• 방사성폐기물학회지
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• 제22권2호
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• pp.211-217
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• 2024

The growing significance of sustainable energy technologies underscores the need for safe and efficient management of spent nuclear fuels (SNFs), particularly via deep geological disposal (DGD). DGD involves the long-term isolation of SNFs from the biosphere to ensure public safety and environmental protection, necessitating materials with high corrosion resistance for DGD canisters. This study investigated the feasibility of a Cu-Ni film, fabricated via additive manufacturing (AM), as a corrosion-resistant layer for DGD canister applications. A wire-fed AM technique was used to deposit a millimeter-scale Cu-Ni film onto a carbon steel (CS) substrate. Electrochemical analyses were conducted using aerated groundwater from the KAERI underground research tunnel (KURT) as an electrolyte with an NaCl additive to characterize the oxic corrosion behavior of the Cu-Ni film. The results demonstrated that the AM-fabricated Cu-Ni film exhibited enhanced corrosion resistance (manifested as lower corrosion current density and formation of a dense passive layer) in an NaCl-supplemented groundwater solution. Extensive investigations are necessary to elucidate microstructural performance, mechanical properties, and corrosion resistance in the presence of various corroding agents to simplify the implementation of this technology for DGD canisters.

• 방사성폐기물학회지
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• 제20권4호
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• pp.489-500
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• 2022

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly ³H (500-3,600 Bq·g^-1), ¹⁴C (7.5-29 Bq·g^-1), ⁵⁵Fe (1.1- 7.1 Bq·g^-1), ⁵⁹Ni (0.60-1.0 Bq·g^-1), ⁶⁰Co (0.74-70 Bq·g^-1), ⁶³Ni (0.60-94 Bq·g^-1), ⁹⁰Sr (0.25-5.0 Bq·g^-1), ¹³⁷Cs (0.64-8.7 Bq·g^-1), and ¹⁵²Eu (0.19-2.9) Bq·g^-1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32-1.1 Bq·g^-1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.720-724
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• 2003

국내의 가동중지 된 TRIGA 연구로의 해체 시 발생하는 다양한 덕트 폐기물의 자체처분을 위한 제염 공정을 개발할 목적으로 TRIGA 연구로에서 직접 인출한 덕트 시편을 대상으로 오염물의 성상을 조사하였고, 이를 통해 적절한 제염공정을 선정하였다 페인트 도막이 입혀진 덕트 내부계통 표면은 주로 Co-60과 Cs-137로 오염이 되어 있었으며, 이들 핵종은 페인트 층 내부뿐만 아니라 덕트 재료인 함석 표면의 아연 도금막까지 침투되어 있음을 알 수 있었다. 이들 표면 오염을 제거하기 위한 공정으로 NaOH와 황산을 교대로 사용하는 두 단계의 화학제염을 제시하였으며, 이들 제염제를 사용하는 제염공정의 적용을 통해 덕트 폐기물을 효과적으로 제염할 수 있었다.

• Nuclear Engineering and Technology
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• 제25권1호
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• pp.63-70
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• 1993

암반동굴 타입의 저준위방사성폐기물 처분장의 보수적인 안전성평가를 처분장 선원항 REPS 모델을 사용하여 수행하였다. 신뢰할만한 핵종별 침출율 예측을 위하여 REPS 모델에서 콘크리트 구조물의 열하시간, 부석의 형태와 부식율. 드럼표면의 부식면적 비, 그리고 핵종의 특성등이 고려되고 있다. 예비평가의 결과로 Cs-137, Ni-63, Sr-90등이 주요한 핵종임을 알 수 있다. 파라메타의 불확실성과 민감도분석을 위하여 라틴하이퍼큐브 샘플링과 Rank Correlation 기법이 사용되었다. 침입자 시나리오를 적용하였을 경우의 예상 피폭선량도 허용치 이하임과 처분장의 환경영향평가에 있어서 비교적 불확실성이 적은 Near Field의 중요성에 대한 인식이 새롭게 강조되어야 할 필요가 있음을 알 수 있었다.

