• Journal of the Korean Society for Nondestructive Testing
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• v.5 no.1
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• pp.34-47
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• 1985

The developing country, KOREA where does not possess the natural resources for traditional energy such as oil and gas, so. The nuclear energy is the most single reliable source available for closing the energy gap. For these reason, It is inavoidable to construct the nuclear power plant and to develop technology related nuclear energy. The rate of operation in large nuclear power facilities depends upon the performance of work system through design and construction, and also the applied technology. Especially, it is the most important element that safety and reliability in operation of nuclear power plant. In view of this aspects, Nuclear power plant is performed severe examinations during preservice and inservice inspection. This study provide an automation of analysis for volumetric examination which is required to nuclear power plant components. It is composed as follows: I. Introduction II. Inservice Inspection of Nuclear Power Plant ${\ast}$ General Requirement. ${\ast}$ Principle and Methods of Ultrasonic Test. ${\ast}$ Study of Flaw Evaluation and Design of Classifying Formula for Flaws. III. Design of Automation for Flaw Evaluation. IV. An Example V. Conclusion In this theory, It is classifying the flaws, the formula of classifying flaws and the design of automation that is the main important point. As motioned the above, Owing to such as automatic design, more time could be allocated to practical test than that of evaluation of defects, Protecting against subjective bias tester by himself and miscalculation by dint of various process of computation. For the more, adopting this method would be used to more retaining for many test data and comparative evaluating during successive inspection intervals. Inspite of limitation for testing method and required application to test components, it provide useful application to flow evaluation for volumetric examination. Owing to the characteristics of nuclear power plant that is highly skill intensive industry and has huze system, the more notice should be concentrated as follows. Establishing rational operation plan, developing various technology, and making the newly designed system for undeveloped sector.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.13-19
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• 2015

Steam generator is one of the most important component of nuclear power plant, and it's integrity and reliability are to be assured to high level by pre-service inspection and in-service inspection. To improve the reliability of steam generator heat exchanger tubes and to secure the management of nuclear power plant safely, KHNP CRI recently has developed eddy current testing system for steam generator. KHNP CRI have performed a series of experimental verification and field application to confirm the performance of the developed ECT system in accordance with ASME Code requirements. The ECT system consists of a remote data acquisition unit, an ECT signal acquisition and analysis software, a water chamber robot controller and a probe push-puller. In this paper, we will details of the developed ECT system and the software and their experimental performance. And also we will report the field applying performance and the issues for further steps.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1181-1188
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• 2021

When abnormal events occur in a nuclear power plant, operators must conduct appropriate abnormal operating procedures. It is burdensome though for operators to choose the appropriate procedure considering the numerous main plant parameters and hundreds of alarms that should be judged in a short time. Recently, various research has applied deep-learning algorithms to support this problem by classifying each abnormal condition with high accuracy. Most of these models are trained with simulator data because of a lack of plant data for abnormal states, and as such, developed models may not have tolerance for plant data in actual situations. In this study, two approaches are investigated for a deep-learning model trained with simulator data to overcome the performance degradation caused by noise in actual plant data. First, a preprocessing method using several filters was employed to smooth the test data noise, and second, a data augmentation method was applied to increase the acceptability of the untrained data. Results of this study confirm that the combination of these two approaches can enable high model performance even in the presence of noisy data as in real plants.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.814-826
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• 2023

Sensor faults in nuclear power plant instrumentation have the potential to spread negative effects from wrong signals that can cause an accident misdiagnosis by plant operators. To detect sensor faults and make accurate accident diagnoses, prior studies have developed a supervised learning-based sensor fault detection model and an accident diagnosis model with faulty sensor isolation. Even though the developed neural network models demonstrated satisfactory performance, their diagnosis performance should be reevaluated considering real-time connection. When operating in real-time, the diagnosis model is expected to indiscriminately accept fault data before receiving delayed fault information transferred from the previous fault detection model. The uncertainty of neural networks can also have a significant impact following the sensor fault features. In the present work, a pilot study was conducted to connect two models and observe actual outcomes from a real-time application with an integrated system. While the initial results showed an overall successful diagnosis, some issues were observed. To recover the diagnosis performance degradations, additive logics were applied to minimize the diagnosis failures that were not observed in the previous validations of the separate models. The results of a case study were then analyzed in terms of the real-time diagnosis outputs that plant operators would actually face in an emergency situation.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2576-2583
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• 2024

