• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2014.11a
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• pp.273-273
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• 2014

Quantum-dot sensitized solar cells (QDSCs) are an emerging class of solar cells owing to their easy fabrication, low cost and material diversity. Despite of the fact that the maximum conversion efficiency of QDSCs is still far less than that of Dye-Sensitized Solar Cells (>12 %), their unique characteristics like Multiple Exciton Generation (MEG), energy band tune-ability and tendency to incorporate multiple co-sensitizers concurrently has made QDs a suitable alternative to expensive dyes for solar cell application. Lead Sulfide (PbS) Quantum dot sensitized solar cells are theoretically proficient enough to have a photo-current density ($J_{sc}$) of $36mA/cm^2$, but practically there are very few reports on photocurrent enhancement in PbS QDSCs. Recently, $Hg^{2+}$ incorporated PbS quantumdots and Cadmium Sulfide (CdS) co-sensitized PbS solarcells are reported to show an improvement in photo-current density ($J_{sc}$). In this study, we explored the efficacy of $In_2S_3$ as an interfacial layer deposited through SILAR process for PbS QDSCs. $In_2S_3$ was chosen as the interfacial layer in order to avoid the usage of hazardous CdS or Mercury (Hg). Herein, the deposition of $In_2S_3$ interfacial layer on $TiO_2$ prior to PbS QDs exhibited a direct enhancement in the photo-current (Isc). Improved photo-absorption as well as interfacial recombination barrier caused by $In_2S_3$ deposition increased the photo-current density ($J_{sc}$) from $13mA/cm^2$ to $15.5mA/cm^2$ for single cycle of $In_2S_3$ deposition. Increase in the number of cycles of $In_2S_3$ deposition was found to deteriorate the photocurrent, however it increased $V_{oc}$ of the device which reached to an optimum value of 2.25% Photo-conversion Efficiency (PCE) for 2 cycles of $In_2S_3$ deposition. Effect of Heat Treatment, Normalized Current Stability, Open Circuit Voltage Decay and Dark IV Characteristics were further measured to reveal the characteristics of device.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.427-446
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• 2009

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical $CO_2$ Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.65-73
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• 2013

Temperature-dependent hydrolysis behaviors of aqueous U(VI) species were investigated with time-resolved laser fluorescence spectroscopy (TRLFS) in the temperature range from 15 to $75^{\circ}C$. The formation of four different U(VI) hydrolysis species was measured at pHs from 1 to 7. The predominant presence of $UO{_2}^{2+}$, $(UO_2)_2(OH){_2}^{2+}$, $(UO_2)_3(OH){_5}^+$, and $(UO_2)_3(OH){_7}^-$ species were identified based on the spectroscopic properties such as fluorescence wavelengths and fluorescence lifetimes. With an increasing temperature, a remarkable decrement in the fluorescence lifetime for all U(VI) hydrolysis species was observed, representing the dynamic quenching behavior. Furthermore, the increase in the fluorescence intensity of the further hydrolyzed U(VI) species was clearly observed at an elevated temperature, showing stronger hydrolysis reactions with increasing temperatures. The formation constants of the U(VI) hydrolysis species were calculated to be $log\;K{^0}_{2,2}=-4.0{\pm}0.6$ for $(UO_2)_2(OH){_2}^{2+}$, $log\;K{^0}_{3,5}=-15.0{\pm}0.3$ for $(UO_2)_3(OH){_5}^+$, and $log\;K{^0}_{3,7}=-27.7{\pm}0.7$ for $(UO_2)_3(OH){_7}^-$ at $25^{\circ}C$ and I = 0 M. The specific ion interaction theory (SIT) was applied for the extrapolation of the formation constants to infinitely diluted solution. The results of temperature-dependent hydrolysis behavior in terms of the U(VI) fluorescence were compared and validated with those obtained using computational methods (DQUANT and constant enthalpy equation). Both results matched well with each other. The reaction enthalpies and entropies that are vital for the computational methods were determined by a combination of the van't Hoff equation and the Gibbs free energy equation. The temperature-dependent hydrolysis reaction of the U(VI) species indicates the transition of a major U(VI) species by means of geothermal gradient and decay heat from the radioactive isotopes, representing the necessity of deeper consideration in the safety assessment of geologic repository.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.248-255
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• 1983

