• Proceedings of the Korean Society of Soil and Groundwater Environment Conference
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• 2001.04a
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• pp.81-84
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• 2001

고준위 방사성 폐기물 영구처분 안전성을 평가하기 위하여 입력 자료로 처분장 주변 각 암 반에서의 지하수 유동 속도 및 유동 시간이 요구된다. 이러한 유동 속도와 시간은 대부분의 경우 단일 값이 요구되지만 고준위 방사성 처분의 경우 지하 매질의 불확실성을 고려하기 위하여 확률론적 분석이 요구된다. 지하수 유동 속도 및 시간이 확률밀도함수로 표시되기 위해서는 기존의 방법에서는 수리 해석의 입력 인자 값들을 변화시키면서 반복적인 계산을 수행하는 방법이 사용되었다. 그러나 이러한 방법론의 한계를 극복하기 위해 최근 섭동 이론을 이용한 adjoint 방법론이 사용되고 있는 바 이를 이용하여 가상 처분장에서의 지하수 유동 속도와 시간을 확률론적으로 해석하였다.

• Proceedings of the Korean Society of Soil and Groundwater Environment Conference
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• 2005.04a
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• pp.288-291
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• 2005

방사성폐기물 처분연구의 일환으로 단열 암반지역의 지하수 유동에 있어서 각 단열조의 투수량계수를 알아보기 위하여 연구지역에 설치된 3기의 시추공에서 초음파 검층, 정압주입시험 및 유동차원 해석을 수행하였다. 단열은 방향성, 틈의 크기 등의 그 분포 특성으로 인해 각 시험구간내의 지하수 유동에 있어서, 유동형태 및 단열투수량계수를 좌우하므로 일반적으로 수리특성에 널리 이용되는 다공성매질의 연속체 개념을 통한 해석의 적용에 신중성을 고려할 필요가 있다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.276-282
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• 2003

The geoscientific study on geological disposal for radioactive wastes has established the stepwise site characterization program, methods and investigation technology. However the intrinsic properties of geological material such as heterogeneity and scale dependent properties make difficulty on satisfactory understanding of geological conditions. To avoid unnecessary time delay and unexpected extra-cost for site investigation, the accurate and complete site investigation program should be established in a stepwise manner and the QC programs for investigation methods and procedures. Moreover, the technical requirements and preferences for a repository should be distinguished and be assessed at the end of each investigation step.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.385-390
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• 1998

방사성 페이온교환수지 아스팔트고화체를 처분장 등지에서 장기간 저장시 안전성 확보를 위하여 물리적 강도가 높고 고화체내에서 방사성핵종의 침출저항성 및 처리시 감용의 효과가 우수한 고화체 연구가 필요하게 되었다. 실험에 사용된 이온교환수지는 입상형 양이온 교환수지를 대상으로 하였으며 고화매질로서는 도로포장용으로 생산되는 직류아스팔트 60/70을 사용하였다. 고화보조제는 방사성 고체패기물 포장시 사용되어 폐기물로 발생되는 페폴리에틸렌(폐PE) 필름을 사용하였다. 실험결과 고화체의 형태안정성은 PE 함유량이 10 wt% 이상일 때 고화체 형태를 그대로 유지할 수 있으며 압축강도는 414 kPa(60 psi) 이상을 나타내었다. 최적의 운전조건은 이온교환수지, PE 함유량이 건조기준으로 각각 30~50 wt%, 10~25 wt% 이며, 고화온도는 170~20$0^{\circ}C$이다. 고화체의 침출특성은 확산 (diffusion) 으로 해석이 가능하며, 유효확산계수(De)는 Cs, Co의 경우 각각 1.621$\times$$10^{-7}$, 1.186$\times$$10^{-9}$ $\textrm{cm}^2$/day로 나타나고, Leachablity index는 각각 11.7, 13.8로 미국 원자력위원회 (NRC)가 요구하는 기준값 6보다 훨씬 높게 나타났다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.134-142
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• 2005

