Neutron pumping of active medium formed by gadolinium isotopes 155Gd and 156Gd pair: A feasibility study.

The possibility of creating technical means for controlling the processes of accumulation and conversion of the energies of thermal and epithermal neutrons into the energy of monoenergetic photons due to neutron pumping of an active medium consisting of nuclei with long-lived isomeric states was studied in this work. The system under study consisted of an external pulse-periodic source of deuterium-tritium neutrons (PSN) and a subcritical blanket, which included a variable neutron-collimation beam-shaping assembly (vBSA) and an active medium. The vBSA was composed of moderating blocks and selective plates designed to trap and shape a pulsed neutron flux with subsequent conversion of a millisecond signature into monoenergetic photon emission. Gadolinium oxide enriched in 155Gd isotope was used as the active medium, where the heavier one could be at different excited states, the de-excitations of which were accompanied by photon emission. In this research, the possibility of using the conjugate system (i.e., blanket - PSN - vBSA) for converting excess neutron energy accumulated in the inverse state of 156Gd nuclei into photon emission was demonstrated in detail.

Thermal neutron beam optimization for PGNAA applications using Q-learning algorithm and neural network.

As a powerful, non-destructive analysis tool based on thermal neutron capture reaction, prompt gamma neutron activation analysis (PGNAA) indeed requires the appropriate neutron source. Neutrons produced by electron Linac-based neutron sources should be thermalized to be appropriate for PGNAA. As a result, thermalization devices (TDs) are used for the usual fast neutron beam to simultaneously maximize the thermal neutron flux and minimize the non- thermal neutron flux at the beam port of TD. To achieve the desired thermal neutron flux, the optimized geometry of TD including the proper materials for moderators and collimator, as well as the optimized dimensions are required. In this context, TD optimization using only Monte Carlo approaches such as MCNP is a multi-parameter problem and time-consuming task. In this work, multilayer perceptron (MLP) neural network has been applied in combination with Q-learning algorithm to optimize the geometry of TD containing collimator and two moderators. Using MLP, both thickness and diameter of the collimator at the beam port of TD have first been optimized for different input electron energies of Linac as well as for moderators' thickness values and the collimator. Then, the MLP has been learned by the thermal and non-thermal neutron flux simultaneously at the beam port of TD calculated by MCNPX2.6 code. After selecting the optimized geometry of the collimator, a combination of Q-learning algorithm and MLP artificial neural network have been used to find the optimal moderators' thickness for different input electron energies of Linac. Results verify that the final optimum setup can be obtained based on the prepared dataset in a considerably smaller number of simulations compared to conventional calculation methods as implemented in MCNP.

Different applications of telemedicine - assessing the challenges, barriers, and opportunities- a narrative review.

Telemedicine (TLM) is a technique of telecommunication used to create, promote, or accelerate health services. Because of its societal significance, the current study attempted to demonstrate its essential applications in the health sector and the challenges, obstacles, and opportunities that lie ahead. Various studies and reports were received based on the subject of the current study, first using MeSH terms related to the subject in authentic and available international databases. After that, 30-related articles were selected based on the study criteria, and then the required results were extracted from the selected studies. The study results showed that TLM has a significant role in more than 13 major areas of health and treatment, and in most of these areas, it has made the relevant affairs easier for both patients and medical staff. Although TLM has many advantages, it still has obstacles and challenges requiring further studies to manage this technology better. Given the high importance of the TLM in the health sector in most countries worldwide, efforts are needed to promote this technology and remove the obstacles in front of it. Therefore, further evaluations of TLM efficiency in terms of economics, speed of action, effectiveness, and the provision of infrastructure are necessary to overcome the obstacles highlighted based on the results of these studies and improve the efficiency of using this technology.

A multi-moderator neutron spectrometer for use in BNCT studies of the Tehran research reactor.

