• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.87-93
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• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.287-297
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• 2014

For several decades, many countries operating nuclear power plants have been studying the various disposal alternatives to dispose of the spent nuclear fuel or high-level radioactive waste safely. In this paper, as a direct disposal of spent nuclear fuels for deep geological disposal concept, the rod consolidation from spent fuel assembly for the disposal efficiency was considered and analyzed. To do this, a concept of spent fuel rod consolidation was described and the related concepts of disposal canister and disposal system were reviewed. With these concepts, several thermal analyses were carried out to determine whether the most important requirement of the temperature limit for a buffer material was satisfiedin designing an engineered barrier of a deep geological disposal system. Based on the results of thermal analyses, the deposition hole distance, disposal tunnel spacing and heat release area of a disposal canister were reviewed. And the unit disposal areas for each case were calculated and the disposal efficiencies were evaluated. This evaluation showed that the rod consolidation of spent nuclear fuel had no advantages in terms of disposal efficiency. In addition, the cooling time of spent nuclear fuels from nuclear power plant were reviewed. It showed that the disposal efficiency for the consolidated spent fuel rods could be improved in the case that cooling time was 70 years or more. But, the integrity of fuels and other conditions due to the longer term storage before disposal should be analyzed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.40 no.5
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• pp.437-447
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• 2016

In a nuclear power plant, the fuel assembly, which is composed of fuel rods, burns, and the high temperature can generate power. The fuel rod consists of pellets and a cladding that covers the pellets. It is important to understand the pellet-cladding mechanical interaction with regard to nuclear safety. This paper proposes simulation of the PCMI. The gap between the pellets and the cladding, and the contact pressure are very important for conducting thermal analysis. Since the gap conductance is not known, it has to be determined by a suitable method. This paper suggests a solution. In this study, finite element (FE) contact analysis is conducted considering thermal expansion of the pellets. As the contact causes plastic deformation, this aspect is considered in the analysis. A 3D FE module is developed to analyze the PCMI using FORTRAN 90. The plastic deformation due to the contact between the pellets and the cladding is the major physical phenomenon. The simple analytical solution of a cylinder is proposed and compared with the fuel rod performance code results.

• 한국신재생에너지학회:학술대회논문집
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• 2008.05a
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• pp.13-16
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• 2008

Solid oxide fuel cell(SOFC) is an electrochemical energy conversion system with high efficiency and low-emission of pollution. In order to reduce the operating temperature of SOFC system under $800^{\circ}C$, the thickness reduction of YSZ electrolyte to be as thin as possible, e.g., less than 10 ${\mu}m$ are considered with the microstructure control and optimum design of unit cell. Methods for reducing the thickness of YSZ electrolyte have been investigated in coin cell. Moreover, a large unit cell($8cm{\times}8cm$) for SOFC was fabricated using an anode-supported electrolyte assembly with a thinner electrolyte layer, which was prepared by a tape casting method with a co-sintering technique. we studied the design factors such as active layer, electrolyte thickness, cathode composition, etc,. by the coin type of unit cell ahead of the fabrication process of a large unit cell and also reviewed about the evaluation technique of a large size unit cell such as interconnect design, sealing materials and current collector and so forth. Electrochemical evaluations of the unit cells, including measurements such as power density and impedance, were performed and analyzed. Maximum power density and polarization impedance of coin cell were 0.34W/$cm^2$ and $0.45{\Omega}cm^2$ at $800^{\circ}C$, respectively. However, Maxium power density of a large unit cell($5cm{\times}5cm$) decreased to 0.21W/$cm^2$ at $800^{\circ}C$ due to the increase of ohmic resistance. However, It was found that the potential value of a large unit cell loaded by 0.22A/$cm^2$ showed 0.76V at 100hrs without the degradation of unit cell.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.97-102
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• 1984

Criticality analyses of the KSC-1(Korean Shipping Cask-1) spent fuel shipping cask have been performed with the help of KENO-IV Monte Carlo computer code and 19-group CSLIB 19 cross section set which was generated from AMPX modular system. The analyses followed a benchmark calculation which has been made regard to the B & W CX-10 criticality facility in order to validate the Monte Carlo code cross section set described above. The KSC-1 shipping cask seems to be safe in the criticality point of view for the transport of one PWR spent fuel assembly under the normal conditions as well as the hypothetical accident conditions.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2460-2470
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• 2020

To study the thermophysical and neutronic properties of thorium-plutonium fuel, a conceptual design of a hybrid facility consisting of a subcritical Th-Pu reactor core and a source of additional D-D neutrons that places on the axis of the core is proposed. The source of such neutrons is a column of high-temperature plasma held in a long magnetic trap for D-D fusionreactions. This article presents computer simulation results of generation of thermonuclear neutrons in the plasma, facility neutronic properties and the evolution of a fuel nuclide composition in the reactor core. Simulations were performed for an axis-symmetric radially profiled reactor core consisting of zones with various nuclear fuel composition. Such reactor core containing a continuously operating stationary D-D neutron source with a yield intensity of Y = 2 × 10¹⁶ neutrons per second can operate as a nuclear hybrid system at its effective coefficient of neutron multiplication 0.95-0.99. Options are proposed for optimizing plasma parameters to increase the neutron yield in order to compensate the effective multiplication factor decreasing and plant power in a long operating cycle (3000-day duration). The obtained simulation results demonstrate the possibility of organizing the stable operation of the proposed hybrid 'fusion-fission' facility.

