• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.283.1-283
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.490.1-490.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.156.2-156
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• 1999

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• The Journal of Society for e-Business Studies
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• v.6 no.2
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• pp.13-24
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• 2001

The Real-Time Best-Estimate simulator NPA4K is being developed for providing an efficient nuclear power, KALIMER, simulation environment for transient safety analyses using information visualization. The advanced features of NPA4K simulator are the Once-Through Running Environment, Functionalities of displaying the several X-Y Plot on one system, and Multi-thread Processing, The objective of NPA4K simulator is ta realize the Virtual Reality Environment through Network and Internet technology in Nuclear Power Plants.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.215-220
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• 1998

기능분석은 발전소의 안전성에 관련된 중요한 기능들을 계층적으로 규명하고 기능의 목적, 관련되는 발전소변수, 기능간의 연관관계 등을 파악하고자 수행하며 과는 기능할당 운전원과 자동화시스템간의 역할부담을 결정하기 위해 활용된다. 기존의 발전소에 비하여 자동화시스템의 적극적인 활용이 기대되는 KALIMER MMIS는 운전원과 자동화시스템의 역할에 대한 체계적인 분석이 필요하다. 본 연구에서는 KALIMER MMIS 설계시 수행하는 발전소기능분석 및 기능할당에 대한 수행전학을 제시하고, 기능분석을 지원할 수 있는 지원시스템인 FAI(Function Analysis Interface)를 개발하였다.

• Journal of the Korean Society of Visualization
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• v.9 no.4
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• pp.54-62
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• 2011

KAERI has developed a KALIMER-600 which is a pool-type sodium-cooled fast reactor with a 600MWe electric generation capacity. For a SFR development, one of the main topics is an enhancement of the reactor system safety. Therefore, we have a long-term plan to design the large sodium experimental facility to evaluate the reactor safety and component performance. In order to extrapolate a thermal hydraulic phenomena in a large sodium reactor, the thermal hydraulics phenomena is under investigation in a 1/$10^{th}$ water-simulant facility for the KALIMER-600. In this paper, we shortly described the experimental facility setup and the measurement of the isothermal global flow behavior. For the flow field measurement, the PIV method was used in a transparent Plexiglas reactor vessel model at around $20^{\circ}C$ water condition.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.285-290
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• 2007

KALIMER-600 is a sodium cooled fast reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

• Nuclear Engineering and Technology
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• v.35 no.6
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• pp.537-546
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• 2003

The purpose of this paper is to investigate the buckling characteristics of a conceptually designed KALIMER-150(Korea Advanced LIquid MEtal Reactor, 150MWe) reactor vessel and verify the buckling behavior using the reduced scale cylindrical shell structures. To do this, nonlinear buckling analyses using finite element method and evaluation formulae are carried out. From the results, the KALIMER-150 reactor vessel exhibits a dominant bending buckling mode and is significantly affected by the plastic behavior. The interaction effects with the vertical seismic load cause the lateral buckling load to be slightly decrease. From the results of the buckling experiments using reduced scaled cylindrical shell structures, it is verified that the buckling modes such as pure bending, pure shear, and mixed(bending plus shear) mode clearly appear under a lateral load corresponding to the slenderness ratio of cylinder.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.291-296
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• 1998

The DAta management and Communication NETworks (DACONET), Which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor(KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterised as the distributed real-time system architecture with high performance, Future direction, in which advanced technology is being continually applied to Man-Machine interface System Development of Nuclear Power Plants, will be considered for designing data management and communication networks of KALIMER MMIS