• Proceedings of the Korean Vacuum Society Conference
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• 2011.08a
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• pp.149-149
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• 2011

아르곤 기체의 방사세기 또는 그 세기 비는 플라즈마 공정 진단에서 일반적으로 사용된다. 본 실험에서는 100 mTorr 압력 조건하의 유도결합 플라즈마(13.56 MHz)에서 E-H 모드 전이 영역, rf 바이어스(12.5 MHz) 전력 인가 및 N2 혼합 시 단순화한 충돌-방사 모델에 기초한 광방사 세기비 방법을 적용하여 플라즈마 변수를 진단하였다. 개발 프로그램 기반의 분광기를 사용하여 아르곤 기체의 특정 파장(750.4, 751.5 그리고 811.5 nm)들을 관측하였고, 동일한 조건하에서 정전 탐침법으을 이용하여 전자 에너지 분포함수의 변화도 측정 하였다. 맥스웰 전자 에너지 분포를 가정하는 일반적인 경우와 비교하여 볼 때 실제적인 전자 에너지 분포함수의 측정은 전자의 가열 메커니즘에 대한 상세한 정보를 제공함과 동시에 플라즈마 재흡수에 대한 보정을 가능하게 해준다. 광방사 세기비법에 의해 측정된 결과에 의하면, 750.4 nm/751.5 nm는 높은 에너지(>13.08 eV)의 전자들의 유효 전자온도에 대한 정보를 나타내는 반면 811.5 nm/750.4 nm는 아르곤 준안정 준위 밀도(1s5)에 대한 정보를 제공하게 된다. 수행된 실험 조건하에서, 측정된 준안정 준위 밀도는 E-H 모드 전이 영역에서 최대값을 나타내었고 바이어스 전압 및 N2 기체 혼합 비율이 증가함에 따라 감소하는 결과를 얻었다. 유효 전자온도의 경우 광방사 세기비법과 정전 탐침법 모두 같은 결과를 보여 주었는데, E-H 모드 전이 영역에서는 전자온도는 거의 일정하였고 바이어스 전압 및 N2 기체 혼합 비율이 증가함에 따라 전자온도는 증가하였다. 이러한 실험 결과는 방전 모드 전이, 바이어스 인가 그리고 혼합 기체 사용하는 공정 플라즈마를 이해하는데 있어 이들 변수의 진단이 중요한 요소임을 보여준다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.147-154
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• 1996

외국에서는 도로건설시 성토관리용으로 '방사성동위원소(Radioisotope : 이하 RI)를 이용한 습윤밀도 및 함수량 측정장비'가 많이 사용되고 있는 추세이다. 국내에서는 1989년 제정된 건설 기술관리법 감리전문회사건립 등록기준에 밀도\ulcorner함수량 측정기의 보유가 명시되어 있고 시공성 향상 차원에서 도입될 예정이다. 그러나 국내의 토양에 맞게 제작되어지지 않았고 사용상의 방사선 관련 제약 때문에 기대만큼 실제 사용이 되지 않고 있다.$^{(1)}$ 연구발표내용은 국내에서 제작, 실험하여 만들 RI계기개발의 기초가 되는 내용이며, 만들어질 계기의 부품과 계산과정 등을 예측하는 것이다.

• Journal of Radiation Protection and Research
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• v.40 no.4
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• pp.216-222
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• 2015

