• Title/Summary/Keyword: nuclear waste disposal

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A review of the features, events, and processes and scenario development for Korean risk assessment of a deep geological repository for high-level radioactive waste

  • Kibeom Son;Karyoung Choi;Jaehyeon Yang;Haeram Jeong;Hyungdae Kim;Kunok Chang;Gyunyoung Heo
    • Nuclear Engineering and Technology
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    • v.55 no.11
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    • pp.4083-4095
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    • 2023
  • Currently, various research institutes in Korea are conducting research to develop a safety case for deep geological repository for high-level radioactive waste (HLW). In the past, low and intermediate-level waste (LILW) was approved by a regulatory body by performing a post-closure safety assessment, but HLW has different disposal characteristics and safety objectives are different. Therefore, in the case of HLW, safety assessment should be performed based on these changed conditions, and specific procedures are also under development. In this paper, the regulatory status of prior research institutes, feature, event and process (FEP) and scenario development cases were investigated for well-organized FEP and scenario development methodologies. In addition, through the results of these surveys, the requirements and procedures necessary for the FEP and scenario development stage during the safety assessment of repository for HLW were presented. These review results are expected to be used to identify the overall status of previous studies in conducting post-closure risk assessment for HLW repository, starting with identifying regulatory requirements, the most basic element.

Systems Engineering Approach for the Reuse of Metallic Waste From NPP Decommissioning and Dose Evaluation (금속해체 폐기물의 재활용을 위한 시스템엔지니어링 방법론 적용 및 피폭선량 평가)

  • Seo, Hyung-Woo;Kim, Chang-Lak
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.15 no.1
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    • pp.45-63
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    • 2017
  • The oldest commercial reactor in South Korea, Kori-1 Nuclear Power Plant (NPP), will be shut down in 2017. Proper treatment for decommissioning wastes is one of the key factors to decommission a plant successfully. Particularly important is the recycling of clearance level or very low level radioactively contaminated metallic wastes, which contributes to waste minimization and the reduction of disposal volume. The aim of this study is to introduce a conceptual design of a recycle system and to evaluate the doses incurred through defined work flows. The various architecture diagrams were organized to define operational procedures and tasks. Potential exposure scenarios were selected in accordance with the recycle system, and the doses were evaluated with the RESRAD-RECYCLE computer code. By using this tool, the important scenarios and radionuclides as well as impacts of radionuclide characteristics and partitioning factors are analyzed. Moreover, dose analysis can be used to provide information on the necessary decontamination, radiation protection process, and allowable concentration limits for exposure scenarios.

Behaviour of Uranyl Phosphate Containing Solid Waste During Thermal Treatment for the Purpose of Sentencing and Immobilisation: Preliminary Results

  • Foster, Richard Ian;Sung, Hyun-Hee;Kim, Kwang-Wook;Lee, Keunyoung
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.18 no.3
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    • pp.407-414
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    • 2020
  • Thermal decomposition of the uranyl phosphate mineral phase meta-ankoleite (KUO2PO4·3H2O) has been considered in relation to high temperature thermal sintering for the immobilisation of a uranyl phosphate containing waste. Meta-ankoleite thermal decomposition was studied across the temperature range 25 - 1200℃ under an inert N2 atmosphere at 1 atm. It is shown that the meta-ankoleite mineral phase undergoes a double de-hydration event at 56.90 and 125.85℃. Subsequently, synthetically produced pure meta-ankoleite remains stable until at least 1150℃ exhibiting no apparent phase changes. In contrast, when present in a mixed waste the meta-ankoleite phase is not identifiable after thermal treatment indicating incorporation within the bulk waste either as an amorphous phase and/or as uranium oxide. Visual inspection of the waste post thermal treatment showed evidence of self-sintering owing to the presence of glass former materials, namely, silica (SiO2) and antimony(V) oxide (Sb2O5). Therefore, incorporation of the uranium phase into the waste as part of waste sentencing and immobilisation via high temperature sintering for the purpose of long-term disposal is deemed feasible.

