• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.223-224
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• 2004

고준위폐기물 혹은 사용후 핵연료의 처분용기 재질은 각 국의 처분개념과 처분공의 주위 환경에 따라 달라질 수 있다. 용기의 후보재질로는 탄소강, 스텐레스 강, 구리, 니켈, 티탄 혹은 이들의 합금이 주로 고려되고 있으나, 국내에서는 아직 선정되지 않았다. 국내 처분환경에서 이들 재질의 부식특성을 조사하고자 모의 화강암 지하수를 가해 만든 겔 상태의 경주 벤토나이트에 탄소강, 스텐레스 강, 구리 시편을 넣고, $70^{\circ}C$, 아르곤 분위기에서 530일 경과한 후 시편의 표면 변화 (그림 1) 및 무게 감소를 측정하였다. 철 부식시편은 검정색의 철 화합물 층으로 덮여 있었으며, 구리표면에는 노란색의 부식층이 형성되었는데, 이를 XRD로 분석한 결과 $Cu_2O$로 판명되었다. 그러나 $700^{\circ}C$에서 각각 0, 24, 96시간동안 예민화시킨 스텐레스 강 시편들은 모두 초기상태 그대로 광택을 유지하고 있었으며, XRD에서 다른 화합물의 형성을 발견할 수 없었다 (그림 2). 시편의 무게 감소가 균일부식에 기인한 것으로 가정하여 환산한 결과, 구리와 스텐레스 강 모두 0.3~0.4 $\mu\textrm{m}/yr$의 부식속도를 나타내었다. 그러나 구리는 부식생성물이 표면에 부착되어 있기 때문에 실제 부식두께는 이 값보다 더 클 것으로 생각된다. 용기가 초기 530일간과 같은 속도로 처분용기의 부식이 진행된다면 한국에서 기준처분 개념으로 삼고 있는 50mm 두께의 내부식성 외벽 금속용기는 적어도 만년이상 견딜 수 있을 것으로 추정된다. 한편, 검정색 부식층을 제거한 무게감소로부터 계산한 철의 부식속도는 구리의 약 30배에 해당하였다. 금속 재질의 정확한 부식 거동을 파악하기 위해서는 보다 장기간의 실험이 요구된다. 시험법 선정에 각계(규제기관, 학계, 발전소 현장 및 산업계 등) 전문가로부터 기술자문회의를 통하여 자문 의견을 받기로 하였다. 특히 현재 폐기물 인수 기술기준치가 설정된 국가의 시험법을 심층 있게 검토하기로 하였다.검토하기로 하였다. 혹은 수성주변 환경이 지배하는 산악이나 구릉지에서 흔히 나타나는 침엽수-낙엽활엽수의 혼합림 식생상태를 잘 대변해 주고 있는 것으로 판단된다. 끝으로, 의림지 호저 퇴적층 중에서 인위적인 교란흔적이 없는 암회색 유기질 니층에 대한 탄소연대측정 결과, 제1호공 12번 시료에서 950$\pm$40 years B.P을 얻었으며, 제3-1호공에서도 아래로 내려가면서 8, 10, 11번 시료에 대하여 500$\pm$30 years B.P, 650$\pm$30 years B.P, 800$\pm$40 years B.P의 연대측정 결과를 획득하였다. 이상과 같은 의림지 호저 퇴적층의 형성환경과 형성시기 연구를 통하여 의림지의 제방축조의 최초시기를 해석해 보면, 의림지의 제방은 적어도 과거 약 827년 전에서 866년 전에는 이미 축조되어 있었음을 알 수 있다. 과거 제천 일대에 살았던 옛사람들이 의림지 하류의 곡지중앙과 고기 충적선상지에 대한 관계용 용수조달의 필요성에 부응하여 상류부 곡지하천의 자연입지 환경을 최대한 이용하여 축조한 것으로 판단된다..준비하였다.전류를 구성하는 주요 입자의 에너지 영역(75~l13keV)에서 가장 높은(0.80) 상관계수를 기록했다. 넷째, 회복기 중에 일어나는 입자들의 유입은 자기폭풍의 지속시간을 연장시키는 경향을 보이며 큰 자기폭풍일수록 현저했다. 주상에서 관측된 이러한 특성은 서브스톰 확장기 활동이 자기폭풍의 발달과 밀접한 관계가 있음을 시사한다.se that were all low

• Journal of the Korean Geotechnical Society
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• v.35 no.11
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• pp.7-14
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• 2019

