Volume 111, Issue 3

Theoretical predictions were made using TALYS-1.95(G) and EMPIRE 3.2 reaction-model codes for 69 Ge, 67 Ge, and medically used 68 Ge, 67 Ga, 68 Ga, 65 Zn, 64 Cu radionuclides produced in the interaction of α -projectile with 66 Zn-target at 10–40 MeV α -energies. Pearson’s statistical coefficients showed moderate to strong positive correlations between the theoretically predicted and experimentally measured production cross sections for radionuclides with practical medical applications. Furthermore, the present results indicated that a medium-sized cyclotron and a single α  +  66 Zn system (projectile + target system) might be an option for optimized production of 68 Ge, 68 Ga, 67 Ga, 65 Zn, and  64 Cu radionuclides.

In this study, an efficient method for targetry and production of zirconium-89 radioisotope ( 89 Zr) with Cyclone-30 cyclotron was developed. The preparation of a highly pure compressed yttrium oxide target material and design of a target made by copper for better heat transfer was performed. Electrodeposition of target with gold was done to prevent the entry of metallic impurities (copper, zinc and other trace metal elements). Nuclear reaction cross sections for optimization of production with new target and irradiation parameters of the target were evaluated. The prepared 89 Zr in the form of [ 89 Zr] Zr-oxalate had high radionuclidic purity (>99.9%) and a low chemical impurity concentration (<0.1 ppm for copper and zinc elements). The yield of 89 Zr radioisotope production via the reaction of 89 Y(p,n) 89 Zr was measured to be 77 ± 9.5 MBq/μAh (time of irradiation = 3, the current 20–30 µA). [ 89 Zr] Zr-oxalate specific-activity was in the range 2.319641 × 10 4 –3.479443 × 10 4  MBq/mmol of Oxalate.

Significant amount of uranous nitrate is required for reducing Pu(IV) into inextractable Pu(III) for partitioning of U(VI) and Pu(IV) present in the loaded organic phase in PUREX process. Experiments have been conducted for the preparation of uranous nitrate by reducing uranyl nitrate present in nitric acid solution using hydrogen over Pt/SiO 2 catalyst. The effect of process variables such as U(VI) concentration, H 2 pressure, nitric acid concentration, catalyst quantity, temperature, mixing speed, and hydrazine concentration (for stabilizing U(IV)) on the yield of U(IV) was studied. The process variables were optimized by a two-step statistical approach namely design of experiments. The initial screening of process variables and determination of important variables that affect the production of U(IV) was determined by definitive screening design (DSD) methodology. The DSD yielded three variables affecting the U(VI) reduction to a significant extent were U(VI) concentration, H 2 pressure and mixing speed. These significant variables were further optimized using five-level full factorial central composite design (CCD) methodology for understanding the intricate interactions between the variables and the combined effect of all variables at a time influencing the U(VI) reduction. A second-order polynomial equation derived from CCD was subjected to analysis of variance (ANOVA) for estimating the validity of the model and statistical significance of the terms involved in the polynomial. The results revealed that the model can predict the yield of U(IV) generation with 95% confidence in the proposed experimental range.

Adsorption of molybdenum was carried out on two materials; potassium doped γ -Al 2 O 3 (KA) and bentonite iron oxide composite (BIOC) with purpose of using in production of chromatographic 99m Tc generators, for KA absorbent, different ranges of potassium loadings have been considered, for BIOC solid the used weight ratio was (Bentonite/Fe = 2/1). Samples were characterized by XRD and WDXRF. Adsorption of Mo x O y z− species was carried out at pH 2.5 and 25 °C, effects of contact time and pH were studied. The Mo adsorption data onto used materials were well fitted using Freundlich and Langmuir models. Doped alumina samples showed higher adsorption capacities (∼90 mg g −1 ) than the usual alumina used in generator technology (∼60 mg g −1 ), whereas adsorption capacities of BIOC solid were in the range of (∼75 mg g −1 ).

In this study, two separate hydrogel dosimeters were prepared for low-dose measurement. The first system is based on Congo Red (CR) dye and a blend of natural polymers (gelatin/CMC) while the second system is based on 2,6-dichlorophenol indophenol (DCP) blue dye and a hybrid polymer blend (PVA/gelatin/CMC). The sensitivity of Congo red (CR)–(gelatin/CMC) gel dosimeter and 2, 6 – di choro phenol indophenols(DCP)-(PVA/gelatin/CMC) gel dosimetry systems were examined by studying the comparative results between the two different systems through the change in optical properties upon exposure to different gamma radiation doses, which showed the sensitivity of the second system (DCP)-(PVA/gelatin/CMC) to radiation is more than the first system(CR)–(gelatin/CMC) (CR) and (DCP)dyes have absorbance crest at 492 and 622 nm respectively. With the increase of the radiation dose, the decomposition of the dyes increases, and thus the color of the prepared hydrogel changes from orange to colorless in the case of (CR) dye at the dose (0–3.5 kGy) and from blue to colorless in the case (DCP) dye at the absorbed doses ranged from (0–1000 Gy). Dose-response curves, sensitivity, FTIR analysis, and the relative stability of the prepared hydrogel before and after irradiation were studied.

