• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.199-212
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• 2007

If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or dispersal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions. For the tests we examined the failure of bolts and the deformation of flange to evaluate a loss or dispersal of radioactive material and measured the shielding thickness using a ultrasonic thickness gauge to assess a loss of shielding integrity. The strains and accelerations acquired from tests were compared with those by analyses to verify the impact dynamic analysis. The analytic results were larger than the those of test so that the analysis showed the conservative results. Finally, we evaluated the safety of the type IP-2 transport package under the stacking test condition using a finite element analysis. Under the stacking test condition, the maximum Tresca stress of the shielding material was 1/3 of the yielding stress. Two kinds of a type IP-2 transport package were safe for the free drop test condition and the stacking test condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.117-123
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• 2012

In order to dispose of the LILW(low and intermediate level radioactive waste) stored at KAERI, the radwaste drum assay system will be introduced to evaluate the radioisotopes inventory of stored drums. At present, the construction project of the dedicated assay facility to operate it and carry out routine maintenance of that equipment has been conducting at the radwaste treatment facility. Since that facility will be constructed in front of a 1st radwaste storage facility as well as the radwaste drums to be assayed and the transmission source in the radwaste drum assay system are in that facility, they could act as the radioactive sources and then, would affect the dose rate at the inside and the outside of the facility. Therefore, the radiation shielding should be evaluated through the concrete wall near to the radioactive sources whether the wall thickness is sufficient against the regulations. In this study, the radiation safety for the concrete wall around the radiation controlled area in the radwaste drum assay facility was evaluated by the MCNP code. From the evaluation results, the thickness of those concrete walls which are under consideration of about 30 cm was enough to shield the radiation from the radioactive sources.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.64-72
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• 2004

The dryer (ten per unit) are operating to remove tritium in PHWR(Pressurized Heavy Water Reactor). There are coming out heavy water adsorbent from operating the dryer (95 drums for ten year per unit) The amount of radioactivity of heavy water adsorbent almost exceed ninety times more than disposal limit-in-itself showed by The Ministry of Science and Technology. It has to be disposed whole radioactive waste products, however there are problems of increase at the expense of their permanent disposal. In this research, We have studied how to remove kinds of nuclear materials and amount of tritium with in heavy water adsorbent. As the result we could develop disposal equipment and apply it. D20 adsorbent have to contain below Gamma nuclide O.3Bq/g and tritium 100Bq/g "The Regulation for disposal of the radioactivity wastes" showed by The Ministry of Science and Technology. There fore. So as to remove amount of tritium and kinds of nuclear materials (DTO) we needed a equipment. Also we have studied how to remove effectively radioactivity with in Adsorbent. As cleaning heavy water adsorbent and drying on each condition (temperature for drying and hours for cleaning). Because there is something to return heavy water adsorbent by removing impurities within adsorbent when it is dried o high temperature. After operating, we have been applying this research to the way to dispose heavy water adsorbent. Through this we could reduce solid waste products and the expense of permanent disposal of radioactive waste products and also we could contribute nuclear power plant run safely. According to the result we could keep the best condition of radiation safety super vision and we could help people believe in safety with Radioactivity wastes control for harmony with Environment.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• Journal of the Korean Institute of Gas
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• v.2 no.1
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• pp.74-82
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• 1998

The grain boundary etching method as a technique for assessing degradation of structural materials used at elevated temperature has received much attention since it is simple, inexpensive and easy to apply to real plant components. In this study, the technique is applied to some aged petroleum and chemical plant components such as reactors and drums. As a degradation parameter, intersection number ratio ($N_i/N_o$), is employed. The intersection number ratio ($N_i/N_o$) is defined as the ratio of intersection number ($N_i$) obtained from 5-minute picric acid etched surface to the number ($N_o$) obtained from Nital etched surface. In order to study degradation level, several relationships were measured such as the correlation between shift in ductile brittle transition temperature, $({\Delta}DBTT)_{sp}$ and intersection number ratio, ($N_i/N_o$) and the correlation between the measured ($N_i/N_o$) values and Larson-Miller Parameter values.

• Progress in Medical Physics
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• v.31 no.4
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• pp.194-204
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• 2020

This study investigated and analyzed the Korean and international status of radioactive waste management for medical linear accelerators (linacs) and proceed prior research to suggest radiation safety regulations and guidelines for the safe use of radiation. We analyzed the number of linacs installed in the radiation oncology departments of 103 institutions. In addition, we analyzed the procedures and standards for disposal in Korea and foreign countries. For foreign countries, we analyzed the status based on reports from the United States, Japan, Europe, and Canada. A total of 182 linacs are installed in Korea and 95% of them use more than 10 MV of energy. In Korea, standards for managing radioactive waste from a linac, disposal procedures, and clearance criteria have yet to be established. Therefore, radioactive waste is disposed of in different ways depending on the hospitals where they originate. Japan, the US, and Canada have recommended clearance levels and procedures for linacs. Other countries have provided management guidelines for research or large-scale accelerators, but not for medical purposes. In this study, we investigated the management of radioactive waste from medical linacs in Korea and abroad. Several foreign countries have suggested a clearance level and criteria for disposing of waste storage drums. For the safe management of medical linacs, it is necessary to establish safety management regulations. In Korea, standards for disposal, such as radiation or dose limits, are required for medical linacs. A system for clearance when disposing at a medical institution should be created.

