• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.92-96
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• 1986

The $^{93}Nb(n,n\alpha)^{89m}Y$, $^{93}Nb(n,{\alpha})^{90m}Y$ and $^{93}Nb(n,2n)^{92m}Nb$ cross sections at a neutron energy of 14.6 MeV have been measured relative to the $^{27}Al(n,p)^{27}Mg$ and $^{27}Al(n,{\alpha})^{24}Na$ cross sections. A small accelerator utilizing $T(D,n)^4He$ reaction was used as a neutron source and the neutron energy spread is about 0.4MeV at the sample. All induced activities were measured with a 70cc HPGe detector in the same geometry.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.331-341
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• 2016

To evaluate the physicochemical and radiological properties of raw materials and by-products in domestic distribution, about 220 samples with 16 species were prepared. We measured the energy spectrum and the chemical content, such as U, Th, and K, using a $LaBr_3$ scintillation detector and ED-XRF. In addition, HPGe detector was used to analyze the radioac-tivity of $^{234}Th$, $^{234}mPa$, and $^{214}Bi$ in uranium decay series and $^{228}Ac$, $^{212}Pb$, and $^{208}Tl$ in thorium decay series, and $^{40}K$. The correlation between characteristic variables, such as the count rate in several ROIs, chemical content, and radioactivity, was assessed to infer the radioactivity of natural radionuclides through a rapid screening method. Based on the results, a characteristic database for raw material and by-product in domestic distribution was established and it will provide useful information in the analysis procedure and improve the accuracy and reproducibility in the analysis of natural radionuclides.

• Journal of Radiation Protection and Research
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• v.36 no.4
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• pp.200-208
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• 2011

The neutron fluence measurement and evaluation technology is very important for material irradiation test. The most essential technology in this study is the neutron irradiation evaluation method using a fluence monitor. The fluence monitors were fabricated with metal wires of the purity ${\geq}$ 99.9%, whose dimensions were 0.1mm diameter, about 3 mm length, and around 150-200 ${\mu}g$ mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. After irradiation tests, radiation activities were measured with the high purity germanium (HPGe) detector. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1781-1795
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• 2024

This study investigated the levels of radioactivity in soil surrounding a phosphate fertilizer factory in Egypt, aiming to assess potential risks to the population exposed to radiation. Concentrations of ²³⁸U, ²²⁶Ra, ²³²Th, and ⁴⁰K were measured in soil samples collected from two subsites: one near the factory (subsite 1) and another further away (subsite 2). Two different systems were used for measuring radioactivity, a high-purity gamma ray spectroscopy system with an HPGe detector for gamma-emitting isotopes and a CR-39 solid nuclear track detector for alpha-emitting radon gas. Subsite 1, located close to the factory, displayed significantly elevated levels of ²²⁶Ra compared to global background levels (514 and 456 Bq/kg vs. 35 Bq/kg). Additionally, the concentrations of ²³⁸U (241.06 Bq/kg vs. global average 35 Bq/kg), ²³²Th (16.15 Bq/kg vs. global average 30 Bq/kg), and ⁴⁰K (146.36 Bq/kg vs. global average 400 Bq/kg) were all above global averages. Furthermore, a high concentration of radon gas (337.06 μSv/y) was measured at subsite 1. The strong positive correlation observed between ²²⁶Ra and ²³⁸U (0.96256) provides further evidence of potentially elevated radioactivity levels near the factory. In contrast, subsite 2, situated farther from the factory, exhibited natural radioactive background levels within international limits. Quantitative analysis revealed that gamma ray absorbed doses for ²²⁶Ra and ²³²Th exceeded global averages in some samples. Specifically, ²²⁶Ra doses ranged from 7.8 to 46.26 ppm (exceeding the 20 ppm global average in some cases), and ²³²Th doses ranged from 1.98 to 9.14 ppm (exceeding the 10 ppm global average in some cases). The concentration of ⁴⁰K, however, remained within the global range (0.07%-0.69 %). The observed imbalances in the ratios of Th/U (0.17-0.24 Bq/kg and 0.73-0.24 ppm) and U/Ra (0.81-0.73 Bq/kg and 0.73-0.17 ppm), both of which are significantly lower than their respective global averages of 4 and 2.4, point towards the presence of fertilizer-derived contamination. This conclusion is further supported by the high phosphate concentrations detected in the samples. Overall, this study suggests that radioactive contamination near the phosphate fertilizer factory significantly exceeds global background levels and international limits in some cases. This raises concerns about potential risks posed to surrounding agricultural land and crops.

• Restorative Dentistry and Endodontics
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• v.13 no.1
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• pp.69-78
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• 1988

The study was designed to establish a more nearly quantitative method for assessing the marginal leakage of dental restorations. 27 Class V cavities with $45^{\circ}$ bevel joint were prepared and classified into 2 groups. One group was filled with Scotchbond and silux. The other group was filled with glass ionomer cement, Scotchbond and silux. After finishing, all specimens were subjected manually to 100 thermal cycles at $0^{\circ}C$ and $100^{\circ}C$ water-bath. They were soaked in a samarium nitrate solution for 3 hours, irradiated with flux of $6{\times}10^{12}$ neutrons/$cm^2$/sec for 11 hours, cooled for 200 hours, counted with the HPGE detector and the tracer uptake was determined by comparison with a standard of samarium (10 ${\mu}g$). The following results were obtained. 1. Both of the two groups showed a considerable amounts of marginal leakage. 2. The group filled without glass ionomer cement base showed more marginal leakage than the group filled with glass ionomer cement base. 3. Neutron Activation Analysis produced a good quantitative method to measure the marginal leakage and samarium was appropriate to measure the marginal leakage of resin restorations using neutron activation analysis.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.452-461
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• 1998

