• Nuclear Engineering and Technology
• /
• v.26 no.4
• /
• pp.536-547
• /
• 1994

The state feedback optimal control techniques are used in designing the reactor control system. The mathematical plant model with the temperature feedback effects is established from the one delayed neutron group point kinetics equation and the singly lumped thermal-hydraulic balance equations, and is expressed in terms of state variables. The LQR (Linear Quadratic Regulator) control system is designed, being followed by the LQG (Linear Quadratic Gaussian) design to determine the optimal conditions of rod movement for the desired reactor power responses. And two different servo control schemes, the ordinary feedback system and the order increased regulating system, are proposed for the purpose of input tacking. The general control characteristics such as stability margins and output responses are discussed. Comparing each other, it is found that the order increased regulating system has far better control characteristics than the ordinary feedback system.

• International Journal of Reliability and Applications
• /
• v.5 no.1
• /
• pp.37-46
• /
• 2004

Coil spring of automobile suspension system is very important to safety and dynamics of passenger car and requires a highly advanced quality control during manufacturing processes. Surface cracks on the coil spring rod produced by mechanical machining and heat treatment may cause a severe accident and large cost to the manufacturer. In order to detect surface cracks of the rod, guided wave technique is applied for a fast total volumetric inspection. Pochhammer equation is studied to investigate the dispersion characteristics of the guided wave in the spring rod and optimal wave modes sensitive to the surface crack are selected experimentally to design the experimental arrangement for the generation of guided wave. Rod samples with different size of artificial axial EDM notch on the surface ranging from 50${\mu}{\textrm}{m}$ to 1 mm are examined by guided wave and inspection results discussed.

• Journal of Advanced Marine Engineering and Technology
• /
• v.28 no.1
• /
• pp.136-144
• /
• 2004

In general a hydraulic system which uses a single rod hydraulic as an actuator is modeled as a nonlinear system and reveals uncertain Parameter characteristics such as the density variation of hydraulic oil and is subject to load variations and severe disturbances during operation. A variable design-parameter fuzzy PID controller is adopted to solve these undesirable internal and external problems and its effectiveness is verified through computer simulations for control performance and real time control possibility.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.26 no.5
• /
• pp.552-558
• /
• 2016

A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod within a reactor core to control the reactivity of the core. The CRDM for Jordan Research and Training Reactor with 5MW power has been designed and fabricated based on the HANARO’s experience through KAERI and DAEWOO consortium. This paper describes the seismic test results to demonstrate the operability, the drop performance and the structural integrity of CRDM during or after seismic excitations. The seismic tests are carried out under 5 OBE and 1 SSE loads at three Test Rigs simulating the reactor structure and the pool top. From the tests, the CRDM is smoothly driven without a malfunction of stepping motor under OBE load. The pure drop time under OBE and SSE loads is measured as 1.169s and 1.855s to meet the design requirement. Also, it is found that the CRDM maintains the structural integrity without a change of the function and natural frequency before and after seismic loads.

• Proceedings of the KIEE Conference
• /
• 2002.07b
• /
• pp.638-640
• /
• 2002

This paper introduces the design method of a Control Rod Drive Mechanism that consists of 3 coils -lift, movable gripper and stationary gripper coil. The vertical attraction forces of the lift, movable gripper, and stationary gripper armatures are calculated by FEM, then the dynamics with full load is demonstrated.

• Proceedings of the Korean Institute of Intelligent Systems Conference
• /
• 1995.10b
• /
• pp.145-152
• /
• 1995

This paper presents a novel SOFLIC(self organizing fuzzy logic intelligent controller) for reactor rod control system in nuclear power plant. The output of fuzzy controller is gener ated by using two signal : the error between reference and average temperature, and the error between reference and neutron flux-converted temperatures. Flexibility of the controller is enhanced by using self-organizing feature and the controller respond to variation of system parameter with more precision. performances of the SOFLIC and PID are simulated with the model developed for a nuclear power plant. The SOFLIC is superior to PID : SOFLIC provides more rapid load following capability. more robustiness for variation in process dynamics and minimization of engineer's mistakes in controller design.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2000.11a
• /
• pp.639-642
• /
• 2000

• Nuclear Engineering and Technology
• /
• v.29 no.2
• /
• pp.167-174
• /
• 1997

The method that the operating condition of Control Rod Drive Mechanism (CRDM) can be monitored without mounting sensors within CRDM housing was developed, and by using this developed method the closed-loop controller for the CRDM was designed which can optimize the performance and maximize the reliability of CRDM operation. Neural network is utilized as pattern recognition engine in detecting CRDM actuation. In this paper, most problems in previous open loop system are resolved. The control algorithms for closed-loop system ore developed and implemented within the hardware of timing controller based on microprocessor. All functions in the timing controller ore verified by means of real time CRDM simulator. The results show that the timing controller performs its intended functions properly.

• Nuclear Engineering and Technology
• /
• v.49 no.1
• /
• pp.141-147
• /
• 2017

Mechanical shim is an advanced technology for reactor power and axial offset control with control rod assemblies. To address the adverse accuracy impact on the ex-core power range neutron flux measurements-based axial offset control resulting from the variable positions of control rod assemblies, the lead-lag-compensated in-core self-powered vanadium detector signals are utilized. The prompt ${\gamma}$ current of self-powered detector is ignored normally due to its weakness compared with the delayed ${\beta}$ current, although it promptly reflects the flux change of the core. Based on the features of the prompt ${\gamma}$ current, a method for configuration of the lead-lag dynamic compensator is proposed. The simulations indicate that the method can improve dynamic response significantly with negligible adverse effects on the steady response. The robustness of the design implies that the method is of great value for engineering applications.

• Nuclear Engineering and Technology
• /
• v.41 no.3
• /
• pp.307-318
• /
• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.