• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Journal of the Korean Vacuum Society
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• v.15 no.3
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• pp.325-330
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• 2006

In order to verify the integrity of the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER), the fabricated Cu/SS mock-up is tested in the JAEA Electron Beam Irradiation Test Stand (JEBIS). To fabricate the Cu/SS mock-up, CuCrZr and 316L authentic stainless steel (SS316L) are used for Cu alloy and steel, respectively The hot isostatic pressing (HIP) is used as a manufacturing method with a $1050^{\circ}C$ and 150 MPa. The high heat flux (HHF) test is performed using an electron beam with a heat flux of $5MW/m^2$ and a cycle of 15-sec on time and 30-sec off time. The temperature measurement in the HHF test shows good agreement with the results obtained from ANSYS code analysis, which is used for determining the HHF test conditions.

• Proceedings of the Korean Vacuum Society Conference
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• 2009.08a
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• pp.271-271
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• 2009

• Proceedings of the Korean Vacuum Society Conference
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• 2006.02a
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• pp.232-232
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• 2006

• Proceedings of the Korean Vacuum Society Conference
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• 2008.08a
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• pp.186-186
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• 2008

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.905-906
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• 2005

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.662-669
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• 2010

To develop the manufacturing methods for the blanket first wall (FW) of the International Thermonuclear Experimental Reactor (ITER) and to verify the integrity of the joint, Be/Cu mockups were fabricated and tested at the KoHLT-1 (Korea Heat Load Test facility), a graphite heater facility located at the Korea Atomic Energy Research Institute (KAERI). Since Be and Cu joining is the focus of the present study, the fabricated mockups had a CuCrZr heat sink joined with three Be tiles as an armor material, unlike the original ITER blanket FW, which has a stainless steel structure and coolant tubes. Hot isostatic pressing (HIP) was carried out at $580^{\circ}C$ and 100 MPa for 2 hours as the method for Be/Cu joining. Three interlayers, namely, $1{\mu}mCr/10{\mu}mCu$, $1{\mu}mTi/0.5{\mu}mCr/10{\mu}mCu$, and $5{\mu}mTi/10{\mu}mCu$ were applied as a coating to the Be tiles by a physical vapor deposition (PVD) method. A shear test was performed with the specimens, which were fabricated by the same methods as those used to fabricate the mockups. The average values were 125 MPa to 180 MPa, and the samples with the $1{\mu}mCr/10{\mu}mCu$ interlayer showed the lowest value. No defect or delamination was found in the joints of the mockups by the developed ultrasonic test using a flat-type probe with a 10 MHz frequency and a 0.25 inch diameter. High heat flux (HHF) tests were performed at $1.0\;MW/m^2$ heat flux for each mockup using the given conditions, and the results were analyzed by ANSYS-CFX code. For the test criteria, an expected fatigue lifetime about 1,000 cycles was obtained by analysis with ANSYS-mechanical code. Mockups using the interlayers of $1{\mu}mTi/0.5{\mu}mCr/10{\mu}mCu$ and $5{\mu}mTi/10{\mu}mCu$ survived up to 1,100 cycles over the required number of cycles. However, one of the Be tiles in the other two mockups using the $1{\mu}mCr/10{\mu}mCu$ interlayer was detached during the screening test, and others were detached by discharge after 862 cycles. The integrity of the joints using the proposed interlayers was proven by the HHF test, but the other interlayer requires more study before it can be used for the joining of Be to Cu. Moreover, it was confirmed that the measured temperatures agreed well with the analysis temperatures, which were used to estimate the lifetime and that the developed facility showed its capability of the long time operation.