• Journal of the Korean Institute of Gas
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• v.25 no.3
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• pp.1-8
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• 2021

Among the flare systems that handle discharged substances from safety valves, the knockout drum was a key facility for safety, but the installation standards were not clear, so it was necessary to review the standards acceptable to the workplace and regulatory agencies. After investigating the domestic and foreign technical standards of knockout drums and the deficiencies of previous studies, research was first conducted on the degree of mass discharge, the installation location of the intermediate knockout drum, and the effect of changes in the composition of the discharged material. As a result of the study under the process simulation conditions, the gas-liquid separation of the knockout drum was completed in a small amount of less than 7,500kg/hr. However, when more than that was released, the gas-liquid separation effect was small even with the addition of an intermediate knockout drum. In addition, when the composition ratio of the material easily condensed was increased (molar fraction 10%), the gas-liquid separation effect of the knockout drum increased in the case of mass release. The gas-liquid separation effect was analyzed to be greater when the knockout drum was installed adjacent to the stack than the knockout drum was installed adjacent to the process equipment.

• Journal of Radiation Protection and Research
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• v.20 no.4
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• pp.237-243
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• 1995

An estimate is made on the potential generation rate of H: from radiolysis of the Paraffin-wax encapsulant Proposed for the solidified liquid concentrate wasteform. The results show that the radiolytic Production of $H_2$ from paraffin-wax-encapsulated waste is dominated by the radiation energy released from $^{60}Co$. The radiolytic production of $H_2$ will proceed at an initial rate equivalent to aproximately $4.4{\times}10^2cm^3yr^1$ in 200 litre drums that are partly filled with 120 litres of encapsulated waste. The gas production rate will fall to a value of $7.2cm^3yr^1$ after 100 years. The lower flammable limit for $H_2$ in air will be reached in about 25 years and the lower explosive limit for $H_2$ in air would not be reached in 1000years. The timescale in which these safety-related limits are reached is strongly dependent on the level of filling of each waste drum. A reduction of the air space inside each drum would reduce the time required to reach the lower flammable limit.

• Journal of Environmental Health Sciences
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• v.39 no.3
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• pp.197-210
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• 2013

Several US veterans stationed in Korea have told the press that around 250 fifty-five gallon drums of herbicides, including Agent Orange, were buried at Camp Carroll in 1978. Joint Korean-US Agent Orange investigation and environmental and health studies were started in July 2011. Korean soldiers and military personnel who served in Vietnam during 1964-1973 or near the demilitarized zone in Korea during 1967-1970 were exposed to dioxincontaminated Agent Orange. The joint Korean-US Agent Orange investigation team found that herbicides, pesticides, solvents and other chemicals -not Agent Orange- were buried at Camp Carroll. However, there remains the possibility that Agent Orange was stored and buried at Camp Carroll or other military camps in Korea. Adverse health effects have not been clearly explained despite a number of health studies among veterans in Korea with potential Agent Orange exposure. Although the Korean government has been compensating veterans and military personnel with 18 presumptive-service-connected-diseases and their offspring with three diseases, there are many veterans, military personnel and civilians who require the government°Øs support. The environmental study on contaminated sites and health studies among veterans and civilians were initiated three or four decades after possible Agent Orange contamination and exposure. Several toxic chemicals, including dioxin-contaminated Agent Orange, could remain in the environment and could have hazardous effects on the health of exposed people for more than several decades. Further environmental investigations and health studies are needed to ensure public safety and health, and government support should be guaranteed for people potentially exposed to these toxic chemicals.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.109-116
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• 2003

The amount of radioactive waste is expected to be increased continuously because of the rapid growth of the domestic nuclear industry, full power operation of the HANARO reactor and the increased research activities of the nuclear fuel cycle. Accordingly the efforts are focused to achieve the handling of radioactive waste in safe and reduce the volume of radioactive waste. The PIEF is carrying out the PIE (post irradiation examination) of spent fuel rods related to the identification of cause defect and evaluation of integration safety. This study describes the technologies and experiences of compaction, shredding and cutting of the solid radioactive waste used in the PIE. The quantity of the high level waste was reduced by 1/12 using the 100-ton compressor installed in hot-cell. Also middle and low level waste was reduced by 1/8 using the 60-ton compressor installed in intervention area. Plastic drums were shredded by crusher to be compacted in the ratio of 1/5, used filters in the ratio of 1/6 and the number of drum is also reduced by cutting procedure for the non-volatile materials such as metal.