In this study, natural radioactivity levels (226Ra, 232Th, and 4 K) of some medicinal plant samples with known anti-oxidative properties, which are frequently consumed by animals and humans, were obtained from Ankara province and its surroundings (Mamak, Kızılcahamam, Beypazarı, Kahramankazan, and Polatlı districts) were determined using a thallium-doped sodium iodide NaI(Tl) gamma spectrometry. By using the determined natural radioactivity concentrations in the collected plant samples, the number of radiological doses that people could be exposed by consuming these plants was calculated. As a result of the study, 226Ra, 232Th, and 4 K radioactivity concentration ranges of the plant samples were found be 14.69 ± 1.27–59.08 ± 3.12 Bq kg−1, 1.78 ± 0.04–50.05 ± 2.76 Bq kg−1 and 207.24 ± 34.09–826.13 ± 25.40 Bq kg−1, respectively. The highest 226Ra, 232Th, and 4 K activity concentrations were measured in Astragalus densifolius subsp. ayashensis (Kahramankazan), Astragalus kochakii (Kahramankazan) and Rumex patientia (Patience Dock) (Kahramankazan) plants, respectively. The lowest 226Ra, 232Th and,4 K activity concentration plants were determined respectively as Rumex patientia (Mamak), Lavandula angustifolia (Kızılcahamam), and Astragalus acikirensis (Polatlı). The establishment and routine repetition of environmental radioactivity monitoring programs in each region are important for human and animal health, and the results of this study gain importance for Ankara and its surroundings in terms of environmental health.

An analytical method is used to describe isotope production at an electron accelerator. The key characteristics that determine the total target activity and its distribution have been established. The expressions for the reaction yield depend explicitly on the irradiation regime and parameters of the giant dipole resonance. The model predictions for the bremsstrahlung spectrum and yield of the reference reactions are in good agreement with the results of simulation and experiment.

Radioisotope 94mTc is a favorable positron emitter radioisotope in nuclear medicine that can be used for PET. Production of 94mTc via indirect reaction 92Mo(α, 2n)94Ru → 94mTc was predicted by SRIM, TALYS,and GEANT4 codes. SRIM and GEANT4 codes were employed to calculate the range of alpha particles into enriched 92Mo and to estimate the thickness of the target. Further, the GEANT4 and the TALYS codes were used to investigate the cross-section and production yield of the 92Mo(α, 2n)94Ru reaction. Four model methods were utilized to compute the cross-section by TALYS-1.95 code. The simulated results were compared with experimental values that reported by F.O. Denzler and found to be in good agreement between them. This observation indicates that the SRIM, the GEANT4 and the TALYS codes are suitable tools for prediction of producing radioisotopes via various reactions.

A bilateral comparison to determine the activity concentration of the same 125I solution was organized. As electron-capture radionuclide with a rather high atomic number, 125I must be regarded as difficult to measure. The situation is partly exacerbated by the fact that some established standardization methods, like photon-photon coincidence counting, can no longer be applied due to the unavailability of appropriate equipment and expertise.One aim of this work is to compare modern liquid scintillation counting methods for the standardization of 125I. Both participating metrology institutes have used their custom-built triple-to-double-coincidence ratio (TDCR) counters and the determined activity concentrations are in excellent agreement even though the ways to analyze the data and to compute counting efficiencies were widely independent. The results also agree with the outcome of 4π-γ counting that was carried out at LNHB.In both laboratories, the measurements were complemented by measurements with several secondary standardization methods which even allow to establish a link to the CCRI(II)–K2.I-125(2) comparison started in 2004. A good agreement between the TDCR results and the key comparison reference value of the 2004/2005 comparison was obtained.

