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Gamma radiation indicators are appropriate tools for monitoring visually whether or not the irradiation process has been carried out properly. Among chemical radiation indicators available worldwide, a few are suitable for monitoring low... more
Gamma radiation indicators are appropriate tools for monitoring visually whether or not the irradiation process has been carried out properly. Among chemical radiation indicators available worldwide, a few are suitable for monitoring low dose ranges (especially for blood irradiation, below 50 Gy). Addressing this scope, PVA-Fricke gel was proposed in this work. Irradiation of the prepared PVA-Fricke gel samples was performed by Co-60 gamma cell unit up to a dose of 80 Gy. Color change of the samples was observed from orange to purple proportional to increasing absorbed dose. Prepared samples were divided into three groups, kept at different environmental conditions, to investigate stability of the gel against temperature and light. Results revealed that the irradiated samples kept at dark and refrigerator were stable for seven days. Optical absorbance measurement of the samples also estimated pre- and post-irradiation color stability. The gel can be easily used to identify processed and unprocessed products in blood irradiation. Although the gel is designed to be a qualitative indicator, it is also a good quantitative dosimeter for gamma rays.
In this research, the governing dynamic equations of the Bushehr NPP core are studied and modeled using Matlab (Simulink) software. The point kinetic equation with the temperature feedbacks and the fuel-coolant energy balance equations in... more
In this research, the governing dynamic equations of the Bushehr NPP core are studied and modeled using Matlab (Simulink) software. The point kinetic equation with the temperature feedbacks and the fuel-coolant energy balance equations in the time domain were used for this purpose. The model is validated against the rod drop accident data available in BNPP-1 FSAR, and they agreed. Then, this time-domain model is used to find the maximum movement speed of the control rods. For this goal, linear and non-linear rod movement equations have been modeled. In this regard, the maximum withdrawal speed of the working bank (H10) with a worth of 1.1 dollars has been investigated. Using the linear CR model, a speed limit of 9 cm.s-1 has been obtained to prevent the initiation of a reactor trip. The maximum speed using the non-linear model of the CR was found out to be dependent on its initial position. Thus, in three positions of the H10 bank: 100%, 80%, and 50% of the length inside the reactor, the maximum withdrawal speed values were valuated 11.5, 7.7, and 4.4 cm.s-1 respectively. According to the results, among the reactor parameters including power, period, and fuel temperature, which are monitored by the reactor protection system to initiate the reactor trip, the reactor power is the limiting factor for specifying the maximum withdrawal speed. This study is performed using time domain analysis, and the obtained results are consistent with the results reported in the previous research using Laplace transform approach.
‎Fricke gel dosimeters obtained by modifications on standard Fricke dosimeter presents some advantages like easy preparation‎, ‎tissue equivalence‎, ‎good reproducibility and dose mapping‎. ‎In this work‎, ‎dose response characteristics... more
‎Fricke gel dosimeters obtained by modifications on standard Fricke dosimeter presents some advantages like easy preparation‎, ‎tissue equivalence‎, ‎good reproducibility and dose mapping‎. ‎In this work‎, ‎dose response characteristics of Gelatin Fricke gel dosimeters was investigated and compared with Fricke agarose gel dosimeters in terms of sesitivity‎. ‎After prepration of three different formulation of Gelatin Fricke gel dosimeters and gamma irradiation of the samples‎, ‎a spectrophotometer was applied to measure the optical absorbance of the samples‎. ‎Results indicate a linear dose range response of 10 to 30 Gy‎, ‎as well as increased gelatin concentrations cause the sensitivity of the dosimeter to detereorate with a 80% reduction of dose response for a change in gelatin concentration from 3 to 8 weight percent‎. ‎Obtained coefficient variation verifies the good repeatability of the gel response‎. ‎The gel dosimeter has no dose rate dependence‎. ‎Comparison of the most sensitive Gelatin Fricke gel sample with the prepared Fricke agarose gel samples confirm that Fricke agarose dosimeter is more sensitive than Gelatin Fricke gel dosimeter‎.
Today, with the development of nuclear technology and radiation therapy equipment, radiation protection is important. This study aimed to design heavy concrete with high compressive strength and effective protection against neutron and... more
Today, with the development of nuclear technology and radiation therapy equipment, radiation protection is important. This study aimed to design heavy concrete with high compressive strength and effective protection against neutron and gamma rays. In this study, 11 types of concrete with different mixing designs including 88 samples were made. In these samples, iron ore aggregates galena, limonite, hematite, polypropylene fibers, nanoparticles, micro-particles of silicon, and B4C powder have been used. Concrete quality coefficient, compressive strength, gamma, and neutron attenuation coefficients were measured for all samples. Also, the neutron attenuation coefficient for all samples was calculated using the Monte Carlo simulation (MCNPX) code and compared with the experimental values. The density, neutron attenuation coefficient, and compressive strength of concrete samples varied from 2.37 to 3.17 g.cm-3, from 0.0162 to 0.0306 cm2.g-1, and from 48.0 to 81.3 MPa respectively. The linear gamma attenuation coefficient and gamma-ray tenth value layer (TVL) were obtained from 0.148 to 0.398 cm-1 and 15.74 to 5.85 cm respectively. These results showed that the highest neutron and gamma attenuation coefficients were obtained for concrete containing 70% galena iron ore and 20% boron carbide and the highest compressive strength belonged to sample G15 containing 15% galena iron ore and 1.8% boron carbide. G70 was the best concrete regarding the quality factor, defined as the product of multiplying the compressive strength and linear attenuation coefficients of neutron and gamma.
Nuclear-pumped lasers (NPL) are lasers that excite the active laser environment caused by nuclear reaction. Such lasers need ionizing radiation shielding for mixed neutron and gamma fields. In this work, a shielding system for NPL was... more
Nuclear-pumped lasers (NPL) are lasers that excite the active laser environment caused by nuclear reaction. Such lasers need ionizing radiation shielding for mixed neutron and gamma fields. In this work, a shielding system for NPL was designed which using 10B(n,α)7Li. In this simulation, we have used MCNPX 2.6.0 Monte Carlo code and the thermal neutron flux as 1×1016 n.cm-2.s-1 for excitation reaction. Such a large neutron flux can be obtained from a reactor source or a heavy ion accelerator. For this work, 10B fuel is covered on the surface of a rectangular cube aluminum shell by using the Monte Carlo method. In the design of the shielding, combinations with different materials have been used with various arrangements in three layers. According to the simulation, the arrangement of Fe2-B-BPE-Pb is a suitable protection compound for such lasers.
Nowadays, a very particular type of nuclear reactors has become fascinating not only for most nuclear communities but also for the prominent energy suppliers to fix the global warming effects worldwide. They are Small Modular Reactors... more
Nowadays, a very particular type of nuclear reactors has become fascinating not only for most nuclear communities but also for the prominent energy suppliers to fix the global warming effects worldwide. They are Small Modular Reactors called SMRs. Usually, SMRs can are classified according to the seven different categories. They include PWRs (especially iPWRs), BWRs, PHWRs, GCR, LMFBR, MSR, and MMRs. Although many different plans have been proposed worldwide, only a few well-established or successive developing action plans are among many innovative conceptual designs. This paper briefly presents a comparison study reviewing the last advances and challenges. The proposed roadmap is strongly correlated and depends on the technology readiness and documentation, technology availability, safety and reliability, design, and construction feasibility for different countries. A new graded approach Phenomenological Identification Ranking Table (PIRT) has been developed and proposed to choose the most profitable and compatible action plan dependent on the situation. Finally, the best feasible designs are compared and proposed against the lack of First-of-A-Kind (FOAK). Furthermore, different options are proposed for different priorities and preferences based on the available nuclear infrastructures. Studies are very profitable to save money and time and develop a strategic action plan for newcomers and developing countries. On the other hand, some exceptional designs have extraordinary advantages for industrial countries and even more for the future of nuclear energy worldwide. Therefore, the proposed roadmap covers short-term, mid-term, and long-term strategies for developing countries and newcomers in the nuclear reactor industry.
