil). The results obtained, considering the dose due to photons for both phantoms indicate that organs located inside the thoracic cavity received higher dose, being the bronchi, heart and esophagus more affected, due to their anatomical positioning. Clinical data describe the development of bronchiolitis, esophagitis and cardiomyopathies with decreased cardiopulmonary function as one of the major effects of lung cancer treatment. In the Regina phantom, the second largest dose was in the region of the breasts with 615.73 mSv/Gy, while in the Rex the dose was 514.06 mSv/Gy, event related to the difference of anatomical structure of the organ. A qualitative analysis was performed between the dose deposition profile of the treatment planning system and the simulated treatment through the tmesh command and a similar profile of dose distribution was verified throughout the patient's body.

Assessment and estimation of 65Zn production yield via neutron induced reaction on natZnO and natZnONPs

Zinc-65 has been of great interest in medical, biomedical, agricultural, and industrial applications due to its suitable half-life and decay properties. The ⁶⁵Zn was produced via neutron irradiation on natural zinc oxide and natural zinc oxide nanoparticles targets in Tehran Research Reactor (TRR) at a thermal neutron flux of 4.5 × 10¹³ n cm^-2 s^-1 for 30 min. The excitation function of ⁶⁴Zn(n,γ)⁶⁵Zn reaction was calculated via the TALYS-1.8 code. The MCNPX code was used to calculate the thermal neutron distribution. The ⁶⁵Zn theoretical production yield was estimated using calculated cross sections and the calculated thermal neutron distribution. The obtained experimental data and simulated value of production yield for ⁶⁵Zn were in reasonable agreement.

Radiation shielding parameters of some antioxidants using Monte Carlo method

In this paper, radiation shielding parameters such as mass attenuation coefficients and half value layer (HVL) of some antioxidants are investigated using MCNPX (version 2.4.0). The validation of the generated MCNPX simulation geometry for antioxidant structures is provided by comparing the results with standard WinXcom data for radiation mass attenuation coefficients of antioxidants. Very good agreement between WİNXCOM and MCNPX was obtained. The results from the validated geometry were used to calculate the shielding parameters of different antioxidants. The radiation attenuation properties of each antioxidant were compared with each other. The results showed that, on average, the highest and the lowest radiation mass attenuation coefficients were observed on hesperidin and delphinidin chloride, respectively. It can be concluded that Monte Carlo simulation is a strong tool and an alternate method where experimental investigations are not possible and a standard simulation setup can be used in further studies for different biological structures. It can also be concluded that the obtained results from this study are very useful for radiology and radiotherapy applications where antioxidants are frequently used.

Comparison of simulated and experimental values of self-absorption correction factors for a fast and credible adjust in efficiency curve of gamma spectroscopy

Self-absorption correction factors are fundamental in spectroscopy to correct the efficiency of the samples detection whose density is different from the radioactive standard. Mathematical simulations have been widespread as a tool to facilitate the procedure of correction factors calculation. In this paper, LabSOCS was used to calculate the self-absorption correction factor for some geometries and the values found were compared to those obtained in MCNP and experimental values. The percentage deviations found for the self-absorption correction factor calculated by LabSOCS were below 1.6% when compared to experimental values. Deviations were below 1.9% in the curve extrapolation of the experimental procedure found in literature. Results obtained show that the deviations increase proportionally to the difference between the density values of the radioactive standard and the sample. High percentage deviations were also noticed in simulations whose samples had high densities, complex geometries and low energy gamma-rays.

Neutronic design study of accelerator driven system (ADS) for Jordan subcritical reactor as a neutron source for nuclear research

In this paper, a preliminary neutronic design study of an accelerator driven subcritical system for Jordan Subcritical Assembly (JSA) is presented. The conceptual design of coupling the JSA core with proton accelerator and spallation target is investigated, and its feasibility as a neutron source for nuclear research, and possibly for target irradiation and isotope production evaluated. 3D MCNPX model of the JSA reactor, the accelerator beam, and the Pb target was developed, based on actual reactor parameters. MCNPX calculations were carried out to estimate the absolute radial and axial neutron flux in the reactor, and to calculate the multiplication factor K_eff and heat generated in the reactor. Numerical results showed an enormous increase in the neutron flux, by seven orders of magnitude, compared to the current JSA core design using Pu-Be source. In this research the results obtained are discussed and compared with those of the JSA, and do confirm the feasibility of utilizing the JSA as a viable nuclear research facility with adequate neutron flux.