Various types of radionuclides have different atmospheric dispersion characteristics, such as buoyancy and gravitational deposition phenomenon of light gas and heavy particles, respectively. Gaussian plume model was widely used to describe atmospheric dispersion behaviors of radioactive effluents, particularly for the purpose of engineering environmental impact assessment or nuclear emergency support. Nonetheless, buoyancy and gravitational deposition were rarely reported in previous work for tritium in particular, which might cause a deviation in evaluating near-surface concentration distribution and radiation dose to the public. Based on the multi-form tritium case, we made a quantitative description for the buoyancy and gravitational deposition phenomenon and discussed the feasibility of developing an improved Gaussian plume model to predict near-surface concentration distribution. Firstly, tritium concentration distribution near to the surface was predicted by using computational fluid dynamics method (CFD) and standard Gaussian plume model to reach consistency without consideration of buoyancy and gravitational deposition effects. Secondly, effects of buoyancy and gravitational deposition were identified by species transport model for gaseous tritium and discrete phase model for droplet tritium with integrating the buoyancy force caused by density variation of gaseous tritium and gravitational force of droplet tritium with enough size. Thirdly, buoyancy and gravitational deposition correction factors were obtained to modify the standard Gaussian plume model. Lastly, predictive results by improved Gaussian plume model were compared with CFD method. It was proved the improved correction method enables higher accuracy in predicting the atmospheric concentration distribution of gaseous pollutants with density variation or particles with gravitational deposition properties.

In the context of age dating of 90Sr, the selective adsorption of zirconium ions from the mixture with strontium and yttrium by adsorbents based on TiO2 with a chemically modified surface was investigated. The general features of the separation process of strontium, yttrium, and zirconium in batch conditions were determined. Inductively Coupled Plasma Mass Spectrometry (ICP-MS) was used to analyze the initial and residual concentrations of the studied cations. Separation of 90Zr and 90Sr from a liquid source containing 90Sr–90Y using adsorbents based on TiO2 was performed for the first time. The ratio of 90Zr/90Sr was measured, and the age of liquid 90Sr–90Y sources was determined. In addition, we studied the age dating of 90Sr–90Y sources using a combination of liquid-scintillation counting of 90Sr and ICP-MS measurement. The results of both methods – the method of age-dating with the chemical separation of isotopes and the combination of LSC and ICP-MS analysis – agree very well and thus serve for cross-validation. Moreover, the combination of the two methods increases the confidence in the age-dating results of 90Sr–90Y sources.

The concentrations of U-238, Th-232, and K-40 measured with a portable gamma-ray spectrometer in cutting samples of two exploratory wells allowed the radiometric characterization of Cretaceous Rancheria sub-basin rocks and the definition of twelve (12) zones according to paleo-redox facies. The Th/Ua (authigenic uranium) and Th/U ratio greater than 7 (seven) indicates paleo-redox conditions related to oxygenation changes and detrital material input during deposition and terrestrial freshwater environment. However, the Lagunitas, Aguas Blancas, La Luna, and Molino formations characterize by facies from sub-oxic (dioxic) to anoxic redox. In the Aguas Blancas and Molino Formations, pyrite and the high uranium measurements determine an anoxic and euxinic environment. The high values of both + uranium and authigenic uranium in the La Luna and Molino formations relate to the conservation of organic matter, which is an essential factor in the generation of hydrocarbons. The abrupt changes of both K/U and Th/U identify potential sequential or genetic limit surfaces (e.g., maximum flooding surfaces), limiting such zones. Based on radiometric behavior, this research permitted identifying eight Cretaceous to Miocene unconformities in the area, three of them reported here for the first time.