Using the experimental data in nuclear computing to verify the calculation methods and tools based on numerical and statistical methods has many benefits such as illustrating the quality, ensuring the capabilities, and computer codes... more
Using the experimental data in nuclear computing to verify the calculation methods and tools based on numerical and statistical methods has many benefits such as illustrating the quality, ensuring the capabilities, and computer codes validating. Simulation by computer tools is also applicable in the safety analysis of research reactors. In this research, the computer tool (MCNPX 2.7.0: 2011) was verified against the experimental data of neutron flux and spectrum on the sample position of the Tehran Research Reactor (TRR) neutron imaging system by the neutron activation method. To determine the benchmark specifications, the simulation of the system was done at the first step by considering a well-defined facility geometric, material specification and reactor core configuration, fuel elements, and radiation facility (beam tubes and collimator, reactor core, and neutron imaging components). Then the flux and neutron spectrum at the sample position were calculated. In the second step, a set of In (bare and covered by cd) and Au foils and a set of Au, Ni, Ti, and Zr, were placed and exposed almost in front of the reactor E beam tube. The neutron energy spectrum was unfolded by calculating the saturation activity of each foil by SAND-II code, and the neutron flux was calculated. A comparison of the results obtained in two steps shows a relatively good and acceptable agreement (Max. 30% deviation) between the flux and the shape of the flux profile obtained from calculations and experimental data.
In many human diseases and health cases, therapy of blood transfusion becomes necessary. In spite of the necessity, there are some risks associated with blood used in blood transfusion process. The TA-GVHD (transfusion-associated... more
In many human diseases and health cases, therapy of blood transfusion becomes necessary. In spite of the necessity, there are some risks associated with blood used in blood transfusion process. The TA-GVHD (transfusion-associated graft-versus-host-disease) is a problem when a blood transfusion occurs. The blood irradiation with gamma rays in blood bags can eliminate this risk. It should be mentioned that Co-60 sources are widely used for such blood irradiators. The present work investigates Co-60 production yield inside the external irradiation boxes of Tehran Research Reactor (TRR) using MCNPX code. 10-rod and 4-rod Co-59 assemblies were modeled at different external irradiation boxes to investigate their negative reactivity impact on TRR core as well Co-60 buildup rate during 3 years operation of the nuclear core at 4 MW power. The obtained results from MCNPX code showed a 4-rod assembly in linear form could obtain the highest specific activity (Ci.g-1) inside the external irradiation box faced to the core center. The computational results showed about 8 kCi of Co-60 is produced at the optimized irradiation position after 3 years TRR operation at 4 MW power.
By expanding the applications of GEM detectors, a newer pattern of such detectors was introduced in 2004, named THGEM detectors. In this work, a sample of an X-ray detector was designed and constructed using 2cm×2cm THGEMs domestically... more
By expanding the applications of GEM detectors, a newer pattern of such detectors was introduced in 2004, named THGEM detectors. In this work, a sample of an X-ray detector was designed and constructed using 2cm×2cm THGEMs domestically produced with a thickness of 250 μm, a hole diameter of 300 μm and a pitch of 500 μm, for the first time. The triple THGEM detector working in Ar/CO2 gas mixture was characterized. Influence of gas pressure and gas mixture on gain of the detector was investigated. Results show the detector operated in a stable mode with no discharges. The gain of the detector increased with high voltage across the THGEM electrodes exponentially. This verified the performance of a detector as a proportional counter. Also, the detector’s gain is maximum at Ar/CO2 (80/20) gas mixture and voltage of 700 V applied to each multiplier. The detector is promising for localization applications such as particle physics experiments.
Dielectric barrier discharge (DBD) plasma is used for various applications. DBD is also one of the most efficient and low-cost methods for active fluid flow control. In this study, a detailed physical model of DBD in atmospheric pressure... more
Dielectric barrier discharge (DBD) plasma is used for various applications. DBD is also one of the most efficient and low-cost methods for active fluid flow control. In this study, a detailed physical model of DBD in atmospheric pressure at 1 kV DC voltage is developed with COMSOL Multiphysics software. Argon gas is also used as a background gas and electrodes are assumed to be copper. Plasma parameters such as electron and ion density, electric field, potential, and temperature for different gap distances of electrodes (1.0 mm, 0.9 mm, 0.8 mm) and different dielectric types (Quartz, Silica Glass, Mica). The results of the simulation show that the longitudinal distance of the grounded electrodes to the power electrodes has a direct influence on parameters such as electron temperature, and electron and ion density which are the main factors of fluid flow control. These parameters have the maximum value when Mica is used as a dielectric and the lowest value when Silica Glass is utilized.
Indirect drive inertial confinement fusion (ICF) holds promise for achieving practical energy generation through controlled fusion reactions. However, the efficiency of ICF is constrained by the Be ablator material used to contain the... more
Indirect drive inertial confinement fusion (ICF) holds promise for achieving practical energy generation through controlled fusion reactions. However, the efficiency of ICF is constrained by the Be ablator material used to contain the fuel. To overcome this limitation, researchers have proposed doping Be with various elements. In this study, we investigate the effects of Na and Br dopants, incorporated at concentrations of 4.86% and 2.1%, respectively, using a one-dimensional MULTI-IFE hydrodynamic code. This code serves as a numerical tool dedicated to analyzing Inertial Fusion Energy microcapsules, facilitating the examination of the Be ablator's performance in indirect drive ICF. Our results indicate that the addition of a beryllium layer doped with Na and Br significantly enhances the target gain, elevating it from the break-even value (G ≈ 1) to approximately G ≈ 12. Furthermore, we delve into the impact of these dopants on the plasma fuel conditions during the implosion, shedding light on the underlying physics of the system. These findings demonstrate that Na and Br doping in the Be ablator represents a viable approach for improving the efficiency of indirect drive ICF, potentially paving the way for the development of practical fusion energy systems.
The present work is concerned on neutron flux increasing in Tehran Research Reactor (TRR). TRR is a 5 MW pool-type research reactor with plate type fuels in which the light water is used as both the coolant and moderator. The main goal of... more
The present work is concerned on neutron flux increasing in Tehran Research Reactor (TRR). TRR is a 5 MW pool-type research reactor with plate type fuels in which the light water is used as both the coolant and moderator. The main goal of this paper is reaching to the average thermal neutron flux of the order of 1014 #/cm-2.s-1 in the central irradiation box. Combination of the TRR power upgrading with the compact core can enable us to reach a neutron flux higher than 1.5 × 1014 #/cm-2.s-1 without violating the neutronic and thermal-hydraulic safety criteria. The compact core, with 19 and 5 standard and control fuel elements respectively, is used as a base for the neutronic analyses. Compact core with 26 fuel assemblies fulfilled all neutronic and operation criteria. Considering thermal hydraulic aspect from previous study lets TRR to be upgraded to 8.5 MW, resulting in neutron thermal flux greater than 1.5 × 1014.
