• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.147-151
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• 1997

Fuel management simulation have been performed for CANFLEX-0.9％ RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor.

• The KSFM Journal of Fluid Machinery
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• v.18 no.6
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• pp.5-11
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• 2015

In a pressurized water reactor (PWR), control rod assembly (CRA) falls into the guide tubes of a fuel assembly due to gravity for scram. Various theoretical approaches and numerical analyses have been performed because its shape is simple and its design was completely developed several decades ago. A control rod assembly for a sodium-cooled faster reactor (SFR) which is geometrically more complicated is being actively developed in Korea nowadays. Drop time and impact velocity of a CRA are important parameters with respect to reactivity insertion time and the mechanical robustness of a CRA and a guide duct. In this paper, computational method considering simultaneously the equation of motion for rigid body and the Navier-Stokes equations for fluid is suggested and verified by comparison with theoretical analysis results. Through this valuable CFD analysis method, drop time and impact velocity of initially designed SFR CRA are evaluated before performing scram tests with it.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.

• Journal of the Korean Institute of Intelligent Systems
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• v.5 no.3
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• pp.52-63
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• 1995

A neuro-fuzzy algorithm is presented for nuclear reactor power control in a pressurized water reactor. Automatic reacotr power control is complicated by the use of control rods because of highly nonlinear dynamics in the axial power shape. Thus, manual shaped controls are usually employed even for the limited capability during the power maneuvers. In an attempt to achieve automatic shape control, a neuro-fuzzy approach is considered because fuzzy algorithms are good at various aspects of operator's knowledge representation while neural networks are efficinet structures capable of learning from experience and adaptation to a changing nuclear core state. In the proposed neuro-fuzzy control scheme, the rule base is formulated based ona multi-input multi-output system and the dynamic back-propagation is used for learning. The neuro-fuzzy powere control algorithm has been tested using simulation fesponses of a Korean standard pressurized water reactor. The results illustrate that the proposed control algorithm would be a parctical strategy for automatic nuclear reactor power control.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.289-290
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• 2002

Stainless steel ball bearings are used in the control element drive mechanism and driving mechanisms such as step motor and gear boxes for the integral nuclear reactor, SMART. The bearings operate in pressurized pure water (primary coolant) at high temperature and should be lubricated with only this water because it is impossible to supply greases or any additional lubricant since the whole nuclear rector system should be perfectly sealed and the coolant cannot contain ingredients for bearing lubrication. Temperature of water changes from room temperature to about 120 degree Celsius and pressure rises up to 15MPa in the nuclear reactor. It can be anticipated that the frictional characteristics of the ball bearings changes according to the operating conditions, however little data are available in the literature. It is found that friction coefficient of 440C stainless steel itself does not change sharply according to temperature variation from the former research, and the friction coefficient is about 0.45 at low speed range. In this research frictional characteristics of the assembled ball bearings are investigated. A special tribometer is used to simulate the axial loading and the bearing operating conditions, temperature and pressure in the driving mechanism in the nuclear reactor. Highly purified water is used as lubricant ‘ and the water is heated up to 120 degree Celsius and pressurized to 15MPa. Friction force is monitored by the torque transducer.

• Transactions of the Korean hydrogen and new energy society
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• v.13 no.2
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• pp.101-109
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• 2002

This work was focused on the thermal decomposition of methane into hydrogen and carbon black without emitting carbon dioxide. Extensive experimental investigation on the thermal decomposition of methane has been carried out using a continuous flow reaction system with tubular reactor. The experiments were conducted at the atmospheric pressure condition in the wide range of temperature ($950-1150^{\circ}C$) and flow rate (250 - 1500 ml/min) in order to study their dependency on hydrogen yield. During the experiments the carbon black was successfully recovered as an useful product. Undesirable pyrocarbon was also formed as solid film, which was deposited on the inside surface of tubular reactor. The film of pyrocarbon in the reactor wall became thicker and thicker, finally blocking the reactor. The design of an efficient reactor which can effectively suppress the formation of pyrocarbon was thought to be one of the most important subjects in the thermal cracking of methane.

• Solar Energy
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• v.18 no.3
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• pp.197-203
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• 1998

A 5 ton/day pilot scale two-phase anaerobic digester was constructed and tasted to treat Korean food wastes in Anyang city. The process was developed based on 3 years of lab-scale experimental results on am optimim treatment method for the recovery of biogas and humus. Problems related to food waste are ever Increasing quantity among municipal solid wastes(MSW) and high moisture and salt contents. Thus our food waste produces large amounts of leachate and bed odor in landfill sites which are being exhausted. The easily degradable presorted food waste was efficiently treated in the two-phase anaerobic digestion process. The waste contained in plastic bags was shredded and then screened for the removal of inert material such as fabrics and plastics, and subsequently put into the two-stage reactors. Heavy and light inerts such as bones, shells, spoons and plastic pieces were again removed by gravity differences. The residual organic component was effectively hydrolyzed and acidified in the first reactor with 5 days space time at pH of about 6.5. The second, methanization reactor part of which is filled with anaerobic fillters, converted the acids into methane with pH between 7.4 to 7.8. The space time for the second reactor was 15 days. The effluent from the second reactor was recycled to the first reactor to provide alkalinities. The process showed stable steady state operation with the maximum organic rate of 7.9 $kgVS/m^3day$ and the volatile solid reduction efficiency of about 70%. The total of 3.6 tons presorted MSW containing 2.9 tons of food organic was treated to produce about $230m^3$ of biogas with 70% of methane and 80kg humus. This process is extended to full scale treating 15 tons of food waste a day in Euiwang city and the produced biogas is utilized for the heating/cooling of adjacent buildings.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.17-26
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• 1996

The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by residual spacer grid spring deflection or rod-to-grid gap. In order to evaluate the impact of fuel design parameters on the fretting wear-induced damage, a simulation methodology of the in-reactor fuel rod supporting conditions as a function of burnup has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. Based on the parametric studies on these key parameters, it is found that the initial spring deflection, the spring force relaxation rate and cladding creepdown rate are in the order of the impact on the in-reactor fuel rod supporting conditions. Application of this simulation methodology to the fretting wear-induced failure experienced in a commercial plant indicates that this methodology can be utilized as an effective tool in evaluating the capability of newly developed cladding materials and/or new spacer grid designs against the fretting wear-induced damage.

• Proceedings of the Materials Research Society of Korea Conference
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• 1998.08a
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• pp.96.1-96
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• 1998

• 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.