A Study on Improvement of Scaling Factor Prediction Using Artificial Neural Network

  • Lee, Sang-Chul;Hwang, Ki-Ha;Kang, Sang-Hee;Lee, Kun-Jai
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2003.11a
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    • pp.534-538
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    • 2003
  • Final disposal of radioactive waste generated from Nuclear Power Plant (NPP) requires the detailed knowledge of the natures and quantities of radionuclides in waste package. Many of these radionuclides are difficult to measure and expensive to assay. Thus it is suggested to the Indirect method by which the concentrations of DTM (Difficult-to-Measure) nuclide is decided using the relation of concentrations (Scaling Factor) between Key (Easy-to-Measure) nuclide and DTM nuclide with measured concentrations of Key nuclide. In general, scaling factor is determined by using of log mean average (LMA) and regression. These methods are adequate to apply most corrosion product nuclides. But in case of fission product nuclides and some corrosion product nuclides, the predicted values aren't well matched with the original values. In this study, the models using artificial neural network (ANN) for C-14 and Sr-90 are compared with those using LMA and regression. The assessment of models is executed in the two parts divided by a training part and a validation part. For all of two nuclides in the training part, the predicted values using ANN are well matched with the measured values compared with those using LMA and regression. In the validation part, the accuracy of the predicted values using ANN is better than that using LMA and is similar to or better than that using regression. It is concluded that the predicted values using ANN model are better than those using conventional model in some nuclides and ANN model can be used as the complement of LMA and regression model.

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A Prediction of Specific Heat Capacity for Compacted Bentonite Buffer (압축 벤토나이트 완충재의 비열 추정)

  • Yoon, Seok;Kim, Geon-Young;Baik, Min-Hoon
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.15 no.3
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    • pp.199-206
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    • 2017
  • A geological repository for the disposal of high-level radioactive waste is generally constructed in host rock at depths of 500~1,000 meters below the ground surface. A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is indispensable to assure the disposal safety of high-level radioactive waste, and it can restrain the release of radionuclides and protect the canister from the inflow of groundwater. Since high temperature in a disposal canister is released to the surrounding buffer material, the thermal properties of the buffer material are very important in determining the entire disposal safety. Even though there have been many studies on thermal conductivity, there have been only few studies that have investigates the specific heat capacity of the bentonite buffer. Therefore, this paper presents a specific heat capacity prediction model for compacted Gyeongju bentonite buffer material, which is a Ca-bentonite produced in Korea. Specific heat capacity of the compacted bentonite buffer was measured using a dual probe method according to various degrees of saturation and dry density. A regression model to predict the specific heat capacity of the compacted bentonite buffer was suggested and fitted using 33 sets of data obtained by the dual probe method.

An Influence Analysis on the Gap Space of an Engineered Barrier for an HLW Repository (고준위폐기물처분장 공학적방벽의 갭 공간이 미치는 영향 분석)

  • Yoon, Seok;Lee, Changsoo;Kim, Min-Jun
    • Journal of the Korean Geotechnical Society
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    • v.37 no.4
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    • pp.19-26
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    • 2021
  • The high-level radioactive waste (HLW) produced from nuclear power plants is disposed in a rock-mass at a depth of hundreds meters below the ground level. Since HLW is very dangerous to human being, it must be disposed of safely by the engineered barrier system (EBS). The EBS consists of a disposal canister, backfill material, buffer material, and so on. When the components of EBS are installed, gaps inevitably exist not only between the rock-mass and buffer material but also between the canister and buffer material. The gap can reduce water-retarding capacity and heat release efficiency of the buffer material, so it is necessary to investigate properties of gap-filling materials and to analyze gap spacing effect. Furthermore, there has been few researches considering domestic disposal system compared to overseas researches. In this reason, this research derived the peak temperature of the bentonite buffer material considering domestic disposal system based on the numerical analysis. The gap between the canister and buffer material had a minor effect on the peak temperature of the bentonite buffer material, but there was 40% difference of the peak temperature of the bentonite buffer material because of the gap existence between the buffer material and rock mass.

Forecasting a Gyeongju's Local Society Change Using Urban Dynamics Model (도시동태모델을 이용한 경주 지역사회변화 예측)

  • Lee, Young-Chan
    • Korean Management Science Review
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    • v.25 no.3
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    • pp.27-43
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    • 2008
  • This study analyzes the changes of Gyeongju local society because of setting up low and intermediate level radioactive waste disposal site by using urban dynamics model. Specifically, after examining 'Gyeongju Long-Term Development Plan' announced in 2007, I establish the number of industries, population, gross local product, residents' income, and the long term employment condition as essential change-causing factors in Gyeongju local society based on the Big3 government project, and forecast it by using 'Gyeongju long-Term Development Plan' and all sorts of statistical data. In this stage, I assume 3 scenarios(basic, optimistic, and pessimistic view) to estimate the changes of local society more exquisitely, and scenarios are composed through mediation about variables of a growth rate and an inflow or outflow rate. The result shows that Gyeonaju local society would have growing changes by 2020. The essential change-causing factors are as follows. The case of population is estimated that it starts going down at the level of approximately 270 thousand by 2009, starts going up continuously after 2009, the year of completion of low and intermediate level radioactive waste disposal site, and increases from the level of about 300 thousand as minimum to 340 thousand as maximum in 2020. The estimates of other cases are made that the number of Industries has about 10 thousand increases, gross local product has almost 6 trillion increases, nominal gross national income doubles, as well as residences have approximately 280 thousand increases, and also made that employment condition also improves continuously, and diffusion ratio of house starts going up but the amount of supplies is a little bit insufficient in the long view.