High-energy heavy ions produced by accelerators are used in industrial and medical applications. Recently carbon ions have been used in the treatment of cancerous tumors. Heavy ions by the spallation process will activate the soft tissue... more
High-energy heavy ions produced by accelerators are used in industrial and medical applications. Recently carbon ions have been used in the treatment of cancerous tumors. Heavy ions by the spallation process will activate the soft tissue components before tumors. In this research by GEANT4 toolkit and MCNPX code simulation were tried to calculate the secondary particles and radioactive elements produced in the soft tissue around tumors by the carbon ions spallation process. In the MCNPX code, the F8 tally card with the FT8 command was used to extract the activation and spallation information of secondary particles in the Z1=1 to Z2=25 atomic numbers range. It was shown that a wide range of radioactive elements was produced in healthy tissues in carbon therapy. addition to produced secondary particles, the Be-10 and C-14 radioactive elements were produced in high-energy carbon ions in soft tissue. Also, the GEANT4 toolkit result of produced secondary particles dosimetry was shown that the secondary particles dose per carbon ion is between 1.66 to 33.54 nGy for carbon ion energy between 1140 to 5160 MeV. The tail for 3480, 4080, and 5160 MeV of carbon ion energy are 0.12,1.01 and 11 cm respectively. The carbon ion beam divergence increases with beam energy and achieve to 33 mm for 5160 MeV carbon ion.
The neutron transmutation doping method is widely used in various fields, such as solar cells, hybrid cars, etc. The Silicon doping process can provide direct commercial income for nuclear research reactors. In this study, we aim to find... more
The neutron transmutation doping method is widely used in various fields, such as solar cells, hybrid cars, etc. The Silicon doping process can provide direct commercial income for nuclear research reactors. In this study, we aim to find the optimal location for silicon doping in the thermal column nose of the Tehran research reactor. For this purpose, computational MCNPX and ORIGEN2 codes were used to calculate the neutronic and radioactivity parameters of the silicon ingot. The important parameters such as the thermal to fast neutron ratio, heat deposition by gamma and neutron, and the radioactivity level of the silicon ingot and the produced radioisotopes have been determined to obtain the optimal irradiation channel. The results showed that the irradiation channel placed in the thermal column at a distance of 90 cm from the center of the TRR core has optimal conditions for the implementation of silicon doping. The channel provides a thermal neutron flux in order of 1.721012 n.cm-2.s‎-1‎ which is the least acceptable value to achieve a proposed neutron fluence during the operation cycles of TRR reactor. Also, the channel has the least possible heat deposition inside the silicon ingot of about 191 W. In addition, the thermal to fast neutron flux ratio of about 311 is enough higher than the determined IAEA limit for NTD.
‎Fast neutron irradiation is one of the most strategic radiation applications of research reactors‎. ‎Usually‎, ‎it is performed around the reactor core containing lower neutron flux‎. ‎In this paper‎, ‎a hybrid object has been introduced... more
‎Fast neutron irradiation is one of the most strategic radiation applications of research reactors‎. ‎Usually‎, ‎it is performed around the reactor core containing lower neutron flux‎. ‎In this paper‎, ‎a hybrid object has been introduced and analyzed to enhance irradiating applications of the fast neutrons in the core of a Material Testing Reactor (MTR)‎. ‎The tool includes an old-type low-consumed HEU control fuel element‎, ‎a dry channel‎, ‎and a Cd filter‎. ‎It is supposed to be installed at the internal neutron trap (D4 positions) of TRR core configuration‎. ‎Calculating results are very promising for using the proposed tool to increase neutron fluxes‎, ‎reduce thermal and epi-thermal neutron fluxes‎, ‎and shift the neutron spectrum toward the fast neutron region (hardening effect) at the chosen irradiating location‎. ‎Primary safety parameters are also checked and passed successfully‎. ‎Furthermore‎, ‎there are also some other presented safety items which must be checked carefully and conservatively in order to refabricate and install such a irradiating tool in an in-core location of a MTR‎.
‎In the present work‎, ‎the eigenvalue and eigenvector has been obtained by the Bohr Hamiltonian for even-even nuclei‎. ‎The competition between γ-stable and γ-rigid collective motions has been created in the presence of the rigidity... more
‎In the present work‎, ‎the eigenvalue and eigenvector has been obtained by the Bohr Hamiltonian for even-even nuclei‎. ‎The competition between γ-stable and γ-rigid collective motions has been created in the presence of the rigidity parameter‎. ‎The β-part of the collective potential has been chosen to be equal to the generalized Hulthen potential‎, ‎while the γ-angular part of the problem is associated with Ring-shaped potential around the γ=π/6 and the Harmonic oscillation around the γ=0‎. ‎In both cases‎, ‎the effect of rigidity and free parameters on energy spectrum of Os-180‎, ‎Dy-162‎, ‎Gd-160‎, ‎Ru-100‎, ‎Pd-114‎, ‎and Xe-124 nuclei have been investigated‎. ‎Also‎, ‎the rates of B(E2) transition have been calculated and compared with experimental data‎. ‎This model has an appropriate description of energy spectra for the mentioned nuclei‎.
‎Both of small and medium sized reactors and small modular reactors are called SMRs‎. ‎They are reviewed and discussed in this paper‎, ‎particularly integral Pressurized Water Reactors (iPWRs)‎. ‎Studies show that PWRs are the most... more
‎Both of small and medium sized reactors and small modular reactors are called SMRs‎. ‎They are reviewed and discussed in this paper‎, ‎particularly integral Pressurized Water Reactors (iPWRs)‎. ‎Studies show that PWRs are the most interested‎, ‎designed and constructed nuclear reactor type worldwide‎. ‎Some innovative small modular PWRs like the MASLWR‎, ‎NuScale‎, ‎CAREM-25‎, ‎SMART and ACP-100 have several outstanding characteristics to be promisingly recognized as near term options of the next generation of small modular PWRs‎. ‎They have several inherently safety features and improved passive safety system‎. ‎They require smaller infrastructure and capital costs‎. ‎They can be also developed rapidly in different and independent modular unites even for remote area or outlands without required infrastructure or electrical grids‎. ‎It should be noted that new modern economy strategies like the Return of Investment (ROI) issues may advice medium or large reactors rather than small units for developed and industrial countries while small modular plans can be much more interesting and accessible for new comers or even developing countries‎. ‎Finally‎, ‎multi-applicability is an appropriate solution to develop expensive nuclear power plants economically as well as multi-purpose research reactors (especially by means of small modular iPWRs)‎.
‎The analysis deals with the assessment of best estimate code RELAP5/SCDAP mod3.4 in the simulation of double-ended loss coolant accident as a LBOCA‎, ‎4 in break as a SBLOCA an SBO accident with considering except accumulator water where... more
‎The analysis deals with the assessment of best estimate code RELAP5/SCDAP mod3.4 in the simulation of double-ended loss coolant accident as a LBOCA‎, ‎4 in break as a SBLOCA an SBO accident with considering except accumulator water where no core cooling water systems are available‎. ‎The reference plant is SURRY nuclear power plant as a Westinghouse three-loop nuclear power plant‎. ‎In order to mitigation accident‎, ‎the in-vessel retention strategy was investigated for the prevention of lower plenum failure‎. ‎It has been concluded that during the SBLOCA‎, ‎LBLOCA conditions bottom of active fuel is uncovered at 6340 s and 2160 s‎, ‎respectively‎. It occurred for two times at 11650 s and 15608 s in SBO‎. ‎At 6792 s and 57002 s in the LBLOCA and SBO due to reaching melting point and in the SBLOCA at 15215 s due to lower plenum creep rupture‎, ‎failure of the reactor pressure vessel occurred‎. ‎The results show that hydrogen production in the SBO is more than the other two cases‎. ‎For the prevention of the lower plenum failure‎, ‎the in-vessel molten material retention strategy is investigated as a passive system‎. ‎The results show that lower plenum heat flux can be kept below the critical heat flux and its integrity is preserved in two cases of this analysis‎.
