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RISS 인기검색어
- 검색키워드
- 저자 : Hyungju Yun (검색결과 21 건)
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Hyungju Yun,Manho Han,Seo-Yeon Cho,Jihye Seo,Hyeonwoo Oh 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1
The criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of PWR SNF disposal canister: the KBS-3 type canister and the small standardized transportation, aging and disposal (STAD) canister. The criticality analyses were carried out for four cases as follows: (1) the calculation of isotopic compositions within a SNF using a depletion assessment code and (2) the calculation of the effective multiplication factor (keff) value using a criticality assessment code. Firstly, the KBS-3 type canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU was modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.74407, 0.74102, and 0.73783, respectively. Secondly, the STAD canister was modelled. The SNFs contained in the STAD canister were assumed to be the enrichment of 4.0wt% and the burnup of 45,000 MWD/MTU. The keff values for the cooling times of 40, 50, and 60 years were estimated to be 0.71448, 0.70982, and 0.70743, respectively. Thirdly, the KBS-3 canister with four SNFs of which the enrichment was 4.5wt% and the burnup was 55,000 MWD/MTU was modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.73366, 0.72880, and 0.72634, respectively. Finally, the calculations were carried out for the STAD canister containing four SNFs of the enrichment of 4.5wt% and the burnup of 55,000 MWD/MTU. The keff values for the cooling times of 40, 50, and 60 years were 0.70323, 0.69946, and 0.69719, respectively. Therefore, all of four cases met the performance target with respect to the keff values, 0.95. The STAD canister showed lower keff values than the KBS-3 canister. This appears to be the neutron absorber plate installed in the STAD canister although the distance among the four SNFs in the STAD canister was shorter than the KBS-3 canister.
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Hyungju Yun,Min-Seok Kim,Manho Han,Seo-Yeon Cho 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2
The thermal evaluations for the conceptual design of the deep geological repository considering the improved modeling of the spent fuel decay heat were conducted using COMSOL Multiphysics computational program. The maximum temperature at the surface of a disposal canister for the technical design requirement should not exceed 100°C. However, the peak temperature at the canister surface should not exceed 95°C considering the safety margin of 5°C due to several uncertainties. All thermal evaluations were based on the time-dependent simulation from the emplacement time of the canister to 100,000 years later. In particular, the heat source condition was set to the decay heat rate and axial decay heat profile of the PLUS7 fuel with 4.0wt% U-235 and 45 GWD/MTU. The thermal properties of the granitic rock in South Korea were applied to the host rock region. For the reference design case, the cooling time of the SNF was set to 40 years, the distance between the deposition holes 8 meters and that between the deposition tunnels 30 meters. However, the peak temperature at the canister surface at 10 years was 95.979°C greater than 95°C. This design did not meet the thermal safety requirement and needed to be modified. For the first modified case, when the distance between the deposition tunnels was set to 30 meters, three cooling time cases of 40, 50 and 60 years and five distances of 6, 7, 8, 9 and 10 meters between the deposition holes were considered. The design with the distances of 9 and 10 meters between the deposition holes for the cooling time of 40 years and all five distances for 50 and 60 years were less than 95°C. For the second modified case, when the distance between the deposition holes was set to 8 meters, three cooling time cases of 40, 50 and 60 years and five distances of 20, 25, 30, 35 and 40 meters between the deposition tunnels were considered. The design with the distances of 35 and 40 meters between the deposition tunnels for the cooling time of 40 years, the distances of 25, 30, 35 and 40 meters for 50 years and all five distances for 60 years were less than 95°C. As a result, the peak temperature at the canister surface decreased as the cooling time and the distance between the deposition holes and the tunnels increased.
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Yun, Hyungju,Park, Kwangheon,Choi, Wooyong,Hong, Ser Gi Pergamon Press 2017 Annals of nuclear energy Vol.110 No.-
<P><B>Abstract</B></P> <P>It is important to evaluate an accurate depletion uncertainty resulting from the limitation of a computer code on accuracy in nuclear depletion calculations from the viewpoint of burnup credit with axial burnup distributions. In this work, the bias and bias uncertainty in k<SUB>eff</SUB> resulting from the uncertainty in the calculation of isotopic concentrations are assessed using the Monte Carlo sampling method for a generic 32 PWR-assembly burnup credit cask loaded with the PLUS7 assemblies discharged from Hanbit Nuclear Power Plant Unit 3 in South Korea. In addition, an efficient method with the two-way ANOVA is suggested to obtain reasonably accurate values of total depletion uncertainties resulting from the isotopic uncertainties with reduction of Monte Carlo criticality calculations. Finally, the reasonable total depletion uncertainties in k<SUB>eff</SUB> for 10, 30, and 40 GWD/MTU were efficiently estimated to be 0.011866, 0.013175, and 0.014262, respectively, with 100 Monte Carlo criticality calculations, which have about 1.6, 2.1, and 2.4% discrepancies, respectively, from the ones obtained with 200 calculations. The one for 50 GWD/MTU was efficiently estimated to be 0.030811 with 120 Monte Carlo criticality calculations, which has about 3.6% discrepancy from the one obtained with 200 calculations. Thus, these uncertainties showed good agreements with the uncertainties obtained from 200 nuclear criticality calculations.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Depletion uncertainties for GBC-32 cask with PLUS7 used fuels were evaluated. </LI> <LI> The Monte Carlo sampling method and statistical F-tests of ANOVA were used. </LI> <LI> Depletion uncertainties with a reduced number of calculations were determined. </LI> <LI> The depletion uncertainties for 10 to 50 GWD/MTU were 1.187 to 3.081% Δk. </LI> </UL> </P>
