• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.66-67
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• 2018

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.441-446
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• 1996

A preliminary conceptual design of the DUPIC fuel fabrication plant with production capacity of 400 MTHE/yr is presented. Capital and operating costs are also included. The levelized unit fabrication cost (LUC) for a reference mode was estimated at $509/kgHE, and sensitivity of some variable parameters to this reference was analysed.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.335-340
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• 1998

In this paper a prototype of Requirement Tracking and Verification System(RTVS) for a Distributed Control System was implemented and tested. The RTVS is a software design and verification tool. The main functions required by the RTVS are managing, tracking and verification of the software requirements listed in the documentation of the DCS. The analysis of DCS software design procedures and inter(aces with documents were performed to define the user of the RTVS, and the design requirements for RTVS were developed.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.05a
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• pp.39-40
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• 2017

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.425-431
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• 1996

KT-2 poloidal-field (PF) system is designed to cope the up-down symmetric double-null (DN) and asymmetric single-null (SN) discharges with typical plasma parameters, in which three sets of "design-basis" scenarios - the ohmic heating (OH), the 5MW and the high bootstrap (HIBS) baseline modes - are applied. The power and energy demand for each cases are also deduced. The peak power and the maximum energy requirements for the KT-2 magnet system, incorporating the PF and the toroidal-field (TF) coils, are proven to be 123MW and 1601MJ, respectively when it is driven in DN configuration. The KT-2 PF system is capable of achieving the machine mission of creating a 500kA heated plasma with a current flattop of $\geq$20 seconds.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1059-1070
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• 2016

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.918-919
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1433-1434
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.831-832
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.382-383
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• 2009