We study the theoretical and simulated shielding properties of 69P 2 O 5 –10Gd 2 O 3 –10BaO–10ZnO–1Er 2 O 3 and 69P 2 O5–10GdF 3 –10BaO–10ZnO-1Er 2 O 3 glasses. We exploit advanced simulation (MCNPX) and Phy-X, an online Photon Shielding and Dosimetry software, in the 0.001–15 MeV energy range, to calculate the linear attenuation coefficient and mass attenuation coefficients of the samples. The obtained mass attenuation coefficient is then used to calculate the critical radiation shielding characteristics, such as the half-value layer, mean free path, tenth value layer, and effective atomic number. The calculated LAC values range from 108.759 to 0.097 and from 84.124 to 0.087 cm −1 while HVL values range from 0.006 to 7.164 and from 0.008 to 8.013 cm, and MFP values vary from 0.009 to 10.335 and from 0.012 to 11.561 cm for ZBaOEr and ZBaFEr, respectively, with energy ranging from 0.015 to 15 MeV. We find that the results obtained from MCNPX and Phy-X are consistent with each other, based on the computed relative difference. The glass sample containing Gd 2 O 3 exhibits better shielding properties than the one with GdF 3 , over the entire energy range. Our examined glass samples are thick enough to substantially lessen the intensity of incident X-rays and gamma rays.

The radiation-shielding attributes of six glass samples with BaO-TiO 2 -SiO 2 -GeO 2 compositions were examined. Using the Phy-X software, the radiation shielding factors for the proposed glasses were reported. The transmission factor (TF) was calculated for various glass thicknesses, and the findings showed that the TF reduces as the glass thickness increases. The transmission of the photon decreased for BTGS55 from 0.83 to 0.49 at 0.284 MeV and from 0.93 to 0.73 at 0.826 MeV as the sample thickness increased from 0.4 to 1.6 cm. The findings of the radiation protection efficiency (RPE) indicated that a thick glass sample absorbs more photons, suggesting that less radiation passes through the glass, improving RPE. Furthermore, the BTGS0 glass has a higher RPE than the BTGS55 glass, indicating that the glass is more effective at blocking incoming radiation when it contains more GeO 2 . The BTGS55 sample has the lowest RPE and the greatest TF, whereas the BTGS0 glass has the highest RPE and lowest TF. The tenth value layer (TVL) was also obtained, and it was found that TVL rises with energy, peaking at 1.33 MeV and varied from 9.437 to 11.519 cm. The minimal TVL, in contrast, is estimated to range between 3.245 and 3.890 cm and it occurs at 0.284 MeV BTGS0 has the lowest TVL after the addition of GeO 2 , which lowers the TVL.

The study of radiation shielding properties for a material is an important part of research in scientific world. It is expressed in terms of the linear linear coefficients (LAC) which can also be used to obtain some other parameters. The gamma shielding properties of glass with addition of boron oxides in different amounts were studied, thereby obtaining linear attenuation coefficients (LAC), mean free path (mfp), Exposure Buildup factor (EBF) which were calculated using Phy-X/PSD code. It was found that increasing boron oxide content of 0, 2, 4, 6, 8 (w %) in glass increased the gamma attenuation coefficients.

In this investigation, the photons shielding factors for tungsten tellurite glasses with the y WO 3 –(90 −  y )TeO 2 –10Na 2 O ( y  = 05, 10, 15 and 20 mol%), were reported. The penetration and attenuation factors for this system at various energies were reported using the Phy-X/PSD program. With increasing energy, it is observed that the linear attenuation coefficient (LAC) values, which range from 1.087–0.234 cm −1 (for TWN1) to 1.354–0.248 cm −1 (for TWN4), decrease exponentially. The LAC values were found to increase with the addition of WO 3 from 1.087 to 1.354 cm −1 at 0.245 MeV and from 0.515 to 0.586 cm −1 at 0.444 MeV. Additionally, the greater potential for photon interactions at higher WO 3 concentrations was indicated by the findings of the effective atomic number ( Z eff ) calculation. According to the obtained results, the maximal Z eff occurred at 0.284 MeV, which is equivalent to 32.53 for TWN1 and 36.89 for TWN4. The half value layer (HVL) for the samples under consideration between 0.245 and 1.458 MeV has been determined using the Phy-X/PSD. The HVL results demonstrated that at 0.284 MeV, more gamma rays are shielded whereas the potential of photon shielding decreases as energy increases. The tenth value layer (TVL) increased with rising energy and decreased with rising WO 3 concentrations. TVL for TWN4 is the lowest (1.701 cm at 0.245 MeV and 9.284 cm at 1.458 MeV).

Hookah smoking has become very popular in Iraq among women and men. Hookah tobacco contains natural radioactive elements, such as radon, radium, and uranium, as well as toxic elements, such as polonium, which are released during the combustion of tobacco and are inhaled by smoking. Most reviews focus on hookah tobacco, and only a few have investigated the blood of hookah smokers. In this study, a CR-39 detector was used to measure radon, radium, and polonium concentrations and conduct risk assessments in female hookah smokers of different ages. The results show that the concentrations of radon-222, polonium-218, and polonium-214 varied between 61.62 and 384.80, 5.45–33.64 on the wall of the can, and 2.43–15.00 Bq/m 3 on the surface of the detector, respectively. The effective radium-226 concentration varied between 4.52 and 56.31 Bq/kg. The absorbed effective dose varied between 1.55 and 9.71 mSv/y, which is within the recommended limit (3–10 mSv/y) by International Commission on Radiological Protection (ICRP). The average case of lung cancer 107.91 cppp, which exceeds the European Union (EU) limit (96.9–104.8 cppp). The rates of radon activity and radon exhalation from the intake of a natural radionuclide due to hookah smoking in a female’s blood were calculated and discussed. This study aimed to establish preliminary results on the risks of radioactivity concentrations and assess the dose in the blood of women who smoke hookah and assess the possibility of developing cancer.