The Tehran Research Reactor is the only appropriate and available neutron source in Iran for clinical boron neutron capture therapy (BNCT). One of the requirements for BNCT is to carefully evaluate and measure the therapeutic neutron beam (epithermal neutrons) as well as the fast and thermal neutron components for successful treatment. In this research, a multi-moderator neutron spectrometer (MMNS) with LiI(Eu) scintillator as neutron counter was proposed for these measurements in the range of 10-11 eV to 15 MeV. The results confirmed promising precision of the designed MMNS for the epithermal spectrum; however, the angular dependency of the therapeutic beam due to any probable change in the beam-shaping assembly should be considered.

A feasibility study of gamma ray source finder development for multiple sources scenario based on a Monte Carlo simulation.

In this research, a rapid and accurate gamma finder system which can find orphan gamma-ray sources has been designed based on NaI(Tl) detector. By proposing complementary methods to our previous researches, an attempt has been made to provide an approach to solve the problem of 3D localization in multiple orphan gamma-ray sources scenario. Based on our previous research, the new conceptual design has been proposed and simulated using Monte Carlo MCNPX2.6 code. Afterward, in order to identify its key characteristics and features, the proposed design has been tested in several different scenarios (multiple gamma sources with various activities in different distances).

Nedis-Serpent simulation of a neutron source assembly with complex internal heterogeneous structure.

The hybrid use of Nedis-2m and Serpent 2.1.30 codes to predict the radiation characteristics (i.e., neutron yield and energy spectrum) of an Am-Be source with a fine-grained mixture of americium dioxide (AmO2) and beryllium (Be) core was studied with a focus on the grain size influence on the simulation results. The study showed that the fine-grained structure of the source core would decrease the number of alpha particles participating in the nuclear reactions with 17,18O and 9Be nuclei, which softened the neutron energy spectrum and reduced the neutron yield. The simulations also confirmed that the source core made of the stable crystals of AmBe13 intermetallic alloy would improve the neutron yield to maximum 50% compared to the core made of AmO2. Moreover, a source with a variable neutron yield was proposed with a heterogeneous core of AmO2 rods embedded in Be. The neutron energy spectrum of heterogeneous source resembled the energy spectrum of Deuterium-Tritium (D-T) neutrons which were generated in a long magnetic trap with high-temperature plasma. The subcritical irradiation facility assembled from the nth number of heterogeneous Am-Be source can be used to study the properties of materials and the equipment operating in the epithermal and fast neutron spectra. The use of a heterogeneous Am-Be assembly, as a basic element of an irradiation installation, simplifies the handling and operation procedures because it is easily disabled by removing the Be layer, or by inserting a sheet of the appropriate size and material between the Be and Am rod.

A Monte Carlo study of neutron contamination in presence of circular cones during stereotactic radiotherapy with 18 MV photon beams.

High-energy photons are being used to treat different kinds of cancer, but it may increase the rate of secondary cancers due to the neutron contamination as well as over exposing of patients and medical staffs in radiation therapy Takam, Bezak, Marcu, and Yeoh, 2011, Radiation Research, 176, 508-520. Due to some difficulties in experimental measurements of neutron contamination, Monte Carlo method is an efficient tool to investigate dose parameters and characteristics in new techniques. The 18-MV photon beam of linac and circular cones have been simulated by MCNP5 code. Various parameters of photon and neutron including mean energy, flux, KERMA, the number of particles crossing a surface at a distance of 100 cm (SSD = 100 cm) as well as the change in photon and neutron spectrum as well as in intensity through the transmission in the circular collimators have been investigated. The results of this study show that the use of a circular collimator decreases neutron dose in the central axis, which is an advantage, but neutron contamination inducing small neutron dose is distributed all over the space. On the surface of phantom, photon dose rate is approximately equal to 3.41E7 (mGy/mA.min) for different collimators, but the neutron dose rate is 1.64E2 (mGy/ mA.min), 2.03E2 (mGy/ mA.min) and 2.52E2 (mGy/mA.min) for diameters of 12, 20 and 40 mm, respectively and it decreases by decreasing the diameter of the collimator. The neutron dose rate decreases from 9.68E7 and 9.74E7 (mGy/min.mA) for open field size 33 cm2 and 55 cm2 to 1.64E2 (mGy/min.mA), 2.02E2 (mGy/min.mA) and 2.52E2 (mGy/min.mA) for collimator diameter of 12 mm, 20 mm and 40 mm. It can be concluded that the use of circular collimators has an advantage of reducing neutron dose in the central axis. It should be mentioned that the off-axis neutron dose surrounding the collimator can be eliminated using an external neutron shield without perturbing the treatment field.