The compacted bentonite buffer is one of the major components of an engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLW), and it is considered the best candidate for the buffer material. The buffer is located between disposal canisters and near-field rock mass, and it interrupts the release of radionuclide from disposal canisters and protect them from the penetration of groundwater. At initial disposal condition, degree of saturation of the compacted bentonite buffer decreases because of high thermal quantities released from the disposal canisters. However, the degree of saturation of the compacted bentonite buffer gradually increases caused by inflow of groundwater. The saturated and unsaturated behavior of the buffer is a very important input data since it can determine the safety performance of EBS. Therefore, this paper investigated water retention capacity (WRC) for the Korean compacted bentonite buffer. The WRC of the compacted bentonite buffer was derived by measuring volumetric water content and water suction when temperature variation was between 24℃~125℃ considering decrease of degree of saturation with respect to temperature increase. The WRC was also derived with the same volumetric water content under the room temperature condition, and it showed 1~15% larger water suction than high temperature condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.55-61
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• 2013

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls using a chlorination method was demonstrated for complete conversion of Zr into $ZrCl_4$. A chlorination reaction was performed by reacting Zry-4 hulls for 8 h under a 70 cc/min $Cl_2$ + 70 cc/min Ar flow at $380^{\circ}C$. The initial weight of the reactant (51.7 g) decreased to 0.49 g after 8 h of operation, which is only 0.95wt% of the initial weight. The weight of the total reaction products was 121.7 g with a high Zr purity of 99.80wt%. Fe and Sn were identified as major (0.18wt%) and minor (0.02wt%) impurities of the reaction products, respectively. It was also shown that Zr exhibited a high recovery ratio of 96.95wt% with a relatively small experimental loss of 2.34wt%. Observation of the reaction residues revealed that the chlorination reaction was dominant along the longitudinal direction, and surface oxide layers remained as reaction residues. The high purity and recovery ratio of Zr proposed the feasibility of the chlorination technique as an effective hull waste treatment method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.33-39
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• 2010

The electrolytic reduction of a spent oxide fuel involves a liberation of the oxygen in a molten LiCl electrolyte, which results in a chemically aggressive environment that is too corrosive for typical structural materials. Accordingly, it is essential to choose the optimum material for the processing equipment that handles the high molten salt. In this study, hot corrosion studies were performed on bare as well as coated superalloy specimens after exposure to lithium molten salt at $675^{\circ}C$ for 216 h under an oxidizing atmosphere. The IN713LC superalloy specimens were sprayed with an aluminized NiCrAlY bond coat and then with an $Y_2O_3$ top coat. The bare superalloy reveals an obvious weight loss due to spalling of the scale by the rapid scale growth and thermal stress. The chemical and thermal stability of the top coat has been found to be beneficial for increasing to the corrosion resistance of the structural materials for handling high temperature lithium molten salts.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.361-370
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• 1993

Low power physics tests should be peformed for the domestic pressurized light water reactors (PWRs) after refueling. The tests are peformed to ensure that operating characteristics of the core are consistent with predictions and that the core can be operated as designed. But in some low power physics tests, slow but steady reactivity increasing phenomena were noticed after step reactivity insertion by the control rod movement. These reactivity increasing phenomena are due to the low flux level and the gamma background because an uncompensated ion chamber (UIC) is used as the ex-core neutron detector. The gamma background may affect the results or the lour power physics tests. The aims or this paper are to analyze the grounds of such phenomena, to simulate a reference bank worth measurement test and to present a resolution quantitatively. In this study, the gamma background level was estimated by numerically solving the point kinetics equations accounting the gamma background effect. The reactivity computer check test was simulated to verify the model. Also, an appropriate neutron flux level was determined by simulating the reference bank worth measurement test. The determined neutron flux level is approximately 0.3 of the nuclear heating flux. This level is about 3 times as high as the current test upper limit specified in the test procedure. Then, the findings from this work were successfully applied to Kori unit 4 cycle 7 and Yonggwang unit 1 cycle 7 physics tests.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.397-410
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• 2018

When considering the long-term safety assessment of spent-nuclear fuel management, americium is one of the most radio-toxic actinides. Although spectroscopic methods are widely used for the study of actinide chemistry, application of those methods to americium chemistry has been limited. Herein, we purified $^{241}Am$ to obtain a highly pure stock solution required for spectroscopic studies. Quantitative and qualitative analyses of purified $^{241}Am$ were carried out using liquid scintillation counting, and gamma and alpha radiation spectrometry. Highly sensitive absorption spectrometry coupled with a liquid waveguide capillary cell and time-resolved laser fluorescence spectroscopy were employed for the study of Am(III) hydrolysis and oxalate (Ox) complexation. $Am^{3+}$ ions under acidic conditions exhibit maximum absorbance at 503 nm, with a molar absorption coefficient of $424{\pm}8cm^{-1}{\cdot}M^{-1}$. $Am(OH)_3(s)$ colloidal particles formed under near neutral pH conditions were identified by monitoring the absorbance at around 506-507 nm. The formation of ${Am(Ox)_3}^{3-}$ was detected by red-shifts of the absorption and luminescence spectra of 4 and 5 nm, respectively. In addition, considerable enhancements of the luminescence intensities were observed. The luminescence lifetime of ${Am(Ox)_3}^{3-}$ increased from 23 to 56 ns, which indicates that approximately six water molecules are replaced by carboxylate ligands in the inner-sphere of the Am(III). These results suggest that ${Am(Ox)_3}^{3-}$ is formed through the bidentate coordination of the oxalate ligands.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.179-189
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• 2014

