• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.235-241
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• 1995

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor Power Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.544-551
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• 2020

Reactor control is a standout amongst the most vital issues in the nuclear power plant. In this paper, the optimal design of variable structure controller (VSC) based on the lightning search algorithm (LSA) is proposed for a nuclear reactor power system. The LSA is a new optimization algorithm. It is used to find the optimal parameters of the VSC instead of the trial and error method or experts of the designer. The proposed algorithm is used for the tuning of the feedback gains and the sliding equation gains of the VSC to prove a good performance. Furthermore, the parameters of the VSC are tuned by the genetic algorithm (GA). Simulation tests are carried out to verify the performance and robustness of the proposed LSA-based VSC compared with GA-based VSC. The results prove the high performance and the superiority of VSC based on LSA compared with VSC based on GA.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2009-2016
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• 2020

A nuclear power plant is a large complex system with tens of thousands of components. To ensure plant safety, the early and accurate diagnosis of abnormal situations is an important factor. To prevent misdiagnosis, operating procedures provide the anticipated symptoms of abnormal situations. While the more severe emergency situations total less than ten cases and can be diagnosed by dozens of key plant parameters, abnormal situations on the other hand include hundreds of cases and a multitude of parameters that should be considered for diagnosis. The tasks required of operators to select the appropriate operating procedure by monitoring large amounts of information within a limited amount of time can burden operators. This paper aims to develop a system that can, in a short time and with high accuracy, select the appropriate operating procedure and sub-procedure in an abnormal situation. Correspondingly, the proposed model has two levels of prediction to determine the procedure level and the detailed cause of an event. Simulations were conducted to evaluate the developed model, with results demonstrating high levels of performance. The model is expected to reduce the workload of operators in abnormal situations by providing the appropriate procedure to ultimately improve plant safety.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.339-348
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• 2001

NSSS Control Systems automatically mitigate transient conditions and leads to a stable plant condition without operator actions when a transient occurs during normal power operation. In this paper, the function and performance of NSSS control systems were examined and evaluated by comparing the predicted results with the measured data for the selected events. Loss of a Main Feedwater Pump and Load Rejection to House Load Operation events were selected for the evaluation among the transient tests peformed during the Power Ascension Test (PAT) of UCN unit 4. The overall schematic control actions of NSSS control systems can be evaluated easily through the observation of these two typical events. The selected events were analyzed by the KISPAC computer code[l] which had been used in developing the control logic and determining the control setpoints during the plant design. Additionally, the performance of FWCS during low power operation was evaluated. The result of evaluation showed that the NSSS control systems were designed properly and the performance of the NSSS control systems was excellent and also the computer code had a good prediction capability.

• Proceedings of the KIEE Conference
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• 2000.07d
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• pp.2495-2497
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• 2000

Level control in the steam generator of a nuclear power plant is important process. But, the low power operation of nuclear power plant causes nonlinear characteristics and non minimum phase characteristics (swell and shrink), change of delay. So, we can't lead good results with conventional PID controller. Particularly, the design of controller with constraints is necessary. This paper introduces MPC(Model Predictive Control) with constraints and designs a good performance MPC controller in spite of the input constraints and nonlinear characteristics, non-minimum phase characteristics

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.423-428
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• 1995

Simulation models have been developed to predict the overall behavior of the CANDU plant systems during normal operational transients. For real time simulation purpose, simplified thermal hydraulic models are applied with appropriate system control logics, which include primary heat transport system solver with its component models and secondary side system models. The secondary side models are mainly used to provide boundary conditions for primary system calculation and to accomodate plant power control logics. Also, for the effective use of simulation package, hardware oriented basic simulation network has been established with appropriate graphic display system. Through validation with typical plant power maneuvering cases using proven plant performance analysis computer code, the present simulation package shows reasonable capability in the prediction of the dynamic behavior of plant variables during operational transients of CANDU plant, which means that this simulation tool can be utilized as a basic framework for full scope simulation network through further improvements.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.17 no.2
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• pp.102-108
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• 2003

