• Journal of Animal Environmental Science
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• v.2 no.1
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• pp.1-11
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• 1996

This study was conducted to analyze the present status and the levels of mechanization and automation for raising, feeding, water supply, propagation, health management, ventilation and heat control, data analysis, and etc, and to establish the guide of the future study on development of the optimum production system models of swine facility from the results of this analysis. The scheme of the future study on the development of the optimum production system model of swine facility was established as follows : 1. A Korean and environmental control type slatted windowless swine housing model would be developed according to the following basis : \circled1 Boars, gilts and sows, delivery sows should be raised individually and piglets, growing pigs, and finishing pigs should be raised in group. \circled2 The arrangement of furrowing house were two rows of furrowing crates facing the center aisle. 2. The environmental control system and waste management system that are suitable to Korean and environmental control type slatted windowless swine housing model would be developed. 3. An electronic identification device would be developed. 4. The automatic individual wet feeding system by electronic identification device and computers would be developed. 5. The individual management system would be developed, which could manage individually the breeding pigs by the electronic identification device. 6. An expert system would be developed, which could manage the health and data base of pigs.

• Journal of the Korea Concrete Institute
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• v.26 no.1
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• pp.71-78
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• 2014

The long term integrity of concrete cask is very important for spent nuclear fuel dry storage system. However, there are serious concerns about early deterioration of concrete cask from creaking and corrosion of reinforcing steel by chloride ion because the cask is usually located in seaside, expecially by combined deterioration such as chloride ion and heat, carbonation. This study is to investigate the relation between temperature and chloride ion diffusion of concrete. Immersion tests using 3.5% NaCl solution that were controlled in four level of temperature, i.e. 20, 40, 65, and $90^{\circ}C$, were conducted for four months. The chloride ion diffusion coefficient of concrete was predicted based on the results of profiles of Cl- ion concentration with the depth direction of concrete specimens using the method of potentiometric titration by $AgNO_3$. Test results indicate that the diffusion coefficient of chloride ion increases remarkably with increasing temperature, and there was a linear relation between the natural logarithm values of the diffusion coefficients and the reciprocal of the temperature from the Arrhenius plots. Activation energy of concrete in this study was about 46.6 (W/C = 40%), 41.7 (W/C = 50%), 30.7 (W/C = 60%) kJ/mol under a temperature of up to $90^{\circ}C$, and concrete with lower water-cement ratio has a tendency towards having higher temperature dependency.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.147-157
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• 2004

The production of $^{14}C$ occurs in the Moderator(MOD), Primary Heat Transport System (PHTS), Annulus Gas System(AGS) and Fuel in the CANDU reactor. Among the four systems, The MOD system is the largest contributor to $^{14}C$ production(approximately 94.8%). $^{14}C$ is distributed of $^{14}CO_2$, $H_2^{14}CO_3$, $H^{14}{CO_3}^-$ and $^{14}{CO_3}^{2-}$ species as a function of the pH of water. Of these species, $H_2^{14}CO_3$ and $H^{14}{CO_3}^-$ form are predominant because the pH of MOD system is ＞ 5. In this paper, adsorption-desorption characteristics of bicarbonate ion (${HCO_3}^-$) by IRN 150 resin was investigated. ${HCO_3}^-$ ion existed in neutral condition(app. pH 7)was reacted with ion exchange resin (IRN-150) and saturated with it. Then $NaNO_3$ and $Na_3PO_4$ solutions selected as extraction materials were used to make an investigation into feasibility of ${HCO_3}^-$ extraction from resin saturated with ${HCO_3}^-$. Desorption of $CO^{2+}$ and $Cs^+$ ion by $Na^+$ ion was not occurred, and desorption of ${HCO_3}^-$ ion by ${NO_3}^-$ and ${PO_4}^{3-}$ was occurred slowly. Also, the status of ion exchange which is used in Wolsong NPPs and generation of spent resin yearly were surveyed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.389-396
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• 2018

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.

• Journal of the Korean Ceramic Society
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• v.41 no.3
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• pp.221-228
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• 2004

Basic properties of artificial lightweight aggregate by using waste dusts and strength properties of LWA concrete were studied. Bulk specific gravity and water absorption of artificial lightweight aggregates varied from 1.4 to 1.7 and 13 to 16%, respectively. Crushing ratio of artificial lightweight aggregate was above 10% higher than that of crushed stone or gravel. As a result of TCLP leaching test, the leaching amount of tested heavy metal element was below the leaching standard of hazardous material. Slump, compressive strength and stress-strain properties of LWA concrete made of artificial lightweight aggregate were tested. Concrete samples derived from LWA substitution ratio of 30 vol% and W/C ratio of 45 wt% showed the best properties overall. Thermal insulation and sound insulation characteristics of light weight concrete panel with the optimum concrete proportion were tested. Average overall heat transmission of 3.293W/㎡$^{\circ}C$ was observed. It was higher by about 15% than those of normal concrete made by crushed stone. Sound transmission loss of 50.9 ㏈ in frequency of 500 ㎐ was observed. It was higher by about 13% than standard transmission loss.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.137-144
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• 2005

Silica in primary system combines with an alkali grammatical particle metal and forms the zeolite layer which is hindering the heat transfer on the surface of the cladding. Zeolite layer becomes the cause of the damage in this way. The problems of the NPP's primary system have been issued steadily by EPRI. Through a series of experiments of the laboratory scale, we confirmed the applicability of NF membrane for silica removal, as silica rejection rate of NF membrane is about $60\;{\sim}\;70\%$ and boron rejection rate is about $10\;{\sim}\;20\%$. We accomplished a site experiment about four NF membranes manufactured by FilmTec and Osmonics Inc. In experiment using 400L of SFP water, when operation pressure is $10kg_{f}/cm^2$, we confirmed that the silica rejection rate of NF90-2540 manufactured by FilmTec Inc. is about $98\%$, boron rejection rate is about $43\%$. The silica rejection rate of NF270-2540 is about $38\%$, boron rejection rate is about $3.5\%$. Afterward, through additional experiments, such as long term characteristic experiments, we are going to design a optimum NF membrane system for silica removal.

