Abstract At present, Maximum Temperature Crystal Sensors (MTCS) are of great interest in the design and optimization of engines and turbines with high efficiency. These sensors are made in very small dimensions and can measure the maximum temperature in the desired location, for example inside the turbine. It is worth mentioning that a high fluence of fast neutrons is required to produce these sensors. The MTCS method is based on the high temperature phenomenon to restore changes made in the crystal lattice caused by neutron irradiation. Calculations of the maximum fast neutron flux for the diamond sample were performed using MCNP6 with the simulation of the core of the Tehran Research Reactor (TRR). Activity calculations were done using ORIGEN2, dose rate calculations through MCNP6, and temperature calculations with ANSYS software at different points of the sample. The highest fast neutron flux was calculated, and the activity of the sample decreased significantly after 40 days, with the dose rate at a distance of one meter reaching 1.4 µSv/h. Temperature calculations showed that the maximum temperature at the lowest speed of the channel was 402 K, which would not cause any thermal problems. Additionally, using the X-Ray diffraction (XRD) method, the lattice number and unit cell volume were calculated for the sample. The results obtained suggest that the diamond sample, from the perspective of neutronics, dosimetry, thermohydraulics, and changes in crystal lattice structure, can be recommended as a crystal sensor in the Tehran Research Reactor.

Abstract In clinical trials, Computed Tomography (CT) is widely used for diagnosis and treatment guidance. With the increasing use of CT in clinical practice, the issue of high radiation dose has become a significant concern. One way to reduce the dose in CT is by utilizing sparse view imaging. However, sparse view imaging often leads to artifacts in the reconstructed images due to the lack of data. This paper aims to examine and evaluate image reconstruction methods to introduce effective algorithms for sparse view studies. Common image reconstruction algorithms such as Maximum Likelihood Expectation Maximization (MLEM), Algebraic Reconstruction Technique (ART), and Filtered Back Projection (FBP) were reviewed. FBP and MLEM algorithms perform well when there is complete data, but due to the high speed of the FBP algorithm, it is best suited for such cases. However, when data is limited, FBP performs poorly, leading to a comparison between the ART and MLEM algorithms. The results indicate that MLEM performs better in sparse view studies. Quantitative parameters such as Peak Signal-to-Noise Ratio (PSNR), Root Mean Square Error (RMSE), and Structural Similarity Index (SSIM) were evaluated to assess the results. The findings suggest that FBP and MLEM algorithms perform better when data is complete, while MLEM algorithm excels in sparse view studies.

Abstract When performing reactor core neutron calculations, the main cost lies in cell calculations to generate macroscopic cross-sections of fuel assemblies (FAs). The development and utilization of a cross-section library based on the reactor type is the primary procedure to reduce calculation costs. In some cases, when there is no library available for specific analyses, it becomes necessary to conduct cell calculations. Examples of such cases include investigating FAs vibration neutron noise or coolant asymmetric distribution around FAs due to bowing. In this study, an Artificial Neural Network (ANN) was used instead of high-volume cell calculations to significantly reduce the time required for calculations in the VVER-1000 reactor. For criticality calculations at Beginning of Cycle (BOC), the calculation time was reduced from 42 minutes using 32 computational cores to just 1.3 minutes. Additionally, the results achieved were of acceptable accuracy, with the difference in the critical boric acid value obtained using ANN compared to cell calculation as a reference being less than 1%, and the power difference being less than 5%.

Abstract In this paper, quantum phase transitions and nuclear structure were investigated based on quantum entanglement. One of the measures for entanglement between two bodies is the von Neumann entropy, which is considered a suitable criterion for examining entanglement. The entanglement entropy of s-d bosons within the framework of the Interacting Boson Model (IBM) is investigated using the consistent-Q formalism and semiclassical approximation. This research presents a method for deriving the entanglement entropy in the dynamical symmetry limits of the IBM, including transition regions between different shapes. By utilizing the entanglement effect, the quantum phase transition was described, and the entanglement behavior of s and d bosons was investigated in various regions of the Castan triangle. A method for determining entropy in the IBM model using the semiclassical approximation was presented. The results showed that entanglement entropy values are sensitive to phase changes and can be a powerful tool for detecting quantum phase transitions in nuclei.

