• Nuclear Engineering and Technology
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• v.6 no.2
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• pp.89-96
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• 1974

The cross-section ratios of the nuclear isomeric pairs $^{80}$ B $r^{m, g}$, sup 81/S $e^{m, g}$, $^{104}$ R $h^{m, g}$, $^{116}$ I $n^{m, g}$ and $^{134}$ C $s^{m, g}$ through the radiative thermal neutron capture process have been studied. The experimental values of these ratios obtained by the activation method have been compared with the calculated ones deduced from the modified Huizenga-Vandenbosch method. Agreement between these values within 30% could be attained by controlling the spin cut-off parameter and gamma-ray multiplicity.

• Korean Journal of Food Science and Technology
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• v.4 no.2
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• pp.61-71
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• 1972

The selenium content of a wide variety of Korean food was determined by neutron activation analysis. Most fruits and vegetables contained quantities of selenium less than $0.4{\mu}g/g$. Grain products varied widely in their selenium content with $0.5{\mu}g/g$ and barley cereal as high as $0.7{\mu}g/g$. Dried milk powder sample ranged from $0.7\;to\;0.15{\mu}g/g$. Chicken muscle contained about $0.7{\mu}g/g$. The content of sea food was generally higher, ranging from $0.3\;to\;3.65{\mu}g/g$. These values suggest that a diet well balanced in other nutrients is probably also nutritionally adequate with regard to selenium, although possible effects of cooking and biological availability remain to be investigated.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.462-468
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• 2022

The selection and effective utilization of peak-fitting software for conventional gamma-ray spectrum analysis is significant for accurate determination of the mass fraction of elements, particularly in complex peak regions. Majority of the peak-fitting programs can derive similar peak characteristics for singlet peaks, but very few programs can deconvolute multi-peaks in a complex region. The deconvolution of multi-peaks requires special peak-fitting functions, such as left and right-skew distributions. In the this study, 843.76 keV (²⁷Mg) peak area from the complex region (840 keV-850 keV) determined and compared using four different peak-fitting programs, namely, GammaVision, Genie2000, HyperLab, and HyperGam. The 843.76 keV peak interfered with 841.63 keV (^152mEu) and 846.81 keV (⁵⁶Mn). The total Mg concentration was determined through k₀-instrumental neutron activation analysis by applying the isotopic interference correction factor ²⁷Al(n,p)²⁷Mg through the simultaneous determination of Al concentration. HyperLab and HyperGam peak-fitting programs reported consistent peak areas, and resultant concentrations agreed with the certified values of matrix-certified reference materials.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1708-1717
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• 2023

The grade analysis of lead-zinc ore is the basis for the optimal development and utilization of deposits. In this study, a method combining Prompt Gamma Neutron Activation Analysis (PGNAA) technology and machine learning is proposed for lead-zinc mine borehole logging, which can identify lead-zinc ores of different grades and gangue in the formation, providing real-time grade information qualitatively and semi-quantitatively. Firstly, Monte Carlo simulation is used to obtain a gamma-ray spectrum data set for training and testing machine learning classification algorithms. These spectra are broadened, normalized and separated into inelastic scattering and capture spectra, and then used to fit different classifier models. When the comprehensive grade boundary of high- and low-grade ores is set to 5%, the evaluation metrics calculated by the 5-fold cross-validation show that the SVM (Support Vector Machine), KNN (K-Nearest Neighbor), GNB (Gaussian Naive Bayes) and RF (Random Forest) models can effectively distinguish lead-zinc ore from gangue. At the same time, the GNB model has achieved the optimal accuracy of 91.45% when identifying high- and low-grade ores, and the F₁ score for both types of ores is greater than 0.9.

• Journal of the Korean Chemical Society
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• v.21 no.6
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• pp.427-433
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• 1977

A neutron activation method has been developed for the simultaneous determination of chromium, iron, lanthanum, scandium and zinc in river-water samples. The sample is sealed in the silica ampoule without pretreatment and irradiated for a week at a thermal neutron flux of $1{\times}10^{13}n{\cdot}cm^{-2}{\cdot}sec^{-1}$. After cooling for about two days, the elements in the sample are sequentially extracted at different pH by 0.1M oxine-chloroform solution. The organic layers are checked by Gamma-ray spectrometry with $″3\;{\times}\;3″$ NaI (T1) detector connected to a 800-channel pulse hight analyzer. The ppb concentration of the elements in most of river-water samples could be determined by this method. The tracer study for the quantitative separation of the elements was also carried out.