The absolute intensity for the 803-keV γ ray of 210Po was evaluated by α-γ coincidence technique. A liquid sample with a known amount of 210Po embedded in scintillation fluid was measured in a coincidence-based system that comprises a Liquid Scintillator (LS) detector and a High-Purity Germanium (HPGe) detector. A photo-reflector assembly that contains the 210Po sample provides 100% efficiency for detecting the α particles. The combination between the HPGe and the LS detectors allows to reject non-coincident α-γ events while maintaining high resolution γ spectroscopy. Consequently, the faint 803-keV photopeak from 210Po could be observed in a background-free environment, and its intensity could be evaluated with good accuracy. Sample measurements were carried out over nine months to gather statistics and verify the reliability of the experimental procedure. The absolute intensity of the 803-keV line was found to be (1.22 ± 0.03) × 10−5, in excellent agreement with the adopted value in a recent data compilation and consistent with previous experimental works.

Boron neutron capture therapy (BNCT) is a promising cancer treatment that uses energetic ions released from 10B(n, α)7Li reactions. Accurate assessment of neutron energy spectra is important for simulation-based evaluation of neutron doses during BNCT. In this study, a proof-of-concept study was conducted for a neutron spectrometry technique that involves the use of a water phantom, which is commonly used for quality assurance in BNCT, as a moderator. The technique involves applying unfolding to the count rate distribution of the thermal neutron counter measured within the phantom to derive the energy spectrum. We performed experiments using a spherical 3He proportional counter in neutron fields generated by 252Cf and 241Am-Be sources. The results demonstrated that the spectrometer reasonably reproduced neutron spectra and showed the potential of using a water phantom as a moderator for such a technique.

Radiation protection is crucial for the safe utilization of ionizing radiation and minimizing the harmful effect upon exposure, hence some standards have been defined by some relevant organizations for the safe uses of radiation. One of the parameters relevant to the calculation of gamma ray shielding is the half-value layer (HVL), which is normally calculated using the knowledge of linear attenuation coefficient (μ). In this research, an attempt has been made to directly calculate HVL without the knowledge of μ via Monte Carlo simulation technique. For this purpose, in the Monte Carlo N‐Particle eXtended (MCNPX) code, F1, F5 and Mesh Popul sequences tallies were defined and the optimal structure for the least measurement error was introduced. The MCNPX calculated values showed reasonable agreement with the experimental findings. According to the obtained results, it is suggested that in order to reduce the error of HVL calculations, in exchange for the MCNPX code, the values of the R parameter and the radiation angle of the source should be considered according to the calculations introduced in this plan. Because the results show that by considering the measurement error between 6 and 20%, the code output can be cited in different energy ranges.

165Er is a pure Auger-electron emitter with promising characteristics for therapeutic applications in nuclear medicine. The short penetration path and high Linear Energy Transfer (LET) of the emitted Auger electrons make 165Er particularly suitable for treating small tumor metastases. Several production methods based on irradiation with charged particles of Er and Ho targets can be found in the literature. In this paper, we report on the study of 165Er indirect production performed via the 166Er(p,2n)165Tm → 165Er reaction at the 18 MeV Bern medical cyclotron. Despite the use of highly enriched 166Er2O3 targets, several Tm radioisotopes are produced during the irradiation, making the knowledge of the cross sections involved crucial. For this reason, a precise investigation of the cross sections of the relevant nuclear reactions in the energy range of interest was performed by irradiating Er2O3 targets with different isotopic enrichment levels and using a method based on the inversion of a linear system of equations. For the reactions 164Er(p, γ)165Tm, 166Er(p,n)166Tm, 166Er(p, γ)167Tm, 167Er(p,3n)165Tm, 167Er(p, γ)168Tm, 168Er(p,2n)167Tm and 170Er(p,3n)168Tm, the nuclear cross section was measured for the first time. From the results obtained, the production yield and purity of the parent radioisotope 165Tm were calculated to assess the optimal irradiation conditions. Several production tests with solid targets were performed to confirm these findings.

In this study, we investigated the impact of chemical etching on the light output and energy resolution of LYSO scintillators using simple, affordable laboratory equipment. We found that etching with phosphoric acid at temperatures between 180 °C and 190 °C improved the light output and energy resolution compared to mechanically polished crystals, even after minimal etching times. Our results show that with 7.5 min of chemical etching, the light output increase rate is 45.7%, and the relative energy resolution improvement is 12%.

