• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1250-1253
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• 2024

Activity measurement of the primary circuit water of fission reactors is one method that can provide early detection of a damaged fuel assembly in the reactor core. This is an important aspect in the safe operation of the reactor and for radiation protection of staff. Radionuclides in the primary circuit water are produced by the activation of stable nuclides and the fission of fissile nuclides, mainly the isotope ²³⁵U. In the LVR-15 research reactor, measurement of the activity of the primary circuit water has been regularly undertaken since 1996. A water sample is taken from the primary circuit every week and the activities are measured four days later using gamma spectrometry. The results of these long-term measurements from 1996 to 2022 are presented. The activity time dependences of the individual radionuclides are discussed in relation to fuel assembly damage and for events connected to contamination of the water by objects inserted into the primary circuit during experiments carried out near the reactor core.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• Journal of the korean society of oceanography
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• v.33 no.4
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• pp.196-204
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• 1998

$^{14}$C uptake experiments were carried out in and around the cold water mass in the southern part of the Korean East Sea in August and October 1995 to assess spatial and seasonal variability of primary productivity and its relation to physical and chemical factors. The cold and high saline water mass in the bottom layer extended upward to the surface layer and developed along the eastern coast of Korea in August. Chlorophyll-a concentration was maintained high in the cold water mass through August to October and its maximum concentration was 6.3 ${\mu}$g 1$^{-1}$ at Stn. 209-4 in August. Primary productivity and daily primary productivity ranged from 0.29 to 8.02 mgC m$^{-3}$ hr$^{-1}$ and from 58.3 to 63.1 mgC m$^{-2}$ d$^{-1}$, respectively, throughout the study period. Primary productivity of the cold water mass was higher than that of offshore waters in both summer and autumn seasons. P$_{max}$ and I$_{max}$ of the cold water mass in August were higher than those in October, except Stn. 208-5. These results suggest that high primary productivity in the cold water mass may be established by the upwelled nutrients and light adaptaion to convected phytoplankton due to upwelling of the bottom waters.

• Corrosion Science and Technology
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• v.22 no.4
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• pp.265-272
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• 2023

The aim of this study was to develop a new in-situ observation method and instrument in micro-scale to investigate the mechanism of stress corrosion cracking (SCC) initiation of Ni-base alloys in a high temperature water environment of pressurized water reactors (PWRs). A laser confocal microscope (LCM), an autoclave with diamond window view port, and a slow strain-rate tester with primary water circulation loop system were components of the instrument. Diamond window, one of the core components of the instrument, was selected based on its optical, chemical, and mechanical properties. LCM was used to observe the specimen in micro-scale, considering the experimental condition of a high-temperature primary water environment. Using in-situ method and instrument, it is possible to observe oxidation and deformation of specimen surface in micro-scale through the diamond window in a high-temperature primary water in real-time. The in-situ method and instrument developed in this work can be utilized to investigate effects of various factors on SCC initiation in a high-temperature water environment.

• Journal of Korean society of Dental Hygiene
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• v.7 no.4
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• pp.509-519
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• 2007

The purpose of this study was to take inforamtion to increase support on water fluoridation primary school teachers in Jinhae. The survey was done by self-administration method to primary school teachers 88 at non-fluoridated area in Jinhae. The date was analysed using the SPSS 10.1 program. The results are as follows : The mean percentage of the respondents using the tap water and purified water for drinking was 51.1% and 33.3%, respectively. Recognition rate of water fluoridation was 64.8%, 33.3% of the respondents knew about water fluoridation from mass mesdia. 62.5% of the respondents agreed with water fluoridation. 76.1% of the respondents knew that the purpose of water fluoridation was the prevention of dental caries. 72.7% of the respondents agreed with continuous water fluoridation. 70.5% of the recognized about safety of water fluoridation. 72.7% of the respondents agreed with decision of professional group for the performance of water fluoridation. In order to secure residents' support for the continuation of the adjusted water fluoridation program and to reinforce the education for safety and economical efficiency for the residents through systemic education.