‎It is well-known that response function of organic scintillation detectors does not appear with photopeaks‎. ‎Instead‎, ‎their dominant feature is a continuum‎, ‎usually called the Compton edge that innately exposes the resolution... more
‎It is well-known that response function of organic scintillation detectors does not appear with photopeaks‎. ‎Instead‎, ‎their dominant feature is a continuum‎, ‎usually called the Compton edge that innately exposes the resolution characteristics of detection system‎. ‎While‎, ‎accurate characterization of Compton edge is crucial for calibration purposes‎, ‎it is also in charge of elaborating the energy resolution of detector‎. ‎This paper presents a simple method for accurate characterization of the Compton edge in organic scintillation detectors‎. ‎The method is based on the fact that differentiating the response function leads to accurate estimation of the constituting functions‎. ‎The differentiation method‎, ‎in addition to the location of the Compton edge‎, ‎gives insights into the parameters of the folded Gaussian function which could lead to depict the energy resolution‎. ‎Moreover‎, ‎it is observed that the uncorrelated noise in the measurement of the response function does not impose significant uncertainties in the evaluations‎, ‎so it could preserve its functionality even in lower-quality measurements‎. ‎By simulation of the bounded electrons and considering the Doppler effects‎, ‎we are able to demonstrate‎ -‎the first ever‎- ‎estimation for intrinsic Doppler resolution of an organic plastic scintillator‎. ‎Even though‎, ‎this possibility is an immediate result of benefiting the presented method for analysis of the Compton continua‎.
Kinetic and neutronic parameters play an important role in analysis of reactors dynamic behavior. Some of these parameters include: effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial... more
Kinetic and neutronic parameters play an important role in analysis of reactors dynamic behavior. Some of these parameters include: effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial distributions, effective delayed neutron fraction (βeff) and prompt neutron lifetime (lp ). In this work, Monte Carlo modeling and analysis of Isfahan MNSR is performed for calculation of the kinetic and neutronic parameters of using MCNPX2.6 code, slope fit and perturbation methods. Relative differences between results of the MCNPX2.6 code in calculation of the ρ and βeff and the reference values are about 0.5% and 2.1%, respectively. The relative differences between the results of the slope fit and perturbation methods and MCNPX2.6 code in calculation of the parameter with the reference values are about 17.6%, 4.8% and 29.19%, respectively. Therefore, the results of these research show that the MCNPX2.6 code is suitable for calculation of the reactor kinetic parameters such as the βeff, while the perturbation method is a simple and convenient method for calculating the .
Today thorium based fuels are being investigated as an alternative fuel technology. However, the majority of thorium fuel research studies are limited to reactor physics investigations, which leaves a gap for dose evaluation and shielding... more
Today thorium based fuels are being investigated as an alternative fuel technology. However, the majority of thorium fuel research studies are limited to reactor physics investigations, which leaves a gap for dose evaluation and shielding concerns of such spent fuels. The present work investigates thorium oxide fuel assemblies in Tehran research reactor. The fuel gamma dose rates are calculated at different burnups and cooling times. A comparison between the reactor routine fuel and the thorium oxide fuel is conducted to reveal the thorium-based fuel application shielding challenges. The obtained results showed that inverse to U3O8-Al routine fuel the spent ThO2 gamma dose rates are completely dependent to the burnup values. In addition, for transporting the spent ThO2 fuel with the routine transport casks there is needed to be waited for the higher cooling times than U3O8-Al transportation time or construction of thicker transport casks is needed for transportation of the thorium-based spent fuels at shorter times.
‎By the rapid development of imaging systems such as PET/CT for diagnosis of cancer‎, ‎the protection of staff and public has become a main health concern‎. ‎Due to serious and irreversible harms of ionization radiations‎, ‎protection of... more
‎By the rapid development of imaging systems such as PET/CT for diagnosis of cancer‎, ‎the protection of staff and public has become a main health concern‎. ‎Due to serious and irreversible harms of ionization radiations‎, ‎protection of all those who are exposed is the main concern of health issues‎. ‎The main basis of the calculation of the shielding design in the medical imaging systems is that the absorbed dose should not exceed the allowed limit‎. ‎In this study‎, ‎the current shielding status of the PET/CT installations in Tehran's Shariati hospital was investigated using the MCNPX Monte Carlo code to ensure that the dose limits for both the controlled and uncontrolled area are not violated‎. ‎The proposed simulation method was benchmarked with a validated analytical method‎. ‎Shariati hospital provides services to four patients every day‎, ‎leading to a dose rate in the range of 2.6 × 10-6 to 9.35 × 10-3 mSv/week‎. ‎The minimum dose rate in this range represents the value behind the door of the waiting room (public uncontrolled area)‎, ‎while the maximum in this range corresponds to the value behind the glass of the scanner room (operator controlled area)‎. ‎The simulation results for 8 patients/day in this center showed that the dose rate behind the wall of the injection room will increase from 4.88 ×10-6 mSv/week to 2.81 × 10-2 mSv/week‎, ‎which is well below the recommended levels‎. ‎This indicates that the present shielding is adequate for up to four more patients per day‎.
‎Spallation process is the most significant process for neutron generation in industry and medicine‎. ‎This process has been used in the subcritical reactor core‎. ‎In this research‎, ‎we study the neutronic behavior of non-fissionable... more
‎Spallation process is the most significant process for neutron generation in industry and medicine‎. ‎This process has been used in the subcritical reactor core‎. ‎In this research‎, ‎we study the neutronic behavior of non-fissionable and fissionable spallation targets consists of U-238‎, ‎Th-232‎, ‎Lead Bismuth Eutectic (LBE) and W-184 materials in cylindrical and conic shapes using MCNPX code‎. ‎Neutronic parameters consist of spallation neutron yield‎, ‎deposition energy‎, ‎and angular spectrum of the neutron output‎. ‎The gas production rate and residual mass spectrum were investigated‎. ‎The results of this research indicate that the shape of the target must be selected based on target material and operational purposes‎. ‎The number of neutrons per energy unit is stable at energies higher than 1 GeV‎, ‎and the rate of change in neutron generation has been reduced after that‎. ‎Furthermore‎, ‎hydrogen is the principal factor in swelling of spallation target and consists of about 88% of gas production‎. ‎It was found that a target of LBE provides the most favorite parameters for both neutronic and physical properties‎
‎The used metering technique in this study is based on the dual energy (Am-241 and Cs-137) gamma ray attenuation‎. ‎Two transmitted NaI detectors in the best orientation were used and four features were extracted and applied to the... more
‎The used metering technique in this study is based on the dual energy (Am-241 and Cs-137) gamma ray attenuation‎. ‎Two transmitted NaI detectors in the best orientation were used and four features were extracted and applied to the model‎. ‎This paper highlights the application of Adaptive Neuro-fuzzy Inference System (ANFIS) for identifying flow regimes and predicting volume fractions in gas-oil-water multiphase systems‎. ‎In fact‎, ‎the aim of the current study is to recognize the flow regimes based on dual energy broad-beam gamma-ray attenuation technique using ANFIS‎. ‎In this study‎, ‎ANFIS is used to classify the flow regimes (annular‎, ‎stratified‎, ‎and homogenous) and predict the value of volume fractions‎. ‎To start modeling‎, ‎sufficient data are gathered‎. ‎Here‎, ‎data are generated numerically using MCNPX code‎. ‎In the next step‎, ‎ANFIS must be trained‎.
