• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.02a
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• pp.314-314
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• 2010

The design of the blanket and shield play a key role in determining the size of a reactor since it has an impact on the various reactor components. The blanket should produce enough tritium for tritium self-sufficiency and the shield should provide sufficient protection for the superconducting TF coil. Neutronic optimization of the blanket and the shield is necessary, and we coupled the system analysis with a neutronic calculation to account for the interrelation of the blanket and shield with the plasma performance of a reactor system in a self-consistent manner. By using the coupled system analysis code, the operational space for a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil is investigated with an spect ratio in the range of 1.5 - 2.5. The minimum major radius which satisfies all the physics and engineering requirements increases with the magnetic field at the magnetic axis. A required inboard shield thickness is mainly determined by the requirement on the protection of the TF coil against radiation damage. It is shown that to have a fusion power bigger than 3,000 MW in the LAR tokamak with a superconducting TF coil, a major radius bigger than 4.0 m is required.

• Proceedings of the Korean Institute of Intelligent Systems Conference
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• 2006.05a
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• pp.96-98
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• 2006

축산폐수는 축사가 대부분 상수원보다 상류지역에 산재하고 있어 이를 효과적으로 관리하기 어려우나, 연속 회분식 반응기(Sequencing Batch Reactor, SBR)는 장치가 간단하고 경제성이 우수하여 축산폐수처리에서 효율적으로 적용될 수 있다. 본 연구에서는 DO(Dissolved Oxygen)과 ORP(Oxidation-Reduction Potential)을 이용하여 지식기반 고장진단 시스템을 제안하였다. 실시간으로 얻어진 ORP, DO값들을 전처리하여, [ORP], [DO]외에 [ORP DO]합성data와 ORP, DO의 특징백터의 합에서 얻어진 fusion data의 총 4개의 data set을 이용하여 각각에 대한 진단과 분류성능을 검토하였다. 이 값을 이용하여 FCM (fuzzy C-mean) 클러스터링 한 후, K-PCA과 LDA로 차원축소시켜 특징백터를 추출하였다. 그리고 Hamming distance로 test data와 특징백터의 거리를 계산하여 각 class를 F1에서 F8까지 분류하였다. 그 결과 데이터를 그대로 이용하는 것 보다 차분데이터형태로 이용하는 것이 우수했으며 그 중 fusion 데이터의 결과가 다른 것들보다 향상된 결과를 보였다. 그리고 K-PCA와 LDA를 결합한 결과가 다른 방법에 비해 우수한 결과를 보였으며 fusion method를 이용한 최고인식율은 98.02%를 나타내었다.

• 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.

• Nuclear Engineering and Technology
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• v.3 no.1
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• pp.33-36
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• 1971

With the use of the significant structure theory of liquid, the fusion criteria has been successfully explained. To test the theory of fusion, the excess volume upon melting has been calculated for some liquids such as simple liquids and fused salts. The results obtained show good agreements between theory and experiment. The theoretical study on the fused salt may be useful to understand the properties and structure of high temperature liquids in the atomic reactor.

• Proceedings of the Korean Vacuum Society Conference
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• 2009.02a
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• pp.195.1-195.1
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• 2009

• Journal of Welding and Joining
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• v.31 no.5
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• pp.41-46
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• 2013

Different microstructures in the weld zone of a metal structure such as a fusion zone or heat affected zone are formed as compared to the parent material. Thus, the mechanical properties in the weld zone are different from those in the parent material. As the basic data for reliably understanding the structural characteristics of a welded PCHE specimen to be made of Alloy 617, the mechanical properties in the weld zone and parent material for a Alloy 617 plate are measured using an instrumented indentation technique in this study.

• Journal of Welding and Joining
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• v.31 no.2
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• pp.37-42
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• 2013

Different microstructures in the weld zone of a metal structure such as a fusion zone or heat affected zone are formed as compared to the parent material. Thus, the mechanical properties in the weld zone are different from those in the parent material. As the basic data for reliably understanding the structural characteristics of welded PCHE prototype made of SUS316L, the mechanical properties in the weld zone and parent material for a SUS316L plate are measured using an the instrumented indentation technique in this study.

• The Transactions of the Korean Institute of Power Electronics
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• v.20 no.2
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• pp.130-136
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• 2015

This study describes the structure and operation modes of vertical stabilization (VS) converter for international thermonuclear experimental reactor (ITER) and proposes a protection method. ITER VS converter supplies voltage (${\pm}1000V$)/current (${\pm}22.5kA$) to superconducting magnets for plasma current vertical stabilization. A four-quadrant operation must be achieved without zero-current discontinuous section. The operation mode of the VS converter is separated in 12-pulse mode, 6-pulse mode and circulation current mode according to the magnitude of the load current. Protection measures, such as bypass and discharge, are proposed for abnormal conditions, such as over current, over voltage, short circuit, and voltage sag. VS converter output voltage is controlled to satisfy voltage response time within 20 msec. Bypass operation is completed within 60 msec and discharge operation is performed successfully. The feasibility of the proposed control algorithm and protection measure is verified by assembling a real controller and implementing a power system including the VS converter in RTDS for a hardware-in-loop (HIL) facility.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.13-18
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• 2003

Reduced activation ferritic steel, JLF-1 steel (Fe-9Cr-2W-V-Ta), is one of the promising candidate materials for fusion reactor applications. Fracture toughness ($J_IC$) and tensile tests were carried out at room temperature and elevated temperature ($400^{\circ}C$). Two types of CT specimen were prepared to examine the effect of rolling direction on the fracture toughness of JLF-1 steel. Four types of tensile specimen were also prepared to investigate the property by the rolling direction and welding. The Micro Vickers hardness was measured at various distances of a cross section of the TIG joints of JLF-1 steel according to the heating history of each position. Finally, the fracture surface was observed by scanning electron microscopy (SEM).