Abstract The Talmessi deposit is located approximately 30 km west of Anarak city and 200 km east of Isfahan, in the Anark district within the structural zone of Central Iran. The main host rock for mineralization in the Talmessi deposit is andesite to andesite-basalt. Mineralogical studies conducted using a field emission electron microscope indicate the presence of at least four stages of mineralization in the Talmessi deposit. The first stage involves fissure-filling copper sulphide mineralization, which is related to Eocene volcanism. The second stage presents itself in the form of nickel-copper vein arsenide mineralization, with the primary mineral being nickeline, accompanied by minor amounts of chalcocite and pyrite. In the third stage, nickel, cobalt, and copper arsenides (such as domeykite, rammelsbergite, safflorite, skutterudite, koutekite, and kutinaite) are deposited, along with minor amounts of copper sulfide (chalcocite), as well as uranium minerals including uraninite, brannerite, and coffinite. During the final stage, as oxidant fluids infiltrate, the minerals from the previous stages are replaced by nickel arsenates (such as xanthiosite), copper arsenates, and secondary uranium minerals. Primary uranium minerals undergo a transformation into secondary uranium minerals, such as uranium arsenate (chadwickite), copper-uranium arsenate (cuprosklodowskite, zeunerite), calcium-uranium silicate (uranophane), uranium carbonate (rutherfordine), calcium-uranium carbonate (sharpite), and calcium-uranium phosphate (autunite).

Abstract The anodic pulse oxidation method has been utilized for the decontamination of 7075 aluminum. This method involves the application of high-intensity electric current for a short period, effectively removing oxides and other impurities from the alloy's surface. The results indicate that by adjusting parameters such as frequency, voltage, and time, precise control over decontamination can be achieved, leading to enhanced mechanical properties and surface quality of aluminum 7075. Consequently, the anodic pulse oxidation method proves to be a rapid and efficient technique for decontamination and property enhancement of aluminum alloys. This method can serve as a dependable and effective solution across various industries, particularly in nuclear industries.

Abstract In the research conducted on the potential use of the Tehran Research Reactor (TRR) for domestic fuel testing, a compact core configuration has been developed. This configuration includes a central neutron flux trap channel to achieve the desired linear heat rate for fuel testing. In this study, a safety analysis of loading the fuel irradiation capsule in the neutron flux trap channel is performed from a thermal-hydraulics perspective under steady-state conditions. The differences in coolant channels between the irradiation capsule and adjacent fuel plates cause a change in coolant flow distribution in the core after loading the fuel irradiation capsule in the neutron flux trap channel. The study investigates the effect of loading the fuel irradiation capsule on heat removal from the fuel plate adjacent to the neutron flux trap channel, which is the hottest fuel plate in the core. ANSYS Fluent software is used for the analysis, allowing for simultaneous simulation of rod-type fuels within the irradiation capsule and the fuel plates adjacent to the neutron flux trap channel. The simulation results show that the maximum clad temperature of the fuel plates adjacent to the irradiation capsule will be approximately 370K, which is lower than the saturation temperature under the reactor operating pressure and below the maximum permissible clad temperature to avoid corrosion. The maximum fuel temperature is 375.9K, which falls within permissible limits. These results indicate safe operation of the TRR core from a thermal-hydraulics perspective if the fuel irradiation capsule is loaded in the flux trap channel.

Abstract In this paper, aluminum 5052 (AA 5052), one of the candidate alloys for nuclear fuel cladding, has been selected for study regarding its corrosion behavior in an electrolyte containing NaCl and Na₂SO₄. Potentiodynamic polarization and electrochemical impedance methods were used to investigate the corrosion of AA 5052. The morphology and characterization of corrosion products were carried out using Scanning Electron Microscopy (SEM), X-ray Diffraction (XRD), and Fourier-Transform Infrared Spectroscopy (FTIR) methods. In the Na₂SO₄ electrolyte, the corrosion resistance of the sample improved over time, with the corrosion current decreasing from 10^-5.4 mA/cm² after 1 hour to 10^-6 mA/cm² after 11 days. AA 5052 showed a wider passivation potential range (ΔE_pass) and less passivation current (i_pass) after 11 days. In the NaCl electrolyte, pitting corrosion, elimination of the passivation zone, and an increase in the number and size of pits were observed with increased time. These results confirm that the time has an influence on the passivation behavior of AA 5052, primarily dependent on the chemical composition of the electrolyte (presence or absence of chlorine and sulfate anions). Additionally, the application of tensile stress showed a destructive effect on the passive layer, as the passivation behavior of AA 5052 was eliminated.