The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.167-172
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• 2010

CANDU spent fuel is to be disposed of at repository finally rather than recycled because of its low fissile nuclide concentration. But the difficult situation of finding a repository site can not help introducing a interim storage in the short term. It is required to find an optimum timing of geological disposal of CANDU spent fuels related to the interim storage operation period. The major factors for determining the disposal starting time are considered as safety, economics, and public acceptance. Safety factor is compared in terms of the decay heat and non-proliferation. Economics factor is compared from the point of the operation cost, and public acceptance factor is reviewed from the point of retrievability and inter-generation ethics. This paper recommended the best solution for the disposal starting time by analyzing the above factors. It is concluded that the optimum timing for the CANDU spent fuel disposal is around 2041 and that the sooner disposal time, the better from the point of technical and safety aspects.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.1-11
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• 1989

Numerous losses of decay heat removal capability have occurred at U during stutodwn while its significance to safety is needless to say. A study is carried out as an attempt to assess what could be done to lower the frequency of these events and to mitigate their consequences in the unlikely event that one occurs. The shutdown risk model is developed and analyzed using Event/Fault Tree for the typical pressurized water reactor. The human cognitive reliability (HCR) model, two-stage bayesian approach and staircase function model are used to estimate human reliability, initiating event frequency and offsite power non-recovery probability given loss of offsite power, respectively. The results of this study indicate that the risk of a Pm at shutdown is not much lower than the risk when the plant is operating. By examining the dominant accident sequences obtained, several design deficiencies are identified and it is found that some proposed changes lead to significant reduction in core damage frequency due to loss of cooling events.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.457-469
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• 2016

During the late phase of severe accidents of light water reactors, a porous debris bed is expected to develop on the bottom of the flooded reactor cavity after breakup of the melt in water. The geometrical configuration, i.e., internal and external characteristics, of the debris bed is significant for the adequate assessment of the coolability of the relocated corium. The internal structure of a debris bed was investigated experimentally using the DAVINCI (Debris bed research Apparatus for Validation of the bubble-Induced Natural Convection effect Issue) test facility. Particle sedimentation under the influence of a two-phase natural convection flow due to the decay heat in the debris bed was simulated by dropping various sizes of particles into a water vessel with air bubble injection from the bottom. Settled particles were collected and sieved to obtain the particle mass, size distribution in the radial and axial positions, and the bed porosity and permeability. The experimental results showed that the center part of the particle bed tended to have larger particles than the peripheral area. For the axial distribution, the lower layer had a higher fraction of larger particles. As the sedimentation progressed, the size distribution in the upper layers can shift to larger sizes because of the higher vapor generation rate and stronger flow intensity.

• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Journal of the Korean Geotechnical Society
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• v.35 no.7
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• pp.41-50
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• 2019

Gyeongju bentonite is a buffer material primarily considered in Korea and it is highly compacted as a part of an engineered barrier system (EBS) of high-level radioactive waste repository. The compacted bentonite undergoes swelling stress by groundwater penetration and thermal stress by decay heat from a canister. Therefore, the mechanical properties of the compacted bentonite buffer material is crucial for the performance assessment of EBS. This paper aims to evaluate deformation properties of Gyeongju compacted bentonite using seismic methods. Two sets of compacted bentonite specimens were prepared having dry densities of $1.59g/cm^3$ and $1.75g/cm^3$ with water contents of 10.6% and 8.7%. Free-free resonant column tests were performed to measure constrained and unconstrained compression wave velocities. With the measured wave velocities, Young's modulus ($E_{max}$) and constrained modulus ($M_{max}$), material damping ratio ($D_{min}$), and Poisson's ratio at small strain were determined. As results, this paper evaluates the deformation properties of Gyeongju compacted bentonite and compares them with the results of previous researches.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.