• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.87-92
• /
• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.

• Journal of Radiation Protection and Research
• /
• v.37 no.2
• /
• pp.63-69
• /
• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• Journal of Technology Innovation
• /
• v.22 no.4
• /
• pp.145-164
• /
• 2014

Nuclear fusion is one of the most promising options for generating large amounts of carbon-free energy in the future. Major countries such as China, EU, and Japan have established a national plan for DEMO construction and they are implementing it. Korea has started a nuclear fusion research and development by the KSTAR project started in 1995. There are matured needs for a full-scale research and development initiatives to ensure competition with the major countries for DEMO as well as achieve the final goal to commercialize fusion energy. In this paper, we apply the TRL and AHP methods in order to identify the key technologies to conduct DEMO R&D. We propose the priorities of future R&D on DEMO by deriving a core technology in the field. At first, we review the scientific theory of fusion and trend of progress of DEMO activities in major countries. For previous studies, we review TRL and AHP methods to examine the technology classification system of DEMO and identify key technologies. We apply TRL method to identify readiness level of DEMO technologies and AHP to compensate shortcoming of TRL. The key technologies of DEMO to be secured from a synthesis result of the TRL and AHP are burning plasma, plasma facing material, structural material, high frequency heating, neutral particle beam, safety, plasma diagnostic, and simulation technologies.

• 한국가스학회:학술대회논문집
• /
• 2005.10a
• /
• pp.141-145
• /
• 2005

DME (Di-Methyl Ether) is a new clean fuel and an environmental-friendly energy resource, also is recently increasing with an alternative interest because of the industrial use. DME has been shown to have excellent properties as a diesel fuel giving emission level better than ULEV standard. So it has been attracting considerable as an alternative diesel fuel. In this study, we carried out simulation of separation and purification process of demo plant for 101on per day DME production, which cause the effect that is important in productivity, from operation results of pilot plant for 50kg per day DME production. The liquefied stream, which was separated by gas-liquid separator after DME reactor, includes $CO_2$, DME, Methanol and $H_2O$. We established three distillation columns for separation and purification of the stream. $CO_2$ was extracted from the stream by first distillation column, DME was extracted by second column and Methanol was extracted by third column. We investigated and analyzed the effect which the actual operation variables cause in efficiency of process and optimized process, finally we got the DME of purity $100\%$.

• 한국신재생에너지학회:학술대회논문집
• /
• 2010.11a
• /
• pp.131.2-131.2
• /
• 2010

서부 발전 태안화력발전소에 건설 예정인 IGCC Demo plant의 설계 자료를 근거로 석탄 가스화기의 정상 상태 전산모사를 PRO/II를 사용하여 수행하였다. 석탄을 PRO/II가 받아들일 수 있는 성분으로 바꾼 후 가스화기를 버너와 가스화기 본체의 두 부분으로 나누어 모델링하였다. 버너는 단열조건의 Gibbs Reactor로 모델링하였다. 모사 결과 산소가 완전 소진될 때까지 반응이 진행되는 것을 확인하였다. 가스화기는 char gasification 반응은 kinetic reaction equation으로, gas phase reaction은 equilibrium reactor로 모사하는 알고리듬을 개발 하였으나 PRO/II의 기능에 한계가 있어 간단한 Gibbs Reactor로 모사하였다. 가스화기는 membrane wall에 의하여 냉각되는 것을 고려하여 $1550^{\circ}C$의 균일한 온도에서 반응이 일어나는 것으로 고려하였다. 전산 모사 결과 주요 성분의 조성이 실제 syngas의 조성과 5% 정도 오차가 있는 것으로 나타났다.

• Nuclear Engineering and Technology
• /
• v.52 no.2
• /
• pp.323-327
• /
• 2020

In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% ⁶Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687℃. As regards the tolerable temperature of beryllium (650℃), the design of blanket module is done in the way that beryllium temperature never exceed 600℃. The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.

• Particle and aerosol research
• /
• v.9 no.4
• /
• pp.239-246
• /
• 2013

The purpose of this study is to determine the feasibility of dry-type desulfurization process for actual application to coal-fired power plant. We used actual exhaust gas from Facility Y, Plant #2 to fabricate a demo-scale testing device to attempt to improve the efficiency of desulfurization. A spout-bed circulating dry scrubber convergence system connecting turbo reactor with bag filter was devised, then analyzed for performance characteristics of $SO_2$ removal for Ca/S mole ratio, superficial gas velocity, and ammonia injection, and for secondary reaction characteristics of the non-reactive sorbent at the bag filter. As a result, the installation of spout-bed circulating dry scrubber convergence system showed better economy and efficiency for removing sulfur than the existing wet/semidry-type desulfurization process. In addition, the best efficiency for desulfurization occurred when connected to the bag filter, with differential pressure maintained at 150 $mmH_2O$.

• Korean Chemical Engineering Research
• /
• v.56 no.2
• /
• pp.169-175
• /
• 2018

This study deals with a process for the recovery of hydrogen isotopes from methane ($CQ_4$) and water ($Q_2O$) containing tritium in the nuclear fusion exhaust gas (Q is Hydrogen, Deuterium, Tritium). Steam Methane Reforming and Water Gas Shift reactions are used to convert $CQ_4$ and $Q_2O$ to $Q_2$ and the produced $Q_2$ is recovered by the subsequent Pd membrane. In this study, one circulation loop consisting of catalytic reactor, Pd membrane, and circulation pump was applied to recover H components from $CH_4$ and $H_2O$, one of $CQ_4$ and $Q_2O$. The conversion of $CH_4$ and $H_2O$ was measured by varying the catalytic reaction temperature and the circulating flow rate. $CH_4$ conversion was 99% or more at the catalytic reaction temperature of $650^{\circ}C$ and the circulating flow rate of 2.0 L/min. $H_2O$ conversion was 96% or more at the catalytic reaction temperature of $375^{\circ}C$ and the circulating flow rate of 1.8 L/min. In addition, the amount of $CQ_4$ generated by Korean Demonstration Fusion Power Plant (K-DEMO) in the future was predicted. Then, the treatment process for the $CQ_4$ was proposed and HAZOP (hazard and operability) analysis was conducted to identify the risk factors and operation problems of the process.

• Nuclear Engineering and Technology
• /
• v.41 no.4
• /
• pp.455-476
• /
• 2009

The history of nuclear fusion research in Korea is rather short compared to that of advanced countries. However, since the mid-1990s, at which time the construction of KSTAR was about to commence, fusion research in Korea has been actively carried out in a wide range of areas, from basic plasma physics to fusion reactor design. The flourishing of fusion research partly owes to the fact that industrial technologies in Korea including those related to the nuclear field have been fully matured, with their quality being highly ranked in the world. Successive pivotal programs such as KSTAR and ITER have provided diverse opportunities to address new scientific and technological problems in fusion as well as to draw young researchers into related fields. The frame of the Korean nuclear fusion program is now changing from a small laboratory scale to a large national agenda. Coordinated strategies from different views and a holistic approach are necessary in order to achieve optimal efficiency and effectiveness. Upon this background, the present paper reflects upon the road taken to arrive at this point and looks ahead at the coming future in nuclear fusion research activities in Korea.