Abstract Liquid-liquid extraction is one of the most common methods for purifying thorium from multicomponent aqueous solutions. In this study, the ability of different organic solvents, such as tributyl phosphate (TBP), di-2-ethylhexyl phosphoric acid (DEHPA), and di(2,4,4-trimethylpentyl) phosphinic acid (Cyanex 272), to separate thorium from a multicomponent aqueous nitrate solution containing uranium, vanadium, and iron elements was investigated. The experimental results showed that the extraction efficiency of elements for tributyl phosphate and di-2-ethylhexyl phosphoric acid is uranium > thorium > iron > vanadium, and for Cyanex 272 is thorium > uranium > vanadium > iron, respectively. Additionally, the separation factor of thorium compared to uranium, vanadium, and iron by Cyanex 272 was calculated as 63.4, 384.4, and 615.6, respectively. The investigation of various process parameters on the solvent extraction process revealed that at a contact time of 15 minutes, a pH of the aqueous phase equal to 0.5, a ratio of aqueous to organic phase equal to 1, and a concentration of Cyanex 272 equal to 0.1 M, the highest percentage of thorium extraction is achieved, with extraction efficiencies of thorium, uranium, vanadium, and iron at 98.6%, 37.8%, 10.6%, and 11.1%, respectively. Furthermore, different mineral acids, including nitric acid, sulfuric acid, and hydrochloric acid solutions, were evaluated for the stripping process from the loaded organic phase. The results demonstrated that utilizing a 4 M sulfuric acid solution could achieve thorium stripping efficiency above 91% from the loaded organic phase.

Abstract Bakherz Mountain is situated in the northeast of Sangan city, at the eastern end of the Darone fault. Iron in this area has been extensively formed under the influence of granitoid masses from the upper Eocene to the lower Oligocene within the metamorphosed sedimentary unit in the form of skarn and hydrothermal deposits. Structural investigation of veins, fractures, movement mechanisms of faults, and displacement of units indicates two types of compressive tectonic regimes along N070 and N350 directions. The initial stage of mineralization occurred in the northeast-southwest direction in the Ferezeneh fault zone (N040). In the northern and middle parts of this fault, two shear-tensile zones were created, leading to the rise of iron-bearing fluids and hydrothermal iron mineralization in the northern part (Ferezeneh mine) and skarn-type iron mineralization in the middle part (Madan-Joe). Subsequently, the intersection of the Ferezeneh and Som-Ahani faults (N300) resulted in the creation of a shear-tensile zone and the ascent of iron-bearing fluids, leading to skarn-thermal iron mineralization in the Som-Ahani mine. Lastly, in the southern part of the Ferezeneh fault, uranium has been deposited in a fault zone approximately 2 km in length, within the Skarn unit. Changes in the geo-structural regime and the creation of suitable spaces in the shear-tensile zones have facilitated the flow of fluids, skarn formation, and the deposition of iron and uranium in separate phases and different directions. Structural studies will aid in the optimal extraction of mineral reserves and the identification of new resources.

Abstract This paper aims to produce uranium with 20% enrichment for research reactor fuel directly from natural uranium with waste xw≤0.3% through a cascade. This purpose has utilized five strategies, including single-optimized tapered, square, 2-section squared-off, 3-section squared-off, and 4-section squared-off cascades. To design and optimize the cascades, five software codes, "STC-PSOA," "SQC-PSOA," "2SQC-PSOA," "3SQC-PSOA," and "4SQC-PSOA," have been developed, with the PSO optimization algorithm employed for optimization. In the first strategy, it is not possible to achieve uranium with a 20% enrichment level using one arrangement and one step. The maximum product enrichment level attainable is 3.73%. However, in the other four strategies, uranium with a 20% enrichment level can be produced in two separation steps with a single arrangement. Notably, when employing the strategy of using 4-section squared-off cascades, the product production is 11.1%, 3.1%, and 1.0% higher compared to square, 2-section, and 3-section squared-off cascades, respectively.

Abstract Mutation breeding has proven to be efficient for developing new cultivars adapted to normal and stress conditions, including salinity. Mutant populations developed through gamma irradiation of Tarom Mahalli, Hassani, and Anbarboo local rice varieties were evaluated and screened for agronomic traits for 10 years in saline paddy fields of Mazandaran. These traits included plant height, early maturity, yield, yield components, and relative tolerance to salinity stress. The salinity levels of both soil and irrigation water varied between 4-11 and 2-9 dS/m, respectively, in different years and locations. The tolerance indices of the mutants were either 3 or 5, with 13 mutants showing lower stress sensitivity indices than their respective parents. Most mutants reached 50% flowering in fewer days than their parents, with selected mutants (2310, 2212, and 133) flowering between 80 and 90 days after sowing. The average plant height of selected mutants (M5-M7) in the saline paddy fields of Mazandaran ranged between 110-130 cm. All selected mutants produced a higher number of panicles per hill compared to Tarom Mahalli. The average yield of rice mutants varied between 3000 and 4500 kg/ha in different locations and years, with mutants 2212 and 2310 showing the highest yields in Bahnamir and Fereydoonkenar. These mutants exhibited at least a 35% increase in yield compared to Tarom Mahalli. The amylose percentage of the mutants ranged between 18-23%. Based on these results, mutants 2212, 2310, and 133 were selected for cultivation in paddy fields with saline or brackish water.