New source models to represent the irradiation process in panoramic gamma irradiator

The use of gamma irradiation technologies generates a number of complex scientific and technical problems; for example, the target is manually loaded onto turntables and is rotated during the entire irradiation process and the MCNPX three-dimensional geometry simulation is kept static. For this, it is necessary to introduce additional approaches. In this paper, two new methodologies are proposed for the simulation of irradiation process in panoramic gamma irradiator. The study was performed at the gamma irradiation facility at the Nuclear Technology Development Center of the National Nuclear Energy Commission, Brazil. The source can be reproduced with a homogenized geometry. Validation of MCNPX calculations of gamma doses were performed by thorough comparison with the experimental measurements. The contribution of this proposed source models has opened new lines of research. The results of this study showed that the proposed source models effectively represent the irradiation process.

Assessment and estimation of 67Cu production yield via deuteron induced reactions on natZn and 70Zn

⁶⁷Cu radioisotope is a beta particle-emitting nuclide used in radioimmunotherapy (RIT) as well as for imaging, tracer kinetic studies and dosimetry. ⁶⁷Cu can be produced by bombarding ^natZn with deuterons. In this study, the physical yields of ⁶⁷Cu via ^natZn(d,x)⁶⁷Cu reaction channel as well as via subreactions of ⁶⁸Zn(d,2pn)⁶⁷Cu, ⁶⁷Zn(d,2p)⁶⁷Cu, ⁷⁰Zn(d,2p3n)⁶⁷Cu, ⁶⁸Zn(d,x)⁶⁷Ni(T_1/2=21s)→⁶⁷Cu and ⁷⁰Zn(d,x)⁶⁷Ni(T_1/2=21s)→⁶⁷Cu in the ^natZn target have been calculated by using the MCNPX-2.6, TALYS-1.8 and SRIM codes. Also, the total cross sections for production of ⁶⁷Cu from ^natZn(d,x)⁶⁷Cu reaction channel in the energy range of 15-45MeV have been estimated by TALYS code. The best reaction to produce ⁶⁷Cu radionuclide in a carrier free form was chosen with deuteron energy around 30MeV on ⁷⁰Zn thick target. Good agreement between the calculated results and the experimental values shows that the employed methods can be used for prediction and production estimation in cyclotron.

Evaluation of the scattered radiations of lead and lead-free aprons in diagnostic radiology by MCNPX

BACKGROUND

Over the past few years, because of high attenuation and lightweight, non-toxic, lead-free aprons (LFAs) have been replaced by lead aprons (LAs). Lots of studies declared that this fact was based on the interactions of diagnostic X-ray with material such as the photoelectric effect (PE) and Compton scattering. These studies have demonstrated that in these types of aprons, due to the presence of different K-edge absorption, PE has a wide absorption in various metals with divers K-edges. The measurement geometry in most of these studies was narrow beam geometry, i.e. a collimated source and a collimated detector with a large source-detector distance.

OBJECTIVE

The present study intended to evaluate the attenuation of radiology scattered radiations in LAs and LFAs in both narrow and broad beam geometries, which is a more realistic situation, in order to check whether or not the higher attenuation is valid.

METHODS

In this study, a lead apron contains (Pb + EPV) and two non-lead compounds of (W + Sn + EPVC) with different weight percent (Wt%) were evaluated in the energy range of diagnostic radiology (100 kVp). The MCNPX code was applied to simulate broad - and narrow-beam measurement geometries. The evaluations have been performed in three situations: 1st) the same density thickness of LA and LFAs 2nd) same line thickness of LA and LFAs 3rd) considering the thickness of LFAs which has the same attenuation with LAs i.e. lead equivalent thickness for LFAs in the narrow beam. Finally, the x-ray transmission ratio (I/I_0) of LAs and LFAs was compared in 100 kVp for three mentioned conditions.

RESULTS

Our results indicated that LFAs had more radiation attenuation rather than LA in the 1st and 2nd conditions with both geometries. However, LFAs had lower attenuation in comparison to LAs in the 3rd condition with broad beam geometry. More importantly, the transmission ratio (I/I_0) of LFAs in the broad beam condition was more significant than LA.

CONCLUSION

The scattered radiations produced by LFAs are more than LAs because of the production of characteristic radiations resulted from K-edge absorption in composited aprons. Consequently, the LFAs should be evaluated in both narrow and broad beam situation using the lead equivalent thickness of LFAs to make sure that the non-lead aprons do not increase the radiation dose of the user.

Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core

This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

Voxel dosimetry: Comparison of MCNPX and DOSXYZnrc Monte Carlo codes in patient specific phantom calculations

INTRODUCTION

Dose evaluation with two Monte Carlo codes using patient specific voxel phantom is presented in this paper. We employ both MCNPX and DOSXYZnrc to perform dosimetry for mathematical voxel phantoms generated by our in-house developed voxel phantom generator and EGSnrc/CTCreate respectively.