‎According to the modeling results‎, ‎the proposed ANFIS can correctly recognize all the three different flow regimes‎, ‎and other ANFIS networks can determine volume fractions with MRE of less than 2% according to the recognized regime‎, ‎which shows that ANFIS can predict the results precisely‎.
‎Collision of protons with background gas and beamline wall in proton therapy causes the creation of secondary particles‎, ‎e.g. neutrons‎, ‎which results in more difficulties in curing the tumors‎. ‎In the present simulation-based... more
‎Collision of protons with background gas and beamline wall in proton therapy causes the creation of secondary particles‎, ‎e.g. neutrons‎, ‎which results in more difficulties in curing the tumors‎. ‎In the present simulation-based study‎, ‎the optimum diameter of proton beamline was determined to minimize the production of secondary particles in the presence of electric field with the magnitude of 50 kV/m‎, ‎perpendicular equal magnetic fields of 0.7 T‎, ‎and background gas of argon under Bounce boundary conditions via finite element method‎. ‎The results showed that the optimum diameter of the beamline for minimization of the secondary particles in the spot scanning proton therapy in the aforementioned conditions was 7 mm‎. ‎Also‎, ‎the values of drift velocities of protons were plotted in different time steps of 10 ns to 50 ns for the optimized size of the beamline‎. ‎Due to few interactions of forwarding particles with background gas‎, ‎the results showed that the forwarding particles in the propagation direction have greater velocities than those of rear particles‎. ‎The results can be used in spot scanning proton therapy for curing the localized cancers‎.
‎The Jacobian-Free Newton-Krylov (JFNK) method has been widely used in solving nonlinear equations arising in many applications‎. ‎In this paper‎, ‎the JFNK solver is examined as an alternative to the traditional power iteration method... more
‎The Jacobian-Free Newton-Krylov (JFNK) method has been widely used in solving nonlinear equations arising in many applications‎. ‎In this paper‎, ‎the JFNK solver is examined as an alternative to the traditional power iteration method for calculation of the fundamental eigenmode in reactor analysis based on even-parity neutron transport theory‎. ‎Since the Jacobian is not formed the only extra storage required is associated with the workspace of the Krylov solver used at every Newton step‎. ‎A new nonlinear function is developed for the even-parity neutron transport equation utilized to solve the eigenvalue problem using the JFNK‎. ‎This Newton-based method is compared with the standard iterative power method for a number of multi-groups‎, ‎one and two dimensional neutron transport benchmarks‎. ‎The results show that the proposed algorithm generally ends with fewer iterations and shorter run times than those of the traditional power method‎.
‎As one of the most clinically relevant parameters in proton radiotherapy‎, ‎the range of incident particles can be measured either by counting the number of protons or through depth-dose evaluation in the target‎. ‎In the latter‎, ‎the... more
‎As one of the most clinically relevant parameters in proton radiotherapy‎, ‎the range of incident particles can be measured either by counting the number of protons or through depth-dose evaluation in the target‎. ‎In the latter‎, ‎the range is defined as the depth in the target at the distal 80% point of the Bragg peak‎. ‎In this work‎, ‎a highly accurate analytical model was employed to predict depth-dose distribution‎, ‎and hence the range‎, ‎in a desired target‎. ‎Aiming to study the effect of energy spread on the range‎, ‎proton beams with initial Gaussian distributions have been considered‎. ‎For our arbitrary tested energies‎, ‎the results show that the more the width of energy distribution increases‎, ‎the more the Bragg peaks shift in depth‎, ‎by about‎ -‎0.25% to‎ -25%, ‎compared with those of monoenergetic beams‎. ‎Furthermore‎, ‎it was found that for different widths of initial energy spectrum‎, ‎keeping the mean energy the same‎, ‎the range remains unchanged‎. ‎It was also shown that the results corresponding to utilizing analytical range determination for proton beams of different incident energies in stack of materials deviate from those of Monte Carlo simulations by less than 1.7%‎. ‎The results are encouraging‎, ‎although accurate modeling of analytical proton dose distribution in the presence of tissue inhomogeneities is still an unsolved problem‎.
The role of saturation property of cold nuclear matter is‎ ‎examined in order to describe the steep falloff phenomenon of the‎ ‎measured fusion cross sections at energies far below the Coulomb‎ ‎barrier for 58Ni+54Fe colliding system‎.... more
The role of saturation property of cold nuclear matter is‎ ‎examined in order to describe the steep falloff phenomenon of the‎ ‎measured fusion cross sections at energies far below the Coulomb‎ ‎barrier for 58Ni+54Fe colliding system‎. ‎For this aim‎, ‎the‎ ‎double-folding microscopic approach which is modified by modeling‎ ‎the repulsive core effects in the nucleon-nucleon interactions is‎ ‎used to calculate the nuclear interaction potential‎. ‎Moreover‎, ‎the‎ ‎theoretical values of the fusion cross section‎, ‎S factor‎, ‎and‎ ‎the logarithmic derivative are computed using the coupled-channel‎ ‎technique‎, ‎including couplings to the low-lying 2+ and 3-‎ ‎states in target and projectile‎. ‎The results obtained reveal that‎ ‎the corrective effects of cold nuclear matter can be responsible for‎ ‎the description of the fusion hindrance phenomenon in our chosen system‎.
‎Distinguishing naturally occurring radioactive (e.g. ceramics‎, ‎fertilizers‎, ‎etc.) from unauthorized materials (e.g. high enriched uranium‎, ‎Pu-239‎, ‎etc.) to reduce false alarms is a prominent characteristic of radiation monitoring... more
‎Distinguishing naturally occurring radioactive (e.g. ceramics‎, ‎fertilizers‎, ‎etc.) from unauthorized materials (e.g. high enriched uranium‎, ‎Pu-239‎, ‎etc.) to reduce false alarms is a prominent characteristic of radiation monitoring port‎. ‎By employing the energy windowing method for the spectrum correspond to the simulation of a plastic scintillator detector using the MCNPX Monte Carlo code together with an artificial neural network‎, ‎the present work proposes a method for distinguishing naturally occurring materials and K-40 from four unauthorized sources including high enriched uranium and Pu-239 (as special nuclear materials)‎, ‎Cs-137 (as an example of dirty bombs)‎, ‎and depleted uranium‎.
‎In this work‎, ‎dynamic responses of a WWER-1000 reactor in reactivity insertions are studied using a coupling method‎. ‎The ANSYS-CFX is implemented for thermal hydraulic study of the core and the point kinetic equation (PKE) is coupled... more
‎In this work‎, ‎dynamic responses of a WWER-1000 reactor in reactivity insertions are studied using a coupling method‎. ‎The ANSYS-CFX is implemented for thermal hydraulic study of the core and the point kinetic equation (PKE) is coupled as a FORTRAN subroutine‎. ‎For transient analysis of the core‎, ‎the thermal feedback of the fuel is added to coolant‎, ‎and numerical solver of cylindrical heat transfer for obtaining the irradiated fuel rod temperature profile is also included‎. ‎In order to investigate the irradiation effect‎, ‎the fuel and gap properties in burnup with appropriate correlations could be calculated‎. ‎Using memory management system (MMS) and data transfer arrays, coupling between numerical subroutines is carried out‎. ‎It is shown that the dynamic response of the core depends on burnup‎, ‎and the response could be varied in time‎. ‎In addition‎, ‎the coupling method is reliable for other dynamic calculations‎.
