• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.348-359
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• 1997

This paper reviews cooling aspects of the diverter system in Tokamak fusion devices with primary emphasis on the critical heat flux (CHF) issues for oater-cooled designs. General characteristics of four (4) coolant options for diverter cooling gases, oater, liquid metal, and organic liquid - are discussed first, focusing on the comparison of advantages and disadvantages of those options. Then results of recent studies on the high-heat-flux CHF of water at subcooled high-velocity conditions are reviewed to provide a general idea on the feasibility of the water-cooled diverter concept for future Tokamak fusion reactors. Water is assessed to be the most viable and practical coolant option for diverters of future experimental Tokamaks.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1998.05a
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• pp.229-234
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• 1998

우리나라는 현재 고유의 액체금속로 모델인 KALIMER(Korea Advanced Liquid Metal Reactor)의 개발을 통해 에너지 자원의 이용 효율 증대와 사용후 핵연료 및 초 장수명핵종 소멸처리 문제 등과 같은 에너지 안보와 환경 문제를 동시에 해결하려 하고 있다. 한편 KALIMER의 개바이 그 의미를 갖기 위해서는 고려중인 개념들이 기술적인 측면에서 기능성, 제작성, 안전성, 운용성, 독창성 등이 우수해야 할 뿐만 아니라 경제성이 확보되어야 한다. (중략)

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.558-563
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• 1997

A thermal sizing code, named as HSGSA (Helical coil Steam Generator Sizing Analyzer), for a sodium heated helical coil steam generator is developed for KALIMER (Korea Advanced LIquid MEtal Reactor) design. The theoretical modeling of the shell and tube sides is described and relevant correlations are presented. For assessment of HSGSA, a reference plant design case is compared to the calculational outputs from HSGSA simulation.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.953-968
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• 2009

It is important to accurately predict the temperature and density distributions in large stratified enclosures both for design optimization and accident analysis. Current reactor system analysis codes only provide lumped-volume based models that can give very approximate results. Previous scaling analysis has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by jets modeled using integral techniques. This allows very large reductions in computational effort compared to three-dimensional CFD simulation. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was developed to implement such ideas. This paper summarizes major models for the BMIX++ code, presents the two-plume mixing experiment simulation as one validation example, and describes the codes' application to the liquid salt buffer pool system in the AHTR (Advanced High Temperature Reactor) design. Three design options have been simulated and they exhibit significantly different stratification patterns. One of design options shows the mildest thermal stratification and is identified as the best design option. This application shows that the BMIX++ code has capability to provide the reactor designers with insights to understand complex mixing behavior with mechanistic methods. Similar analysis is possible for liquid-metal cooled reactors.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.240-250
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• 1997

Liquid sodium is widely used as a coolant of LMR(Liquid Metal Reactor) because of its physical and nuclear properties. However, the liquid sodium is very chemically reactive with oxygen and water so that the study on the sodium fire plays an important role in the LMR safety analysis. In this study, a sodium fire model is suggested to analyze the sodium pool fire where both the flame and the reaction products are considered. And also, sodium pool fire analysis computer code, SOPA, is developed. The sensitivity study on the experimental parameters such as the thermal radiation from flame to atmospheric gas, the vessel cooling and the duration of sodium spill was performed. The results showed good agreements with experimental data in the literature.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.803-810
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• 2022

The small modular liquid-metal fast reactor (SMFR) is an important component of advanced nuclear systems. SMFRs exhibit relatively low breeding capability and constraint space for control rod installation. Consequently, control rods are deeply inserted at beginning and are withdrawn gradually to compensate for large burnup reactivity loss in a long lifetime. This paper is committed to investigating the impact of control rod compensation operation on core neutronics characteristics. This paper presents a whole core fine depletion model of long lifetime SMFR using OpenMC and the influence of depletion chains is verified. Three control rod position schemes to simulate the compensation process are compared. The results show that the fine simulation of the control rod compensation process impacts significantly the fuel burnup distribution and absorber consumption. A control rod equivalent position scheme proposed in this work is an optimal option in the trade-off between computation time and accuracy. The control position is crucial for accurate power distribution and void feedback coefficients in SMFRs. The results in this paper also show that the pin level power distribution is important due to the heterogeneous distribution in SMFRs. The fuel burnup distribution at the end of core life impacts the worth of control rods.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11b
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• pp.683-689
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• 1996

