• Nuclear Engineering and Technology
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• v.5 no.4
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• pp.279-285
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• 1973

A method for the simultaneous determination of mercury, bromine, arsenic and cadmium in biological samples is described. Following neutron activation and a simple distillation of volatile compounds, mercury and bromine were determined by gamma-ray spectrometry. Arsenic and cadmium were further separated by cation exchange separation and determined similarly. Determination limits for mercury, bromine, arsenic and cadmium were 0.001$\mu\textrm{g}$, 0.003$\mu\textrm{g}$, 0.001$\mu\textrm{g}$ and 0.02$\mu\textrm{g}$, respectively. The method has been applied to the determination of mercury, bromine, arsenic and cadmium in rice and fish samples. Analysis of a standard kale powder yielded the values of 0.046$\mu\textrm{g}$/g for mercury, 24.5$\mu\textrm{g}$/g for bromine 0.17 $\mu\textrm{g}$/g for arsenic and 0.50$\mu\textrm{g}$/g for cadmium.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.31 no.4
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• pp.198-202
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• 2018

In this work, we have investigated the effect of a 30-min thermal anneal at $550^{\circ}C$ on the electrical characteristics of neutron-irradiated 4H-SiC MOSFETs. Thermal annealing can recover the on/off characteristics of neutron-irradiated 4H-SiC MOSFETs. After thermal annealing, the interface-trap density decreased and the effective mobility increased in terms of the on-characteristics. This finding could be due to the improvement of the interfacial state from thermal annealing and the reduction in Coulomb scattering due to the reduction in interface traps. Additionally, in terms of the off-characteristics, the thermal annealing resulted in the recovery of the breakdown voltage and leakage current. After the thermal annealing, the number of positive trapped charges at the MOSFET interface was decreased.

• Journal of the Korean Recycled Construction Resources Institute
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• v.7 no.4
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• pp.316-322
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• 2019

A probabilistic approach considering uncertainties was employed to investigate the effects on the material characteristics and strength of nuclear bio-logical shield concrete walls, when exposed to long-term radiation during the entire service life. Time-dependent compressive and tensile strengths were estimated by conducting the neutron fluence analysis. For the contaminated concrete, individual compressive and tensile failure probabilities can be possibly evaluated by not only establishing limit-state function withthe predefined critical values but also performing Monte Carlo Simulation. Nuclear power plant types similar to the Kori Unit 1, which was shut off permanently in 2017 after the 40-year operation, were herein selected for an illustrative purpose. Consequently, the probability-based performance assessment and prediction of contaminated concrete walls were well demonstrated.

• Analytical Science and Technology
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• v.15 no.6
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• pp.509-513
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• 2002

In this study, instrumental neutron activation analysis was used to assess the concentration level of inorganic trace elements in Korean women blood serum. It was found out that high concentration of Na and Cl incurs analytical interference, but 12 elements of Br, Ca, Cl, Co, Cr, Cs, Fe, K, Na, Rb, Se, Zn can be determined under the condition of interference minimization. Serum samples collected from 63 women were analyzed and the concentration level and range of the elements were evaluated. NIST SRMs were analyzed simultaneously for analytical quality control. The average values of Na and Cl determined in serum samples are around 3000 mg/L, Ca is 100 mg/L and K is 200 mg/L. Besides, Br, Se and Zn have concentration level of 6.0, 0.1 and 1.0 mg/L, respectively. It was found that there is no significant difference between the present values and reported values.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.40-47
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• 1989

For the effective design of a beam forming structure of the negative-ion-based neutral beam injector, a computer program based on a particle simulation model is developed for the calculation of charged particle motions in the electrostatic fields. The motions of negative ions inside the acceleration tube of a multiple-aperture triode are computed at finite time steps. The electrostatic potentials are obtained from the Poisson's equation by the finite difference method. The successive overrelaxation method is used to solve the matrix equation. The particle and force weighting methods are used on a cloud-in-cell model. The optimum design of the beam forming structure has been studied by using this computer code for the various conditions of elctrodes. The effects of the acceleration-deceleration gap distance, the thickness of the deceleration electrode and the shape of the acceleration electrode on beam trajectories are exmined to find the minimum beam divergence. Some numerical illustrations are presented for the particle movements at finite time steps in the beam forming tubes. It is found in this particle simulation modelling that the shape of the acceleration electrode is the most significant factor of beam divergence.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.3
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• pp.151-164
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• 2004