The precise measurement of the (n, 2n) reaction cross-sections of xenon isotopes is of great significance for the diagnosis of the Inertial Confinement Fusion fuel area density ρ R. The preparation of xenon samples and how they can be efficiently adsorbed to obtain an optimal reaction cross section are extremely difficult in existing experiments. This work aims to use the 93Nb(n,2n)92mNb reaction cross section as the standard to calculate the cross-sections and isomeric ratio for 134Xe(n,2n)133m,gXe and 136Xe(n,2n)135m,gXe reactions. This work also aims to explore the feasibility of adsorbent selection, the method of adsorption sample and the feasibility of the method used in the laboratory. In this paper, the Geant4 program was used to simulate and calculate the adsorption factors of activated carbon and polymer for standard 152Eu multi-line sources, so as to determine the most suitable adsorption material. Next, the GammaVision analysis software was used to observe whether the characteristic γ-rays released by other isotope reaction products of xenon in the sample and packaging material had an effect on the characteristic γ-rays of the target product. Then, the basic principle of activation method and the relative measurement principle are used to calculate the cross-sections and isomeric ratio for 134Xe(n,2n)133Xe and 136Xe(n,2n)135Xe reactions. Finally, the experimental results are compared and analyzed with the existing data in EXFOR database to verify whether the experimental method is correct.

Since 2001, Nuclear and Energy Research Institute IPEN-CNEN has produced weekly ultrapure iodine-123, using a manual irradiation system, fully developed in IPEN. Iodine-123 radiopharmaceuticals have been produced and distributed to hospitals and clinics of nuclear medicine, where several diagnostic imaging procedures for thyroid, brain and cardiovascular functions are performed. Due to the short half-life and emission of low-energy photons, this radioisotope becomes suitable for diagnosis in children. In the present work, the technical and constructive aspects of a new fully automated irradiation system, dedicated to 123I routine production, employing enriched xenon-124 gas as the target material is presented. This new system consists of a target, a water and helium cooling system, a cryogenic system, an electric power system, and a control and process monitoring unit, composed of supervisory software, connected to a programmable logic controller via personal computer. In this new concept, there is no need for human intervention during radioisotope production, reducing the possibility of eventual failures or incidents involving radioactive material. By using this new system, a specific yield of 2.70 mCi/μAh per irradiation was achieved in validation runs, and after three years of routine production of iodine-123, the system showed reliability and resilience.

A theoretical-experimental study of the interaction of electron beams with 3 filaments conventionally used for 3D printing is presented in this paper. Pieces of polylactic acid (PLA), acrylonitrile butadiene styrene (ABS) and thermoplastic polyurethane (TPU) are studied using Monte Carlo simulation with Geant4 and experimental measurements with plane-parallel ionization chambers and radiochromic films. Using different printing parameters and computed tomography, the presence of air gaps and the uniformity in the bolus density made with the different materials are evaluated. The main parameters in the Percentage Depth Dose (PDDs) are determined, the manufacturing process is standardized and the printing profiles are generated for each of the materials in order to obtain uniform attenuation characteristics in the pieces and improve adaptation to irregular anatomical areas.

The U.S. Environmental Protection Agency established the maximum contaminant level limit for radon concentration in drinking water as 11.1 Bq L−1. A new device based on the bubbling method with a 290 mL sample bottle was designed for intermittent continuous measurement of water radon concentration. A STM32 is used to control the switch of the water pump and the valves. The Water-Radon-Measurement software written in C# is to connect RAD7 and calculate the water radon concentration automatically.

