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Lee, Hyeong-Yeon;Eoh, Jae-Hyuk;Lee, Yong-Bum
- Transactions of the Korean Society of Mechanical Engineers A
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- v.37 no.5
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- pp.665-671
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- 2013
In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.
https://doi.org/10.3795/KSME-A.2013.37.5.665 인용 PDF KSCI

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.
https://doi.org/10.5407/JKSV.2011.9.4.054 인용 PDF KSCI

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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