‎The aim of this research was determination of the required time for coagulation of in vivo cut bleeding treated by non-thermal atmospheric pressure plasma‎. ‎To meet this‎, ‎an atmospheric pressure plasma jet device was designed and... more
‎The aim of this research was determination of the required time for coagulation of in vivo cut bleeding treated by non-thermal atmospheric pressure plasma‎. ‎To meet this‎, ‎an atmospheric pressure plasma jet device was designed and constructed‎. ‎Helium was used as working gas‎. ‎The electrical parameters and optical emission spectrum of helium plasma were measured‎. ‎The averaged treatment time to coagulate the incision bleeding on the mouse liver was obtained 8.6 μs‎, ‎and the average time of naturally incision bleeding coagulation was 10 min‎.
‎In this paper‎, ‎the effect of anode's insert material on spatial distribution of X-ray emission zone of plasma focus device was studied‎. ‎Anode's insert materials were fabricated out of aluminum‎, ‎zinc‎, ‎tin‎, ‎tungsten and lead‎.... more
‎In this paper‎, ‎the effect of anode's insert material on spatial distribution of X-ray emission zone of plasma focus device was studied‎. ‎Anode's insert materials were fabricated out of aluminum‎, ‎zinc‎, ‎tin‎, ‎tungsten and lead‎. ‎For each insert material at the constant operating voltage of 21 kV‎, ‎the image of pinhole camera which monitors the surface and the top of anode was recorded at the various pressures of 0.3‎, ‎0.6‎, ‎0.9 and 1.2 mbar‎. ‎The results indicated that the X-ray emission zone above the anode surface not only includes thermal radiation of plasma‎, ‎but also depends on anode's insert materials‎. ‎This zone could be due to the passage of high energy electrons from the vapor of anode's material above the anode's surface‎.
‎A second shutdown system (SSS) is designed for the Tehran Research Reactor (TRR) completely independent and diverse from the existing First Shutdown System (FSS)‎. ‎Given limitations‎, ‎specifications‎, ‎and requirements of the reactor‎,... more
‎A second shutdown system (SSS) is designed for the Tehran Research Reactor (TRR) completely independent and diverse from the existing First Shutdown System (FSS)‎. ‎Given limitations‎, ‎specifications‎, ‎and requirements of the reactor‎, ‎the design of SSS is based on the injection of liquid neutron absorber‎. ‎The plan has the ability to satisfy the major criterion of required negative reactivity worth‎, ‎to transfer the reactor to subcritical state in needed time‎, ‎with necessary shutdown margin and for the required duration‎. ‎Design calculations are performed using the stochastic code MCNPX2.6.0‎, ‎deterministic code PARET and Pipe Flow Expert software‎. ‎The ORIGEN2 code and HotSpot health physics code are also used for simulation of environmental pollution release‎. ‎The SSS chambers cause a decrease of about 5% and 15% in total and thermal neutron flux‎, ‎respectively‎. ‎To demonstrate the SSS role in enhancing reactor safety‎, ‎the probable accident of core meltdown is investigated‎. ‎As a consequence of this accident‎, ‎the radioactive pollution in and out of reactor containment is released‎. ‎Without existing the SSS and in case of failure of FSS‎, ‎the residents within 58000 m2 of the reactor perimeter would receive about 1 mSv which is more than the annual limit of absorbed dose for the community‎.
‎In this paper‎, ‎dose uniformity ratio in irradiation cell of GC-220 is specifiedutilizing an analytical method based on the multipole moment expansion‎. ‎In this method‎, ‎the values of monople‎, ‎dipole and quadrupole moments for... more
‎In this paper‎, ‎dose uniformity ratio in irradiation cell of GC-220 is specifiedutilizing an analytical method based on the multipole moment expansion‎. ‎In this method‎, ‎the values of monople‎, ‎dipole and quadrupole moments for source arrangements of GC-220 are calculated by numerical integrating‎. ‎Appling these values‎, ‎the dose uniformity ratio in the irradiation cell of GC-220 is calculated equal to 1.92‎. ‎Monte Carlo simulation is applied to validate calculations‎. ‎There is a relative difference about 12% between the results obtained from the analytical calculation and Monte Carlo simulation‎, ‎which confirm the used method‎. ‎In comparison with Monte Carlo methods‎, ‎this method is not time consuming‎, ‎so‎, ‎this method can be used for the conceptual designing and the source load planning of irradiators‎.
‎Numerical solution of the multi-group static forward and adjoint neutron diffusion equation (NDE) using the Finite Elements Method (FEM) is investigated in detail‎. ‎A finite element approach based on the generalized least squares method... more
‎Numerical solution of the multi-group static forward and adjoint neutron diffusion equation (NDE) using the Finite Elements Method (FEM) is investigated in detail‎. ‎A finite element approach based on the generalized least squares method is applied for the spatial discretization of the NDE in 3D-XYZ geometry‎. ‎A computer code called GELES was also developed based on the described methodology covering linear or quadratic tetrahedral elements generated via the mesh generator for an arbitrary shaped system‎. ‎A number of test cases are also studied to validate the proposed approach‎. ‎Moreover‎, ‎to assess the output dependency to the number of elements‎, ‎a sensitivity analysis is carried out at the end‎.
‎Utilizing radioactive stents is a usual method for treatment of advanced esophageal cancer‎. ‎It is necessary to investigate the dose distribution of radioactive esophageal stents before the clinical use‎. ‎This study presents a... more
‎Utilizing radioactive stents is a usual method for treatment of advanced esophageal cancer‎. ‎It is necessary to investigate the dose distribution of radioactive esophageal stents before the clinical use‎. ‎This study presents a dosimetric comparison between three radioactive esophageal stents‎: ‎I-125 seed-loaded stent‎, ‎iodine-eluting stent and double-layered iodine-eluting stent‎. ‎Depth-dose and angular dose distributions were carried out using Geant4 toolkit‎. ‎Moreover‎, ‎the effect of interval distance between two adjacent seeds on the dose distribution was investigated‎. ‎Esophageal stents loaded with I-125 seeds seems to be better than iodine-eluting stents‎, ‎with the distance less than 15 mm between two adjacent seeds‎.
‎18F-FDG PET/CT is commonly used for evaluation and diagnostic of many types of cancer‎, ‎such as; tumor diagnosis‎, ‎treatment monitoring‎, ‎and radiation therapy planning‎. ‎Accurate diagnostic is needed in meticulous patient... more
‎18F-FDG PET/CT is commonly used for evaluation and diagnostic of many types of cancer‎, ‎such as; tumor diagnosis‎, ‎treatment monitoring‎, ‎and radiation therapy planning‎. ‎Accurate diagnostic is needed in meticulous patient preparation‎, ‎including restrictions of diet and activity and management of blood glucose levels in diabetic patients‎, ‎as well as an awareness of the effect of medications and environmental conditions‎. ‎All of these conditions play important roles toward obtaining good-quality images‎, ‎which are essential for accurate interpretation‎. ‎This article introduces the new synthesis and quality control method for obtaining the best quality FDG which is used as radiopharmaceutical‎. ‎All the reactions are carried out and completed in one reaction vessel without any replacement‎. ‎The paper is including details of synthesis‎, ‎quality control and transportation step‎. ‎It is the first time that the alkaline FDG synthesis is introducing by details in Iran‎.
