• Journal of the Korean Society of Groundwater Environment
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• v.5 no.2
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• pp.116-121
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• 1998

For the application of study on pollution and conservation of environment determination of 33 elemental concetrations in different sediment samples were carried out using instrumental neutron activation analysis (INAA). For verification and evaluation of the analytical method, three standard reference materials (two NIST SRMs and one NRCC CRM) were chosen and the accuracy and precision of the analysis were estimated by comparison to the certified values. Under the optimum condition, the analytical procedure to apply a practical sample was estimated. Neutron irradiation of sample was done at the irradiation facilities (neutron flux, 1-3${\times}$10$\^$13/n/$\textrm{cm}^2$$.$s) of the TRIGA MARK-III and HANARO research reactor in the Korea Atomic Energy Research Institute. In addition, analysis of two IAEA's sediment was performed according to the pre-established analytical method. The analytical results of elements such as Al, As, Co, Cr, Fe, Sb and Zn by INAA were intercompared with those of WD-XRF, ICP-MS and AAS, and are relatively agreed with each other.

• Journal of Radiation Protection and Research
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• v.4 no.1
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• pp.21-28
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• 1979

A calculation is presented of air-scattered gamma rays using the modified single-scattering approximation. The air-scattered tissue dose rates are calculated for a general purpose taking into account (a) the buildup and exponential attenuation, (b) the energy spectrum at the position of question and (c) the geometrical scattering volume in three dimensions. These calculations have been further modified to render them applicable to a typical field irradiation facility which is surrounded by a shield wall and in which the source is fitted with a beam collimating device. The results of the calculation include the energy spectra, angular distribution and tissue does rates at source-receiver separation distances of from 35m to 300m. The comparison shows that the present method developed may be generally adequate for the gamma-ray air-scattering problems in field irradiation facilities if energy and angular distribution at the shield are unimportant.

• Journal of Radiation Protection and Research
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• v.42 no.3
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• pp.141-145
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• 2017

Background: The concrete walls inside the vaults of cyclotron facilities are activated by neutrons emitted by the targets during radioisotope production. Reducing the amount of radioactive waste created in such facilities is very important in case they are decommissioned. Thus, we proposed a strategy of reducing the neutron activation of the concrete walls in cyclotrons during operation. Materials and Methods: A polyethylene plate and B-doped Al sheet (30 wt% of B and 2.5 mm in thickness) were placed in front of the wall in the cyclotron room of a radioisotope production facility for pharmaceutical use. The target was Xe gas, and a Cu block was utilized for proton dumping. The irradiation time, proton energy, and beam current were 8 hours, 30 MeV, and $125{\mu}A$, respectively. To determine a suitable thickness for the polyethylene plate set in front of the B-doped Al sheet, the neutron-reducing effects achieved by inserting such sheets at several depths within polyethylene plate stacks were evaluated. The neutron fluence was monitored using an activation detector and 20-g on de Au foil samples with and without 0.5-mm-thick Cd foil. Each Au foil sample was pasted onto the center of a polyethylene plate and B-doped Al sheet, and the absolute activity of one Au foil sample was measured as a standard using a Ge detector. The resulting relative activities were obtained by calculating the ratio of the photostimulated luminescence of each foil sample to that of the standard Au foil. Results and Discussion: When the combination of a 4-cm-thick polyethylene plate and B-doped Al sheet was employed, the thermal neutron rate was reduced by 78%. Conclusion: The combination of a 4-cm-thick polyethylene plate and B-doped Al sheet effectively reduced the neutron activation of the investigated concrete wall.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Journal of Korean Society of Water and Wastewater
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• v.26 no.2
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• pp.219-228
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• 2012

Problems due to the taste and odor in drinking water are common in treatment facilities around the world. Taste and odor are perceived by the public as the primary indicators of the safely and acceptability of drinking water, and are mainly caused by the presence of two semi-volatile compounds-2-methylisoborneol(2-MIB) and geosmin. Conventional treatment processes in water treatment plants, such as coagulation, sedimentation and chlorination have been found to be ineffective for the removal of 2-MIB and geosmin. Pulse UV system is a new UV irradiation system that is a non-mercury lamp-based alternative to currently used continuous wave systems for water disinfection. This study shows pulse UV system to be effective in treatment of these two compounds. Geosmin removal efficiency of UV process alone achieved approximately 70% at 10sec contact time. 2-MIB removal efficiency of UV only process achieved approximately 60% at 10sec contact time. The addition of $H_{2}O_{2}$ 7mg/L increased geosmin and 2-MIB removal efficiency upto approximately 94% and 91%, respectively.

