نفیسه سالک

In this study, graphene oxide (GO) adsorbent was prepared by modified Hummer’s method and functionalized with aminomethylphosphonic acid. Its application to the adsorption of Sr ions from aqueous solutions in a batch sorption process was investigated. Adsorbents were characterized by TEM and FT-IR. The TEM images of functionalized graphene oxide showed that graphene sheets are formed in wavy sheets. The FTIR spectrum revealed that graphene oxide was highly functionalized with aminomethylphosphonic acid and oxygenated functional groups were reduced. Response surface methodology investigates pH, adsorbent dosage, and temperature parameters. The quadratic model corresponds well with the experimental data. The results illustrated that there is no systematic error in the experiments. The verification of the model with random experiments showed a low error in the values predicted by the model. The kinetic data were analyzed by Pseudo-first-order, Pseudo-second-order, and Double- exponential kinetic models, and experimental data were well modeled by the pseudo-second-order kinetic model. The research results showed that graphene oxide functionalized with aminomethylphosphonic acid has a good ability to strontium adsorption from aqueous solutions.

In this paper, we have tried to provide details about possible radioactive wastes produced in the fission molybdenum-99 production process using LEU uranium targets in the country and how to manage them throughout the production process. Potential solid, liquid, and gaseous wastes are a combination of low-level (LLW), and medium-level (ILW) wastes that contain enriched uranium and other fission products and actinides. Based on the results of radiation calculations and heat generated by the radionuclides in each waste package, it seems that no high-level waste (HLW) is generated from the 99Mo production facility, as this waste contains long-lived or alpha-emitting radionuclides with more activity of 400 Bq/g which their heat generation is not more than 2 kW/m3. The economic aspects of 99Mo-related waste management are important and can substantially influence the final price of 99Mo and by-products. In the fission 99Mo production process, the recovery of unburnt uranium is highly recommended, preferably the uranium should be segregated from the waste streams within the production facility.

Neutron radiography (NRG) is a non-destructive imaging technique for image generation using neutron radiation. In this paper, feasibility studies of neutron radiography for IECF were performed by Geant4 Monte Carlo code. The effects of the different thicknesses of lead on the image and the detection possibility of cavities with different sizes inside a thick lead have been investigated to understand the performance of the device for neutron radiography purposes. The quality of the images was evaluated in terms of contrast. The simulation results showed the efficiency and limitations of NRG for IECF devices and the potential areas where NRG can be performed.

Given to its favorable radionuclidic characteristics, such as tl/2 = 1.93 days and Eβ(max) = 0.71 MeV and ease of its large-scale production using medium flux research reactors, 153Sm is an attractive radionuclide for various therapeutic applications. 153Sm is a beta and gamma emitter that due to its physical and radiochemical properties is ideal for the treatment of bone metastasis. In this study, 153Sm was produced by thermal neutrons irradiation of 152Sm in the Tehran research reactor with a flux of 2.5 ×1013 n / cm2.s. A two-step electrolysis was used to separate 153Sm with separation efficiency of 88.25% from europium in lithium citrate medium by optimization of voltage (6V), electrolyte concentration (0.2 M), pH (6) and time (30 min).

Investigation and monitoring of drinking water pollution is an important parameter for public health and environmental health studies and it is necessary to study it from different aspects. One of the most important aspects is to study the presence of various radionuclides in the water that emit alpha, beta, and gamma rays when their nuclei decay. The aim of this study was to investigate the radiological quality of water by measuring the concentration of gross alpha and beta activity in water resources of Hormozgan province. For this purpose, screening of 30 water samples from different water sources of Hormozgan province was performed by liquid scintillation method using a 1220 Quantulus liquid scintillation counter (LSC). The results of the analysis showed the activity concentration of gross alpha ranges from 48 mBq/l to 296 mBq/l with a mean of 166.2 mBq/l. the activity concentration of gross beta in the samples ranged from 65 mBq/l to 729 mBq/l with a mean of 460.3 mBq/l. The results showed that the gross alpha and beta radioactivity concentration in all waters was lower than the standard limit of the World Health Organization (WHO) and the Institute of Standards and Industrial Research of Iran, which is 500 mBq/l and 1000 mBq/l, respectively also waters are radiologically healthy.

