• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1013-1021
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• 2024

A series of experiments were performed to produce a fuel source for a molten salt reactor (MSR) through pyroprocessing technology. A simulated LiCl-KCl-UCl₃-NdCl₃ salt system was prepared, and the U element was fully recovered using a liquid cadmium cathode (LCC) by applying a constant current. As a result, the salt was purified with an UCl₃ concentration lower than 100 ppm. Subsequently, the U/RE ingot was prepared by melting U and RE metals in Y₂O₃ crucible at 1473 K as a surrogate for RE-rich ingot product from pyroprocessing. The produced ingot was sliced and used as a working electrode in LiCl-KCl-LaCl₃ salt. Only RE elements were then anodically dissolved by applying potential at - 1.7 V versus Ag/AgCl reference electrode. The RE-removed ingot product was used to produce UCl₃ via the reaction with NH4Cl in a sealed reactor.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.67 no.5
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• pp.611-616
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• 2018

In the cost based pool market in Korea, there is no compensation of reactive power because the fuel cost for reactive power is relatively low compared to that of active power. However, the change of energy paradigm in the future, such as widespread integration of distributed renewable energy source, will prevent the system operator from mandating the reactive power supply without any compensation. Thus, in this study, we propose the reference voltage of the 345 kV transmission system that minimizes the reactive power supply. This is closely related to the economic dispatch of reactive power aiming at minimizing the compensation cost for the reactive power service. In order to verify the effectiveness of the proposed reference voltage, the simulations are performed using the IEEE 14 bus system and the KEPCO real networks. The simulation results show that a voltage lower than the current reference value is recommended to reduce the reactive power supply and also suggest that the current voltage specification for the 345 kV system needs to be reviewed.

• 제어로봇시스템학회:학술대회논문집
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• 1999.10a
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• pp.104-107
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• 1999

Tracking control of an arbitrary reference has been discussed for 7Th order 2-input 2-output non-minimum phase EGR/VGT diesel engines. To meet strict emission regulations and customer demands, the desired set points, the air-fuel ratio and the ERG flow fraction, determined from a static engine data based on engine speed and the desired fueling rate are transformed into the set points for the two sensor measurement outputs. Applying the sliding mode tracking control theory proposed by Jeong and Utkin, two step design was carried out using the bounded solution of an unstable zero dynamics for the given reference signals. This paper shows through simulations how stabilization of unstable zero dynamics and reference tracking can be accomplished simultaneously.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.273-282
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• 2012

A new effective methodology for optimizing the enrichment of low-enriched zones as well as gadolinia fuel ($UO_2/Gd_2O_3$) rod designs in PLUS7 fuel assemblies was developed to minimize the maximum peak power in the core and to maximize the cycle lifetime. An automated link code was developed to integrate the genetic algorithm (GA) and the core design code package of ALPHA/PHOENIX-P/ANC and to generate and evaluate the candidates to be optimized efficiently through the integrated code package. This study introduces an optimization technique for the optimization of gadolinia fuel rod designs in order to effectively reduce the peak powers for a few hot assemblies simultaneously during the cycle. Coupled with the gadolinia optimization, the optimum enrichments were determined using the same automated code package. Applying this technique to the reference core of Ulchin Unit 4 Cycle 11, the gadolinia fuel rods in each hot assembly were optimized to different numbers and positions from their original designs, and the maximum peak power was decreased by 2.5%, while the independent optimization technique showed a decrease of 1.6% for the same fuel assembly. The lower enrichments at the fuel rods adjacent to the corner gap (CG), guide tube (GT), and instrumentation tube (IT) were optimized from the current 4.1, 4.1, 4.1 w/o to 4.65, 4.2, 4.2 w/o. The increase in the cycle lifetime achieved through this methodology was 5 effective full-power days (EFPD) on an ideal equilibrium cycle basis while keeping the peak power as low as 2.3% compared with the original design.

• Environmental Engineering Research
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• v.14 no.3
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• pp.158-163
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• 2009

It is difficult for a burner to achieve an increase in combustibility and a reduction of NOx emission, simultaneously. The reason is because thermal NOx could be reduced at low temperature, while the combustibility should be decreased. To solve this problem, an externally oscillated staging burner was developed, and experiment was conducted according to effective parameters. The combustibility could be improved through the accelerated transfer of heat, mass and momentum obtained by external oscillation. Also, NO is reduced by the decrease of residence time of burning gas in the local highest-temperature spot, which is decreased by the external oscillation and fuel staging. Experiments on variables were conducted to determine the reference flame, and the flame generating the lowest NO concentration was selected. The conditions of reference flame were oscillation frequency 250 Hz, sound pressure 1 VPP, and air ratio 1.1, and NO and CO concentrations were 1ppm and 20 ppm, respectively.