The pin power distribution is determined by analyzing the rotational gamma scanning data for 36 element fuel bundle of HANARO. A fission monitor of Nb$^{95}$ is chosen by considering the criteria of the half-life, fission yield, emitting ${\gamma}$-ray energy and probability. The ${\gamma}$-ray spectra were measured in Korea Atomic Energy Research Institute(KAERI) by using a HPGe detector and by rotating the fuel bundle at steps of 10$^{\circ}$. The counting rates of Nb$^{95}$ 766 keV ${\gamma}$-rays are determined by analyzing the full absorption peak in the spectra. A 36$\times$36 response matrix is obtained from calculating the contribution of each rod at every scanning angle by assuming 2-dimensional and parallel beam approximations for the measuring geometry. In terms of the measured counting rates and the calculated response matrix, an inverse problem is set up for the unknown distribution of activity concentrations of pins. To select a suitable solving method, the performances of three direct methods and the iterative least-square method are tested by solving simulation examples. The final solution is obtained by using the iterative least-square method that shows a good stability. The influences of detection error, step size of rotation and the collimator width are discussed on the accuracy of the numerical solution. Hence an improvement in the accuracy of the solution is proposed by reducing the collimator width of the scanning arrangement.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.13-18
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• 2011

In this study, simulations were done of a 661.6 keV line from a point source of $^{137}Cs$ housed in a lead shield. When increasing the scattering angle from 60 to 120 degrees with a 6061 aluminum alloy target placed at angles of 30 and 45 degrees to the incident beam, the spectra showed that the single scattering component increases and that the multiple scattering component decreases. The investigation of the single and multiple scattering components was carried out using a MCNP5 simulation code. The component of the single Compton scattering photons is proportional to the target electron density at the point where the scattering occurs. The single scattering peak increases according to the thickness of the target and saturates at a certain thickness. The signal-to-noise ratio was found to decrease according to the target thickness. The simulation was experimentally validated by measurements. These results will be used to determine the best conditions under which this method can be applied to testing electron densities or to assess the thickness of samples to locate defects in them.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.818-824
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• 2019

In the present work, we determined the gamma-ray attenuation characteristics of eight different polymers(Polyamide (Nylon 6) (PA-6), polyacrylonitrile (PAN), polyvinylidenechloride (PVDC), polyaniline (PANI), polyethyleneterephthalate (PET), polyphenylenesulfide (PPS), polypyrrole (PPy) and polytetrafluoroethylene (PTFE)) using transmission geometry utilizing the high resolution HPGe detector and different radioactive sources in the energy range 81-1333 keV. The experimental linear attenuation coefficient values are compared with theoretical data (WinXCOM data). The linear attenuation coefficient of all polymers reduced quickly with the increase in energy, at the beginning, while decrease more slowly in the region from 267 keV to 835 keV. The effective atomic number of PVDC and PTFE are comparatively higher than the $Z_{eff}$ of the remaining polymers, while PA-6 possesses the lowest effective atomic number. The half value layer results showed that PTFE ($C_2F_4$, highest density) is more effective to attenuate the gamma photons. Also, the theoretical results of macroscopic effective removal cross section for fast neutrons ($\sum_{R}$) were computed to investigate the neutron attenuation characteristics. It is found that the $\sum_{R}$ values of the eight investigated polymers are close and ranged from $0.07058cm^{-1}$ for PVDC to $0.11510cm^{-1}$ for PA-6.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.325-336
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• 2019

The activity concentrations of $^{226}Ra$, $^{232}Th$, and $^{40}K$ from 102 building materials samples were determined using a high-purity germanium (HPGe) detector. The activity concentrations were evaluated for possible radiological hazards to the human health. The excess lifetime cancer risks (ELCR) were also estimated, and the average values were recorded as $0.42{\pm}0.24{\times}10^{-3}$, $3.22{\pm}1.83{\times}10^{-3}$, and $3.65{\pm}1.85{\times}10^{-3}$ for outdoor, indoor, and total ELCR respectively. The activity concentrations were further subjected to RESRAD-BUILD computer code to evaluate the long-term radiation exposure to a dweller. The indoor doses were assessed from zero up to 70 years. The simulation results were $92{\pm}59$, $689{\pm}566$, and $782{\pm}569{\mu}Sv\;y^{-1}$ for indoor external, internal, and total effective dose equivalent (TEDE) respectively. The results reported were all below the recommended maximum values. Therefore, the radiological hazards attributed to building materials under study are negligible.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4246-4251
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• 2023

The optimized design of a Neutron Activation Analysis (NAA) system, including Delayed Gamma NAA (DGNAA) and Prompt Gamma NAA (PGNAA), has been proposed in this research based on Mevex Linac with 5 MeV electron energy and 50 kW power as a neutron source. Based on the MCNPX 2.6 simulation, the optimized configuration contains; tungsten as an electron-photon converter, BeO as a photoneutron target, BeD₂ and plexiglass as moderators, and graphite as a reflector and collimator, as well as lead as a gamma shield. The obtained thermal neutron flux at the beam port is equal to 2.06 × 10⁹ (# /cm².s). In addition, using the optimized neutron beam, the detection limit has been calculated for some elements such as H-1, B-10, Na-23, Al-27, and Ti-48. The HPGe Coaxial detector has been used to measure gamma rays emitted by nuclides in the sample. By the results, the proposed system can be an appropriate solution to measure the concentration and toxicity of elements in different samples such as food, soil, and plant samples.