• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.327-336
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• 2020

The high-flux advanced neutron application reactor (HANARO) is a research reactor with thermal power of 30 MW applied in various research and development using neutrons generated from uranium fission chain reaction. A degasifier tank is installed in the ancillary facility of HANARO. This facility generates gas pollutants produced owing to internal environmental factors. The degasifier tank is designed to maintain the gas contaminants below acceptable levels and is monitored using an analyzer in the gas sampling panel. If condensate water is generated and flows into the analyzer of the gas sampling panel, corrosion occurs inside the analyzer's measurement chamber, which causes failure. Condensate water is generated because of the temperature difference between the degasifier tank and analyzer when the gas flows into the analyzer. A heating system is installed between the degasifier tank and gas sampling panel to suppress condensate water generation and effectively remove the condensate water inside the system. In this study, we investigated the efficiency of the heating system. In addition, the variations in the pipe temperature and the amount of average condensate water were modeled using a wall condensation model based on the changes in the fluid inlet temperature, outside air temperature, and heating cable-setting temperature.

• Journal of Radiation Protection and Research
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• v.34 no.4
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• pp.195-200
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• 2009

effluents to the environment. The activity of carbon-14, one of the radioactive effluents, in the environment is already high level and its effect on radiation exposure to the public and the environment is insignificant; thus, NPPs did not perform the carbon-14 monitoring in effluents in the past. By the way, effluents of noble gas and particulate radioactive materials originated from nuclear fuels has been continuously reduced due to both the advancement of manufacturing and integrity technology for nuclear fuels and the improvement of operation methods of NPPs. Futhermore, the portion of dose assessment by tritium and carbon-14 to the public has been relatively increased because the lower limit of detection for low-energy beta sources, such as tritium and carbon-14, is low due to the advancement of radiation detection technology. In this paper, the technical background for carbon-14 monitoring in nuclear facilities was investigated using United States technical reports and papers. This paper also reviews whether carbon-14 monitoring is necessary or not based on the investigated documents.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.177-177
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• 2009

사용후 핵연료의 건식처리 시 핵연료 다발을 절단하여 voloxidation 즉 휘발산화처리를 하면 고온에 의해 분리가 가능한 핵분열생성물의 분리와 우라늄의 산화에 의한 부피팽창으로 핵연료가 쪼개져서 입도가 작아지고 또한 핵연료가 피복재에서 쉽게 박리되게 된다. 그 결과 폐기물 처리 시에 발열핵종으로 폐기물의 저준위화시에 분리가 요망되는 Cs-137이 분리되는 장점이 있어 습식 재처리에 있어서도 바람직하다. 건식처리에 있어서는 voloxidation 으로 처리된 피복재에는 금속 지르코늄에 불순물로 함유된 우라늄의 의한 방사화 생성물과 피복재 표변에 부착/침투한 방사화 생성물이 방사능을 갖게 된다. 이러한 부착된 TRU 잔류물은 통상 1% 미만으로 알파핵종의 방사능이 원자로에서 배출시에는 고준위 기준치의 약 100배 수준이었다가 30년 냉각후에는 약 1/10 수준으로 저준위화 된다. 지르코늄 금속중에 불순물로 함유된 우라늄의 방사화로 생기는 방사능은 고준위 기준치의 10% 를 넘지 않아서 피복재의 저준위화시에 고려할 필요가 없다. 발생열은 방출시에 고준위 기준치의 약 30 배 수준에서 5년 냉각후에는 기준치 미만이 되며 30년후에는 1/8000 정도로 저준위화 된다. 사용후 핵연료를 습시처리시에 발생하는 고준위 폐기물 중 약 1/4 가 피복재 (hull) 임을 고려하면 피복재의 저준위화는 사용후 연료의 건식처리에 있어서도 필수적인 과정이다. 특히 미국의 고준위 폐기물 처분장 Yucca Mt.의 포기와 우리의 고준위 폐기불 처분장이 공론화되는 싯점에서 저준위화는 매우 필요한 기술이다. 피복재는 방사성 물질의 침투두께가 0.01mm 미만이 대부분으로 저준위화에는 표면제염에 의한 저준위화가 주로 연구되어왔다. 표면제염에 의한 저준화는 이온 빔, laser에 의한 방법, dry ice 분사에 의한 방법이 시도되었다. 염소기체를 이용하여 지르코늄의 산화막을 제거하고자 하였으나 이 산화막이 안정적이어서 표변의 연마, 아크릴 칼의 사용, 표면을 눌러서 처리하는 등 전처리하여서 염소기체 반응에 의한 표면제거 실험이 가장 효과적임이 실험적 결과이었다. 이러한 전처리로 방사능을 1/100 수준으로 낮춘다고 하더라도 지르코늄 금속중에 불순물로 함유된 우라늄의 방사화에 의해 중저준위 폐기물의 범주에서 벗어나지 않으므로 재활용에는 제한이 있다. 또한 전처리(표면제염)하여 분리되는 고준위는 다른 고준위 염폐기물과 함께 처리하여 발열 핵종을 제거하면 중저준위화가 가능하다. 저준위화 된 hull폐기물에는 지르코늄 금속에 불순물로서 함유되어있는 우라늄에 의한 방사능을 갖는데 이들의 제거나 분리는 지르코늄 합금 피복재 원료물질에 불순물로 함유하는 우라늄의 함량을 낮추는 것과 유사한 문제이다. 현재까지 지르코늄합금 피복재에 우라늄이 불순물로 함유된 것을 사용함으로 원자로내에서 방사화되어서 방사능을 갖게 되는 것은 피할 수가 없다. 따라서 저준위화 처리된 피복재는 장기 보관으로 방사능을 감쇠시켜서 재활용하도록 한다. 처리 방법으로는 초고압 압축저장, 시멘트 고화, 합성암석에 의한 고화법 등으로 장기간 보관 후에 금속으로서 재활용한다.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.382-393
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• 2020