본 논문의 목적은 현재 국내에서 개념설계중인 KALIMER(Korea Advanced LIquid MEtal Reactor) 원자로구조물에 대한 면진성능과 내진여유도를 평가하여 이들 성능을 향상시킬 수 있는 주요 설계변경 부위를 검토하는 것이다. 이를 위하여 ANSYS 범용 유한요소해석코드를 이용하여 원자로구조물에 대한 3차원 유한요소해석모델을 작성하고 이로부터 집중질량 스프링으로 이루어진 지진해석모델을 개발하여 지진해석을 수행하였다. KALIMER 원자로 구조물에 대한 내진평가결과 내진능력(Seismic Capability)은 0.35g로 나타났으며 이는 Reactor Vessel Liner, Separation Plate그리고 Support Barrel의 연결부위의 수직 강성을 증가시키는 설계변경을 통하여 크게 향상될 수 있는 것으로 나타났다.

• Journal of the Earthquake Engineering Society of Korea
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• v.3 no.1
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• pp.75-92
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• 1999

In this paper, the seismic analysis model for seismically isolated KALIMER reactor structures is developed and the modal analysis and the seismic time history analysis are carried out for seismic isolation and non-isolation cases. To check the seismic stress limit according to the ASME Code, the equivalent seismic stress analyses are preformed using the 3-D finite element model. From the seismic stress analysis, the seismic margins are calculated for structural members. The limit of seismic load is defined to show that the maximum input acceleration ensures the structural safety for seismic load. In comparison of seismic responses between seismic isolation and non-isolation cases, the seismic isolation design gives significantly reduced acceleration responses and relative displacements between structures. The seismic margin of KALIMER reactor structure is high enough to produce the limit seismic load 0.8g.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3181-3188
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• 2008

The supercritical $CO_2$ Brayton cycle energy conversion system is presented as a promising alternative to the present Rankine cycle. The principal advantage of the S-$CO_2$ gas is a good efficiency at a modest temperature and a compact size of its components. The S-$CO_2$ Brayton cycle coupled to a SFR also excludes the possibilities of a SWR (Sodium-Water Reaction) which is a major safety-related event, so that the safety of a SFR can be improved. KAERI is conducting a feasibility study for the supercritical carbon dioxide (S-$CO_2$) Brayton cycle power conversion system coupled to KALIMER(Korea Advanced LIquid MEtal Reactor). The purpose of this research is to develop S-$CO_2$ Brayton cycle energy conversion systems and evaluate their performance when they are coupled to advanced nuclear reactor concepts of the type under investigation in the Generation IV Nuclear Energy Systems. This paper contains the research overview of the S-$CO_2$ Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.6
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• pp.418-426
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• 2009

The effect of an inertia moment of a pump flywheel on the thermal-hydraulic behaviors of the KALIMER-600(Korea Advanced LIquid MEtal Reactor) reactor pool during an early-phase of a loss of normal heat sink accident was investigated. The thermal-hydraulic analyses for a steady and a transient state were made by using the COMMIX-1AR/P code. In the present analysis a quarter of the reactor geometry was modeled in a cylindrical coordinate system, which includes a quarter of a reactor core and a UIS, a half of a DHX and a pump and a full IHX. In order to evaluate the effects of an inertia moment of the pump flywheel, a coastdown flow whose flow halving time amounts to 3.69 seconds was supplied to a natural circulation flow in the reactor vessel. Thermal-hydraulic behaviors in the reactor vessel were compared to those without the flywheel equipment. The numerical results showed a good agreement with the design values in a steady state. It was found that the inertia moment contributes to an increase in the circulation flow rate during the first 40 seconds, however to a decrease of it there after. It was also found that the flow stagnant region induced by a core exit overcooling decelerated the flow rate. The appearance of the first-peak temperature was delayed by the flow coastdown during the initial stages after a reactor trip.