• Korean Chemical Engineering Research
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• v.56 no.3
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• pp.388-403
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• 2018

We investigated design requirements of vent lines for Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We developed design requirements of areas of the rupture disks of the steam generator, a diameter of the gas vent line of the sodium dump tank, a diameter of the gas vent line of the water dump tank, a diameter of the water dump line of the steam generator. With the design requirements, we calculated the time to vent fluid inside the steam generator and analyzed the transient pressure behavior, also evaluated the close pressure value of the isolation valve of the water dump line. Our results are expected to be used as basis information to design Sodium-Water Reaction Pressure Relief System of Prototype Generation IV Sodium-Cooled Fast Reactor.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.10a
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• pp.335-336
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• 2010

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1251-1259
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• 2013

In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section III Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.

• Korean Chemical Engineering Research
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• v.55 no.2
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• pp.170-179
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• 2017

We investigated design requirements of feed water drain and hydrogen vent systems for the sodium-water reaction pressure relief system (SWRPRS) of the prototype generation IV sodium cooled fast reactor (PGSFR). We evaluated the areas of the gas vent pipe of the water dump tank and the length of the water drain pipe of the steam generator to rapid drain of the water steam inside the steam generator for the normal and refueling operations, respectively. We also calculated the diameter of the gas vent pipe of the sodium dump tank which met its design pressure.

• Journal of the Korean Society for Nondestructive Testing
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• v.32 no.4
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• pp.439-447
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• 2012

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1217-1225
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• 2010

The possibility of a sodium-water reaction occurring in a conventional single-wall-tube steam generator in an SFR is a major problem. To improve the reliability of a steam generator, a double-wall-tube steam generator that can reduce the possibility of the occurrence of a sodium-water reaction is being developed. Current developments are focusing on improving the heat-transfer capability of a double-wall tube; further, the development of a leak-detection method to detect the occurrence of a sodium-water reaction during the reactor operation is also underway. In this study, new concepts, which will solve the above-mentioned problems, have been developed. Accordingly, a double-wall tube has been designed, fabricated, and mechanically tested for the purpose.

• Nuclear industry
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• v.24 no.9 s.259
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• pp.31-38
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• 2004

• JOURNAL OF ELECTRICAL WORLD
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• s.441
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• pp.76-79
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• 2013

• Nuclear industry
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• v.36 no.4
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• pp.70-79
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• 2016