• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.6
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• pp.1053-1059
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• 2002

The purpose of this study is to suggest how the machining technique of constant control of cutting speed can improve precision machining and tool life in high speed machining using a ball end mill. Cutting speed is changed in machining fee form surfaces such as connecting rod die. So, we don't have supreme surface form and tool life on machining. To solve this problem we should settle on optimal cutting speeds in free form surface machining. And, to improve precision machining, We must execute high speed machining methods to output optimum NC data using developed constant control of cutting speed program after modeling by CAD/CAM. In this paper, a comparison was made of the cutting precision and tool life in conventional cutting and those in connecting rod machining applying the program developed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.134-140
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• 2016

The Control Rod Assembly (CRA) controls the reactor power by adjusting its position in the reactor core during normal operation and should be quickly inserted into the reactor core by free drop under scram condition to shut down chain reactions. Therefore, the drop time of the CRA is one of important factors for the safety of the nuclear reactor and must be experimentally verified. This study presents the drop performance test of the CRA which has been conceptually designed for the Proto-type Generation IV Sodium-cooled Fast Reactor. During the test, the CRA was free dropped from a height of 1 m under different flow rate conditions and its drop time was measured. The results showed that the drop time of the CRA increased as the flow rate increased; the average drop times of the CRA were approximately 1.527 seconds, 1.599 seconds and 1.676 seconds at 0%, 100% and 200% of design flow rates, respectively.

• 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.53 no.12
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• pp.4166-4178
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• 2021

The Control Rod Drive Mechanism (CRDM) is an essential part of the reactor, which realizes the start-stop and power adjustment of the reactor by lifting and lowering the control rod assembly. As a moving part in CRDM, the latch directly contacts with the control rod assembly, and the life of latch is closely related to the service life of the reactor. In this paper, the relationship between the life of the latch and the step stress, friction stress, and impact stress in the process of movement is analyzed, and the optimization methodology and process of latch life based on the approximate model are proposed. The design variables that affect the life of the latch are studied through the experimental design, and the optimization objective of design variables based on the latch life is established. Based on this, an approximate model of the life of the latch is built, and the multi-objective optimization of the life of the latch is optimized through the NSGA-II algorithm.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.81-92
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• 2006

In this study, the core dynamics of a PWR reactor is identified online by a recursive least-squares method. Based on the identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to designing an automatic controller for the thermal power control of PWR reactors. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, this procedure for solving the optimization problem is repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired temperature, as well as minimizing the variation of the control rod positions. In addition, the objectives are subject to the maximum and minimum control rod positions as well as the maximum control rod speed. Therefore, a genetic algorithm that is appropriate for the accomplishment of multiple objectives is utilized in order to optimize the model predictive controller. A three-dimensional nuclear reactor analysis code, MASTER that was developed by the Korea Atomic Energy Research Institute (KAERI) , is used to verify the proposed controller for a nuclear reactor. From the results of a numerical simulation that was carried out in order to verify the performance of the proposed controller with a $5\%/min$ ramp increase or decrease of a desired load and a $10\%$ step increase or decrease (which were design requirements), it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• Proceedings of the KIEE Conference
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• 2005.10b
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• pp.104-106
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• 2005

In this work, a PWR reactor core dynamics is identified online by a recursive least squares method. Based on this identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to design an automatic controller for thermal power control in PWRs. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired one, and the variation of the control rod positions. Also, the objectives are subject to maximum and minimum control rod positions and maximum control rod speed. Therefore, the genetic algorithm that is appropriate to accomplish multiple objectives is used to optimize the model predictive controller. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), is used to verify the proposed controller for a nuclear reactor. From results of numerical simulation to check the performance of the proposed controller at the 5%/min ramp increase or decrease of a desired load and its 10% step increase or decrease which are design requirements, it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.18 no.9
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• pp.92-100
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• 2017

The Helicopter MRA Control Rod System has the important function of controlling the speed, height, and direction of helicoptersby adjusting the main rotor disc. However, the ingress of water into the inner control rod can cause ice damage in the rod during winter operation and also corrosion;these defects need to be rectified. The water flowed into the control rod through the upper side space, and the rod was cracked during icing expansion occurring at low temperature. The corrosion occurred due to the lack of coating process during the manufacturing process. To resolve these problems, the upper rod was sealed to prevent water inflow and a coating process was added to prevent corrosion. These solutions were verified by awaterproof test and a salt fog test. The phenomena, causes and measures were reviewed and the methods of improvement were established and proven. This proposed technology to prevent water infiltration and corrosion will contribute to the safety of rotary wing aircraft.

• Journal of the Korean Society for Precision Engineering
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• v.14 no.12
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• pp.174-181
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• 1997

A proposed hybrid self-tuning control scheme for single rod hydraulic cylinder which has varying loads is presented here. An adaptive controller is developed for the system that use feedforward and P feedback control for simultaneous parameter identification and tracking control. Through experimental results, the performance comparison of the hybrid self-tuning controller with a constant gain P contro- ller clearly shows its superior ability in handling load changes in quiescent states.

• Journal of KSNVE
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• v.8 no.3
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• pp.441-449
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• 1998

A self-tuning control scheme, incorporated with the simplified 1st-order ARMAX(Auto-Regressive Moving Average eXogenous) model, for single rod hydraulic cylinder which has varying dynamic characteristics is presented here. An adaptive controller is developed for the system that uses feedforward and optimal feedback control for simultaneous parameter identification and tracking control. Through experimental results, the performance comparison of the self-tuning controller with a fixed gain proportional controller clearly shows its superior ability in handling load changes in quiescent states.

• Proceedings of the KIEE Conference
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• 2002.07d
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• pp.2646-2648
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• 2002

This paper deals with a comparative evaluation of the reliability of control rod drive mechanisms including their driving methods for nuclear power plants. Basically there exist two types of electromagnetic-jack-type drive mechanisms in commercial use that are called as Control Rod Drive Mechanism and Control Element Drive Mechanism. Each type has corresponding drive sequence to make the movements of insertion and withdrawal. A state-space model is derived for each model graphically. Then the evaluation of the reliability is carried out on the programming tool called SHARPE. The evaluation does not give any meaningful numerical values for the reliability but just shows a relative degree to each other in view of reliability.