MATERIAL AND METHOD

Our case study was a 2.5 × 2.4 × 2.4 cm3 tumor in the middle lobe of right lung of a male patient exposed to 6MV parallel beam. In order to compare these Monte Carlo codes with together gross tumor volume (GTV) and organ at risks (OAR) doses and dose volume histograms (DVH) were calculated.

RESULTS

Comparing the mean absorbed dose results (in Gy) from both codes indicates that gross tumor volume, heart and spinal cord have 2% to 10% difference. The 10% difference between the codes were from the spinal cord region where was not in the therapy beam and it just received the scatter radiation. The dose volume DVH obtained from DOSXYZnrc results demonstrate a milder slope compared with MCNPX DVHs.

CONCLUSION

It was revealed that MCNPX has some advantages in comparison to DOSXYZnrc, but it is important to consider that for equal precision in voxel dosimetry calculation, DOSXYZnrc runs faster than MCNPX and it is a great advantage.

Corrigendum to "Monte Carlo simulations of the secondary neutron ambient and effective dose equivalent rates from surface to suborbital altitudes and low Earth orbit"

A recent paper published in Life Sciences in Space Research (El-Jaby and Richardson, 2015) presented estimates of the secondary neutron ambient and effective dose equivalent rates, in air, from surface altitudes up to suborbital altitudes and low Earth orbit. These estimates were based on MCNPX (LANL, 2011) (Monte Carlo N-Particle eXtended) radiation transport simulations of galactic cosmic radiation passing through Earth's atmosphere. During a recent review of the input decks used for these simulations, a systematic error was discovered that is addressed here. After reassessment, the neutron ambient and effective dose equivalent rates estimated are found to be 10 to 15% different, though, the essence of the conclusions drawn remains unchanged.

Design and fabrication of an in situ gamma radioactivity measurement system for marine environment and its calibration with Monte Carlo method

Simulation, design and fabrication of a sealing enclosure is carried out for a NaI(Tl) 2″×2″ detector, to be used as in situ gamma radioactivity measurement system in marine environment. Effect of sealing enclosure on performance of the system in laboratory and marine environment (distinct tank with 10m(3) volume) were studied using point sources. The marine volumetric efficiency for radiation with 1461keV energy (from (40)K) is measured with KCl volumetric liquid source diluted in distinct tank. The experimental and simulated efficiency values agreed well. Marine volumetric efficiency calibration curve is calculated for 60keV to 1461keV energy with Monte Carlo method. This curve indicates that efficiency increasing rapidly up to 140.5keV but then drops exponentially.

Design of an explosive detection system using Monte Carlo method

Regardless the motivation terrorism is the most important risk for the national security in many countries. Attacks with explosives are the most common method used by terrorists. Therefore several procedures to detect explosives are utilized; among these methods are the use of neutrons and photons. In this study the Monte Carlo method an explosive detection system using a ²⁴¹AmBe neutron source was designed. In the design light water, paraffin, polyethylene, and graphite were used as moderators. In the work the explosive RDX was used and the induced gamma rays due to neutron capture in the explosive was estimated using NaI(Tl) and HPGe detectors. When light water is used as moderator and HPGe as the detector the system has the best performance allowing distinguishing between the explosive and urea. For the final design the Ambient dose equivalent for neutrons and photons were estimated along the radial and axial axis.

Study of a 10B+ZnS(Ag) neutron detector as an alternative to 3He-based detectors in Homeland Security

The response of a scintillation neutron detector of ZnS(Ag) with ¹⁰B was calculated, using the MCNPX Monte Carlo Code. The detector consists of four panels of polymethyl methacrylate (PMMA) and five thin layers of ~0.017cm thick ¹⁰B+ZnS(Ag) in contact with the PMMA. The response was calculated for the bare detector and with different thicknesses of High Density Polyethylene, HDPE, moderator for 29 monoenergetic sources as well as ²⁴¹AmBe and ²⁵²Cf neutrons sources. In these calculations the reaction rate ¹⁰B(n, α)⁷Li and the neutron fluence in the sensitive area of the detector ¹⁰B+ZnS(Ag) was estimated. Measurements were made at the Neutron Measurements Laboratory, Universidad Politécnica de Madrid, LMN-UPM, to quantify the detections in counts per second in response to a ²⁵²Cf neutron source separated 200cm. The MCNPX computations were compared with measurements to estimate the efficiency of ZnS(Ag) for detecting the α that is created in the ¹⁰B(n, α)⁷Li reaction. After validating new models with different geometries it will be possible to improve the detector response trying to achieve a sensitivity of 2.5cps-ng²⁵²Cf comparable with the response requirements for ³He detectors installed in the Radiation Portal Monitors, RPMs. This type of detector can be considered an alternative to the ³He detectors for detection of Special Nuclear Material, SNM.