• 방사선산업학회지
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• 제18권3호
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• pp.249-253
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• 2024

For the safe management and disposal of spent fuel, it is essential to accurately determine the quantities of nuclides present within the spent fuel. In this study, a Gamma/Neutron detection system was developed as a part of basic research to measure the burnup profile of spent fuel, and a performance was evaluated using major nuclides. The prototype of the Gamma/Neutron detection system consisted of a CZT sensor and a ³He chamber. For quantitative evaluation, studies were conducted using calibrated ¹³⁷Cs, ¹³⁴Cs, ¹⁵⁴Eu and ²⁵²Cf sources. In the performance evaluation, a field applicability was verified by analyzing the detection characteristics according to the nuclide.

• 자원환경지질
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• 제29권2호
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• pp.215-225
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• 1996

All the radionuclides in high-level nuclear waste will decay to harmless levels eventually but for some radionuclides decay is so slow that their radiation remains dangerous for times on the order of tens or hundreds of thousands of years. At the present time, the most favorite disposal plan for high-level radioactive waste is a mined geological disposal in which canister enclosing stable solid form of radioactive waste is placed in mined cavities locating hundred meters below the surface. The chief hazard in such disposal is dissolution of radionuclides from the waste in the groundwater that will eventually carry the dissolved radionuclides to surface environments. The hazard from possible escape of the radionuclides through groundwater can be delayed by engineered and geologic barriers. The engineered barriers can become useless by unexpected geologic catastrophe such as volcanism, earthquake, and tectonic movement and by fraudulent work such as careless construction, improperly welded canisters within the first few decades or centuries. As a result, dangerously radioactive waste which is still intensively radioactive is directly exposed to attack by moving groundwater. All the more, it is almost impossible to control repositories for times more than 10,000 years. Therefore, naturally controlled geologic, barriers whose properties will not be changed within 10,000 years are important to guarantee the safety of repositories of high-level radioactive waste. In Sweden and France, the suitability of granite for the mined geological disposal of high-level waste has been studied intensively. According to the research in Sweden and France, granites has the following physio-chemical characteristics which can delay the transportation of radionuclide by groundwater. First, the permeabilities of granites decreases as the depth increases and is $10^{-8}{\sim}10^{-12}m/s$ at depth below 300 m. Second, groundwater at depth below 300 m has pH=7-9 and reducing condition (Eh=-0.1~0.4). This geochemical condition is desirable to prevent both canister and solid waste from corrosion. Third most radionuclides are not transported by low solubilities and some radionuclide with high solubility such as Cs and Sr are retarded by absorption of geologic media through which ground water flows. Therefore, if high-level waste is disposed at depth below 300 m in the granite body which has a low permeability and is geologically stable more than 10,000 years, the safety of repositories from the hazard due to radionuclide escape can guaranteed for more than 10,000 years.

• Journal of Radiation Protection and Research
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• 제16권2호
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• pp.27-39
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• 1991

압축된 국산벤토나이트에서 Sr-85, Cs-237, Co-60 및 Am-241의 확산연구를 수행하였다. 본 실험에서는 원통형으로 압축된 벤토나이트 시료의 중앙부에서 축 방향으로 방사성핵종의 확산이동이 이루어지도록 하여 각 방사성핵종의 확산계수를 측정하였다 그리고 벤토나이트의 열처리 온도와 압축밀도가 확산에 미치는 영향 등을 분석하였다. Sr-85, Cs-137, Co-60 및 Am-241의 겉보기 확산계수는 각각 $1.07{\times}10^{-11},\;6.705{\times}10^{-13},\;1.226{\times}10^{-13},\;1.310{\times}10^{-14},\;m^2/sec$로 측정되었다. 그리고 시료의 압축 밀도를 $1.8g/cm^2$에서 $2.0g/cm^2$으로 증가시켰을 때, Cs-137의 확산계수는 약 1/4로 감소되어 나타났다. 반면, 열처리된 벤토나이트의 경우에는 확산계수가 크게 변하지 많았는데, 이는 $150^{\circ}C$ 이하의 온도에서는 국산 벤토나이트가 방사성핵종의 이동을 지연시킬 수 있는 화학적 방벽으로서 사용할 수 있다는 가능성을 보여준 것이라 생각된다. 그리고 음이온 Cl-36의 화산계수를 이용하여 도출한 각 방사성핵종의 공극확산계수와 표면확산계수를 측정한 겉보기확산계수와 비교해 볼 때, 전체 방사성 핵종의 확산이동에 있어서 표면확산이동이 지배적인 것으로 나타났다.