• Transactions of the Korean Society of Automotive Engineers
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• v.16 no.5
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• pp.157-163
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• 2008

The HCCI combustion mode poses its own set of narrow engine operating by knocking. In order to solve this, inhomogeneity method of mixture and temperature is suggested. The purpose of this research is to get fundamental knowledge about the effect of thermal stratification on HCCI combustion of PRF -Air mixture. The temperature stratification is made by buoyancy effect in combustion chamber of RCM. The analysis items are pressure, temperature of in-cylinder gas and combustion duration. In addition, the structure of flames using the two dimensional chemiluminescence's images by a framing camera are analyzed. Under stratification, the LTR starting time and the HTR starting time are advanced than that of homogeneous. Further, the LTR period of homogeneous conditions became shorter than that of the stratified conditions. With the case of homogeneous condition, the luminosity duration becomes shorter than the case of stratified condition. Additionally, under stratified condition, the brightest luminosity intensity is delayed longer than at homogeneous condition.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.318-329
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• 2016

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed "Burnable absorber-Integrated Guide Thimble" (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.915-923
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• 2015

Pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems in Korea. The Korea Atomic Energy Research Institute, Daejeon, Korea has been studying pyroprocess technology, and the conceptual design of an engineering-scale pyroprocess facility, called the Reference Engineering-scale Pyroprocess Facility, has been performed on the basis of a 10 ton heavy metal throughput per year. In this paper the concept of Reference Engineering-scale Pyroprocess Facility is introduced along with its safety requirements for the protection of facility workers, collocated workers, the off-site public, and the environment. For the identification of safety structures, systems, and components and/or administrative controls, the following activities were conducted: (1) identifying hazards associated with operations; (2) identifying potential events associated with these hazards; and (3) identifying the potential preventive and/or mitigative controls that reduce the risk associated with these accident events. This study will be used to perform a safety evaluation for accidents involving any of the hazards identified, and to establish safety design policies and propose a more definite safety design.

• Journal of Energy Engineering
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• v.13 no.4
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• pp.242-258
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• 2004

The duplex burnable poison absorbers concept was suggested by Korea Atomic Energy Research Institute. This BP rod is composed of inner region of natural U-Gd$_2$O$_3$ and outer shell of enriched UO$_2$-Er$_2$O$_3$. It is expected that this burnable absorber has same reactivity control capability with gadolinia burnable absorber used in extened fuel cycle. In order to evaluate the nuclear feasibility of duplex BPs, the nuclear design characteristics were compared with that of four types of burnable absorbers; gadolinia, erbia, IFBA, dysprosia duplex BP on 24 months fuel cycle for Korean Standard Nuclear Power plants. According to the evaluation results of nuclear characteristics, the duplex BPs were better than other BPs on k-infinitives, reactivity holddown worth (RHW), pin power peaking and moderator temperature coefficient (MTC). The possibility of nuclear core design was also confirmed based on the optimized fuel assemblies which were searched for a sensitivity analysis. Characteristics of core design with duplex BPs was compared with that of reference core with gadolinia BPs for cycle length, power peaking and MTC. The duplex BP core had a little longer cycle length by 4 to 7 days because of increased amount of fissile in enriched uranium at the outer shell of duplex BP In case of power peaking F$\_$Q/ of duplex BP core was reduced from 1.5773 to 1.5335. MTC was also less -0.48 pcm/C than that of reference core. Finally, evaluation of fuel cycle economy was performed for the manufacturing feasibility test and fuel cost evaluation with duplex BPs. Fuel cycle economy of duplex BP core almost was equivalent with that of gadolinia BP core.

• Nuclear Engineering and Technology
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• v.28 no.5
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• pp.440-451
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• 1996

With emphasis on safety, this study addresses for better design condition for the cooling system in a wet-type interim spent fuel storage facility, using a probabilistic safety assessment method. To incorporate the design renovation into the design phase, a simple approach is proposed. By taking the cooling system of a reference design, a fault tree analysis was performed to identify the weak point of the considered system, and then basic factors for design renovation were defined. A total of 21 design alternatives were selected through the combination of the basic factors. Finally, the optimum design alternative for the cooling system is derived by means of the cost and effect analysis based on the estimated cost, system reliability and assumed probabilistic safety criteria. With the assumption that the failure frequency of at-reactor spent fuel cooling system compiles with probabilistic safety criteria for the interim spent fuel cooling system, it was shown that the optimum alternative should have l00% cooling loop redundancy with one pump per cooling loop and a cleanup system installed separately from the main loop. Furthermore, it also should be classified into safety system. The result of this study can be used as a useful basis to identify factors of safety concern and to establish design requirements in the future. The method also can be applied for other nuclear facilities.