Evaluation of boron neutron capture therapy in-phantom parameters by response matrix method.

AIM: Determination of boron neutron capture therapy in-phantom parameters by response matrix (RM) method. MATERIALS AND METHODS: In this study, various in-phantom figures-of-merit including therapeutic gain, advantage depth dose rate, advantage depth, therapeutic depth, treatment time, skin dose rate, and skull dose rate have been analyzed using the RM method. This method is based on the division of neutron/gamma spectrum and calculation of various dose components of each energy group. Summation of these dose responses is equal to the total dose of the whole spectrum. Based on this method, in-phantom parameters could be calculated by a computer program in a very short time. RESULTS: There is a good agreement between direct calculation and RM method. The maximum allowable contaminations of the thermal and fast neutrons in a neutron beam have been calculated by RM method. It was found that these values are 17.4% and 2.6%, for thermal and fast neutron, respectively. CONCLUSION: The results confirm that the RM method is a fast method to evaluate in-phantom parameters without repeating simulations due to change in neutron spectrum and treatment conditions.

An approach to design a 90Sr radioisotope thermoelectric generator using analytical and Monte Carlo methods with ANSYS, COMSOL, and MCNP.

In this study, a 90Sr radioisotope thermoelectric generator (RTG) with power of milliWatt was designed to operate in the determined temperature (300-312K). For this purpose, the combination of analytical and Monte Carlo methods with ANSYS and COMSOL software as well as the MCNP code was used. This designed RTG contains 90Sr as a radioisotope heat source (RHS) and 127 coupled thermoelectric modules (TEMs) based on bismuth telluride. Kapton (2.45mm in thickness) and Cryotherm sheets (0.78mm in thickness) were selected as the thermal insulators of the RHS, as well as a stainless steel container was used as a generator chamber. The initial design of the RHS geometry was performed according to the amount of radioactive material (strontium titanate) as well as the heat transfer calculations and mechanical strength considerations. According to the Monte Carlo simulation performed by the MCNP code, approximately 0.35 kCi of 90Sr is sufficient to generate heat power in the RHS. To determine the optimal design of the RTG, the distribution of temperature as well as the dissipated heat and input power to the module were calculated in different parts of the generator using the ANSYS software. Output voltage according to temperature distribution on TEM was calculated using COMSOL. Optimization of the dimension of the RHS and heat insulator was performed to adapt the average temperature of the hot plate of TEM to the determined hot temperature value. This designed RTG generates 8mW in power with an efficiency of 1%. This proposed approach of combination method can be used for the precise design of various types of RTGs.

Investigation of the gold nanoparticles effects on the prostate dose distribution in brachytherapy: gel dosimetry and Monte Carlo method.