• Proceedings of the SAREK Conference
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• 2007.11a
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• pp.310-315
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• 2007

The FTL has been developed to be able to irradiate test fuels at the irradiation hole(IR1 hole) by considering its utility and user's irradiation requirements. FTL consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). Test condition in IPS such as pressure, temperature and the water quality, can be controlled by OPS. For safety assurance IPS is designed to have dual stainless steel pressure vessel and OPS is composed of main cooling water system, emergency cooling water system, LMP(letdown, make-up, purification) system, etc. FTL Conceptual design was set up in 2001, basic design had completed including a design requirement, basic piping & instrument diagram (P&ID), and the detail design in 2004. In 2005, the development team carried out purchase and manufacture hardware and make a contract for construction work. FTL construction work began on August, 2006 and ended on March, 2007. After FTL development which is expected to be finished by 2008, FTL will be used for the irradiation test of the new PWR-type fuel and can maximize the usage of HANARO.

• Korean Chemical Engineering Research
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• v.54 no.2
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• pp.181-186
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• 2016

Recently, there are many efforts focused on development of more economical non-fluorinated membranes for use in PEMFCs (Proton Exchange Membrane Fuel Cells). In this study, characteristics of sulfonated Poly(ether ether ketone) (sPEEK) were compared according to degrees of sulfonation (DS), relative humidity, cell temperatures at PEMFC operation condition. I-V polarization curve, hydrogen crossover, electrochemical surface area, membrane resistance and charge transfer resistance were measured. sPEEK membrane showed high performance at high DS, high temperature and high relative humidity, in particular, performance of sPEEK membrane decreased largely due to low ionic conductivity at low DS and low relative humidity.

• Macromolecular Research
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• v.16 no.6
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• pp.549-554
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• 2008

Proton conducting crosslinked membranes were prepared using polymer blends of polystyrene-b-poly(hydroxyethyl acrylate)-b-poly(styrene sulfonic acid) (PS-b-PHEA-b-PSSA) and poly(vinyl alcohol) (PVA). PS-b-PHEA-b-PSSA triblock copolymer at 28:21:51 wt% was synthesized sequentially using atom transfer radical polymerization (ATRP). FT-IR spectroscopy showed that after thermal ($120^{\circ}C$, 2 h) and chemical (sulfosuccinic acid, SA) treatments of the membranes, the middle PHEA block of the triblock copolymer was crosslinked with PVA through an esterification reaction between the -OH group of the membrane and the -COOH group of SA. The ion exchange capacity (IEC) decreased from 1.56 to 0.61 meq/g with increasing amount of PVA. Therefore, the proton conductivity at room temperature decreased from 0.044 to 0.018 S/cm. However, the introduction of PVA resulted in a decrease in water uptake from 87.0 to 44.3%, providing good mechanical properties applicable to the membrane electrode assembly (MEA) of fuel cells. Transmission electron microscopy (TEM) showed that the membrane was microphase-separated with a nanometer range with good connectivity of the $SO_3H$ ionic aggregates. The power density of a single $H_2/O_2$ fuel cell system using the membrane with 50 wt% PVA was $230\;mW/cm^2$ at $70^{\circ}C$ with a relative humidity of 100%. Thermogravimetric analysis (TGA) also showed a decrease in the thermal stability of the membranes with increasing PVA concentration.

• Resources Recycling
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• v.20 no.2
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• pp.67-73
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• 2011

In this study, for recovery of renewable noble metal from used stack of fuel cell, synthesis of platinum nano particle is established through effect of platinum solution concentration, pH value, reducing agent and dispersing agent at a volume ratio of 1 mM $H_2PtCl_6$:10 mM $NaBH_4$:8 mM Cl4TABr = 1:0.4:0.4(vol.%), pH4, $50^{\circ}C$, 160 rpm and 10min. Less than 5 nm platinum particles were synthesized using Pt leaching solution from used MEA of stack under same condition of method using simulated Pt solution. The characteristics of synthesized nano particles was illustrated by XPS analysis as the reduction of platinum ions into platinum metals(zero-valent).