Cancer treatment with protons and carbon ions relies on the property of the accelerated charged particles to deposit most of their energy in the vicinity of their range (around the Bragg peak). The level of hydration in a cancer patient's body may vary within hours. Some patients may be heavy to moderately dehydrated, and some may be well and even excessively hydrated. In this research, we aim to estimate the uncertainty of the protons and C-ion ranges because of the different hydration levels of the human body. For the study of the impact of body hydration level on the particle's ranges, we have designed a new phantom model – a homogeneous mixture of an Average HUuman BOdy constituting elements (AHUBO) in three states of hydration: normal (n), dehydrated (d), and excessively hydrated (e) by applying corresponding recalibration in the “atomic-stoichiometry model” due to the water sufficiency/deficiency. The purpose of the study is to estimate the shift in the ranges depending on the hydration level, possibly suggest particle beam energy adjustments to overcome the range uncertainties, to deliver the prescribed dose to the tumour while sparing the healthy tissue. Herein we present the results of the FLUKA-Flair simulations of the therapeutic range of energies of protons (50–105 MeV) and C-ions (30–210 MeV) respectively, into an AHUBO head phantom model at three levels of hydration (normal, dehydrated, and excessively hydrated). The range uncertainty was estimated via the shifts of the Bragg-peaks position for the three different hydration levels. The estimations showed that the range uncertainty (ΔR) due to body hydration for the maximum energy in the range for protons at 105 MeV is about 0.04 mm and for C-ions at 190 MeV/u is about 0.06 mm.

In this study, 5, 10, and 15 wt% ulexite (NaCaB5O6(ΟH)6·5H2O, hydrated sodium calcium borate hydroxide) and 15 wt %Bi2O3 filled high density polyethylene (HDPE) composite materials were fabricated through conventional melt-extrusion processing techniques in the form of a layered structure in order to absorb both neutron radiations and the secondary radiations resulting from neutron induced reactions. In the layered structure, HDPE was used to slow down neutrons, while ulexite and Bi2O3 were used to capture thermal neutrons and secondary gamma radiations, respectively. The properties of ulexite/HDPE and Bi2O3/HDPE composites were investigated through scanning electron microscopy (SEM), X-ray diffraction (XRD), thermogravimetric analysis (TGA). The mechanical properties (tensile and hardness) of composite materials were also investigated. The results showed that the addition of ulexite and Bi2O3 particles does not change the thermal and mechanical properties of the HDPE composites significantly. Total macroscopic cross-section of the composites was determined using a 239Pu–Be (α,n) neutron source, while their linear and mass attenuation coefficients were determined using a137Cs gamma-ray source. The results show that ulexite filled HDPE composites enhance neutron shielding property and Bi2O3 filled HDPE composites provide good shielding performance for gamma rays.

Off-line gamma-ray spectrometry was used to accurately measure the Cumulative fission product yields (CFPYs) of fission products in the 235U (n, f) reaction induced by 2.8 MeV neutrons. The 2.8 MeV quasi-monoenergetic neutron beam was produced by the CPNG-600 Cockcroft Walton accelerator at the China Institute of Atomic Energy （CIAE）and the gamma spectra were measured by the HPGe γ-ray Spectrometer. After fully considering and revising the sources of uncertainty, high-precision CFPYs of 4 fission products were obtained. This study has important applications in reactor design and operation and is conducive to the establishment of an evaluated nuclear database.

52gMn is a promising radionuclide for positron emission tomography (PET). Enriched 52Cr targets are required to minimize formation of 54Mn radioisotopic impurities during production with proton beams. The need for radioisotopically pure 52gMn, accessibility and cost of 52Cr, sustainability of the radiochemical process, and potential for iterative purification of target materials motivate this development of recyclable, electroplated 52Cr metal targets and radiochemical isolation and labeling with resulting >99.89% radionuclidically pure 52gMn. The run-to-run replating efficiency is 60 ± 20%, and unplated chromium from this method is recovered with 94% efficiency as 52CrCl3 hexahydrate. The decay-corrected molar activity of chemically isolated 52gMn for common chelating ligands was 376 MBq/μmol.