• Title/Summary/Keyword: KALIMER-600

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Development of Seismic Analysis Model and Time History Analysis for KALIMER-600 (KALIMER-600 지진해석모델 개발 및 시간이력 지진응답해석)

  • Koo, Gyeong-Hoi;Lee, Jae-Han
    • Journal of the Earthquake Engineering Society of Korea
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    • v.11 no.3 s.55
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    • pp.73-86
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    • 2007
  • In this paper, a simple seismic analysis model of the KALIMER-600 sodium-cooled fast reactor selected to be the candidate of the GEN-IV reactor is developed. By using this model, the seismic time history analysis is carried out to investigate the feasibilities of a seismic isolation design. The developed simple seismic analysis model includes the reactor building, reactor system,, IHTS piping system, steam generator, and seismic isolators. The dynamic characteristics of the simple seismic model are verified with the detailed 3-dimensional finite element analysis for each part of the KALIMER-600 system. By using the developed simple seismic model, the seismic time history analyses for both cases of a seismic isolation and non-isolation design are performed for the artificial time history of a SSE (Safe Shutdown Earthquake) 0.3g. From the comparison of the calculated floor response spectrum, it is verified that the seismically isolated KALIMER-600 reactor building shows a great performance of a seismic isolation and assures a seismic integrity.

DEVELOPMENT OF A SIMPLIFIED MODEL FOR ANALYZING THE PERFORMANCE OF KALIMER-600 COUPLED WITH A SUPERCRITICAL CARBON DIOXIDE BRAYTON ENERGY CONVERSION CYCLE

  • Seong, Seung-Hwan;Lee, Tae-Ho;Kim, Seong-O
    • Nuclear Engineering and Technology
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    • v.41 no.6
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    • pp.785-796
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    • 2009
  • A KALIMER-600 concept which is a type of sodium-cooled fast reactor, has been developed at KAERI. It uses sodium as a primary coolant and is a pool-type reactor to enhance safety. Also, a supercritical carbon dioxide ($CO_2$) Brayton cycle is considered as an alternative to an energy conversion system to eliminate the sodium water reaction and to improve efficiency. In this study, a simplified model for analyzing the thermodynamic performance of the KALIMER-600 coupled with a supercritical $CO_2$ Brayton cycle was developed. To develop the analysis model, a commercial modular modeling system (MMS) was adopted as a base engine, which was developed by nHance Technology in USA. It has a convenient graphical user interface and many component modules to model the plant. A new user library for thermodynamic properties of sodium and supercritical $CO_2$ was developed and attached to the MMS. In addition, some component modules in the MMS were modified to be appropriate for analysis of the KALIMER-600 coupled with the supercritical $CO_2$ cycle. Then, a simplified performance analysis code was developed by modeling the KALIMER-600 plant with the modified MMS. After evaluating the developed code with each component data and a steady state of the plant, a simple power reduction and recovery event was evaluated. The results showed an achievable capability for a performance analysis code. The developed code will be used to develop the operational strategy and some control logics for the operation of the KALIMER-600 with a supercritical $CO_2$ Brayton cycle after further studies of analyzing various operational events.

Water-Simulant Facility Installation for the Sodium-Cooled Fast Reactor KALIMER-600 and Global Flow Measurement (소듐냉각고속로 KALIMER-600 축소 물모의 열유동 가시화 실험장치 구축 및 거시 유동장 특성 측정)

  • Cha, Jae-Eun;Kim, Seong-O
    • 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.

Structural Concept Design of KALIMER-600 Sodium Cooled Fast Reactor (소듐냉각 고속로 KALIMER-600 원자로 구조 개념설계)

  • Lee, Jae-Han;Park, Chang-Gyu;Kim, Jong-Bum;Koo, Gyeong-Hoi
    • 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.

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COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

  • Ha, Kwi-Seok;Jeong, Hae-Yong
    • Nuclear Engineering and Technology
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    • v.44 no.5
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    • pp.535-542
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    • 2012
  • A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

Transient Performance Analysis of the Reactor Pool in KALIMER-600 with an Inertia Moment of a Pump Flywheel (펌프 회전차의 관성모멘트 제공에 의한 KALIMER-600 원자로 풀 과도 성능 분석)

  • Han, Ji-Woong;Eoh, Jae-Hyuk;Lee, Tea-Ho;Kim, Seong-O
    • 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.