• Resources Recycling
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• v.7 no.3
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• pp.17-26
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• 1998

A study on the recovery of silver and nitric acid in the liquid waste resulted from the mediated electrochemical oxidation(MEO) process was conducted. The removal of silver in the concentrated nitric acid solutions was carried out by the electrodeposition. The removal efficiency more than 98% could be obtained in nitric acid concentrations less than 3 M with the current efficiency of nearly 100%. The experimonts on the evaporation for the recovery of nitric acid were performed as well. At the evaporation factor of 25., the degree of nitric acid recovery in 3.5 M nitric acid solution containing 0.5 to 1.0 mol% NaNO, was 80~90% resulting in 2.8~3.1 M nitric acid. The design factors and operating conditions of the distillation tower were analyzed by using MEH model derived by Maphtali-Sandholm with the throughput of 4 kg/hr for the enrichment of dilute nitric acid solution recovered by evaporation to reuse in the MEO process. The distillation column composed of eleven theoretical stages having the overall tray efficiency of 70% are needed to obtain 1.03 kg/h of 12M nitric acid and 2.97 kg/h of water with feed being introduced to the column at tray 6 from the bottom at the reflux ratio of 0.25, the reboiler with the heat load of 2.7 kW, and the condenser with the cooling load of 0.5 kW.

• Resources Recycling
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• v.28 no.4
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• pp.30-36
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• 2019

In this study, Li powder was recovered from the by-product of LNO ($Li_2NiO_2$) process, which is the positive electrode active material of waste lithium ion battery, through the $CO_2$ thermal reaction process. In the process of recovering Li powder, the $CO_2$ injection amount is 300 cc/min. The $Li_2NiO_2$ award was phase-separated into the $Li_2CO_3$ phase and the NiO phase by holding at $600^{\circ}C$ for 1 min. After this, the collected sample:distilled water = 1:50 weight ratio, and after leaching, the solution was subjected to vacuum filtration to recover $Li_2CO_3$ from the solution, and the NiO powder was recovered. In order to increase the purity of Ni, it was maintained in $H_2$ atmosphere for 3 hours to reduce NiO to Ni. Through the above-mentioned steps, the purity of Li was 2290 ppm and the recovery was 92.74% from the solution, and Ni was finally produced 90.1% purity, 92.6% recovery.

• Tunnel and Underground Space
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• v.32 no.1
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• pp.59-77
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• 2022

Currently, the design reference temperature of the buffer material for disposing of high-level radioactive waste is less than 100℃, so if the heat dissipation capacity of the buffer material is improved, the spacings of the disposal tunnel and the deposition hole in the repository can be reduced. First of all, this study tries to analyze the criteria for thermal-hydraulic-mechanical performance of the buffer materials and to investigate the researches regarding the enhanced buffer materials with improved thermal conductivity. First, the thermal conductivity should be as high as possible and is affected by dry density, water content, temperature, mineral composition, and bentonite type. the organic content of the buffer material can have a significant effect on the corrosion performance of a canister, so the organic content should be low. In addition, hydraulic conductivity of the buffer material should be less than that of near-field rock and swelling pressure should be appropriate for buffer materials to function properly. For the development of enhanced buffer materials, additives such as sand, graphite, and graphite oxide are typically used, and a thermal conductivity can be greatly improved with a very small amount of graphite addition compared to sand.

• Journal of Radiation Industry
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• v.17 no.3
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• pp.283-292
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• 2023

Spent nuclear fuel corresponds to high-level radioactive waste that has high decay heat and radioactivity. Accordingly, Spent nuclear fuel withdrawn from the reactor core is primarily stored and managed in a spent nuclear fuel pool in the nuclear power plant to reduce decay heat and radioactivity. In Korea, most nuclear power plant store all spent nuclear fuel in a spent nuclear fuel pool. For wet storage, there are no defense in depth different with reactor core. The study related to spent nuclear fuel pool accident should be carried out to ensure safety. Therefore, it is necessary to analyze previous study cases related to safety of spent nuclear fuel pool and accident cases to build foundational knowledge. The Objective of this study is to analyze study cases of safety assessment and cases for spent nuclear fuel pool accident. For analyzing study cases of safety assessment, possible phenomena when spent nuclear fuel pool accident occurring identified, Subsequently, study cases for safety assessment about each phenomena were investigated, and materials & methods and results for each study are analyzed. For analyzing cases for spent nuclear fuel pool accident, we analyzed accident cases caused by loss of cooling and loss of coolant in spent nuclear fuel pool. Subsequently, causes and change of water level and temperature by each accident case are analyzed. As a result of the analysis on study cases of spent nuclear fuel pool accident, the results of the study conducted by each research institute were vary depending on the computer code, materials & methods of experiment and major assumptions used in the study. As a result of analyzing cases for spent nuclear fuel pool accident, it was found that accident cases for loss of cooling is more than cases for loss of coolant accident. Even though the types of accident in spent nuclear fuel pool were similar, the specific causes were different by each accident case. All the accident cases analyzed did not lead to severe accidents, such as nuclear fuel being exposed to the air. The result of this study will be used as fundamental data for study on spent nuclear fuel pool accident that will be conducted in the future.