• Nuclear industry
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• v.6 no.1 s.35
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• pp.38-39
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• 1986

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2003.06a
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• pp.1458-1462
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• 2003

High temperature and high pressure heavy water flows through the pipes in atomic power plants. The curved parts of pipes are critical parts in that they change the direction of steam flow, and these parts are especially affected by severe wear. Therefore, most pipes in atomic power plants are tested by non-destructive examination by workers who use ultrasonic sensors to measure the wall thickness of pipes. But not only are these pipes located in a very dangerous environment, but the space is also very limited. For the safety of workers, it is necessary to design a device that uses a mobile robot that can inspect curved pipes. This paper presents the design and construction of a small device that can generate the necessary contact forces between ultrasonic sensors and pipe walls in a limited space. And a mobile robot is used in place ortho worker for successful non-destructive examination.

• Journal of the Korean Society of Safety
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• v.31 no.1
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• pp.132-138
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• 2016

The purpose of this study was to develop safety competences, a set of behavioral indicators of each competence and measuring methods of behavioral indicators for preventing human error of nuclear power plants(NPPs). The safety competences and behavioral indicators were derived from the five steps consisted of derivation of preliminary competence items through literature review, content analysis, interview(FGI, BEI), examination of content validity and decision making of final indicators. The results showed that 13 core safety competences and 35 behavior indicators were derived finally. In addition, the methods of measuring safety competences or behavioral indicators such as Behaviorally Anchored Rating Scale (BARS), Behavior Observation Scale (BOS) were developed and suggested.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.797-798
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1321-1322
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• 2005

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.05a
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• pp.387-388
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• 2017

• Corrosion Science and Technology
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• v.20 no.4
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.255-268
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• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.96-104
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• 1995

EPRI URD requires that the reactor be capable of accommodating an unintended CEA drop without initiating a trip and operating at a reduced power with ay single CEA fully inserted. YGN 3 and 4 reactors have 12-finger CEAs, and the CPCS will trip the reactor due to their large reactivities when one of them is dropped at a high power. The ABB-CE reactor power cutback system has been proposed to be used against the 12-Finger CEA drop to avoid the reactor trips. The results of study show that the reactor power cutback can prevent the reactor trips of the 12-Finger CEA drop when the CPCS has enough operating thermal margin (more than 9％ for YGN 3&4 Cycle 1). It is noted, however, that the probability of a 12-Finger CEA drop is very low, less than one per 100 reactor years for YGN 3& and System 80$^{+}$ plants.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.600-606
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• 2019

Studies on the application of the lead rubber bearing (LRB) isolation system to nuclear power plants are being carried out as one of the measures to improve seismic performance. Nuclear power plants with isolation systems require seismic probabilistic safety assessments, for which the seismic fragility of the structures, systems, and components needs be calculated, including for beyond design basis earthquakes. To this end, seismic response analyses are required, where it can be seen that the behaviors of the isolation system components govern the overall seismic response of an isolated plant. The numerical model of the LRB used in these seismic response analyses plays an important role, but in most cases, the extreme performance of the LRB has not been well studied. The current work therefore develops an extreme nonlinear numerical model that can express the seismic response of the LRB for beyond design basis earthquakes. A full-scale LRB was fabricated and dynamically tested with various input conditions, and test results confirmed that the developed numerical model better represents the behavior of the LRB over previous models. Subsequent seismic response analyses of isolated nuclear power plants using the model developed here are expected to provide more accurate results for seismic probabilistic safety assessments.