Abstract In this study, the design and construction of the neutron velocity selector were investigated. Theoretical and experimental studies have shown that it is preferable to construct the blades and the main shaft of the device from carbon fibers to reduce weight, prevent imbalance, and improve performance at high speeds. After designing the other components of the device, manufacturing the parts, and assembling the device, the vacuum system was tested, measuring a vacuum pressure of 3×10^-2 millibars.  In the next phase, the device was mechanically tested at various speeds between 3000 and 11000 rpm, confirming the mechanical stability of the device. The measurement of selected thermal neutrons was carried out using neutrons from a ²⁵²Cf source. The experimental results show that this device is suitable for measuring the energy of neutrons in the range of less than 0.001 eV. This device requires a high-speed motor with a rotational speed of approximately 35,000 rpm and a high-resolution encoder to allow the passage of thermal neutrons with an energy of 0.025 eV, in order to reduce the error in time-of-flight measurements to an acceptable level.

Abstract Today, phytoremediation is recognized as one of the most cost-effective and environmentally friendly methods for removing pollutants from soil. This study focused on investigating the phytoremediation capabilities of Conyza canadensis (L) Cronq in removing uranium from saline soil. Soil samples were collected from the nuclear area of Isfahan, with contamination levels of 100, 300, and 500 mg/kg of uranium. The seeds were germinated in a suitable substrate and then transplanted into pots containing the contaminated soil. The plants were then cultivated in a greenhouse for one month, and the amount of uranium accumulation in various parts of the plant was determined using the ash method. Results indicated that as the concentration of uranium in the soil increased, so did the accumulation of uranium in the plant. The study revealed a direct correlation between the amount of uranium accumulation in the plant and the soil's texture and electrical conductivity (EC). The highest accumulation was found in the plant's stem, reaching 3633 mg/kg. Conyza canadensis (L) Cronq demonstrated the ability to accumulate over 1000 mg/kg of uranium in different parts of the plant, with transfer factors exceeding one. Therefore, Conyza canadensis (L) Cronq shows promise as a suitable plant for phytoremediation of saline soils.

Abstract Exploring and mapping radioactive environments presents potentially challenging and risky tasks. To ensure safety, the use of robots can help mitigate these risks. However, deploying a single robot may not be enough to effectively cover the entire area. Therefore, the technique of using multiple robots or a robot swarm can play a significant role in reducing time and improving the exploration and mapping process. Additionally, multi-robot systems have high reliability, and if a number of robots fail due to reasons such as radiation exposure, the remaining robots can continue the mission. This article focuses on the use of several autonomous robots simultaneously for radiation mapping and the discovery of radioactive sources. The study has examined various approaches, including the effect of the number and absorbed dose of robots. The results indicate that increasing the number of robots can enhance the speed of exploration, but the rate of increase is lower than the rate of the robot number due to the crowding of robots. To achieve optimal exploration, it is necessary to utilize more advanced models, such as movement patterns inspired by nature. This paper presents valuable insights into the coordination of collective robots for searching in radioactive environments, which could open up opportunities for the application of swarm intelligence in nuclear scenarios.

Abstract Gallium-68 (⁶⁸Ga) is a radionuclide widely used for diagnostic purposes in Nuclear Medicine. With the availability of ⁶⁸Ge/⁶⁸Ga generators, the radiopharmaceutical industry is currently focused on developing this generator. The development and application of ⁶⁸Ge/⁶⁸Ga generators remain a hot topic in nuclear medicine. In recent years, Pars Isotope Company in Iran has produced ⁶⁸Ge/⁶⁸Ga generators (Pars-GalluGen generator) that are eluted with 5 ml of 0.6 M hydrochloric acid. Additionally, Pars Isotope has developed the second generation of Pars-GalluGen generators. This new generator is based on a SnO₂-TiO₂ resin matrix and is eluted with 3 ml 0.1 M hydrochloric acid. This research aimed to determine the main characteristics of the second generation of Pars-GalluGen generators and evaluate their quality control. The results obtained indicate that this generator meets all specifications of the European Pharmacopoeia for use and has better performance characteristics, including higher chemical, radiochemical, and radionuclide purity, as well as a higher elution yield compared to the first generation.