The experimental setups for flow visualization and processes identification in laboratory scale (so cal led Flow Lab.) has developed to get ideas and answer fundamental questions of flow and migration in geologic media. The setup was made of a granite block of $50{\times}50cm$ scale and a transparent acrylate plate. The tracers used in this experiments were tritiated water, anions, and sorbing cations as well as an organic dye, eosine, to visualize migration paths. The migration plumes were taken with a digital camera as a function of time and stored as digital images. A migration model was also developed to describe and identify the transport processes. Computer simulation was carried out not only for the hydraulic behavior such as distributions of pressure and flow vectors in the fracture but also for the migration plume and the elution curves.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.293-300
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• 2005

Ultrasonic decontamination of the type 304 stainless steel specimen loosely contaminated with $Eu_2O_3$ powders was investigated. Decontamination factors (DFs) by the three kinds of ultrasonic media such as water, pure PFC (Pefluorocarbon, $C_7F_{16}$) and a mixed solution of $99.9\;vol\%\;PFC\;and\;0.1\;vol\%$ anionic surfactant were determined. The determined DF values were 20, 50 and 200, respectively. This significant difference in the decontamination factors for the different decontamination solution was well explained by the surface tension of the media as well as the interaction between the positively charged surface of $Eu_2O_3$ powders and the anionic surfactant. Ultrasonic decontamination behavior of the loosely contaminated metal specimens such as plate, pipe, welding specimen and crevice specimen in the mixed solution of PFC and anionic surfactant was also investigated. The contaminants were completely removed for the tested specimens except for the longest specimen. For 6-cm long pipe specimen, however, $98.5\%$ of the contaminants were removed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.77-84
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• 2010

Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-$Li_2O$ molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.295-303
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• 2012

Based on the data observed and analyzed on a groundwater flow system in the KURT (KAERI Underground Research Tunnel) site, the transport of radionuclides, which were assumed to be released at the supposed position, was calculated on the time-domain. A groundwater pathway from the release position to the surface was identified by simulating the groundwater flow model with the hydrogeological characteristics measured from the field tests in the KURT site. The elapsed time when the radionuclides moved through the pathway is evaluated using TDRW (Time Domain Random Walk) method for simulating the transport on the time-domain. Some retention mechanisms, such as radioactive decay, equilibrium sorption, and matrix diffusion, as well as the advection-dispersion were selected as the factors to influence on the elapsed time. From the simulation results, the effects of the sorption and matrix diffusion, determined by the properties of the radionuclides and underground media, on the transport of the radionuclides were analyzed and a decay chain of the radionuclides was also examined. The radionuclide ratio of the mass discharge into the surface environment to the mass released from the supposed repository did not exceed $10^{-3}$, and it decreased when the matrix diffusion were considered. The method used in this study could be used in preparing the data on radionuclide transport for a safety assessment of a geological disposal facility because the method could evaluate the travel time of the radionuclides considering the transport retention mechanism.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.93-103
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• 2003

IAEA, FT-04-020, and ANS 16.1, standard leaching test methods, were evaluated comparatively with their test results. Leaching index of $^{60}$ Co and $^{137}$ Cs by ANS 16.1 method for waste forms of paraffin and cement were above 6.0. Their leaching behavior were depending on the type of matrix and leachant. Leachability of $^{60}$ Co for cement waste form was higher in simulated seawater than do-mineralized water, and was higher in de-mineralized water for paraffin waste form. leachability of $^{60}$ Co was contrary to $^{137}$ Cs. Cumulative fraction leached of $^{60}$ Co was higher in order or IAEA ＞ ANS ＞ FT in a cement waste form.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.3
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• pp.201-209
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• 2004

Important operation parameters and performance of a high temperature ceramic candle filter system were evaluated through a series of demonstration tests at a pilot-scale vitrification plant. At the initial period of each test, due to the growth of dust cake on the surface of ceramic candles, the pressure drop across the filter media increased sharply. After that it became stable to a certain range and varied continuously proportion to the face velocity of off-gas. On the contrary, at the initial period of each test, the permeability of filter element decreased rapidly and then it became stable. Back flushing of the filter system was effective under the back flushing air pressure range of 3∼5 bar. Based on the dust concentrations measured by iso-kinetic dust sampling at the inlet and outlet point of HTF, the dust collection efficiency of HTF evaluated. The result met the designed performance value of 99.9%. During the demonstration tests including a hundred hour long test, no specific failure or problem affecting the performance of HTF system were observed.