• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3595-3603
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• 2022

Nuclear safety-class DCS is used for nuclear reactor protection function, which is one of the key facilities to ensure nuclear power plant safety, the maintenance for DCS to keep system in a high reliability is significant. In this paper, Nuclear safety-class DCS system developed by the Nuclear Power Institute of China is investigated, the model of reliability estimation considering nuclear power plant emergency trip control process is carried out using Markov transfer process. According to the System-Subgroup-Module hierarchical iteration calculation, the evolution curve of failure probability is established, and the preventive maintenance optimization strategy is constructed combining reliability numerical calculation and periodic overhaul interval of nuclear power plant, which could provide a quantitative basis for the maintenance decision of DCS system.

• Journal of Institute of Control, Robotics and Systems
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• v.11 no.6
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• pp.550-557
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• 2005

The multivariate statistical analysis methods, using both multiple linear regression(MLR) and partial least square(PLS), have been applied to predict the reactor properties and the product quality of a direct esterification reactor for polyethylene terephthalate(PET) synthesis. On the basis of the set of data including the flow rate of water vapor, the flow rate of EG vapor, the concentration of acid end groups of a product and other operating conditions such as temperature, pressure, reaction times and feed monomer mole ratio, two multi-variable analysis methods have been applied. Their regression and prediction abilities also have been compared. The prediction results are critically compared with the actual plant data and the other mathematical model based results in reliability. This paper shows that PLS method approach can be used for the reasonably accurate prediction of a product quality of a direct esterification reactor in PET synthesis process.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.84-91
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• 2020

The APR1400 reactor stud holes can be stuck due to high temperatures, high pressure, prolonged engagement, and load changes according to pressure changes in the reactor. Threaded surfaces of a stud hole should be cleaned for the sealing of pressure in reactor vessel by removing any foreign materials which may exist in the stud holes. Human workers can access to the stud hole for the cleaning of stud holes manually, but the radiation exposure of human workers is increased. Robot is an effective way to work in hazardous area. So we introduced robot for the cleaning of stud holes. Localization of mobile robots is generally based on odometry, but with increased mileage, position errors can be accumulated. In order to eliminate cumulative error and to ensure stability of its driving, laser sensors and new control algorithm were utilized. The distance between the robot and the wall was measured by laser sensors, and the control algorithm was implemented so as to travel the desired trajectory by using the measured values from sensors. The performance of driving and hole sensing were verified through field application, and mobile robot was confirmed to be applicable to the APR 1400 NPP.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2952-2965
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• 2023

Advanced reactors, such as Small Modular Reactors or existing Nuclear Power Plants, often use Field Programmable Gate Array (FPGA) based controllers in new Instrumentation and Control (I&C) system architectures or as an alternative to existing analog-based I&C systems. Compared to CPU-based Programmable Logic Controllers (PLCs), FPGAs offer better overall performance. However, programming functions on FPGAs can be challenging due to the requirement for a hardware description language that does not explicitly support the operation of real numbers. This study aims to implement the Reactor Trip (RT) functions of the existing analog-based Reactor Protection System (RPS) using FPGAs. The RT equations for Overtemperature delta Temperature and Overpower delta Temperature involve dynamic compensators expressed with the Laplace transform variable, 's', which is not directly supported by FPGAs. To address this issue, the trip equations with the Laplace variable in the continuous-time domain are transformed to the discrete-time domain using the Z-transform. Additionally, a new operation based on a relative value for the equation range is introduced for the handling of real numbers in the RT functions. The proposed approach can be utilized for upgrading the existing analog-based RPS as well as digitalizing control systems in advanced reactor systems.

• Proceedings of the KIEE Conference
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• 2007.10a
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• pp.291-292
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• 2007

This paper deals with the design of control cabinet based on safety PLC for Control Rod Control System(CRCS). The CRCS controls the operation of the CRDMs(Control Rod Drive Mechanisms). The CRDM moves the control rods which regulate the reactor power. vertically in the reactor core. The Control Cabinet in CRCS makes and conveys control signals to the power cabinet which provides power to the CRDM. We designed the Control Cabinet, based on POSAFE-Q, safety PLC. The application programs working in PLC can be programmed by pSET(POSAFE-Q Software Engineering Tool), Identified Development Environment.