Facilities processing raw materials containing naturally occurring radioactive materials (NORM) may give rise to enhanced radiation dose to workers due to chronic inhalation of airborne particulates. Internal radiation dose due to particulate inhalation varies depending on particulate properties, including size, shape, density, and absorption type. The objective of the present study was to assess inhalation dose sensitivity to physicochemical properties of airborne particulates. Committed effective doses to workers resulting from inhalation of airborne particulates were calculated based on International Commission on Radiological Protection 66 human respiratory tract model. Inhalation dose generally increased with decreasing particulate size. Committed effective doses due to inhalation of $0.01{\mu}m$ sized particulates were higher than doses due to $100{\mu}m$ sized particulates by factors of about 100 and 50 for $^{238}U$ and $^{230}Th$, respectively. Inhalation dose increased with decreasing shape factor. Shape factors of 1 and 2 resulted in dose difference by about 18 %. Inhalation dose increased with particulate mass density. Particulate mass densities of $11g{\cdot}cm^{-3}$ and $0.7g{\cdot}cm^{-3}$ resulted in dose difference by about 60 %. For $^{238}U$, inhalation doses were higher for absorption type of S, M, and F in that sequence. Committed effective dose for absorption type S of $^{238}U$ was about 9 times higher than dose for absorption F. For $^{230}Th$, inhalation doses were higher for absorption type of F, M, and S in that sequence. Committed effective dose for absorption type F of $^{230}Th$ was about 16 times higher than dose for absorption S. Consequently, use of default values for particulate properties without consideration of site specific physiochemical properties may potentially skew radiation dose estimates to unrealistic values up to 1-2 orders of magnitude. For this reason, it is highly recommended to consider site specific working materials and conditions and use the site specific particulate properties to accurately access radiation dose to workers at NORM processing facilities.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.43-51
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• 2004

In order to decommission the shielding concrete of KRR(Korea Research Reactor) -1＆2, it must be exactly determined activated level and range by neutron irradiation during operation. To determine the activated level and range, it must be sampled and analyzed the core sample. But, there are difficulties in sample preparation and determination of the measurement efficiency by self-absorption. In the study, the full energy efficiency of the HPGe detector was compared with the measured value using standard source and the calculated one using Monte Carlo simulation. Also. self-absorption effects due to the density and component change of the concrete were calculated using the Monte Carlo method. Its results will be used radioactivity analysis of the real concrete core sample in the future.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.866-871
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• 1995

방사성 폐기물 처분장 주변에서의 지하수 유동에 대한 민감도 분석을 수행하여 안전성 평가측면에서 필요한 성능측도에 미치는 영향을 파악하였다. 각 암반충의 투수계수 및 공극률의 변화에 대한 지하수 유속과 수두의 민감도와 경제 조건을 변경함으로 인해 지하수 유동시간에 미치는 영향을 adjoint 민감도 분석법에 의해 살펴보았다. 민감도 분석 결과, 본 논문에서 고려된 처분부 지의 경우 해수에 접한 경계 면에서는 해수의 밀도를 고려한 경계조건을 사용하는 것이 지하수 유동이 없다고 가정하는 경계조건보다 약간 보수적임을 보여주었고, 투수계수 변화에 따른 지하 수두 및 Darcy 속도는 처분동굴이 위치한 암반의 투수계수 변화에 매우 민감하고 실제적으로 동굴에서 멀리 떨어진 바닥 암반층의 투수계수 변화에는 민감하지 않았다.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.309-317
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• 1997

The measured radioactivity of gamma-emitting radionuclides in each radioactive waste drum using the non-destructive waste assay method is underestimated than real radioactivity in radioactive waste drum because the gamma-rays are attenuated within the medium. Therefore, the measured radioactivity should be corrected for the attenuation of gamma-rays. For the correction of the attenuation of gamma-rays, the attenuation correction method should be applied differently by considering the distribution and density of medium in radioactive wastes drum generated from nuclear power plants. In this study, the model drums were fabricated for simulating five types of radioactive waste drums generated from nuclear power plant and the optimum methods of the attenuation correction were experimentally determined to analyze the activity of radionuclides in the waste drum accurately using the segmented gamma scanning system. With the determination of the attenuation correction methods from the experimental results the transmission method and the average density method for the miscellaneous waste drum, the transmission method and the differential peak absorption method for the shielded miscellaneous waste drum were used to measure the density of medium in waste drums. Also, the average density method and the differential peak absorption method for the spent resin drum, the paraffin solidified drum, and the spent filter drum were used.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.275-284
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• 2006

