• 대한전기학회:학술대회논문집
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• 대한전기학회 2007년도 제38회 하계학술대회
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• pp.1630-1631
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• 2007

This paper deals with the design of control cabinet based on safety PLC for reactor power control system(PCS). The PCS controls the operation of the CEDMs(Control Element Drive Mechanisms). The CEDM moves the CEAs(Control Element Assemblies) which regulates the reactor power, vertically in the reactor core. The Control Cabinet in PCS makes and conveys control signals to the power cabinet which provides power to the CEDM. We designed the Control Cabinet, based on POSAFE-Q, safety PLC. The application programs working in PLC can be programmed by pSET, Identified Development Environment.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.803-810
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• 2022

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

• 대한전자공학회논문지
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• 제4권2호
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• pp.22-28
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• 1967

Argonne Advanced Research Reactor의 동특성을 가지고 원자로의 과도적응답에 대한 문제를 비선형인 Relay와 gear를 중점으로 하고 Analog Computer를 이용하여 자동제어상의 안전도를 검토하였다.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1991년도 한국자동제어학술회의논문집(국내학술편); KOEX, Seoul; 22-24 Oct. 1991
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• pp.81-86
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• 1991

The iterative learning operation has been utilized in the temperature Control of a batch reactor. A generic form of feedback-assisted first-order learning control scheme was constructed and then various design and operation modes were derived through convergence and robustness analysis in the frequency domain. The proposed learning control scheme was then implemented on a bench scale batch reactor with the heat of reaction simulated by an electric heater. The results show a great improvement in the performance of control as the number of batch operations progressed.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1990년도 한국자동제어학술회의논문집(국제학술편); KOEX, Seoul; 26-27 Oct. 1990
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• pp.1277-1283
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• 1990

The iterative learning control synthesized in the frequency domain has been utilized for temperature control of a batch reactor. For this purpose, a feedback-assisted generalized learning control scheme was constructed first, and the convergence and robustness analyses were conducted in the frequency domain. The feedback-assisted learning operation was then implemented in a bench scale batch reactor where reaction heat is simulated using an electric heater. As a result, progressive reduction of temperature control error could be obviously observed as batch operation is repeated.

• 제어로봇시스템학회논문지
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• 제14권3호
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• pp.221-225
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• 2008

This paper presents a self-tuning controller for pressurized water reactor (P.W.R.). This self-tuning controller includes two substantial steps, such as parameter identification and control-law building in each cycle. Extended least square algorithm is used for parameter identification, Kalman filter is used for state estimation, and discrete Riccati equation is used for optimal control. Effectiveness of this algorithm is shown through computer simulation and sensitivity analysis.

• 상하수도학회지
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• 제20권6호
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• pp.829-836
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• 2006

The reaction rate of denitrification is primarily affected by the utilization of organics that are usually limited in the anoxic period in a sequencing batch reactor. It is necessary to add an extemal carbon source for sufficient denitrification. An adaptive model of state-space based on the extended Kalman filter is applied to manipulate the dosage rate of extemal carbon automatically. Control strategies for denitrification have been studied to improve control performance through simulations. The normal control strategy of the constant set-point results in the overdosage of external carbon and deterioration of water quality. To prevent the overdosage of external carbon, improved control strategies such as the constrained control action, variable set-point, and variable set-point after dissolved oxygen depletion are required. More stable control is obtained through the application of the variable set-point after dissolved oxygen depletion. The converging value of the estimated denitrification coefficient reflects conditions in the reactor.

• Journal of Mechanical Science and Technology
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• 제18권12호
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• pp.2080-2094
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• 2004

A model predictive controller is designed to control thermal power in a nuclear reactor. The basic concept of the model predictive control is to solve an optimization problem for finite future time steps at current time, to implement only the first optimal control input among the solved control inputs, and to repeat the procedure at each subsequent instant. A controller design model used for designing the model predictive controller is estimated every time step by applying a recursive parameter estimation algorithm. A 3-dimensional nuclear reactor analysis code, MASTER that was developed by Korea Atomic Energy Research Institute (KAERI), was used to verify the proposed controller for a nuclear reactor. It was known that the nuclear power controlled by the proposed controller well tracks the desired power level and the desired axial power distribution.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2522-2534
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• 2020

It is well known that the model of nuclear reactors features natural nonlinearity, and variable parameters during power tracking operation. In this paper, a disturbance observer-based adaptive sliding mode control (DOB-ASMC) strategy is proposed for power tracking of the pressurized-water reactor (PWR) in the presence of lumped disturbances. The nuclear reactor model is firstly established based on point-reactor kinetics equations with six delayed neutron groups. Then, a new sliding mode disturbance observer is designed to estimate the lumped disturbance, and its stability is discussed. On the basis of the developed DOB, an adaptive sliding mode control scheme is proposed, which is a combination of backstepping technique and integral sliding mode control approach. In addition, an adaptive law is introduced to enhance the robustness of a PWR with disturbances. The asymptotic stability of the overall control system is verified by Lyapunov stability theory. Simulation results are provided to demonstrate that the proposed DOB-ASMC strategy has better power tracking performance than conventional sliding mode controller and PID control method as well as conventional backstepping controller.

• 한국압력기기공학회 논문집
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• 제12권1호
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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.