PURPOSE: In this work, gold nanoparticles (GNPs) were embedded in the MAGIC-f polymer gel irradiated with the 192Ir brachytherapy sources. MATERIAL AND METHODS: At the first plexiglas phantom was made as the human pelvis. The GNPs were synthesized with 15 nm in diameter and 0.1 mM (0.0197 mg/ml) in concentration by using a chemical reduction method. Then, the MAGIC-f gel was synthesized. The fabricated gel was poured into the tubes located at the prostate (with and without the GNPs) locations of the phantom. The phantom was irradiated with 192Ir brachytherapy sources for prostate cancer. After 24 hours, the irradiated gels was read by using Siemens 1.5 Tesla MRI scanner. Following the brachytherapy practices, the absolute doses at the reference points and isodose curves were extracted and compared by experimental measurements and Monte Carlo (MC) simulations. RESULTS: The mean absorbed doses in the presence of the GNPs in prostate were 14% higher than the corresponding values without the GNPs in the brachytherapy. The gamma index analysis (between gel and MC) using 7%/7 mm was also applied to the data and a high pass rate achieved (91.7% and 86.4% for analysis with/without GNPs, respectively). CONCLUSIONS: The real three-dimensional analysis shows the comparison of the dose-volume histograms measured for planning volumes and the expected one from the MC calculation. The results indicate that the polymer gel dosimetry method, which developed and used in this study, could be recommended as a reliable method for investigating the dose enhancement factor of GNPs in brachytherapy.

Design and simulation of betavoltaic angle sensor Based on 6³Ni-Si.

A theoretical design and simulation of betavoltaic angle sensor (beta-AS) based on (63)Ni-Si using MCNP code is presented in this article. It can measure the full angle of 0-360° in the temperature range of 233-353 K. Beta-AS is composed of semicircular (63)Ni as the beta source, which rotates along the circular (four-quadrant) surface of Si as a semiconductor (in p-n structure), so that the change in the source angle in relation to Si surface can be measured based on the changes in V(oc) observed in each quadrant of Si. For better performance, characteristics of Si and (63)Ni have been optimized: N(D) and N(A) values of 8e19 and 4e18 cm(-3) (donor and acceptor doping concentration in Si, respectively), source thickness and activity of 1.5 µm and 18 mCi, respectively. The relation between angle and V(oc) is also investigated. The maximum difference between measured and real values of angle (the worst case, i.e., 0.18° for the angle of 45°) occurs at 233 K. It has been shown that sensitivity of the sensor decreases with an increase of angle. The results also show that the change in activity does not affect the sensitivity.

Gold nanoparticle-based brachytherapy enhancement in choroidal melanoma using a full Monte Carlo model of the human eye.

The effects of gold nanoparticles (GNPs) in 125I brachytherapy dose enhancement on choroidal melanoma are examined using the Monte Carlo simulation technique. Usually, Monte Carlo ophthalmic brachytherapy dosimetry is performed in a water phantom. However, here, the compositions of human eye have been considered instead of water. Both human eye and water phantoms have been simulated with MCNP5 code. These simulations were performed for a fully loaded 16 mm COMS eye plaque containing 13 125I seeds. The dose delivered to the tumor and normal tissues have been calculated in both phantoms with and without GNPs. Normally, the radiation therapy of cancer patients is designed to deliver a required dose to the tumor while sparing the surrounding normal tissues. However, as the normal and cancerous cells absorbed dose in an almost identical fashion, the normal tissue absorbed radiation dose during the treatment time. The use of GNPs in combination with radiotherapy in the treatment of tumor decreases the absorbed dose by normal tissues. The results indicate that the dose to the tumor in an eyeball implanted with COMS plaque increases with increasing GNPs concentration inside the target. Therefore, the required irradiation time for the tumors in the eye is decreased by adding the GNPs prior to treatment. As a result, the dose to normal tissues decreases when the irradiation time is reduced. Furthermore, a comparison between the simulated data in an eye phantom made of water and eye phantom made of human eye composition, in the presence of GNPs, shows the significance of utilizing the composition of eye in ophthalmic brachytherapy dosimetry Also, defining the eye composition instead of water leads to more accurate calculations of GNPs radiation effects in ophthalmic brachytherapy dosimetry.

Design of photon converter and photoneutron target for High power electron accelerator based BNCT.

