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Feasibility study of fusion breeding blanket concept employing graphite reflector
Cho, Seungyon,Ahn, Mu-Young,Lee, Cheol Woo,Kim, Eung Seon,Park, Yi-Hyun,Lee, Youngmin,Lee, Dong Won Elsevier 2015 Fusion engineering and design Vol.98 No.-
<P><B>Abstract</B></P> <P>To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. </LI> <LI> Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. </LI> <LI> A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. </LI> <LI> Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. </LI> <LI> In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. </LI> </UL> </P>
2SF-3 Development Status of Test Blanket Module (TBM)
( Seungyon Cho ),( Mu-young Ahn ),( Dong Won Lee ) 한국공업화학회 2017 한국공업화학회 연구논문 초록집 Vol.2017 No.1
Nuclear fusion reacts with deuterium and tritium as a fuel, and the generated neutrons react with the lithium element to produce tritium and thermal energy. Here, the bred tritium is recovered and reused as fuel, and at the same time, the multiplied heat is recovered by the coolant and used for electricity production. As one of key technologies for the fusion reactor development, development of test blanket module (TBM) for the verification of the thermal energy production for electric production and for the verification of the tritium breeding capability using fusion energy is underway. Through the TBM development, development and verification of system design code and safety analysis, fabrication and characterization of evaluation structural material (ARAA)/functional material, mockup fabrication and performance verification, cooling system verification test, tritium purge gas characteristic evaluation etc. are on-going. In this study the status of the TBM development is presented.
Investigation of technical gaps between DEMO blanket and HCCR TBM
Cho, Seungyon,Ahn, Mu-Young,Choi, Ji-Hyun,Chun, Young-Bum,Im, Kihak,Jin, Hyung Gon,Kim, Chang Shuk,Kim, Dongjun,Kim, Suk Kwon,Ku, Duck Young,Lee, Cheol Woo,Lee, Dong Won,Lee, Eo Hwak,Lee, Youngmin,Par Elsevier 2018 Fusion engineering and design Vol.136 No.1
<P><B>Abstract</B></P> <P>Korea has own programs toward DEMO and fusion reactors that will require further improved DEMO blanket and energy utilization systems. Primary goals of DEMO blanket development in Korea are to develop and verify the integrated blanket design tools; to develop blanket materials, cooling and tritium fuel cycle technologies; and to develop and evaluate fabrication and joining technologies. The concept of helium-cooled ceramic reflector (HCCR) blanket is adopted to be tested in ITER as Test Blanket Module (TBM). Currently, the design and R&D activities are mainly performed through the ITER TBM program in Korea. It is expected to demonstrate the major objectives of the breeding blanket: extraction of heat from burning plasma, tritium self-sufficiency and its integrity within the whole systems considering safety features. Although TBM program will provide very essential data and experience, there will be a considerable technical gap from DEMO blanket. This paper presents the technical gap in the main technical categories of the breeding blanket based on the experience of the development of HCCR TBM and breeding blanket. It is found that still about 60–70% of the DEMO blanket technology can be achieved through the HCCR TBM depending on the technology maturity level, and some ways to DEMO blanket from HCCR TBM are proposed.</P>
Chemical compatibility between ARAA alloy and lithium meta-titanate breeder material
Cho, Seungyon,Park, Yi-Hyun,Chun, Young-Bum,Min, Kyung-Mi,Ahn, Mu-Young,Park, Soon Chang,Lee, Youngmin Elsevier 2017 Fusion engineering and design Vol.124 No.-
<P><B>Abstract</B></P> <P>Chemical compatibility between reduced activation ferritic-martensitic steel, ARAA, and lithium meta-titanate breeder was investigated under operating conditions; high temperature and helium purge gas including a low concentration of hydrogen. ARAA specimens were embedded inside lithium meta-titanate powder and compacted under a load of 200MPa to form disk-shaped samples. The samples were heated at 550°C for up to 1000h in helium with up to 1% hydrogen atmosphere to simulate a fusion breeding blanket environment. The surface of the ARAA specimen was chemically reacted with the lithium meta-titanate breeder to form an oxide layer. The thickness of the oxide layer increased as the heating time increased. The hydrogen in the purge gas made the oxide layer thinner due to the reaction with oxygen in the oxide layer. The growth phenomena of the oxide layer were evaluated.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Chemical compatibility between reduced activation ferritic-martensitic steel, ARAA, and lithium meta-titanate breeder was investigated under operating conditions. </LI> <LI> High temperature and helium purge gas including a low concentration of hydrogen up to 1% to simulate a fusion breeding blanket environment. </LI> <LI> The surface of the ARAA specimen was chemically reacted with the lithium meta-titanate breeder to form an oxide layer. </LI> <LI> The thickness of the oxide layer increased as the heating time increased. </LI> <LI> The hydrogen in the purge gas made the oxide layer thinner due to the reaction with oxygen in the oxide layer. </LI> </UL> </P>
ITER Storage and Delivery System R&D in Korea
Seungyon Cho,Min Ho Chang,Sei-Hun Yun,HyunGoo Kang,Ki-Jung Jung,Hongsuk Chung,Daeseo Koo,Yongkyu Kim,Jaeeun Lee,Kyu-Min Song,Soon-Hwan Sohn,KwangSin Kim,Duk-Jin Kim IEEE 2010 IEEE transactions on plasma science Vol.38 No.3
<P>Korea is supposed to develop the ITER tritium storage and delivery system (SDS), which is one of the main components of the ITER tritium plant. For successful procurement, there are several ongoing R&D activities in the detailed design phase. Investigation of design parameters of the storage and delivery beds has been performed. Small and large-scale mock-ups of ZrCo beds are used to test the capability of desirable rapid delivery and recovery performance and to establish the pertinent procedure of in-bed calorimetry. An experimental apparatus is prepared to develop the integration and verification technologies for the unit processes of the tritium SDS. The performance test of a tritium-compatible metal bellows pump is examined, and the results show a reasonable agreement with the catalog data of the pump. A tritium storage and delivery bed simulator has been developed to simulate various bed operation scenarios under normal and abnormal conditions. A prototype of the SDS simulator is fabricated, and the bed operation scenario generation program to be applied to this simulator is developed. The design requirement of the tritium loading station (TLS) calorimeter is prepared based on a benchmarking mock-up calorimeter, namely, Korea Electric Power Research Institute Tritium Laboratory (KEPTL) calorimeter. Documents for the procurement of the TLS calorimeter will be developed through the experience on the KEPTL calorimeter operation.</P>
조승연(Seungyon Cho),유인근(In-Keun Yu),손수정(Soo Jung Son) 대한기계학회 2006 대한기계학회 춘추학술대회 Vol.2006 No.6
FUSION power offers the potential of an almost limitless source of energy for future generations but it also presents some formidable scientific and engineering challenges. Korea is one of participation countries in ITER project, and study on fusion material is also one of critical issues. A number of materials are necessary to construct the fusion reactor. Especially, structural materials for plasma facing and blanket require reliable properties under severe environments. As a basis on fusion study, material R&D is considerable of interest. Construction of ITER TBM and fabrication data are on progress for next generation nuclear fusion. In this conference, the present state and future prospects for fusion material study will be introduced.