• Transactions of Materials Processing
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• v.14 no.6 s.78
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• pp.565-569
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• 2005

Reactor Vessel is one of the most important structural parts of nuclear power plant. It is manufactured by various steel forgings such as shell, head and transition ring. Head forgings have been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced the open die forging process and manufacturing experience of large head forgings which can be used for the reactor vessel of 1,000MW nuclear power plant.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2005.06a
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• pp.131-139
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• 2005

Reactor Vessel is one of the most important structural part of nuclear power plant. It is manufactured by various steel forgings such as shell, head and transition ring. Head forgings has been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced the development trend of the open die forging process and manufacturing experience of large head forgings which canl be used for the reactor vessel of nuclear power plant.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2005.05a
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• pp.397-400
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• 2005

Reactor Vessel is one of the most important structural part of nuclear power plant. It is manufactured by various steel forgings such as shell, head and transition ring. Head forgings has been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced the open die forging process and manufacturing experience of large head forgings which cant be used for the reactor vessel of 1,000MW nuclear power plant.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.3
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• pp.191-198
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• 2014

In this paper, structural integrity evaluation of reactor pressure vessel bottom head without penetration nozzles in core melting accident has been performed. Considering the analysis results of thermal load, weight of molten core debris and internal pressure, thermal load is the most significant factor in reactor vessel bottom head. The failure probability was evaluated according to the established failure criteria and the evaluation showed that the equivalent plastic strain results are lower than critical strain failure criteria. Thermal-structural coupled analyses show that the existence of elastic zone with a lower stress than yield strength is in the middle of bottom head thickness. As a result of analysis, the elastic zone became narrow and moved to the internal wall as the internal pressure increases, and it is evaluated that the structural integrity of reactor vessel is maintained under core melting accident.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.996-1007
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• 2019

We investigated the influence of the aspect ratio (H/R) of the oxide layer on the reactor vessel heating in three-layer configuration. Based on the analogy between heat and mass transfers, we performed mass transfer experiments to achieve high Rayleigh numbers ranging from $6.70{\times}10^{10}$ to $7.84{\times}10^{12}$. Two-dimensional (2-D) semi-circular apparatuses having the internal heat source were used whose surfaces of top, bottom and side simulate the interfaces of the oxide layer with the light metal layer, the heavy metal layer, and the reactor vessel, respectively. Multi-cell flow pattern was identified when the H/R was reduced to 0.47 or less, which promoted the downward heat transfer from the oxide layer and possibly mitigated the focusing effect at the upper metallic layer. The top boundary condition greatly affected the natural convection of the oxide layer due to the presence of secondary flows underneath the cold light metal layer.

• 제어로봇시스템학회:학술대회논문집
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• 2000.10a
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• pp.288-288
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• 2000

Reactor vessel is one of the most important equipment with regard to the safety of nuclear power plant. Thus nuclear regulation requires its periodical examination by certified inspection experts. Conventional reactor inspection machines are obsolete, hard to handle, and very expensive. To solve these problems we developed robotic reactor vessel inspection system which are small, easy to use for inspection, cost effective, and convenient in operation. This paper describes the main features of Main Control System which is one part of robotic inspection equipment we developed.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1658-1668
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• 2019

Nuclear power generates a large portion of the energy used today and plays an important role in energy development. To ensure safe nuclear power generation, it is essential to conduct an accurate analysis of reactor structural integrity. Accordingly, in this study, a methodology for obtaining accurate structural responses to the combined seismic and reactor coolant loads existing prior to the shutdown of a nuclear reactor is proposed. By applying the proposed analysis method to the reactor vessel internals, it is possible to derive the seismic responses considering the influence of the hydraulic loads present during operation for the first time. The validity of the proposed methodology is confirmed in this research by using the finite element method to conduct seismic and hydraulic load analyses of the advanced APR1400 1400 MWe power reactor, one of the commercial reactors. The structural responses to the combined applied loads are obtained using displacement-based and stress-based superposition methods. The safety of the subject nuclear reactor is then confirmed by analyzing the design margin according to the American Society for Mechanical Engineers (ASME) evaluation criteria, demonstrating the promise of the proposed analysis method.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.21 no.12
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• pp.1098-1103
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20, comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a vibration and stress measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals by the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. This paper presents that APR1400 reactor vessel internals have enough structural integrity against the pulsation of reactor coolant pump as the peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.995-1004
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• 2009

The Advanced Power Reactor 1400 (APR1400) is an evolutionary advanced light water reactor (ALWR) based on the Optimized Power Reactor 1000 (OPR1000), which is in operation in Korea. The APR1400 incorporates a variety of engineering improvements and operational experience to enhance safety, economics, and reliability. The advanced design features and improvements of the APR1400 design include a pilot operated safety relief valve (POSRV), a four-train safety injection system with direct vessel injection (DVI), a fluidic device (FD) in the safety injection tank, an in-containment refueling water storage tank (IRWST), an external reactor vessel cooling system, and an integrated head assembly (IHA). Development of the APR1400 started in 1992 and continued for ten years. The APR1400 design received design certification from the Korean nuclear regulatory body in May of2002. Currently, two construction projects for the APR1400 are in progress in Korea.