Proton therapy of liver tumors can be challenging due to the absorbed dose of produced secondary particles in non-target organs. This study aims to evaluate the absorbed dose of secondary particles during the proton therapy of liver... more
Proton therapy of liver tumors can be challenging due to the absorbed dose of produced secondary particles in non-target organs. This study aims to evaluate the absorbed dose of secondary particles during the proton therapy of liver cancer through the MCNPX Monte Carlo (MC) code by a simplified MIRD-UF standard phantom. At first, a simplified MC model of MIRD-UF standard phantom was simulated using MCNPX. After the proper proton energies calculation ranging from 90 to 120 MeV for 4×4×4 cm3 tumor irradiation, mesh tally type 3 and F6 tally were used to calculate the depth dose profiles as well as the absorbed dose of protons and secondary particles in non-involved organs. The obtained results illustrated that the fluence of internal secondary particles doses was considerably small in comparison with primary protons. Furthermore, most of neutrons and photons doses were absorbed around the liver tissue for all performed proton energies (i.e., 90 and 120 MeV) which non-target organs did not receive a significant high dose. Furthermore, the absorbed dose of secondary photons and neutrons had slight variations in considered normal tissues near the liver. The calculated results in this study indicated that during the proton therapy of liver cancer, the most contribution of the secondary particle doses was absorbed inside the liver tissue. Hence, it can be expected the probable side effects (secondary cancers) associated with the liver cancer proton therapy may be decreased however, the presence of secondary particles should not be ignored.
In recent years, various designs for controlled thermonuclear fusion based on the p11B reaction have been reviewed and optimized. In this article, to innovate in achieving a better power and energy gain of neutron-free p11B fusion... more
In recent years, various designs for controlled thermonuclear fusion based on the p11B reaction have been reviewed and optimized. In this article, to innovate in achieving a better power and energy gain of neutron-free p11B fusion reaction, the improvement of the cross-section and also the kinetic effects. Then, the effects of bremsstrahlung radiation and ion and electron energy exchange rate have been evaluated by introducing relativistic effects and its role on improving fusion energy gain. As a result, the temperature of the electron is kept lower than that of the ion, which improves fuel performance. Finally, it leads to an increase in the number of protons at higher energies compared to the pure Maxwellian distribution and it causes a significant increase in reactivity compared to previous research. Also, the number of alpha particles obtained through calculations coincides with the latest research and leads to an enhancement of approximately 13%. This means that by improving the fusion cross-section of p11B, our calculations show that considering the avalanche effects, the range of achievable energy gain in the temperature range of 300 to 500 keV and the stable characteristic time of 0.64 ps reaches 89 to 111. While in the same temperature range and with the stable characteristic time of 0.74 ps, regardless of the improved cross-section, the energy gain range is 75 to 98.
Hot springs are known as one of the hydrotherapy centers in the world and have been welcomed due to their healing properties. Due to the presence of radon and radioactive elements in hot spring sediments, water and soil, these components... more
Hot springs are known as one of the hydrotherapy centers in the world and have been welcomed due to their healing properties. Due to the presence of radon and radioactive elements in hot spring sediments, water and soil, these components are radioactive. So far, radiation hazards and the annual effective dose of hot spring components in the body organs have not been investigated in Iran. The purpose of this study was to calculate the amount of  U-238, Cs-137, Th-232, and K-40 elements in soil, water, and sediments of Jooshan hot springs in the Kerman region. The presence of these elements causes radiation hazards and an effective annual dose in people who use these hot springs. In addition to the healing properties of hot springs, the high amount of radiation hazard and effective annual dose may cause cancer risk. Experimental results with CsI(Tl) detector showed that the total activities of these elements in soil, water, and sediments of Jooshan hot spring were 95.26±9.76, 52.86±7.27, and 51.61±7.18 Bq.kg-1 respectively. The Jooshan hot spring's radiation hazards were calculated using activity measurement of the radioactive elements in soil, water, and sediments which was less than the permission level. The result of the Monte Carlo simulation with the MCNPX code showed that the effective annual dose of sediment, water, and radon in Jooshan hot spring are 5.43E-06, 3.00×10-3 and 1.16×10-1 mSv.year-1 respectively, which is less than effective annual dose (5 mSv.year-1). The maximum time for treatment by hot spring water is considered equal to one year.
In this work, neutron and gamma shielding were simulated using MCNPX code for an inertial electrostatic confinement Fusion (IECF) device. In this regard, various properties of shields were investigated. Portland reinforced concrete was... more
In this work, neutron and gamma shielding were simulated using MCNPX code for an inertial electrostatic confinement Fusion (IECF) device. In this regard, various properties of shields were investigated. Portland reinforced concrete was considered as the first layer. In addition to being effective in reducing the dosage of fast neutrons, concrete layer was also considerably effective in reducing the dose of gamma rays. As for the second and third layers, we opted for paraffin and boric acid based. These layers were chosen based on parameters such as lethargy, macroscopic slowing down power (MSDP), etc. in order to reduce the speed of epithermal neutrons and then absorb the thermal neutrons, thus reducing the transmitted neutron dosage as much as possible. A layer lead was used after these three layers of shielding to attenuate the gamma ray reaching this layer. In this study, a fusion source based on D-T fuel with homogeneous and isotropic radiation of neutrons was used and then dosimetry was performed for different parts. Afterwards, the thickness of the shielding layers was optimized in such a way that the neutron and gamma doses were reduced according to the standards. We found that it is possible to achieve safe neutron and gamma fluxes and doses by applying about 5 layers of 50 cm thickness. We compared the results of our study with the those of another study done on shielding for the IECF device, which were in good agreement.
In this work, the concentration of tritium in D2O of various degrees of purity was measured. Samples were taken from the Arak heavy water plant and tritium concentrations were determined using a liquid scintillation detector (LSC) based... more
In this work, the concentration of tritium in D2O of various degrees of purity was measured. Samples were taken from the Arak heavy water plant and tritium concentrations were determined using a liquid scintillation detector (LSC) based on tritium decay. In this work, instead of simple distillation, is used the azeotropic distillation method. Absorption and fluorescence spectra were recorded using a Shimadzu UV-2100 spectrometer and an LS50B fluorescence spectrometer. The tritium concentration in the samples varied from 1.75 ± 0.80 to 6.16 ± 1.01 Bq.L-1 in D2O enrichment from 0.35% to 77.50%. The correlation coefficient between tritium concentration and D2O purity in heavy water was obtained as R2 = 0.853. Deviation for 99.8% D2O enriched in heavy water. This was observed from a straight line, leading to a drop in R2. The results of this measurement showed that the tritium concentration did not exceed the value set by the Nuclear Regulatory Commission (NRC).
Neutron detection techniques are widely studied in many articles. Most of this research requires a lot of electronic equipment. In this study, using the Thick Gas electron multiplier (THGEM) detector, a new method for neutron detection is... more
Neutron detection techniques are widely studied in many articles. Most of this research requires a lot of electronic equipment. In this study, using the Thick Gas electron multiplier (THGEM) detector, a new method for neutron detection is proposed to reduce electronic equipment. In the neutron detection system, the converter material is used for converting neutrons to protons that are directed to the THGEM detector. By filling the detector space with noble gas and applying special voltage, THGEM enters to Self-Quenched Streamer (SQS) mode for protons detection. All these steps are examined by simulation, then the detection system is made and is examined in the laboratory. Finally, the simulation results and laboratory results are compared.    The results show that the 1 mm Plexiglas layer is suitable for converting neutrons to protons. The suitable distance between the converter layer and the THGEM detector is 3 cm. Also, the SQS mode happens in the most number of THGEM holes when the THGEM voltage is 980 volt. Investigating an approach to neutron detection by placing THGEM in SQS mode can be useful because, firstly, placing the THGEM detector in SQS mode simplifies electrical circuits and secondly, with this proposed detection system; it is possible to design detectors with different dimensions for neutrons.