• Corrosion Science and Technology
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• v.12 no.2
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• pp.85-92
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• 2013

This work is concerned with an evaluation of dynamic boric acid corrosion (BAC) of low alloy steel for reactor pressure vessel of a pressurized water reactor (PWR). Mockup test method was newly established to investigate dynamic BAC of the low alloy steel under various conditions simulating a primary water leakage incident. The average corrosion rate was measured from the weight loss of the low alloy steel specimen, and the maximum corrosion rate was obtained by the surface profilometry after the mockup test. The corrosion rates increased with the rise of the leakage rate of the primary water containing boric acid, and the presence of oxygen dissolved in the primary water also accelerated the corrosion. From the specimen surface analysis, it was found that typical flow-accelerated corrosion and jet-impingement occurred under two-phase fluid of water droplet and steam environment. The maximum corrosion rate was determined as 5.97 mm/year at the leakage rate of 20 cc/min of the primary water with a saturated content of oxygen within the range of experimental condition of this work.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4481-4490
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• 2022

In this study, a new rupture disk corrosion test (RDCT) method was developed for real-time detection of stress corrosion cracking (SCC) initiation of Alloy 600 in a primary water environment of pressurized water reactors. In the RDCT method, one side of a disk specimen was exposed to a simulated primary water at high temperature and pressure while the other side was maintained at ambient pressure, inducing a dome-shaped deformation and tensile stress on the specimen. When SCC occurs in the primary water environment, it leads to the specimen rupture or water leakage through the specimen, which can be detected in real-time using a pressure gauge. The tensile stress applied to the disk specimen was calculated using a finite element analysis. The tensile stress was calculated to increase as the specimen thickness decreased. The SCC initiation time of the specimen was evaluated by the RDCT method, from which result it was found that the crack initiation time decreased with the decrease of specimen thickness owing to the increase of applied stress. After the SCC initiation test, many cracks were observed on the specimen surface in an intergranular fracture mode, which is a typical characteristic of SCC in the primary water environment.

• Corrosion Science and Technology
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• v.18 no.1
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• pp.33-38
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• 2019

Alloy 600/182 with excellent mechanical/chemical properties have been utilized for nuclear power plants. Although both alloys are known to have superior corrosion resistance, stress corrosion cracking failure has been an issue in primary water environment of nuclear power plants. Therefore, primary water stress corrosion crack (PWSCC) growth rate tests were conducted to investigate crack growth properties of Alloy 600/182. To investigate PWSCC growth rate, test facilities including water chemistry loop, autoclave, and loading system were constructed. In PWSCC crack growth rate tests, half compact-tension specimens were manufactured. These specimens were then placed inside of the autoclave connected to the loop to provide primary water environment. Tested conditions were set at temperature of $360^{\circ}C$ and pressure of 20MPa. Real time crack growth rates of specimens inside the autoclave were measured by Direct Current potential drop (DCPD) method. To confirm inter-granular (IG) crack as a characteristic of PWSCC, fracture surfaces of tested specimens were observed by SEM. Finally, crack growth rate was derived in a specific stress intensity factor (K) range and similarity with overseas database was identified.

• Journal of ILASS-Korea
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• v.27 no.3
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• pp.126-133
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• 2022

The objective of this is to experimentally investigate the effect of mixed jet on the oxygen transfer characteristics with the primary nozzle area ratio of an annular nozzle ejector for the application of a microbial fuel cell. A direct visualization method with a high speed camera system was used to capture the horizontal mixed jet images, and a binarization technique was used to analyze the images. The clean water unsteady state technique was used for the oxygen transfer measurement. The air-water mixed jet discharging into a water tank behaved similar to a buoyancy or horizontal jet with the primary nozzle area ratio. It was found that an optimum primary nozzle area ratio was observed where the oxygen transfer performance reached its maximum value due to the decrease of air volume fraction and the increase of jet length and air bubble dispersion.

• Tribology and Lubricants
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• v.27 no.5
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• pp.240-248
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• 2011

Mechanical face seal installed in primary heat transport pump used for heavy water reactor prevents leakage of working fluid using thin working fluid film between primary seal ring and mating ring. If the leakage of working fluid exceeds the allowable volume, serious accident can be happened by the trouble of primary heat transport pump. The thinner fluid film exists between primary seal ring and mating ring, the less working fluid leaks out. On the other hand, if the thickness of fluid film is not enough, the life of mechanical face seal will be reduced by friction and wear. Therefore appropriate design is necessary to maximize the performance and life of mechanical face seal. In this study, numerical analysis using finite volume method was conducted to investigate the performance of mechanical face seals which have same deep straight groove and 11 different net coning values. As results, equilibrium clearance between primary seal ring and mating ring, leakage volume of working fluid, friction torque on sealing surface and stiffness of working fluid film were obtained. With increasing net coning value, equilibrium clearance and leakage volume increase, and friction torque and stiffness of fluid film decrease.