• Proceedings of the Korean Vacuum Society Conference
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• 2015.08a
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• pp.255-255
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• 2015

국제핵융합실험로(ITER)의 3대 목표 중 하나는 핵융합로 개발을 위한 삼중수소증식블랑켓 개념을 시험하고 검증하는 것이며, 이를 위해 시험증식블랑켓(TBM, Test Blanket Module) 프로그램을 마련, 각국이 참여할 수 있도록 하고 있다. 한국도 2012년 국가핵융합위원회 결정에 따라, EU, 일본, 중국, 인도와 함께 TBM 프로그램에 참여하고 있으며, 2021년 설치를 목표로 헬륨냉각 고체증식재 개념의 HCCR (Helilum Cooled Ceramic Reflector) TBM을 설계, 개발하고 있다. 한국형 TBM은 총 4개의 서브모듈과 하나의 후벽(Back Manifold, BM) 으로 구성되며, 각 서브모듈은 플라즈마와 대면하는 일차벽(First Wall, FW), 증식재와 증배재, 반사재를 담고 있는 증식영역(Breeding Zong, BZ), 냉각재 매니폴드 및 구조물 역할을 하는 측벽(Side Wall, SW) 등의 기능부품으로 구성되어 있다. 냉각재는 8 MPa, $300-500^{\circ}C$의 고온고압헬륨을 사용하고, Li2SiO4 혹은 Li2TiO4 형태의 Li 세라믹 증식재를 사용하며, 중성자 증배를 위해 Be 증배재 및 흑연 반사재를 사용한다 [1-3]. 2015년 2월 개념설계검토(CDR, Conceptual Design Review)를 위해, TBM-shield를 포함한 TBM-set 설계가 완료되었으며, 열수력, 구조, 지진, 전자기, 복합하중에 대한 평가가 진행되었다. 본 논문에서는 이 중 H/He-phase에 시험될 EM-TBM과 D-T phase에 시험될 INT-TBM에 대한 열수력 성능 결과를 소개하였다[5]. 각각의 열부하 조건은 0.17과 $0.3MW/m^2$이며, 중성자 조사는 D-T phase 에서만 고려되었다. 구조재 및 사용된 기능소재별 온도 요건을 정의하고, 성능해석 결과와 비교하였으며, 이를 통해 모든 온도 요건을 만족함을 최종 확인하였다. 이러한 온도 분포는 열응력 평가를 위해 구조해석 입력자료로 활용되었다.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.11 no.1
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• pp.18-30
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• 1999

A heat transfer analysis for the two-dimensional (2-D) steady state using finite difference method (FDM) is performed to predict the thermal behavior of the primary first-wall (FW) system of fusion reactor under various geometric and thermo-hydraulic conditions, such as the beryllium (Be) armor thickness, pitch of cooling tube, and coolant velocity. The FW consists of authentic steel (type 316 stainless steel solution annealed) for cooling tubes, Cu for cooling tubes embedding material, and Be for a protective armor, based on the International Thermonuclear Experiment Reactor (ITER) report. The present 2-D analysis, the control volume discretized with hybrid grid (rectangular grid and polar grid) and Gauss-Seidel iteration method are adapted to solve the governing equations. In the present study, geometric and thermo-hydraulic parameters are optimized with consideration of several limitations. Consequently, it is suggested that the adequate pitch of cooling tube is 22-32mm, the beryllium armor thickness is 10-12mm, and that the coolant velocity is 4.5m/s-6m/s for $100^{\circ}C$ of inlet coolant temperature. The cooling tube should locate near beryllium armor. But, it would be better for locating the center of Cu wall, considering problems of material and manufacturing. Also, 2-D analysis neglecting the axial temperature distribution of cooling tube is appropriate, regarding the discretization error in axial direction.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.401-422
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• 2005

We review here research and development progress achieved in US Plasma Chamber technology roughly over the last decade. In particular, we focus on two major programs carried out in the US: the APEX project (1998-2003) and the US ITER TBM activities (2003-present). The APEX project grew out of the US fusion program emphasis in the late 1990s on more fundamental science and innovation. APEX was commissioned to investigate novel technology concepts for achieving high power density and high temperature reactor coolants. In particular, the idea of liquid walls and the related research is described here, with some detailed examples of liquid metal and molten salt magnetohydrodynamic and free surface effects on flow control and heat transfer. The ongoing US ITER Test Blanket Module (TBM) program is also described, where the current first wall/blanket concepts being considered are the dual coolant lead lithium concept and the solid breeder helium cooled concepts, both using ferritic steel structures. The research described for these concepts includes both thermofluid MHD issues for the liquid metal coolant in the DCLL, and thermomechanical issues for ceramic breeder packed pebble beds in the solid breeder concept. Finally, future directions for ongoing research in these areas are described.