This paper presents a new Data acquisition system that can improve a accuracy and response characteristics by designing a SDAS(Signal Datat Acquisition System). I&C(Instrumentation and Control) system, which it applied for Nuclear Power Plant, affects for safety either directly or indirectly. It should be assessed by the equipment qualification procedure to confirm the functionality during design life before it apply on the nuclear power plant depending on safety classification. This paper proposes data acquisition system that required as an essential measurement equipment in the method and procedure of equipment qualification following of standards and codes of IEEE and Nuclear Regulatory Guide. It provides comparable and experimental studies have been carried out. The presented results from the above investigation show considerably improved accuracy performance in the data acquisition system. This circuit will be able to be used in high performance I&C system.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.888-895
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• 2022

Lately, ocean nuclear power plants have attracted attention as one of diverse uses of nuclear power plants. Because ocean nuclear power plants are movable or transportable, it is necessary to analyze the thermal hydraulics in a moving frame of reference, and computer codes have been developed to predict thermal hydraulics in large moving systems. The purpose of this study is to incorporate a three dimensional dynamic motion model into the SPACE code (Safety and Performance Analysis CodE) so that the code is able to analyze thermal hydraulics in an ocean nuclear power plant. A rotation system that describes three-dimensional rotations about an arbitrary axis was implemented, and modifications were made to the one-dimensional momentum equations to reflect the rectilinear and rotational acceleration effects. To demonstrate the code's ability to solve a problem utilizing a rotational frame of reference, code calculations were conducted on various conceptual problems in the two-dimensional and three-dimensional pipeline loops. In particular, the code results for the three-dimensional pipeline loop with a tilted rotation axis agreed well with the multi-dimensional CFD results.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.4
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• pp.321-329
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• 1998

Structural integrity of nuclear components is important for a safe operation of nuclear power plants. Therefore, nuclear power plants require to perform reliable, periodic inservice inspections. Korea Electric Power Company(KEPCO) operates the entire Korean nuclear power plants. Since nuclear power plant safety and the associated inservice inspection(ISI) are under the plant owner's responsibility, Korea Electric Power Research Institute(KEPRI), the R&D division of KEPCO, has established the ISI NDE Total Support system(TSS) for an efficient performance of ISI tasks, and initiated both key ISI NDE technology development program and traing & qualification system development program for an independent ISI operation. This paper describes details of these programs.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.737-752
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• 2014

As condition-based maintenance (CBM) has risen as a new trend, there has been an active movement to apply information technology for effective implementation of CBM in power plants. This motivation is widespread in operations and maintenance, including monitoring, diagnosis, prognosis, and decision-making on asset management. Thermal efficiency analysis in nuclear power plants (NPPs) is a longstanding concern being updated with new methodologies in an advanced IT environment. It is also a prominent way to differentiate competitiveness in terms of operations and maintenance costs. Although thermal performance tests implemented using industrial codes and standards can provide officially trustworthy results, they are essentially resource-consuming and maybe even a hind-sighted technique rather than a foresighted one, considering their periodicity. Therefore, if more accurate performance monitoring can be achieved using advanced data analysis techniques, we can expect more optimized operations and maintenance. This paper proposes a framework and describes associated methodologies for in-situ thermal performance analysis, which differs from conventional performance monitoring. The methodologies are effective for monitoring, diagnosis, and prognosis in pursuit of CBM. Our enabling techniques cover the intelligent removal of random and systematic errors, deviation detection between a best condition and a currently measured condition, degradation diagnosis using a structured knowledge base, and prognosis for decision-making about maintenance tasks. We also discuss how our new methods can be incorporated with existing performance tests. We provide guidance and directions for developers and end-users interested in in-situ thermal performance management, particularly in NPPs with large steam turbines.