An electron accelerator, ILU-14, with current of 10 mA and 100 kW in power has been considered as one of the options for neutron source in Boron Neutron Capture Therapy (BNCT). The final design of neutron target has been obtained using MCNPX to optimize the neutron production. Tungsten in strip shape and D2O in cylindrical form have been proposed as the photon converter and the photoneutron target, respectively. In addition calculation of heat deposition in the photon target design has been considered to ensure mechanical stability of target. The results show that about 8.37×10(12) photoneutron/s with average energy of 615 keV can be produced by this neutron source design. In addition, using an appropriate beam shaping assembly an epithermal neutron flux of the order of 1.24×10(8) cm(-2) s(-1) can be obtained for BNCT applications.

Radiation shielding design of BNCT treatment room for D-T neutron source.

Recent studies have shown that D-T neutron generator can be used as a proper neutron source for Boron Neutron Capture Therapy (BNCT) of deep-seated brain tumors. In this paper, radiation shielding calculations have been conducted based on the computational method for designing a BNCT treatment room for a recent proposed D-T neutron source. By using the MCNP-4C code, the geometry of the treatment room has been designed and optimized in such a way that the equivalent dose rate out of the treatment room to be less than 0.5µSv/h for uncontrolled areas. The treatment room contains walls, monitoring window, maze and entrance door. According to the radiation protection viewpoint, dose rate results of out of the proposed room showed that using D-T neutron source for BNCT is safe.

Design and construction of a thermal neutron beam for BNCT at Tehran Research Reactor.

An irradiation facility has been designed and constructed at Tehran Research Reactor (TRR) for the treatment of shallow tumors using Boron Neutron Capture Therapy (BNCT). TRR has a thermal column which is about 3m in length with a wide square cross section of 1.2×1.2m(2). This facility is filled with removable graphite blocks. The aim of this work is to perform the necessary modifications in the thermal column structure to meet thermal BNCT beam criteria recommended by International Atomic Energy Agency. The main modifications consist of rearranging graphite blocks and reducing the gamma dose rate at the beam exit. Activation foils and TLD700 dosimeter have been used to measure in-air characteristics of the neutron beam. According to the measurements, a thermal flux is 5.6×10(8) (ncm(-2)s(-1)), a cadmium ratio is 186 for gold foils and a gamma dose rate is 0.57Gy h(-1).

Neutron capture therapy: a comparison between dose enhancement of various agents, nanoparticles and chemotherapy drugs.

The aim of this study is to compare dose enhancement of various agents, nanoparticles and chemotherapy drugs for neutron capture therapy. A (252)Cf source was simulated to obtain its dosimetric parameters, including air kerma strength, dose rate constant, radial dose function and total dose rates. These results were compared with previously published data. Using (252)Cf as a neutron source, the in-tumour dose enhancements in the presence of atomic (10)B, (157)Gd and (33)S agents; (10)B, (157)Gd, (33)S nanoparticles; and Bortezomib and Amifostine chemotherapy drugs were calculated and compared in neutron capture therapy. Monte Carlo code MCNPX was used for simulation of the (252)Cf source, a soft tissue phantom, and a tumour containing each capture agent. Dose enhancement for 100, 200 and 500 ppm of the mentioned media was calculated. Calculated dosimetric parameters of the (252)Cf source were in agreement with previously published values. In comparison to other agents, maximum dose enhancement factor was obtained for 500 ppm of atomic (10)B agent and (10)B nanoparticles, equal to 1.06 and 1.08, respectively. Additionally, Bortezomib showed a considerable dose enhancement level. From a dose enhancement point of view, media containing (10)B are the best agents in neutron capture therapy. Bortezomib is a chemotherapy drug containing boron and can be proposed as an agent in boron neutron capture therapy. However, it should be noted that other physical, chemical and medical criteria should be considered in comparing the mentioned agents before their clinical use in neutron capture therapy.

A feasibility study of the Tehran research reactor as a neutron source for BNCT.

Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10(9) and 1.21×10(9) n/cm(2)/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure.

Design and optimization of beta-cell temperature sensor based on 6³Ni-Si.

