• 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.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.

• Journal of Korean Society for Quality Management
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• v.18 no.2
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• pp.69-80
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• 1990

This paper highlights the economic effects of nuclear power plants standardization in Korea. The major effects of nuclear power plants standardization appear in the reduction of time-related costs. By using this major economic effects of standardization, an optimal plant mix of electric power until the year 2005 is suggested by means of WASP computer model. And the effects between the standardized case and the non-standardized case is compared.

• Korean Institute of Interior Design Journal
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• v.16 no.5
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• pp.81-88
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• 2007

Energy consumption has been increased world widely, and the energy retain is one of the most important economic alternatives. These tendencies expand the nuclear power plants not only quantitatively but also qualitatively. Despite of the increasing construction of nuclear power plants and related facilities, every system in main control room(MCR) has been designed and administered solely based on the safety-first principles because of the specificity of nuclear industry. However, recent main control rooms started with the concept that the operators' performance could be optimized though the organic interrelation between human, machine, and environments. Now, it has been recognised in the scope of Ergonomics and Space Design which acknowledge our living spaces as Man-Environment Interface and this change connotes the MCR spaces should be special spaces rather than ordinary spaces. This research investigated domestic and foreign nuclear power plants' MCRs to suggest basic alternatives which can be applied to future MCR. With the review of characteristics of MCR, an integration of interior design, lighting and Ergonomics was explored and classified as types. Futhermore, the classification of environmental characteristics within the relationships between human, machine, and environments was developed through the case analysis of nuclear power plants. The results of this study will provide a basis of space design for system environments that the high level of safety and function are extremely important.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.32-40
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• 2018

Since 2000, the frequency of earthquakes beyond the 5.0 magnitude quake has been increasing in the Korean peninsula. For instance, the 5.0-magnitude earthquake in Baekryong-do in 2003 has occurred, and recent earthquake with Gyeongju(2016) and Pohang(2017) measured respectively magnitude of 5.2 and 5.8 on the Richter scale. As results, the public concern and anxiety about earthquakes are increasing, and therefore it is necessarily required for social infrastructure to reinforce seismic design and energy production facilities directly related to the national economy and security. This study represents the analysis of seismic performance evaluation methodology such as Seismic Margin Assessment (SMA), Seismic Probabilistic Risk Assessment (SPRA), High Confidence Low Probability Failure (HCLPF) in nuclear power plants in order to develop optimal seismic performance improvement. Current methodologies to evaluate nuclear power plants are also addressed. Through review of the nuclear structure evaluation past and current trend, it contributes to be the basis for the improvement of evaluation techniques on the next generation of nuclear power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.23-29
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• 2011

The renewed global interest in nuclear power has arisen from the need to reduce greenhouse gas emissions and to provide sufficient electricity for a growing global population before the accident at Fukushima Dai-ichi nuclear power plant in Japan. In spite of the safety issues of nuclear power plants raised by the ongoing Japanese nuclear crisis, many countries with nuclear power plants (NPPs) are still implementing license extensions of 10~20 years, and even consideration is being given to the concept of life-beyond-60, a further period of license extension from 60 to 80 years. To solving the materials aging problem is integral to its success. To evaluate the plant aging phenomena, a lot of background information such as materials and environment of the parts of the reactor and plant systems is needed by the experts. Information on degradation mechanisms is also used. In this paper, a materials degradation evaluation program called OnMDE-SYS (On-line Materials Degradation Evaluation System) is introduced. The developed program provides a variety of information on the materials and stressors as well as operational experience to the experts. It is also anticipated that the experts can perform materials degradation assessment on the web directly by referring to domestic and international information about the degradation of a nuclear power plants through OnMDE-SYS.

• Elastomers and Composites
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• v.55 no.3
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• pp.222-228
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• 2020

Rubber materials, used in nuclear power plants, need high heat-oxidation resistance to curing or cracking under a heat aging environment. This is because they are applied to environments with high temperature, high humidity, and radiation exposure. Nuclear radiation causes additional hardening or degradation, therefore, rubber materials need radiation resistance that satisfies the general and any accidental conditions produced in the power plant. Therefore, in this study, we developed a rubber material with excellent heat and radiation resistance for the diaphragm seal of a nuclear steam generator nozzle dam. The rubber material greatly improved the reliability of the steam generator nozzle dam. In addition, 30 inch and 42 inch diaphragm seals were manufactured using the developed rubber material. A nozzle dam was installed in a nuclear power plant and tested under the same conditions as a steam generator to evaluate safety and reliability. In the future, the performance and safety of diaphragm seals developed through field tests of nuclear power plants will be evaluated and applied to currently operating and new nuclear power plants.

• Journal of the Korean Society of Safety
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• v.34 no.5
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• pp.159-166
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• 2019

The seismic PSA is to probabilistically estimate the potential damage that a large earthquake will cause to a nuclear power plant. It integrates the probabilistic seismic hazard analysis, seismic fragility analysis, and system analysis and is utilized to identify seismic vulnerability and improve seismic capacity of nuclear power plants. Recently, the seismic risk of domestic multi-unit nuclear power plant sites has been evaluated after the Great East Japan Earthquake and Gyeongju Earthquake in Korea. However, while the currently available methods for system analysis can derive basic required results of seismic PSA, they do not provide the detailed results required for the efficient improvement of seismic capacity. Therefore, for in-depth seismic risk evaluation, improved system analysis method for seismic PSA has become necessary. This study develops a system analysis method that is not only suitable for multi-unit seismic PSA but also provides risk information for the seismic capacity improvements. It will also contribute to the enhancement of the safety of nuclear power plants by identifying the seismic vulnerability using the detailed results of seismic PSA. In addition, this system analysis method can be applied to other external event PSAs, such as fire PSA and tsunami PSA, which require similar analysis.

• Nuclear Engineering and Technology
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• v.37 no.3
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• pp.259-264
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• 2005

This paper presents a vital area identification (VAI) method based on the current fault tree analysis (FTA) and probabilistic safety assessment (PSA) techniques for the physical protection of nuclear power plants. A structured framework of a top event prevention set analysis (TEPA) application to the VAI of nuclear power plants is also delineated. One of the important processes for physical protection in a nuclear power plant is VAI that is a process for identifying areas containing nuclear materials, structures, systems or components (SSCs) to be protected from sabotage, which could directly or indirectly lead to core damage and unacceptable radiological consequences. A software VIP (Vital area Identification Package based on the PSA method) is being developed by KAERI for the VAI of nuclear power plants. Furthermore, the KAERI fault tree solver FTREX (Fault Tree Reliability Evaluation eXpert) is specialized for the VIP to generate the candidates of the vital areas. FTREX can generate numerous MCSs for a huge fault tree with the lowest truncation limit and all possible prevention sets.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.259-267
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• 2002

NFPA-803 has been referred as the Fire Protection Standard at the Nuclear Power Plants of Pressurized Water Reactor. This Standard has been used as the fire protection regulation, containing prescriptive requirements with deterministic methodology. Recently, with cumulative efforts by the U.S. Nuclear Regulatory Commission and Utilities in America to establish a new Standard, including a quantitative evaluation methodology, NFPA-805, the Performance-Based Standard for FIRE Protection for Light Water Reactor Electric Generating Plants was issued and approved by the American National Standards Institute as an American National Standard with an effective date of February 9, 2001. This paper presents an analysis result from the computer modeling for the fire simulation In addition, it proposes the idea that this kind of analytic method can be available for the facilities design of fire prevention and protection fields, as well as an evaluation for the fire suppression system with a quantitative analysis for the thermal phenomena in fire compartments in Nuclear Power Plants.