The Holmium-166 radionuclide is one of the most effective radionuclides used to treat bone marrow cancer and rheumatoid arthritis. Among the recommended radionuclides used in radiation synovectomy, 166Ho has got much attention due to suitable decay properties such as short half-life, its high beta energy, gamma-ray emission with suitable energy for nuclear imaging, and the possibility of large-scale production in medium flux reactor. One method to deliver 166Ho to the target tissue is via the 166Dy/166Ho-Chitosan in vivo generator. Compared with other similar radiopharmaceuticals, using the in vivo generator to deliver 166Ho, causes minimal non-target tissue exposure and increased absorbed dose in the target tissue. In this work, the absorbed dose of 166Dy/166Ho-Chitosan radio-complex for radio-synovectomy purposes was calculated by GEANT4 and MCNPX. The obtained results were compared with each other. In addition, the dosimetry results of the mentioned radio-complex have been compared with the common radio complexes used for radio-synovectomy.

Holmium-166 is one of the most effective radionuclides used for treatment of bone marrow cancer and rheumatoid arthritis. Using 166Dy/166Ho in vivo generator to deliver 166Ho to the target tissue causes the minimum exposure to the non-target tissues and increased absorbed dose rate in the target tissue compared to other similar radiopharmaceuticals. In this work, theoretical calculations of 166Dy/166Ho in vivo generator production in Tehran Research Reactor were performed through solving the decay chain differential equations simultaneously for natural and enriched dysprosium targets and the obtained results were compared with experimental values. The results showed that there is a good compatibility between the measured values and the results of the theoretical calculations. The outcomes of this study are used to produce 166Dy / 166Ho radionuclide generator.

The peptides radiolabeled with 177Lu are including a wide range of promising radiopharmaceuticals. The NCA form of 177Lu provides a high specific activity that is required in targeted receptor therapy in order to avoid saturation of the targeted receptors. In this study, the outstanding quality of NCA radionuclides especially in the case of 177Lu was evaluated by comparing with CA radionuclide for radiolabeling of DOTATATE peptide and. The NCA 177Lu and CA 177Lu were obtained by chromatography separation of irradiated 176Yb (enrichment: 96.4%) and irradiation of 176Lu (enrichment: 95%) respectively. Radiolabeling of DOTATE with CA and NCA 177Lu was carried out in the same conditions and time. Then, radiolabeling yield was calculated as a function of some parameters like the Ligand: metal ratio and stability of compound after production. The results showed the superiority of NCA 177Lu with respect to CA 177Lu. Not only the particle ratio showed the advantages of the NCA radionuclide, but also the long-term comparison over a period of 14 days revealed the inferior character of CA.177Lu with regard to the high requirements regarding the specific activity set by radio-immunotherapy.

<p>177Lu owing to its favorable nuclidic characteristics, such as tl/2=6/73d  and Eβ(max)= 497 keV and ease of its large-scale production using medium flux research reactors, is an attractive radionuclide for various therapeutic applications. In this study 177Lu with  the radionuclide purity of  >99.9% was obtained by thermal neutron bombardment (4×1013n/cm2.s) of 176Yb target through the 176Yb(n,γ) 177Yb  177Lu process. The method of separation of 177Lu from macro amount of Yb target was based on extraction chromatographic (EXC). The extractant used in resin was di(2-ethylhexyl) orthophosphoric acid (HDEHP). Finally, (177Lu) with a specific activity of 230.1 MBq (6.22 mCi) was prepared without the addiation of any carrier radionuclide. The process provides a separation yield of 75% 177Lu and a decontamination factor of 2000. The whole process to prepare NCA 177Lu takes almost 1.5 hours. The resultant NCA 177Lu can be used for preparation of NCA 177Lu labeled radiotherapeuticals.</p><p> </p>