• Environmental Sciences Bulletin of The Korean Environmental Sciences Society
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• v.1 no.2
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• pp.125-136
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• 1997

Digestion of a municipal wastewater sludge by the anaerobic sequencing batch reactor(ASBR) was investigated to evaluate the performance of the ASBR process at a critical condition of high-solids-content feed. The reactors were operated at an HRT of 10 days with an equivalent loading rate of 0.8-1.5 gVS/L/d at $35^{\circ}C.$ The main conclusions drawn from this study were as follows: 1. Digestion of a municipal wastewater sludge was possible using the ASBR in spite of high concentration of settleable solids in the sludge. The ASBRS with 3- and 4-day cycle period showed almost identical high digestion performances. 2. No adverse effect on digestion stability was observed in the ASBRS in spite of withdrawal and replenishment of $30\%\;or\;40\%$ of liquid contents. A conventional anaerobic digester could be easily converted to the ASBR without any stability problem. 3. Flotation thickening occurred in thicken step of the ASBRS throughout steady state, and floating bed volume at the end of thicken period occupied about $70\%$ of the working volume of the reactor. Efficiency of flotation thickening in the ASBRS could be comparable to that of additional gravity thickening of a completely mixed digester. 4. Solids were accumulated rapidly in the ASBR during start-up period. Solids concentrations in the ASBRS were 2.6 times higher than that in the completely mixed control reactor at steady state. Dehydrogenase activity had a strong correlation with the solids concentration. Dehydrogenase activity of the digested sludge in the ASBR was 2.9 times higher than that of the sludge in the control reactor, and about 25 times higher than that of the subnatant in the ASBR. 5. Remarkable increase in equivalent gas production of $52\%$ was observed at the ASBRS compared with the control reactor in spite of similar Quality of clarified effluent from the ASBRS and control reactor. The increase in gas production from the ASBRS was believed to be combined results of accumulation of microorganisms, higher driving force applied, and additional long-term degradation of organics continuously accumulated.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.99-110
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• 1989

The optimal control of xenon concentration in a nuclear reactor is posed as a linear quadratic regulator problem with state feedback control. Since it is not possible to measure the state variables such as xenon and iodine concentrations directly, implementation of the optimal state feedback control law requires estimation of the unmeasurable state variables. The estimation method used is based on the Luenberger observer. The set of the reactor kinetics equations is a stiff system. This singularly perturbed system arises from the interaction of slow dynamic modes (iodine and xenon concentrations) and fast dynamic modes (neutron flux, fuel and coolant temperatures). The singular perturbation technique is used to overcome this stiffness problem. The observer-based controller of the original system is effected by separate design of the observer and controller of the reduced subsystem and the fast subsystem. In particular, since in the reactor kinetics control problem analyzed in the study the fast mode dies out quickly, we need only design the observer for the reduced slow subsystem. The results of the test problems demonstrated that the state feedback control of the xenon oscillation can be accomplished efficiently and without sacrificing accuracy by using the observer combined with the singular perturbation method.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.57 no.4
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• pp.543-550
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• 2008

This paper proposes an voltage and reactive power control methodology, which is motivated towards implementation in the korea power system. The main voltage control devices are capacitor banks, reactor banks and LTC transformers. Effects of control devices are evaluated by local subsystem's cost computations. This local subsystem is decided by 'Tier' and 'Electrical distance' in the whole system. The control objective at present is to keep the voltage profile within constraints with minimum switching cost. A robust control strategy is proposed to make the control feasible and optimal for a set of power-flow cases that may occur important event from system. This studies conducted for IEEE 39-bus low and high voltage contingency cases indicate that the proposed control methodology is much more effective than PSS/E simulation tool in deciding switching of capacitor and reactor banks.

• Journal of KSNVE
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• v.7 no.5
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• pp.837-845
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• 1997

The dynamic interactions between the reactor vessel and the control element drive mechanisms (CEDMs) of a pressurized water reactor are studied with the simplified mathematical model. The CEDMs are modeled as multiple substructures having different masses and the reactor vessel as a single degree of freedom system. The explicit equation for the frequency responses of the multiple substructure system are presented for the case of harmonic base excitations. The optimum dynamic characteristics of the CEDMs are presented to reduce the dynamic responses of the reactor vessel. The mathematical model and its response equations are verified by finite element analysis for the detailed model of the reactor vessel and the CEDMs for the harmonic base excitations. It is finally shown that the optimal dynamic characteristics of the CEDM presented can be applicable for the aseismic design.