• Nuclear Engineering and Technology
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• 제41권4호
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• pp.391-402
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• 2009

This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.

• 에너지공학
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• 제24권2호
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• pp.154-166
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• 2015

The in-core instrumentation measures core power distribution and coolant temperature in local regions of the core in pressurized water reactors. The installation types are distinguished by the designs of routing paths that exit either through reactor bottom mounted instrument nozzles or through reactor top mounted instrument nozzles. Although each type has unique advantages, it is generally known that top mounted design is more competitive with respect to emphasizing nuclear safety issues and ability to cope with severe accidents. The international nuclear vendors have provided various types of reactors with top mounted design. Nuclear power reactors in Korea, however, only have been designed to be applicable to the use of bottom mounted design, and it has been pointed out that the capabilities of Korean reactors against severe accidents should be further enhanced. The paper deals with technical issues on reactor internal and external design, in-core instrumentation, support assembly, sealing mechanism with nozzles, handling, and analytical issues in order to establish the ways of development.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.411-419
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• 1994

원자력발전소 중대사고시 예상되는 수소생성과 이에 따른 수소연소로 인한 압력증가로 야기되는 격납용기의 파손화률을 몬테카를로 방법을 통하여 계산하였다. 몬테카를로 계산을 수행하기 위해서는 각각의 입력변수들에 대한 적절한 확률분포함수가 요구되는데, 통계적인 처리를 통하여 구하였다. 고리 2호기에 대한 계산을 수행하였으며, 입력변수들에 대한 민감도 분석도 실시하였다. 고리 2호기에서 수소연소로 인한 격납용기의 파손확률은 60% 이하로 계산되었으며, 민감도 분석결과 SFD가 중요한 인자이긴 하지만 다른 인자들도 무시할 수 없는 영향을 미치고 있음이 밝혀졌다.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1206-1218
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• 2016

This article proposes a highly reliable power and communication system that guarantees the protection of essential instruments in a nuclear power plant under a severe accident. Both power and communication lines are established with not only conventional wired channels, but also the proposed wireless channels for emergency reserve. An inductive power transfer system is selected due to its robust power transfer characteristics under high temperature, high pressure, and highly humid environments with a large amount of scattered debris after a severe accident. A thermal insulation box and a glass-fiber reinforced plastic box are proposed to protect the essential instruments, including vulnerable electronic circuits, from extremely high temperatures of up to $627^{\circ}C$ and pressure of up to 5 bar. The proposed wireless power and communication system is experimentally verified by an inductive power transfer system prototype having a dipole coil structure and prototype Zigbee modules over a 7-m distance, where both the thermal insulation box and the glass-fiber reinforced plastic box are fabricated and tested using a high-temperature chamber. Moreover, an experiment on the effects of a high radiation environment on various electronic devices is conducted based on the radiation test having a maximum accumulated dose of 27 Mrad.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.366-372
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• 2020

Using several accessible published data sets, we analyzed the temporal change of ¹³⁷Cs radioactivity (per unit mass of sample) in marine sediments and investigated the effect of the water content of sediment on the ¹³⁷Cs radioactivity, to understand the behavior of ¹³⁷Cs present in marine environments. The ¹³⁷Cs radioactivity in sediments decreased more slowly in the Baltic Sea (near the Chernobyl nuclear power plant) than in the ocean near the Fukushima Daiichi nuclear power plant (FDNPP). The ¹³⁷Cs radioactivity in the sediment near the FDNPP tended to increase as the water content increased, and the water content decreased at certain sampling sites near the FDNPP for several years. Additionally, the decrease in the water content contributed to 51.2% of the average ¹³⁷Cs radioactivity decrease rate for the same period. Thus, it may be necessary to monitor both the ¹³⁷Cs radioactivity and the water content for marine sediments to track the ¹³⁷Cs that was discharged from the sites of Chernobyl and Fukushima nuclear power plants where severe accidents occurred.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2475-2490
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• 2022

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO₂-Zr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1431-1441
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• 2017

The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.66-73
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• 2015

Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the element-birth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.