Pyroprocessing has been developed for the purpose of resolving the current spent nuclear fuel management issue and enhancing the recycle of valuable resources. An electrolytic reduction of the pyroprocessing is a process to reduce oxides into metals using LiCl as an electrolyte and requires a post-treatment process due to the inclusion of residual salt in porous metal products. A vacuum distillation has been adopted for various molten salt systems and could be applied to the post-treatment process of the electrolytic reduction. The residual salt in the metal products includes LiCl, alkali chlorides, and alkaline earth chlorides. In this paper, vapor pressures of chlorides have been estimated and the composition changes on the residual liquid during the vacuum distillation process have been calculated. A model combining a material balance and vapor-liquid equilibrium relations has been proposed under a constant vapor discharging flow rate and liquid composition changes have been calculated using the vapor pressures with respect to a dimensionless time. The behaviors have been compared with temperature and molten salt composition changes to simulate the process condition variation. The distillation of the residual salt has been dominated by LiCl which is the main component of the salt and CsCl of which vapor pressure is higher than that of LiCl would be readily removed. RbCl exhibits similar vapor pressure with LiCl and maintains its composition. However, $SrCl_2$ and $BaCl_2$ of which vapor pressures are much lower than that of LiCl are concentrated with time and expected to be possibly precipitated during the distillation when the initial compositions are increased.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.7-17
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• 2014

Molten salt electrorefining process achieves uranium deposits at cathode using an electrochemical processing of spent nuclear fuel. In order to recover pure uranium from cathode deposit containing about 30wt% salt, the adhered salt should be removed by cathode process (CP). The CP has been regarded as one of the bottle-neck of the pyroprocess as the large amount of uranium is treated in this step and the operation parameters are crucial to determine the final purity of the product. Currently, related research activities are mainly based on experiments consequently it is hard to observe processing variables such as temperature, pressure and salt gas behavior during the operation of the cathode process. Hence, in this study operation procedure of cathode process is numerically described by using appropriate mathematical model. The key parameters of this research are the amount of evaporation at the distillation part, diffusion coefficient of gas phase salt in cathode processor and phase change rate at condensation part. Each of these conditions were composed by Hertz-Langmuir equation, Chapman-Enskog theory, and interphase mass flow application in ANSYS-CFX. And physical properties of salt were taken from the data base in HSC Chemistry. In this study, calculation results on the salt gas behavior and optimal operating condition are discussed. The numerical analysis results could be used to closely understand the physical phenomenon during CP and for further scale up to commercial level.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.75-94
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• 2019

As the present single-layer repository concept requires too large an area for the site of the repository, a multi-layer repository concept has been suggested to improve the disposal density. The effects of the excavation damaged zone around the multi-layer repository constructed in the deep host rock on the temperature distribution in the repository were analyzed. For the thermal analysis of the multi-layer repository, the hydrothermal model was used to consider the resaturation process occurring in the buffer, backfill and rock. The existence of an excavation damaged zone has a significant effect on the temperature distribution in the repository, and the maximum peak temperatures of double-layer and triple-layer repositories can rise to $7^{\circ}C$ and $12^{\circ}C$, respectively depending on the size of the excavation damaged zone and the degree of decrease of the thermal conductivity. The dominant factor affecting the peak temperature in the multi-layer repository is the decrease of thermal conductivity in the excavation damaged zone, and the excavation damaged zone formed around the deposition hole has more significant effects on the peak temperature than does the excavation damaged zone formed around the disposal tunnel.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.41 no.2
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• pp.123-131
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• 2021

The buffer is a key component of an engineered barrier system that safeguards the disposal of high-level radioactive waste. Buffers are located between disposal canisters and host rock, and they can restrain the release of radionuclides and protect canisters from the inflow of ground water. Since considerable heat is released from a disposal canister to the surrounding buffer, the thermal conductivity of the buffer is a very important parameter in the entire disposal safety. For this reason, a lot of research has been conducted on thermal conductivity prediction models that consider various factors. In this study, the thermal conductivity of a buffer is estimated using the machine learning methods of: linear regression, decision tree, support vector machine (SVM), ensemble, Gaussian process regression (GPR), neural network, deep belief network, and genetic programming. In the results, the machine learning methods such as ensemble, genetic programming, SVM with cubic parameter, and GPR showed better performance compared with the regression model, with the ensemble with XGBoost and Gaussian process regression models showing best performance.