In the measurement of gamma rays emitted from the nuclide in the radioactive waste drum, to analyze the nuclide concentration accurately, it is necessary to use the proper calibration standards and to correct for the attenuation of the gamma rays. Two drums having a different density were used to analyze the nuclide concentration inside the drum in this study. After carrying out the system calibration, we measured the gamma rays emitted from the standard source inside the model drum with changing the distance between the drum and the detector. The measured values were corrected with the three kinds of gamma attenuation correction methode, as a results, the error was less than 10 % in the low density drum and less than 25 % in the high density drum. The measured activity in the short distance was more accruable than in the long distance. The transmission correction for the mass attenuation showed good results(very Low error) compared to the mean density and the differential peak correction method.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.503-510
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• 2020

Concrete is one of the most widely used materials as the shielding structures of a nuclear facilities. It is also the most generated radioactive waste in quantity while dismantling facilities. Since the concrete captures neutrons and generates various radionuclides, radiation measurement and analysis of the sample was fulfilled prior to dismantle facilities. An HPGe detector is used in general for the radiation measurement, and effective correction factors such as geometrical correction factor, self-absorption correction, and absolute detector efficiency have to be applied to the measured data to decide exact radioactivity of the sample. Correction factors are obtained by measuring data using a standard source with the same geometry and chemical states as the sample under the same measurement conditions. However, it is very difficult to prepare standard concrete sources because concrete is limited in pretreatment due to various constituent materials and high density. In addition, the concrete sample obtained by core drill is a volumetric source, which requires geometric correction for sample diameter and self absorption correction for sample density. Therefore in recent years, many researchers are working on the calculation of effective correction factors using Monte carlo simulation instead of measuring them using a standard source. In this study we calculated, using Geant4, one of the Monte carlo codes, the correction factors for the various diameter and density of the concrete core sample at the gamma ray energy emitted from the nuclides ¹⁵²Eu and ⁶⁰Co, which are the most generated in radioactive concrete.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.202-206
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• 2003

In this study, a modified bituminization technology has been developed, which needs no grinding of the granular resin waste, and enables the solid form to keep its shape stability as good as that of a cemented solid from Also, the study intended to apply the developed technology to the practical treatment of radioactive resin waste. In the experiment, the granular type resin was used and the straight-run distillation bitumen with penetration rate 60/70 was used as the solidifying agent. The PE was used as the additive. The shape stability increased remarkably with the additive of PE, which act as a binder in the solid form. The shape of the solid form was maintained without failure during the long-term exposure test when the additive content of spent PE is more than 10wt%. The proper ranges of bitumen content, PE content and operating temperature are 30-50wt%, 10-20wt% and $180^{\circ}C$ respectively. The bituminized solid form of radioactive resin waste by the technology of this study has the remarkably superior quality than the conventional solid forms, partially for the shape stability.

• Journal of the Korean Society of Groundwater Environment
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• v.6 no.3
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• pp.111-119
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• 1999

The purpose of this study is the simulation of discontinuous rockmass and identification of characteristics of discontinuity network as a branch of the study on characteristics of groundwater system in discontinuous rockmass for evaluation of safety on disposal site of radioactive waste. In this study the site for LPG underground storage was selected for the similarities of the conditions which were required for disposal site of radioactive waste. Through the identification of hydraulic properties. characteristics of discontinuities and selection of discontinuity model around LPG underground storage facility. the applications of discrete fracture network model were evaluated for the analysis of pathway. The orientation and spatial density of discontinuities are primarily important elements for the simulation of groundwater and solute transportation in discrete fracture network model. In this study three fracture sets identified and the spatial intensity (P$_{32}$) of discontinuities is revealed as 0.85 $m^2$/㎥. The conductive fracture intensity (P$_{32c}$) estimated for the simulation area around propane cavern (200${\times}$200${\times}$200) is 0.536 $m^2$/㎥. Truncated conductive fracture intensity (T-P$_{32c}$) is calculated as 0.26 $m^2$/㎥ by eliminating the fracture with the iowest transmissivity and based on this value the pathway from the water curtain to PC 2. PC 3 analyzed.