Analysis of the statistical properties for the background fractures in the LILW disposal site of Korea (중.저준위 방사성폐기물 처분 부지 내 배경 단열의 통계적 특성 분석)

  • Ji, Sung-Hoon;Park, Kyung-Woo;Kim, Kyoung-Su;Kim, Chun-Soo
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.6 no.4
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    • pp.257-263
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    • 2008
  • We analyzed the statistical properties for the conductive background fractures in the Low and Intermediate Level Waste(LILW) disposal site to conceptualize of its groundwater flow system. The background fractures were classified to fracture sets based on their trends and plunges that were obtained from the borehole logging data, and then the fracture transmissivity distribution was inferred from the fixed interval hydraulic test results. The fracture size distribution of each fracture set was estimated using the fracture density and fracture mapping data. To verify the analyzed results, we compared observed field data to simulated one from the DFN model that was constructed with the analyzed statistical properties of the background fractures, and they showed a good agreement.

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Analysis of the Spent Fuel Cooling Time for a Deep Geological Disposal (심지층 처분을 일한 사용후핵연료 냉각기간 분석)

  • Lee, Jong-Youl;Cho, Dong-Geun;Choi, Heui-Joo;Choi, Jong-Won;Lee, Yang
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.6 no.1
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    • pp.65-72
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    • 2008
  • The purpose of the HLW deep geological disposal is to isolate and to delay the radioactive material release to human beings and the environment for a long time so that the toxicity does not affect to the environment. The main requirements for the HLW repository design is to keep the buffer temperature below $100\;^{\circ}C$ in order to maintain its integrity. So the cooling time of spent fuels discharged from the nuclear power plant is the key consideration factors for efficiency and economic feasibility of the repository. The disposal tunnel/disposal hole spacing, the disposal area and thermal capacity required for the deep geological repository layout which satisfies the temperature requirement of the disposal system is analyzed to set the optimized spent fuels cooling time. To do this, based on the reference disposal concept, thermal stability analyses of the disposal system have been performed and the derived results have been compared by setting the spent fuels cooling time and the disposal tunnel/disposal hole spacing in various ways. From these results, desirable spent fuels cooling time in view of disposal area is derived. The results shows that the time reaching the maximum temperature within the design limit of the temperature in the disposal site is likely shortened as the cooling time of spent fuels becomes short. Also it seems that the temperature-rising and-dropping patterns in the disposal site are of smoothly varying form as the cooling time of spent fuels becomes long. In addition, it is revealed that a desirable cooling time of spent fuels is approximately 40-50 years when spent fuels are supposedly disposed in the deep geological disposal site with its structural scale under consideration in this study.

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A Prediction of Thermal Expansion Coefficient for Compacted Bentonite Buffer Materials (압축 벤토나이트 완충재의 열팽창계수 추정)

  • Yoon, Seok;Kim, Geon-Young;Baik, Min-Hoon
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.16 no.3
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    • pp.339-346
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    • 2018
  • A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is indispensable to assure the disposal safety of high-level radioactive waste. Since the heat generated from spent nuclear fuel in a disposal canister is released to the surrounding buffer materials, the thermal properties of the buffer material are very important in determining the entire disposal safety. Especially, since thermal expansion can cause thermal stress to the intact rock mass in the near-field, it is very important to evaluate thermal expansion characteristics of bentonite buffer materials. Therefore, this paper presents a thermal expansion coefficient prediction model of the Gyeongju bentonite buffer materials which is a Ca-bentonite produced in South Korea. The linear thermal expansion coefficient was measured considering heating rate, dry density and temperature variation using dilatometer equipment. Thermal expansion coefficient values of the Gyeongju bentonite buffer materials were $4.0{\sim}6.0{\times}10^{-6}/^{\circ}C$. Based on the experimental results, a non-linear regression model to predict the thermal expansion coefficient was suggested and fitted according to the dry density.