• Structural Engineering and Mechanics
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• 제87권3호
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• pp.221-229
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• 2023

Urbanization and industrialization have significantly increased the amount of solid waste produced in recent decades, posing considerable disposal problems and environmental burdens. The practice of waste utilization in concrete has gained popularity among construction practitioners and researchers for the efficient use of resources and the transition to the circular economy in construction. This study employed Lytag aggregate, an environmentally friendly pulverized fuel ash-based lightweight aggregate, as a substitute for natural coarse aggregate. At the same time, fly ash, an industrial by-product, was used as a partial substitute for cement. Concrete mix M20 was experimented with using fly ash and Lytag lightweight aggregate. The percentages of fly ash that make up the replacements were 5%, 10%, 15%, 20%, and 25%. The Compressive Strength (CS), Split Tensile Strength (STS), and deflection were discovered at these percentages after 56 days of testing. The concrete cube, cylinder, and beam specimens were examined in the explorations, as mentioned earlier. The results indicate that a 10% substitution of cement with fly ash and a replacement of coarse aggregate with Lytag lightweight aggregate produced concrete that performed well in terms of mechanical properties and deflection. The cementitious composites have varying characteristics as the environment changes. Therefore, understanding their mechanical properties are crucial for safety reasons. CS, STS, and deflection are the essential property of concrete. Machine learning (ML) approaches have been necessary to predict the CS of concrete. The Artificial Fish Swarm Optimization (AFSO), Particle Swarm Optimization (PSO), and Harmony Search (HS) algorithms were investigated for the prediction of outcomes. This work deftly explains the tremendous AFSO technique, which achieves the precise ideal values of the weights in the model to crown the mathematical modeling technique. This has been proved by the minimum, maximum, and sample median, and the first and third quartiles were used as the basis for a boxplot through the standardized method of showing the dataset. It graphically displays the quantitative value distribution of a field. The correlation matrix and confidence interval were represented graphically using the corrupt method.

• 한국건설순환자원학회논문집
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• 제10권3호
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• pp.276-284
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• 2022

이 연구에서는 원전해체 폐콘크리트 미분말을 방사성 폐기물 처분용 고화제로 재활용하기 위한 인수기준 성능 및 레올로지 특성을 분석하였다. 고화제는 방사성 콘크리트 미분말을 모사하여 사용하였으며, 시험체는 증류수, CoCl₂, CsCl 1 mol 수용액을 배합수로 사용하여 제작하였다. 골재 미분말 혼입율 및 혼합수의 종류에 관계없이 재령 28일 압축강도 성능기준 3.45 MPa를 만족하였다. 모든 시험체는 침수강도 기준을 만족하였고 열순환 압축강도는 Plain-50을 제외한 모든 시험체에 대하여 인수기준을 만족하였다. 고화제의 레올로지 특성을 평가한 결과, 골재 혼입율이 증가할수록 항복응력과 소성점도가 감소함을 알 수 있었다. 모든 시험체의 코발트와 세슘에 대한 침출지수는 6 이상으로 인수기준을 만족하였다. 방사성 폐기물 처분용 고화제의 안정적인 성능을 확보하기 위해서는 고화제 내의 골재 성분은 40 % 이하로 사용하는 것이 효과적인 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.847-856
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• 2014

Spent resin waste containing a high concentration of $^{14}C$ radionuclide cannot be disposed of directly. A fundamental study on selective $^{14}C$ stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive $HCO_3{^-}$ ion, as the chemical form of $^{14}C$, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of $NO_3{^-}$ > $HCO_3{^-}{\geq}H_2PO_4{^-}$. Thus the competitive adsorption affinity of $NO_3{^-}$ ion in binary systems appeared far higher than that of $HCO_3{^-}$ or $H_2PO_4{^-}$, and the selective desorption of $HCO_3{^-}$ from the resin was very effective. On one hand, the affinity of $Co^{2+}$ and $Cs^+$ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of $Cs^+$ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process, $NH_4H_2PO_4$ solution was preferable for the stripping of $^{14}C$ from the spent resin.