A theoretical study of the use of a beta-cell as a temperature sensor using MCNP4C Monte Carlo code is presented in this paper. Nickel-63 and silicon were selected as the beta source and semiconductor material, respectively. The maximum open-circuit voltage (VOC) is equal to 0.445 V with doping concentrations of NA=4 × 10(18)#/cm(3) and ND=8 × 10(19)#/cm(3) in the base and the emitter region, respectively, which, depending on the source activity, enables measurement in a wide range of temperature. The effects of the activity and its change over time on VOC were also studied. The results demonstrated that VOC exhibited smaller changes for higher activities. The temperature sensitivity of this sensor ranges from -2.42 mV/K to -3.41 mV/K for source activities from 100 mCi to 0.001 mCi, respectively, so the optimal activity can be determined according to the desired temperature range and sensitivity.

Involvement of four different intracellular sites in chloroacetaldehyde- induced oxidative stress cytotoxicity.

Chloroacetaldehyde (CAA) is a chlorination by-product in finished drinking water and a toxic metabolite of a wide variety of industrial chemicals (e.g. vinyl chloride) and chemotherapeutic agents (e.g. cyclophosphamide and ifosfamide). In this research, the cytotoxic mechanisms of CAA in freshly isolated rat hepatocytes were investigated.CAA cytotoxicity was associated with reactive oxygen species (ROS) formation and glutathione depletion suggesting that oxidative stress contributed to the CAA cytotoxic mechanism. CAA-induced oxidative stress cytotoxicity markers were significantly prevented by antioxidants, ROS scavengers, mitochondrial permeability transition (MPT) pore sealing agents, endocytosis inhibitors, ATP generators and xanthine oxidase inhibitor. In our study the hepatocyte mitochondrial membrane potential was rapidly decreased by CAA which was prevented by antioxidants and ROS scavenger indicating that mitochondrial membrane damage was a consequence of ROS formation. CAA cytotoxicity was also associated with lysosomal membrane rupture. OUR FINDINGS SHOWED THAT AT LEAST FOUR DIFFERENT INTRACELLULAR SOURCES INCLUDING: metabolic enzymes cytochrome P450 and xanthine oxidase, mitochondrial respiratory chain disruption and lysosomal Haber-weiss reaction, were involved in CAA induced ROS formation and other subsequent cytotoxic events. Our other interesting finding was that the lysosomotropic agents prevented CAA induced mitochondrial membrane potential collapse and mitochondrial MPT pore sealing agents inhibited lysosomal membrane damage caused by CAA. It can therefore be suggested that there is probably a toxic interaction (cross-talk) between mitochondrial and lysosomal oxidative stress generating systems, which potentiates each organelle damage and ROS formation in CAA- induced hepatotoxicity.

Dose calculation and in-phantom measurement in BNCT using response matrix method.

In-phantom measurement of physical dose distribution is very important for Boron Neutron Capture Therapy (BNCT) planning validation. If any changes take place in therapeutic neutron beam due to the beam shaping assembly (BSA) change, the dose will be changed so another group of simulations should be carried out for dose calculation. To avoid this time consuming procedure and speed up the dose calculation to help patients not wait for a long time, response matrix method was used. This procedure was performed for neutron beam of the optimized BSA as a reference beam. These calculations were carried out using the MCNPX, Monte Carlo code. The calculated beam parameters were measured for a SNYDER head phantom placed 10 cm away from beam the exit of the BSA. The head phantom can be assumed as a linear system and neutron beam and dose distribution can be assumed as an input and a response of this system (head phantom), respectively. Neutron spectrum energy was digitized into 27 groups. Dose response of each group was calculated. Summation of these dose responses is equal to a total dose of the whole neutron/gamma spectrum. Response matrix is the double dimension matrix (energy/dose) in which each parameter represents a depth-dose resulted from specific energy. If the spectrum is changed, response of each energy group may be differed. By considering response matrix and energy vector, dose response can be calculated. This method was tested for some BSA, and calculations show statistical errors less than 10%.