• Corrosion Science and Technology
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• v.17 no.3
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• pp.123-128
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• 2018

The fabrication of a controlled surface is of great interest because it can be applied to various engineering facilities due to the various properties of the surface, such as self-cleaning, anti-bio-fouling, anti-icing, anti-corrosion, and anti-sticking. Controlled surfaces with micro/nano structures were fabricated using an ion beam focused onto a polypropylene (PP) surface with a fluoridation process. We developed a facile method of fabricating hydrophobic surfaces through ion beam treatment with argon and oxygen ions. The fabrication of low surface energy materials can replace the current expensive and complex manufacturing process. The contact angles (CAs) of the sample surface were $106^{\circ}$ and $108^{\circ}$ degrees using argon and oxygen ions, respectively. X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared (FT-IR) spectroscopy were used to determine the chemical composition of the surface. The morphology change of the surfaces was observed by scanning electron microscopy (SEM). The change of the surface morphology using the ion beam was shown to be very effective and provide enhanced optical properties. It is therefore expected that the prepared surface with wear and corrosion resistance might have a considerable potential in large scale industrial applications.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2006.05a
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• pp.369-370
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• 2006

During demonstrations of a process conditioning spent nuclear fuels, it is necessary to transport and handle Spent fuel road cuts from Post Irradiation Examination facility to Slitting device in The hot cell. the spent fuel pellets which are highly radioactive materials are separated with its clad and are fed into the next conditioning process. For this, a spent fuel rod, 3.5 m long, is cut by 25 cm long which is suitable length for the decladding process. These rod-cuts are packed into the capsule and are moved to the ACPF(Advanced spent nuclear fuel Conditioning Process Facility). In the ACPF, Once the capsule is unloaded in the ACPF, Capsule is taken out one-by-one and installed on the decladding device. In these processes, the crushed spent fuel pellet can be scattered inside the facilities and thus it contaminate the hot cell. In this paper, we developed the specially designed transportation and storage device for spent nuclear fuel capsules.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.253-259
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• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• Journal of the Korean Solar Energy Society
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• v.25 no.2
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• pp.19-26
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• 2005

Concentric evacuated tube solar energy collector has been interested recently since government has driven to install alternative energy systems in new large building. In this paper, testing of the evacuated tube collector is conducted in outdoor during daytime by transient method. The collector thermal efficiencies are plotted in term of $(T_{in}-T_a)/Ic$, where $T_{in}$ is inlet working fluid temperature, $T_a$ is atmospheric temperature and $I_c$ is solar irradiation on the collector surface. The evacuated tube collector efficiency is ranged from 50% to 63% in real outdoor condition. In addition, the total overall heat loss coefficient is found to have an inverse variation to $(T_{in}-T_a)/I_c$ so that the coefficient becomes very high when $(T_{in}-T_a)/I_c$ is small.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1589-1599
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• 2017

Field programmable gate arrays (FPGAs) are getting more attention in safety-related and safety-critical application development of nuclear power plant instrumentation and control systems. The high logic density and advancements in architectural features make static random access memory (SRAM)-based FPGAs suitable for complex design implementations. Devices deployed in the nuclear environment face radiation particle strike that causes transient and permanent failures. The major reasons for failures are total ionization dose effects, displacement damage dose effects, and single event effects. Different from the case of space applications, soft errors are the major concern in terrestrial applications. In this article, a review of radiation effects on FPGAs is presented, especially soft errors in SRAM-based FPGAs. Single event upset (SEU) shows a high probability of error in the dependable application development in FPGAs. This survey covers the main sources of radiation and its effects on FPGAs, with emphasis on SEUs as well as on the measurement of radiation upset sensitivity and irradiation experimental results at various facilities. This article also presents a comparison between the major SEU mitigation techniques in the configuration memory and user logics of SRAM-based FPGAs.