Abstract The HTR-10 reactor is an experimental Pebble Bed Reactor (PBR), where simulating the neutron flux is crucial due to the reactor's multiple advantages, including safety, high efficiency, and modularity. As a result, studying and researching such reactors is a significant focus for countries with nuclear technology, especially for commercialization. Since both the fuel and moderator are in the form of spherical pebbles and are randomly distributed in the core, accurately evaluating the neutron flux is vital. This evaluation involves calculating the minimum height required for the reactor’s criticality, based on the specific arrangement of fuel and moderator. In this study, a new method for arranging the fuel and moderator pellets, while maintaining a ratio of 43:57, has been proposed. The calculations were performed using the MCNP code. Helium gas was chosen as the coolant, and changes in reactivity due to the insertion of control rods were also calculated.

Abstract The timely detection of human immunodeficiency virus (HIV) at the point of care (POC) enables prompt medical intervention, and appropriate treatment, and helps reduce the growth rate and spread of the disease. In this paper, two HIV-1 biosensors are proposed: one based on a perfect 1-D photonic crystal (PC) and the other based on a 1-D PC containing a defect layer. The main characteristics of these biosensors, including sensitivity and Q factor, are analyzed. Additionally, the effect of the irradiation angle on the performance of these two sensors is evaluated. The results show that the sensitivity of the HIV-1 biosensor based on the perfect photonic crystal is independent of the irradiation angle. Meanwhile, the Q factor of this sensor increases with an increase in the irradiation angle, with the Q factor at an angle of 85 degrees being nearly 14 times greater than that at normal incidence. On the other hand, both the sensitivity and Q factor of the biosensor containing the defect layer are dependent on the irradiation angle. The maximum values of these parameters occur at irradiation angles of 65 and 47 degrees, with corresponding values of 1056 nm/RIU and 12683, respectively. Therefore, the biosensor with the defect layer is recommended as a suitable candidate for HIV detection.

Abstract Energy security is one of the most important factors in the sustainable development of societies. Fulfillment of goals such as economic development, increasing social well-being, and improving the quality of life depend on providing energy and guaranteeing stable and reliable, clean, and affordable resources. In recent years, the use of renewable resources has received a lot of attention to diversify their energy portfolio, reduce the harmful effects on the environment, and achieve sustainable energy. From this point of view, with the proper establishment of the nuclear fuel cycle, nuclear energy can be considered as renewable and sustainable energy in the energy portfolio and compensate for the lack of non-renewable energy in the future. The purpose of this paper is to explain the importance of a closed fuel cycle in achieving sustainable and renewable nuclear energy. Also, due to the not-very favorable situation of uranium and thorium resources in Iran, the lack of extensive international cooperation, the existence of sanctions in the provision of raw materials for nuclear fuel production, and the impossibility of guaranteeing the supply of fuel needed for nuclear energy production, the need to pay more attention to establishing a closed fuel cycle and using fast breeder reactors is proven. Paying attention to this important issue will achieve goals such as reducing the need for natural uranium, reducing the cost of nuclear waste, improving the efficiency of nuclear power plants, and producing materials required for the use of radiation in industry, medicine, and health.

Abstract The assessment of the irradiation behavior of nuclear fuel and structural materials is carried out by conducting post-irradiation examination at the hot cell facility. One of the challenges in performing this examination is the safe transfer of cask containing irradiated nuclear fuel and structural materials from the reactor to the hot cell. Therefore, in this study, the safety assessment of the lead cask transfer of a fuel plate, a zirconium fuel cladding, and oxide fuel pellets (UO₂) against the drop accident was conducted using finite element analysis with ANSYS software. The drop of the cask from a height of 9 meters in vertical, horizontal, and inclined directions is one of the most common accidents defined in IAEA Safety Standards No. 6 (SSR-6). In this study, the yield stress of the materials inside the cask was extracted from the references and compared with the maximum stress resulting from the finite element simulation results. Therefore, the safety acceptance limit after an accident was presented as determining the safety factor for each component inside the cask. The simulation results showed that the cask contents are particularly susceptible to serious damage in the drop accident of a 9-meter, especially vertical and horizontal drop. Therefore, a corrective action was taken by incorporating a shock absorber made of polyurethane foam under and around the cask. Finally, a re-simulation with the presence of the shock absorber demonstrated the safety acceptance of the cask against the drop accident.