• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.391-400
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• 1999

A conceptual core design of the 1,300MWe KNGR (Korean Next Generation Reactor) without using soluble boron for reactivity control was developed to determine whether it is technically feasible to implement SBF (Soluble Boron Free) operation. Based on the borated KNGR core design, the fuel assembly and control rod configuration were modified for extensive use of burnable poison rods and control rods. A new fuel rod, in which Pu-238 had been substituted for a small amount of U-238 in fuel composition, was introduced to assist the reactivity control by burnable poison rods. Since Pu-238 has a considerably large thermal neutron capture cross section, the new fuel assembly showed good reactivity suppression capability throughout the entire cycle turnup, especially at BOC (Beginning of Cycle). Moreover, relatively uniform control of power distribution was possible since the new fuel assemblies were loaded throughout the core. In this study, core excess reactivity was limited to 2.0 %$\delta$$\rho$ for the minimal use of control rods. The analysis results of the SBF KNGR core showed that axial power distribution control can be achieved by using the simplest zoning scheme of the fuel assembly Furthermore, the sufficient shutdown margin and the stability against axial xenon oscillations were secured in this SBF core. It is, therefore, concluded that a SBF operation is technically feasible for a large sized LWR (Light Water Reactor).

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.05a
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• pp.705-709
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• 2002

The spacer grids in nuclear fuel assembly locate and align the fuel rods with respect to each other. They provide axial and lateral restraint against an excessive rod motion mainly caused by coolant flow. It is understood that each rod Is supported by multiple spacer grid. In such a case, it is important to determine spacer grid span so as to avoid resonance between the natural frequency of the fuel rods and excitation frequency. Actually dynamic characteristics of the fuel rods can be improved by assigning adequate spacer grid locations. When a dynamic performance of the structure is to be improved, design sensitivity analysis plays an important role as like many structural redesign problems. In this work, a shape design concept, different from conventional design, was applied to the problem. According to the theory shape can be a design parameter and optimal shape design can be found. This study concentrates on eigenvalue design sensitivity of the fuel rod supported by multiple spacer grids to determine optimal spacer grids positions.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.15-20
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• 1996

We present the result of physical studies for three integral-type burnable absorbers of gadolinia, erbia and IFBA, in the hexagonal fuel assembly. The analysis of nuclear characteristics for gadolinia and IFBA cases shows that the spectrum hardening of hexagonal fuel assembly compared to rectangular one leads to smaller reactivity hold-down worth(RHW) and less change of MTC in the negative direction per insertion of one burnable absorber rod. Erbia case, on the other hand, exhibits reversed trend in RHW and MTC due to the enhanced absorption of large resonance of Erbium at 0.5 eV It turns out to be that Erbia performs best in terms of minimizing the peak pin power and maintaining as more negative MTC as practically attainable during the entire operational phase, and IFBA provides the least residual reactivity penalty at EOC. Therefore, we take Erbium as the suitable burnable absorber and provide optimal designs of 60, 120, 180, 240 and 300 erbia-shimmed hexagonal fuel assemblies with regard to minimizing the peak pin power.

• Macromolecular Research
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• v.16 no.6
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• pp.524-531
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• 2008

The mitigation or elimination of methanol crossover for perfluorosulfonic acid fuel cell membranes has been investigated extensively for direct methanol fuel cell applications with the aim of increasing the electrochemical performance and enhancing the utilization rate of methanol. Self-assembly modifications by applying an oppositely charged polyelectrolyte to Nafion membranes were attempted in order to block or reduce methanol crossover while maintaining the other advantageous properties of Nafion membranes. It was reported that anionic polyallylamine hydrochloride (PAH) was the most efficient polyelectrolyte in reducing methanol crossover, and considerable cell performance was obtained even at a methanol feed concentration of 10 M.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1175-1183
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• 2010

A spacer grid assembly is one of the most important structural components in a Light Water Reactor(LWR) nuclear fuel assembly. In the case of the Zircaloy spacer grid assembly, the primary design consideration is to ensure that lateral dynamic crush strength of the spacer grid assembly is sufficient to resist design basis loads and thereby prevent seismic accidents, without a significant increase in the hydraulic head loss for the reactor coolant in the reactor core. In this study, factors affecting the lateral dynamic crush strength of a spacer grid assembly were analyzed by performing lateral dynamic crush tests and finite element analyses. Further, an effective and economical method to enhance the lateral dynamic crush strength of the spacer grid assembly is proposed.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2012.05a
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• pp.1019-1020
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• 2012

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.163.2-163.2
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• 2005

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2012.10a
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• pp.719-720
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• 2012

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.149-156
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• 2016

The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the $3{\times}3$ short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.12
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• pp.815-824
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• 2016

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermal-hydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermal-hydraulic technology and the commercialization.