The engineered barrier system of high-level radioactive waste disposal must maintain its performance in the long term, because it must play a role in slowing the rate of leakage to the surrounding rock mass even if a radionuclide leak occurs from the canister. In particular, it is very important to clarify gas dilation flow phenomenon clearly, that occurs only in a medium containing a large amount of clay material such as a bentonite buffer, which can affect the long-term performance of the bentonite buffer. Accordingly, DECOVALEX-2019 Task A was conducted to identify the hydraulic-mechanical mechanism for the dilation flow, and to develop and verify a new numerical analysis technique for quantitative evaluation of gas migration phenomena. In this study, based on the conventional two-phase flow and mechanical behavior with effective stresses in the porous medium, the hydraulic-mechanical model was developed considering the concept of damage to simulate the formation of micro-cracks and expansion of the medium and the corresponding change in the hydraulic properties. Model verification and validation were conducted through comparison with the results of 1D and 3D gas injection tests. As a result of the numerical analysis, it was possible to model the sudden increase in pore water pressure, stress, gas inflow and outflow rate due to the dilation flow induced by gas pressure, however, the influence of the hydraulic-mechanical interaction was underestimated. Nevertheless, this study can provide a preliminary model for the dilation flow and a basis for developing an advanced model. It is believed that it can be used not only for analyzing data from laboratory and field tests, but also for long-term performance evaluation of the high-level radioactive waste disposal system.

• The Journal of Engineering Geology
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• v.24 no.4
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• pp.535-550
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• 2014

The origins and varieties of natural radioactive materials, including uranium and radon-222, were examined in a drilled borehole extending to a depth of 120 m below the surface in the Goesan area. In addition to core samples, eight groundwater samples were collected at different depths, using a double packer system and bailer, and their geochemical characteristics were determined. Most of the rock samples from the drilled core consisted of granite porphyry, with sedimentary rocks (slate, carbonate, or lime-silicates) and pegmatite occurring in certain sections. The pH of samples varied from 7.8 to 8.4, and the groundwater was of a Na-$HCO_3$type. Uranium and thorium concentrations in the core were < 0.2-14.8 ppm and 0.56-45.0 ppm, respectively. Observations by microscope and an electron probe microanalyzer (EPMA) showed that the mineral containing the natural radioactive materials was monazite contained in biotite crystals. The uranium, which substituted for major elements in the monazite, appeared to have dissolved and been released into the groundwater in a shear zone. Concentrations of Radon-222 in the borehole showed no close relationship with levels of uranium. The isotopes of noble gases, such as helium and neon, would be useful for analyzing the origins and characteristics of the natural radioactive materials.

• Journal of Radiation Protection and Research
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• v.4 no.1
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• pp.5-13
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• 1979

A simple method for determining the airborne concentration of radon daughter products has been developed, which is based on gross alpha counting of the air filter collections at several time intervals after completion of air sampling. The concentration of each nuclide is then obtained from an equation involving the alpha disintegrations, the sampling time, and the known numerical coefficients. The state of radioactive disequilibrium is also investigated. The atmosphere sampled in the TRIGA Mark-III reactor room was largely in disequilibrium. The extent of radioactive disequilibrium between radon daughter products seems likely depend on sampling times associated with turbulence conditions. The data obtained here will certainly provide useful information on the evaluation of internal exposure and calibration of effluent monitoring instruments.