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Yun, Jae Hoon,Ahn, Hyungju,Lee, Phillip,Ko, Min Jae,Son, Hae Jung American Chemical Society 2017 Macromolecules Vol.50 No.19
<P>We developed donor acceptor (D-A)-type random polymers based on 3,3'-difluoro-2,2'-bithiophene with various relative amounts of 5,6-difluoro-4,7-bis(5-bromo-(2-decyltetradecyl)thiophen-2-y1)-2,1,3-benzothiadiazole (2FBT) and 5,6-difluoro-4,7-bis(5-bromo-(2-octyldodecyl) thiophen-2-y1)-2- (3,4-dichlorobenzyloxybutyl)-2H-benzo [d][1,2,3]triazole (DCB-2FBTZ). Introducing small relative amounts of DCB-2FBTZ into the polymer was found to effectively enhance its solar cell performance, resulting in a power conversion efficiency of 9.02%, greater than the 7.29% that resulted from the PFBT-FTh copolymer. Moreover, when the active area of the BHJ film was increased to 1 cm(2), the solar cell reproducibly showed a high performance, here with an efficiency of 8.01% even when the thickness of the active layer was 313 nm. Our studies revealed that including the DCB-2FBTZ group in the polymer simultaneously improved the solution processability and crystallinity of the polymer. These improvements resulted in the formation of highly homogeneous BHJ films throughout large areas with only minor amounts of defects resulting from overaggregation and hence with appropriate morphologies for effective charge generation and transport.</P>
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Cha, Yun Jeong,Gim, Min-Jun,Ahn, Hyungju,Shin, Tae Joo,Jeong, Joonwoo,Yoon, Dong Ki American Chemical Society 2017 ACS APPLIED MATERIALS & INTERFACES Vol.9 No.21
<P>Lyotropic chromonic liquid crystals (LCLCs) have been extensively studied because of the interesting structural characteristics of the linear aggregation of their plank-shaped molecules in aqueous solvents. We report a simple method to control the orientation of LCLCs such as Sunset Yellow (SSY), disodium cromoglycate (DSCG), and DNA by varying pulling speed of the top substrate and temperatures during shear flow induced experiment. Crystallized columns of LCLCs are aligned parallel and perpendicular to the shear direction, at fast and slow pulling speeds of the top substrate, respectively. On the basis of this result, we fabricated an orthogonally patterned film that can be used as an alignment layer for guiding rodlike liquid crystals (LCs) to generate both twisted and planar alignments simultaneously. Our resulting platform can provide a facile method to form multidirectional orientation of soft materials and biomaterials in a process of simple shearing and evaporation, which gives rise to potential patterning applications using LCLCs due to their unique structural characteristics.</P>
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Thermal Analysis of Engineered Barrier System for Long-Term Safety of a Deep Geological Repository
Seo-Yeon Cho,Hyungju Yun,Mi-Seon Jeong,Min-Seok Kim 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2
In order to ensure the long-term safety of a deep geological repository, the performance assessment of the Engineered Barrier System (EBS) considering a thermal process should be performed. The maximum temperature at the side wall of a disposal canister for the technical design requirement should not exceed 100°C. In this study, the thermal modelling was conducted to analyze the effects of the thermal process from a disposal canister to the surrounding near-field host rock using the PFLOTRAN code. The mesh was generated using the LaGriT code and the material properties were assigned by applying the FracMan code. Initial conditions were set as the average geothermal gradient (25.7°C/km) and an average surface temperature (14.7°C) in Korea. The highest temperature was observed at the middle of the canister side wall. The temperature of the buffer was lower than that of the canister, and the temperature increase of the deposition tunnel and the host rock was insignificant due to the lower effect of the heat source. The result of the thermal evolution of the EBS represented the highest thermal effects in the vicinity of the canister. In addition, the thermal effects were largely decreased after 10 years of the entire simulation period. It demonstrated that the model took 3 years to heat up the buffer around the canister. The temperature at the canister side wall increased until 3 years and then decreased after that time. This is because that the radioactive decay heat from the heat source was emitted enough to raise the overall temperature of the EBS by 3 years. However, the decay heat rate of the canister decreased exponentially with the disposal time and then its decay heat was not emitted enough after 3 years. In conclusion, the peak temperature results of the EBS were lower than 70°C to meet the technical design requirement.
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Analysis of Characteristics of Spent Fuels on Long-Term Dry Storage Condition
Yoon, Suji,Park, Kwangheon,Yun, Hyungju Korean Radioactive Waste Society 2021 방사성폐기물학회지 Vol.19 No.2
Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.
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Source Term Evaluation Using ORIGEN-S and TRITON for Spent Nuclear Fuel Disposal
Seo-Yeon Cho,Hyungju Yun,Mi-Seon Jeong 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2
In order to dispose of spent nuclear fuel (SNF) in deep geological repository, source term evaluation considering its specification, enrichment, burnup, cooling time should be performed. In this study, the measured values of Takahama-3 pressurized water reactor SNF (WH 17×17) samples were analyzed with SCALE 6.1/ORIGEN-S and TRITON code calculation results for validation. Unlike the ORIGENS code, TRITON code calculations differed from two-dimensional neutron flux distribution by using the multi-group cross-section library. Both calculation results from ORIGEN-S and TRITON code showed higher errors in 234U, 239Pu, and 241Pu compared to other actinide nuclides. In the case of axial locations of fuel rods in fuel assembly, fuel rods located at the edge of the fuel assembly presented increased errors due to nuclear reaction cross-section. Overall, the ORIGEN-S predictions informed more accurate agreement with the measured results compared with TRITON results. Especially to 235U, 239Pu, and 240Pu radionuclides, ORIGEN-S errors were denoted more than twice as low as the TRITON results. Comparing the calculation results with experimental results implied that the ORIGENS code was more accurate code than the TRITON code for source term evaluation.
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