Primary standardization of radioactivity is related to the direct measurement of activity in radioactive decay. A large variety of primary standardization techniques have been developed in the past years. The photon-photon coincidence... more
Primary standardization of radioactivity is related to the direct measurement of activity in radioactive decay. A large variety of primary standardization techniques have been developed in the past years. The photon-photon coincidence counting is one of the methods for activity determination. This method is particularly applied for the standardization of I-125 using the detection of X-ray and gamma-ray coincident counting. In this paper, a 2D photon-photon coincidence digital system with two similar ‎‎2'' × 2''‎‎ NaI(Tl) detectors for absolute activity measurement is developed. The system is established based on a 100 MHz CAEN waveform digitizer (DT5724) which directly records the pre-amplifier output signals of the two NaI(Tl) detectors. The sampled signals was transformed to trapezoidal signals using pulse height analyzer firmware and coincidence events were recorded in a list file. The list file was analyzed offline using a Matlab code to realize correlated gama lines of Co-60 source.  The Volkovitsky formulas were used for the activity calculation and the details of the experimental setup were also discussed. Standardization of  the two Co-60 standard sources was performed using this system. Results are in good agreement with the reference activity of Co-60 sources. The presented formula can be modified for absolute calibration of the other medical radioisotopes. The technique can be generalized for absolute activity measurement of I-125 which uses for ophthalmic plaque radiation therapy.
Digital spectroscopy with Silicon Photomultiplier (SiPM) array coupled on CsI(Tl) crystal, needs some consideration to achieve desirable energy resolution. an array of SiPMs must be used with large scintillation crystal, therewithal... more
Digital spectroscopy with Silicon Photomultiplier (SiPM) array coupled on CsI(Tl) crystal, needs some consideration to achieve desirable energy resolution. an array of SiPMs must be used with large scintillation crystal, therewithal signal from whole SiPMs is not ideal. Also silicon pixels are arranged in arrays at regular intervals. The distances between the pixels and the interference of the cross-talk of adjacent pixels are undesirable factors for energy resolution when the light is received by the crystal and transmitted to the SiPM array. On the other hand, due to the advantages of the SiPM array, there is a need to improve the energy resolution in order to be able to be applied. In this paper, an attempt is made to obtain the desired energy resolution by using the digital spectroscopy method and digital filters and properly shaping the output pulse of the preamplifier by the trapezoidal filter method. In this case, it will be possible to use it in different applications such as spectroscopy by the detector.
Actinide concentration and activity analysis of the nuclides resulted from the burnup (depletion) process during nuclear reactor operation lifetime is an essential problem in reactor design. Inventory and the corresponding activities of... more
Actinide concentration and activity analysis of the nuclides resulted from the burnup (depletion) process during nuclear reactor operation lifetime is an essential problem in reactor design. Inventory and the corresponding activities of the Tehran Research Reactor (TRR) are evaluated using different methods and compared with each other. WIMS-CITATION, ORIGEN, and MCNP codes are used for plate type inventory calculations. The important actinides, fission products, and fissile inventory ratio of TRR have been calculated at different burnup steps. It is worth noting to mention that knowing the value of inventory helps us for safe handling of the spent fuels and to have a perfect design for transport cask of spent fuels. In this paper, the fuel isotope inventories were calculated for the first and 83rd core configuration of the Tehran Research Reactor, which is named “Core01” and “Core83” respectively. The calculations were first performed using WIMS-D5 and CITATION neutronic codes and then the results are compared with that of ORIGEN and MCNPX code. The total radioactivity of the TRR core at the end of the reactor core life (Core83) is estimated to be 6.47×105 Ci.
We study the low-energy deuteron-deuteron elastic scattering using the cluster effective field theory formalism up to next-to-leading order (NLO). For this purpose, we initially focus on determination of the unknown effective field theory... more
We study the low-energy deuteron-deuteron elastic scattering using the cluster effective field theory formalism up to next-to-leading order (NLO). For this purpose, we initially focus on determination of the unknown effective field theory coupling constant values using the phase shift analysis and available differential cross section data. The differential cross section versus center of mass energy and scattering angle are plotted up to NLO in the suggested power counting and compared to the available experimental data. Our effective field theory results show
good consistency with the present data.
In the present paper, the dechanneling and the energy loss of protons at the energy interval of 1400 to 2200 keV along the {100} and the {110} planar directions of Si were studied by the simulation of the measured channeling Rutherford... more
In the present paper, the dechanneling and the energy loss of protons at the energy interval of 1400 to 2200 keV along the {100} and the {110} planar directions of Si were studied by the simulation of the measured channeling Rutherford back-scattering spectra based on the exponential dechanneling function with a parameter λ. This parameter is proportional to the dechanneling rate and represents the mean distance that ions travel along the channel before escaping from the channel. The Levenberg-Marquardt algorithm was used to set the best values of the channeling to random energy loss ratio, and the mean channeling distance. The experimental results are well reproduced by this simulation. The data analyzed in this energy range did not show any particular trend with regard to energy dependence of the parameters. The differences between both the planar channels in the Si crystal and their influence on the energy loss ratio and dechanneling of proton ions are described.
In this paper, the spectrophotometric properties of a colored Nickel-based solution complex (Nickel nitrate hexahydrate and Methyl Orange (MO)) were investigated as a stable chemical dosimeter for using in radiation processing of... more
In this paper, the spectrophotometric properties of a colored Nickel-based solution complex (Nickel nitrate hexahydrate and Methyl Orange (MO)) were investigated as a stable chemical dosimeter for using in radiation processing of agricultural products. Its simple synthesis method as well as low cost made it a suitable dosimeter for use in radiation processing. The variation of absorbed dose was applied to measure the absorbed dose. The maximum absorbance for the solution was observed at 460 nm. This solution was irradiated at three different concentrations of Ni(No3)2.6H2O and MO by Co-60 gamma-ray. Also the variation of the absorbance as a function of PH of the solution was investigated. The results showed the solution absorbance decreases with an increase in doses, and this solution can be used as a routine dosimeter and has a linear response in the 50 to 1500 Gy range with acceptable stability in environmental conditions up to 40 days before and after irradiation.
In this research, the effect of ions produced in deuterium plasma on Tungsten (W) and Aluminum (Al) plates has been investigated using a plasma focus device with the specifications of (C=10.4 μF, V=23 kV, E=2.75 kJ). The W samples used... more
In this research, the effect of ions produced in deuterium plasma on Tungsten (W) and Aluminum (Al) plates has been investigated using a plasma focus device with the specifications of (C=10.4 μF, V=23 kV, E=2.75 kJ). The W samples used because it is one of the key elements in the Tokamak device. Because we wanted to put the W samples at the distance from the anode top with maximum plasma produced ions, we should find the optimum place. Due to the high cost of W samples, we used Al samples to find the optimal conditions. The samples were irradiated at 8 cm distance from the anode top with deuterium ions produced by a plasma focus device. The sample analyses were done by the SEM and EDX methods. The sample irradiation by deuterium plasma ions caused a lot of damages and bubble formation on the sample surfaces. The analyses showed the extent of surface damage and the number of ions deposited on the surface. The number of damages on the Al surface was much higher than W. Bubbles were formed on the surface were due to the impact of deuterium ions on the W and Al samples. Also, the deuterium ion energy was measured with a Faraday cup as about 50 keV.

And 50 more