• Journal of the Korean Chemical Society
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• v.8 no.2
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• pp.88-91
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• 1964

A rapid and simple method is described here for the determination of scandium in monazite by neutron activation analysis. The sample is irradiated for 20 hours at the neutron flux of $10^{12}$ thermal neutrons/$cm^2$/sec in the TRIGA MARK Ⅱ reactor, after which the sample is decomposed by fusion with concentrated sulfuric acid. The scandium-46 together with scandium carrier are separated from the irradiated sample by precipitating with ammonia, and are extracted by solvent extraction of the thiocyanate complex into ether. The induced radioactivity is measured by gamma scintillation spectrometry using the Multichannel Pulse Height Analyzer connected with 2"${\times}$2" NaI(Tl). The chemical yield is determined gravimetrically by precipitating scandium with mandelic acid. In order to check the efficiency of scandium separation and the errors from interfering activities of the other elements, scandium was separated by the cation exchange resin column, and the results from both samples were compared each other, which showed that the chemical procedure used in this work was as selective as the ion-exchange method with respect to scandium separation. The scandium contents in Korean monazite were found to be about 12 p. p. m.

• Journal of Radiation Protection and Research
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• v.1 no.1
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• pp.1-9
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• 1976

A suggestion has been made for neutron dosimetric techniques using activation and threshold detectors in criticality accidents. Neutron dosimetrical parameters, namely, the fission spectrum-averaged cross-sections of some threshold reactions and fluence-to-dose conversion factors have been calculated by the use of an electronic computer. It appears that detectors having comparatively high threshold energy give more fine information on spectral deformation in criticality accidents, while detectors with low threshold energy are of usefulness for measuring fast neutron fluence regardless of fissioning types. Unexpectedly it is found that the fission spectrum-averaged cross sections of the $^{32}S(n,\;p)^{32}P$ reaction is not sensitive to analytical forms of fission neutron spectrum: the modified Cran-berg and Maxwellian forms. In addition, the fluence-to-dose conversion factors seem to be insensitive to both spectral functions and fissioning types.

• Journal of Radiation Protection and Research
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• v.45 no.3
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• pp.130-141
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• 2020

Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd₂O₃, B₄C, polyethylene (PE), and borated (5 wt% ^natB) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B₄C and Gd₂O₃ were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1- or 0.4-cm-thick B₄C or Gd₂O₃ behind 10 cm-thick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cm-thick Gd₂O₃+10 cm-thick PE or 0.1 cm-thick B₄C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B₄C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.

• Journal of Radiation Protection and Research
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• v.28 no.2
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• pp.109-116
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• 2003

Based on the necessity to evaluate the activation products inventory during decommissioning lot domestic nuclear power plants, a preliminary estimation of the activation products inventory for Kori unit 1, which is getting close to the end of lifetime, was carried out with ANISN and ORIGEN2 code. In order to calculate neutron nux using ANISN code, the reactor was divided into 9 zones from core to bioshield concrete for radial direction. Also :he cross-section of main nuclides were calibrated with neutron flux in the reactor pressure vessel(RPV) region. The results showed that 95 % of tile total radioactivity in RPV from reactor shutdown to 10 years came from the nuclides of $^{55}Fe,\;^{59}Ni,\;^{63}Ni\;and\;^{60}Co$. And the total radioactivity with cooling of more than 50 years after decommissioning was no more than 0.2 % of at the time of shutdown. Considering the weight of RPV is 210 tons, the total radioactivity of RPV reached to $5.25{\times}10^{6}GBq$ at shutdown time. As compared with the total radioactivity of bioshield concrete at reactor shutdown time, the radioactivity after tooling more than 10 years was below 1 %.

• Journal of Radiation Protection and Research
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• v.47 no.2
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• pp.99-106
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• 2022

Background: The radionuclide inventory calculation codes such as ORIGEN and FISPACT collapse neutron reaction libraries with energy spectra and generate an effective one-group cross-section. Since the nuclear cross-section data, energy group (g) structure, and other input details used by the two codes are different, there may be differences in each code's activation inventory calculation results. In this study, the calculation results of neutron-induced activation inventory using ORIGEN and FISPACT were compared and analyzed regarding radioactive waste classification and worker exposure during nuclear decommissioning. Materials and Methods: Two neutron spectra were used to obtain the comparison results: Watt fission spectrum and thermalized energy spectrum. The effective one-group cross-sections were generated for each type of energy group structure provided in ORIGEN and FISPACT. Then, the effective one-group cross-sections were analyzed by focusing on ⁵⁹Ni, ⁶³Ni, ⁹⁴Nb, ⁶⁰Co, ¹⁵²Eu, and ¹⁵⁴Eu, which are the main radionuclides of stainless steel, carbon steel, zircalloy, and concrete for decommissioning nuclear power plant (NPP). Results and Discussion: As a result of the analysis, ¹⁵⁴Eu and ⁵⁹Ni may be overestimated or underestimated depending on the code selection by up to 30%, because the cross-section library used for each code is different. When ORIGEN-44g, -49g, and -238g structures are selected, the differences of the calculation results of effective one-group cross-section according to group structure selection were less than 1% for the six nuclides applied in this study, and when FISPACT-69g, -172g, and -315g were applied, the difference was less than 1%, too. Conclusion: ORIGEN and FISPACT codes can be applied to activation calculations with their own built-in energy group structures for decommissioning NPP. Since the differences in calculation results may occur depending on the selection of codes and energy group structures, it is appropriate to properly select the energy group structure according to the accuracy required in the calculation and the characteristics of the problem.