The changes of microstructural and mechanical properties were evaluated for the high burn-up fuel claddings after the neutron irradiation of $1.8{\sim}3.1{\times}10^{20}n/cm^2$ (E>1.0 MEV) in HANARO research reactor. After the irradiation, the spot-type dislocations (a-type dislocations) were easily observed in most claddings, and the density of the dislocations was different depending on the grains and was higher at grain boundaries than within grains. As the final annealing temperature increased, the density of spot-type dislocations increased and the line-type dislocations (c-type dislocations) which was perpendicular to the <0002> direction, appeared sporadically in some claddings. However, the types of precipitates in the fuel claddings after the irradiation were not changed from that in unirradiated claddings. The mechanical properties including the hardness, strength and elongation after the irradiation were changed due to the formation of spot-type dislocations. That is, the increase in hardness and strength as well as the decrease in elongation after the irradiation was occurred simultaneously with increasing the final annealing temperature. Owing to the Nb contribution to the formation of spot-type dislocation during the irradiation, the increase in hardness and strength in higher Nb-contained Zr alloys after the irradiation was higher than that in lower Nb-contained Zr alloys.

• Polymer(Korea)
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• v.35 no.6
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• pp.531-536
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• 2011

Homopolymer distribution in block copolymer/homopolymer blends was investigated as a function of homopolymer concentration and homopolymer molecular weight. The deuterated poly(methyl methacrylate) or polystyrene was blended with a deuterated polystyrene-poly(methyl methacrylate) diblock copolymer up to a concentration of 20 wt%. Samples were characterized by small-angle X-ray scattering (SAXS), neutron reflectivity and transmission electron microscopy. The block copolymer with a thin-film geometry formed alternating lamellar microdomains oriented parallel to the substrate surface. By adding the homopolymer, the microdomain structure was significantly disturbed. As a consequence, a poorly ordered morphology appeared when the homopolymer concentration exceeded 15 wt%. Increasing the homopolymer concentration and/or the homopolymer molecular weight caused the microdomains to swell less uniformly, resulting in segregation of the homopolymer toward the middle of the microdomains.

• Nuclear Engineering and Technology
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• v.8 no.1
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• pp.1-8
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• 1976

For tile rapid production of fluorine-18 for medical use, the up-to-date methods of separation such as the recoil separation, the alumina column chromatography, and the distillation are reviewed. The amount of the residue, the gamma emitting impurity, and the tritium content in the product obtained by each separation method are determined. The product obtained by the nuclear recoil separation or by the alumina column chromatography is inferior to that obtained by the distillation in the purity point of view. Thus, the separation by the distillation is tile most effective especially in the case of using a natural lithium carbonate target. Carrier free fluorine-18 of about 2 mCi can routinely be produced by irradiating 7g of the natural lithium carbonate under the neutron flux of about 1$\times$10$^{13}$ n/$\textrm{cm}^2$/sec for 3 hrs, and subsequent separation by the distillation. The over-all processing time is 35-40n1in.

• Corrosion Science and Technology
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• v.20 no.5
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• pp.308-314
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• 2021

Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10¹⁵ n/cm². As helium(He) gas can be generated by Ni only at 1/10⁶ level of 5 × 10²¹ n/cm², He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.121-126
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• 2019

The radionuclide inventory prediction of a nuclear power plant can help establish decommissioning plan by providing information of radiation environment. Accumulated radionuclides in reactors and related facilities after reactor shutdown can be divided into neutron activated materials and contaminated materials. Among the neutron activated radionuclides, $^{36}Cl$ and $^{41}Ca$ are important from the viewpoint of disposal because of its long half-life and physiochemical characteristics. In this research, we calculated the radionuclides of $^{36}Cl$ and $^{41}Ca$ in bioshielding concrete by estimating the neutron flux and cross section using the MCNPX. And we evaluated the inventories of $^{36}Cl$ and $^{41}Ca$ using the activation calculation code ORIGEN2.