In this paper, the similarity criterion and dimensionless conversion method combined with the elasticity condition and Hooke's law are used to derive the functional relationship of the maximum effective value of the vibration velocity between the prototype pump and the model pump. The seawater circulation pump of a nuclear power plant is used as the prototype pump, and the model pump is obtained by performance conversion and choosing the appropriate scale, and the vibration state of the model pump under different flow rates is measured and analyzed. The vibration data of the model pump through the function relationship to find out the vibration parameters of the prototype model pump, and compare with the vibration data of the seawater circulation pump in reality. It can be seen that with the increase of flow rate, the maximum effective value of the vibration velocity of both model and prototype decreases and then increases, and the relative error is small, the maximum value is 7.7757%. Therefore, it can be considered that the functional relationship of model pump converted to prototype pump derived in this paper can be used to analyze the vibration of the actual seawater circulation pump of coastal nuclear power plant.

• Proceedings of the KIEE Conference
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• 2001.11c
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• pp.201-203
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• 2001

Safety is emphasized in nuclear power plant system for its characteristic. One of the most important systems to guarantee the safety of nuclear power plant is the PPS(Power protection system). The measurement control system includes PPS, which is similar to the human nervous system, is very important factor not only for the safety but also for the operation. Therefore, developing a technology to level up the performance of measurement system and reliability of the plant will bring cost-effective and safe generation of nuclear power. In this study, I researched composition of digitized PPS and algorithm for cyclic inspection.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.1721-1726
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• 2004

Heat rate is a representative index to estimate the performance of turbine cycle in nuclear power plant. Accuracy of heat rate calculation is dependent on the accuracy of measurement for plant status variables. Uncertainty of heat rate can be modeled using uncertainty propagation model. We developed practical estimation model of heat rate uncertainty using the propagation and regression model. The uncertainty model is used in the performance analysis system developed for the operating nuclear power plant.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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• pp.197-198
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• 2012

If the mechanical development be applied to the Nuclear Power Plant Structures instead of the standard hook development, the problem of overcrowding re-bars in the anchorage zone can be solved and the construction quality of the concrete work will be improved. But there are some problems in applying it to the NPP structures because of the restriction on the re-bar yield strength and diameter. After the performance evaluation test for the mechanical development, we can develop the new design equation of the mechanical development length in order to solve the limitation and apply it to NPP structures.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.861-869
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• 2023

This article proposes a novel approach to measure the performance of Safety-Critical Systems (SCS). Such systems contain multiple processing nodes that communicate with each other is modeled by a Petri nets (PN). The paper uses the PN for the performance evaluation of SCS. A set of ordinary differential equations (ODEs) is derived from the Petri net model that represent the state of the system, and the solutions can be used to measure the system's performance. The proposed method can avoid the state space explosion problem and also introduces new metrics of performance, along with their measurement: deadlock, liveness, stability, boundedness, and steady state. The proposed technique is applied to Shutdown System (SDS) of Nuclear Power Plant (NPP). We obtained 99.887% accuracy of performance measurement, which proves the effectiveness of our approach.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.40-47
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• 2013

The capacity factor of nuclear power plant in Korea is the highest level in the world. However, the integrity assessment of nuclear power plant is depended on foreign country. Especially, most eddy current testing systems for inspecting steam generators in nuclear power plant are currently imported from USA, Canada, and so on. Therefore, the eddy current testing system can react more active and adaptive from economic and managerial standpoint for actual nuclear power plants in Korea is required. In this paper, an eddy current testing system for inspecting steam generators in nuclear power plants is introduced. Frequency generator, analog circuit, analog digital converter circuit, and digital control circuit are composed in eddy current testing system. A benchmarking of acquisition system and acquisition software, eddynet 11i made by Zetec, and modifications are carried out based on the test environment of Korea nuclear power plants. Finally, all eddy current apparatus are integrated to inspect steam generator tubes in nuclear power plants.