• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.480-481
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.269-270
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• 2018

• Proceedings of the Korea Society for Simulation Conference
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• 2004.05a
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• pp.101-108
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• 2004

KEPRI has developed an Nuclear Steam Supply System(NSSS) thermal-hydraulics simulation program (called ARTS-KORI), based on the best-estimate system code, RETRAN, as a part of the development project for the KORI unit 1 Nuclear Power Plant Simulator. A number of code modifications, such as simplifications and removing of discontinuities of the physical correlations, were made in order to change the RETRAN code as an nuclear Steam Supply System thermal-hydraulics engine in the simulator. Some simplified models and a backup system were also developed. This paper briefly presents the results of non-integrated standalone test of ARTS-KORI.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.963-970
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• 2018

The U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 needs to perform a comprehensive vibration assessment program for reactor internals during preoperational and startup testing for nuclear power plants and extended power uprate. Although the measurement program is one of the core programs, it is rarely carried out except for a first-of-a-kind or a unique design. This article describes measurement results of vibration and stress for the comprehensive vibration assessment program for an APR-1400 reactor internals. The measurement was performed at an upper guide structure during the pre-core hot functional test of Shin Kori unit 4 reactor internals because the Shin Kori unit 3 and 4 are the first construction project for the APR-1400, and the upper guide structure assembly was to design change compared with the valid prototype. We confirmed that all measured results are within the test acceptance criteria. It means that the structural integrity of the APR-1400 reactor internals was secured for the flow-induced vibration.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.91.1-91
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.205.1-205
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• 2001

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.670-686
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• 1995

This paper provides an improvement on our previous study ［1］ for multi-fault diagnosis in real time in large-scale systems. In the method, fault propagation probability(FPP) and fault propagation time(FPT) in a fuzzy sense are additively used to describe the fault propagation model(FPM) in more practical manner. A modified fault diagnosis procedure is also given. This method is applied for diagnosis of the primary system in the Kori nuclear power plant unit 2 under a transient condition in case of unit value of FPP on each branch of the FPM.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.141-160
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• 2021

Unit 1 of the Kori Nuclear Power Plant (NPP) and Unit 1 of the Wolsong NPP are being prepared for decommissioning; their decommissioning is expected to generate large amounts of intermediate-level, low-level, and very low level Waste. Mixed waste containing both radioactive and hazardous substances is expected to be produced. Nevertheless, laws and regulations, such as the Korean Nuclear Safety Act and Waste Management Act, do not define clear regulatory guidelines for mixed waste. However, the United States has strictly enforced regulations on mixed waste, focusing on the human health and environmental effects of its hazardous components. The U.S. Nuclear Regulatory Commission and the U.S. Department of Energy regulate the radioactive components of mixed waste under the Atomic Energy Act. The U.S. Environmental Protection Agency regulates the hazardous waste component of mixed waste under the Resource Conservation and Recovery Act. In this study, the laws, regulations, and authorities pertaining to mixed waste in the United States are reviewed. Through comparison and analysis with waste management laws and regulations in Korea, a treatment direction for mixed waste is suggested. Such a treatment for mixed waste will increase the efficiency of managing mixed waste when decommissioning NPPs in the near future.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.20-26
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• 2023

The defected baffle former bolts of Kori unit 1 were withdrawn to analyze the cause of damage and gamma-ray measurement is being scheduled. Prior to that, in order to calculate the specific radioactivity value of the baffle former bolt, a radioactivity calculation method considering the actual operation history of the nuclear power plant is introduced and the calculation results are shown. In particular, the radioactivity calculation method considering the operation history is obtained by defining the monthly contribution factor from the actual monthly operation history. As a result, the results considering operation history are 16-28% lower than the general radioactivity calculation results. These results can contribute to establish a reasonable but economical strategy when planning nuclear power plant decommissioning.

• Corrosion Science and Technology
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• v.22 no.3
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• pp.214-219
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• 2023

Considering the case in the United States where most nuclear power plants with an initial design life of 40 years continue to operate until 60 or 80 years after undergoing material soundness evaluation, it is time to plan a more robust long-term operation strategy for nuclear power plants in Korea. There are some reports that SRO/LRO might be formed when Alloy 690 is heat treated for 10,000 hours to 100,000 hours at 360 to 450 ℃. The possibility of LRO formation in Alloy 690 steam generator tubings of Kori nuclear power plant unit 1 (Kori-1) was investigated using existing research papers. The mechanism in which SRO/LRO occurred was also surveyed. Alloy 690 was found to be more likely to cause ordering than Alloy 600 in terms of alloy composition. The ordering could be evaluated through changes in material properties. However, it is difficult to evaluate it from a microstructural point of view. The likelihood of LRO in Alloy 690 of the Kori-1 plant operated at 320 ℃ for 19 years seemed to be low in terms of time and exposure temperature.