• Journal of Radiation Protection and Research
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• v.25 no.2
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• pp.81-87
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• 2000

XOQ_DW code is currently used to assess the atmospheric dispersion fur the routine releases of radioactive gaseous effluents at Yonggwang nuclear power plants. This code was developed based on XOQDOQ code and an additional code is required to assess the atmospheric dispersion for potential accidental releases. In order to assess the atmospheric dispersion fer the accidental releases, XOQAR code has been developed by using PAVAN code that is based on Reg. Guide 1.145. The terrain data of XOQ_DW code inputs and the relative concentrations (X/Q) of XOQ_DW code outputs are used as the inputs of the XOQAR code through the interface with XOQ_DW code. By using this code, the maximum values of X/Q at exclusion area and low population zone boundaries except for sea areas were assessed as $1.33{\times}10^{-4}$ and $7.66{\times}10^{-6}$ sec/$m^3$, respectively. Through the development of this code, a rode system is prepared for assessing the atmospheric dispersion for the accidental releases as well as the routine releases. This developed code ran be used for other domestic nuclear power plants by modifying the terrain input data.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2001.11a
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• pp.41-41
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• 2001

포항가속기의 초고진공 세정기술은 초고진공 영역에서의 기체방출을 최소화하기 위한 것으로 발생한 오염원을 추적하여 큰 오염원 부터 시작하여 단계적으로 진행해 미세한 오염원 을 제거해 나가는 방법을 적용하고 있다. 이러한 극청정한 진공표면을 얻기 위해서는 표면과 오염물질 사이의 결합에너지를 극복해야 한다. 오염물 제거 방법으로는 물리.전기.화학적인 방법을 모두 적용하며 그리스 및 절삭유를 비롯해 흡착된 탄화수소와 불순물 성분 또는 산소나 황과 같은 반응성 원소와의 화합물 등을 효과적으로 제거한다. 또한, 세정 과정 에서 생성될 수 있는 수소, 불규칙한 산화물, 질화물, 염화물, 그리고 탄수화물을 최소화하여 초고진공 영역에 도달할 수 있는 방법을 제공한다. 본 논문에서는 포항가속기 연구소의 초고진공 환경을 확보하기 위한 화학세정 설비 및 응용기술, 주요 진공 구성재료의 표면 분석 결과를 소개하고자 한다.

• Journal of Radiation Protection and Research
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• v.9 no.1
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• pp.26-32
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• 1984

The exposure doses by the radioactive gaseous effluents from nuclear power plants are investigated in the two cases of normal operation and hypothetical accident. Gaussian equation is adapted in the normal operation as the diffusion model of effluents for long period, which uses annual average meteorological data. But the real time models have been used in the case of accidents which analyze the changes of wind direction and speed. In this study the annual exposure doses by the normal operation of Kori unit 1 during $1977{\sim}1982$ were calculated on the basis of the atmospheric diffusion factor by the Gaussian straight line model. And the image processing technique was suggested as the effective method through the wind tunnel experiments to get the characteristic value of atmospheric diffusion coefficient required especially in the accidents of nuclear power plants.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.767-774
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• 1995

Behavior of radionuclides such $^{60}$ Co, $^{54}$ Mn and $^{137}$ Cs in the incineration Process was Studied by trial burns of simulated wastes with radio-isotope tracers. Behavior of nonvolatiles, $^{60}$ Co and $^{54}$ Mn, was similar to that of particulate matters in the process. Decontamination factors(DFs) for $^{60}$ Co and $^{54}$ Mn were 4.7$\times$10$^{5}$ and 6.2$\times$10$^{5}$ , respectively. Behavior of semivolatile radio-isotope, $^{137}$ Cs, was temperature dependent. DFs for $^{l37}$Cs at In different incineration temperature of 85$0^{\circ}C$ and $700^{\circ}C$ were 2.8$\times$10$^3$ and 2.6$\times$10$^4$, respectively. Trial bums of dry active waste(DAW) transported from nuclear power station(NPS) Kori 3,4 were also performed. DF for gross $\beta$/${\gamma}$ radioactivity in DAW was 1.1$\times$10$^{5}$ . This was a little higher than the estimated value, which was calculated from the tracer test results and nuclides distribution in the DAW. Average emission concentration was 0.019 Bq/N $m^3$, which could meet the maximal permissible concentration(MPC) in stack emission.n.