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Integrated design and performance analysis of the KO HCCR TBM for ITER
Lee, Dong Won,Jin, Hyung Gon,Lee, Eo Hwak,Yoon, Jae Sung,Kim, Suk Kwon,Lee, Cheol Woo,Ahn, Mu-Young,Cho, Seungyon Elsevier 2015 Fusion engineering and design Vol.98 No.-
<P><B>Abstract</B></P> <P>To develop tritium breeding technology for a Fusion Reactor, Korea has participated in the Test Blanket Module (TBM) program in ITER. The He Cooled Ceramic Reflector (HCCR) TBM consists of functional components such as First Wall (FW), Breeding Zone (BZ), Side Wall (SW), and Back Manifold (BM) and it was designed based on the separate analyses for each component in 2012. Based on the each component analysis model, the integrated model is prepared and thermal-hydraulic analysis for the HCCR TBM is performed in the present study. The coolant flow distribution from BM and SW to FW and BZ, and resulted structure temperatures are obtained with the integrated model. It is found that the non-uniform flow rate occurs at FW and BZ and it causes excess of the design limit (550°C) at some region. Based on this integrated model, we will perform the design optimization for obtaining uniform flow distribution for satisfying the design requirements.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Integrated analysis is performed with the conventional CFD code (ANSYS-CFX). </LI> <LI> Overall pressure drop and coolant flow scheme are investigated. </LI> <LI> Manifold design is being performed considering flow distribution. </LI> </UL> </P>
FIRST ATLAS DOMESTIC STANDARD PROBLEM (DSP-01)FOR THE CODE ASSESSMENT
김연식,최기용,강경호,박현식,조석,백원필,김경두,SUK K. SIM,EO-HWAK LEE,SEYUN KIM,김주성,TONG-SOO CHOI,CHEOL-WOO KIM,SUK-HO LEE,SANG-IL LEE,KEO HYOUNG LEE 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.1
KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for AccidentSimulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basisaccidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed andsuccessfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from theATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermalhydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. Forthe first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 wasselected by considering its technical importance and by incorporating comments from participants. Twelve domesticorganizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. ThisATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participantsprior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was alsoused by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison resultsbetween the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations forcode users as well as for developers are suggested.
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>
Preliminary analysis for thermal-fatigue test of HIP joined W and ferritic-martensitic steel mockup
Lee, Dong Won,Kim, Suk Kwon,Shin, Kyu in,Jin, Hyung Gon,Lee, Eo Hwak,Yoon, Jae Sung,Moon, Se Yeon,Hong, Bong Guen Elsevier 2015 Fusion engineering and design Vol.98 No.-
<P><B>Abstract</B></P> <P>For the application to plasma facing component (PFC) in a fusion reactor, joining methods between tungsten (W) and ferritic-martensitic steel (FMS) have been developed and three W/FMS mockups were fabricated by HIP (hot isostatic pressing) joining method. Because the high heat flux test should be performed over the thermal lifetime of the mockup to confirm the integrity of joining technology, test conditions are found by performing a thermal-hydraulic and thermo-mechanical analysis with the conventional codes such as ANSYS-CFX and ANSYS-mechanical, respectively. From the analysis, the heating and the cooling conditions are determined, for 1.0MW/m<SUP>2</SUP> heat flux, to be 30s heating and 30s cooling with given test facility cooling system. And the test cycle number for thermal-fatigue testing is determined to be 2500 cycles because the estimated thermal-lifetime of the mockup is about 2324 cycles from the results of elastic-plastic analysis. The high heat flux test with KoHLT-EB will be performed with these test conditions in the near future.</P> <P><B>Highlights</B></P> <P> <UL> <LI> W/FMS mockup is fabricated to develop the joining method for fusion reactor PFC. </LI> <LI> Thermal-hydraulic analysis is performed with ANSYS-CFX considering the cooling system of KoHLT-EB. </LI> <LI> Heating and cooling time is determined and temperature information for thermo-mechanical analysis is prepared. </LI> <LI> Thermo-mechanical analysis is performed to obtain the stress and strain distribution, and thermal lifetime is estimated using the existing strain to number of cycle to failure curve. </LI> </UL> </P>
Progress of Functional Components Design and Analysis of a Korean HCCR TBM in ITER
Dong-Won Lee,Hyung Gon Jin,Kyu In Shin,Eo Hwak Lee,Suk-Kwon Kim,Jae Sung Yoon,Mu-Young Ahn,Seungyon Cho IEEE 2014 IEEE transactions on plasma science Vol.42 No.5
<P>Korea has developed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) for testing in a ITER, which consists of functional components to distribute the He coolant to each region such as the first wall (FW), breeding zone (BZ), side wall (SW), and back manifold (BM). In this paper, the detailed design of each component is introduced as follows: 1) FW considering cooling under a structural material temperature limit (550 °C); 2) BZ layer for obtaining tritium breeding ratio and cooling with a breeder, reflector, and multiplier pebbles; 3) SW considering the flow distribution to BZ and internal pressure; 4) BM for uniform flow to FW cooling channels; and 5) He purge line in BZ considering a purge gas distribution in BZ. From the performance analysis of each functional component using the CFD code, ANSYS-CFX with the results of nuclear heating from a neutronic analysis, the results show that the design requirements of KO HCCR TBM were satisfied.</P>
Scoping Study on In-Vessel LOCA of a Korean TBS in ITER
Hyung Gon Jin,Dong Won Lee,Eo Hwak Lee,Suk Kwon Kim,Jae Sung Yoon,Moo Yung Ahn,Seung Yon Cho IEEE 2014 IEEE transactions on plasma science Vol.42 No.3
<P>Korea has designed a helium cooled ceramic reflector (HCCR) based test blanket system (TBS) for an ITER. An in-vessel loss of coolant accident is one eight selected reference accidents in the Korean TBS. This accident is initiated by a single or multiple rupture of the test blanket module first wall cooling channels, causing a plasma disruption, and pressurization of the vacuum vessel (VV). In this type of accident, the governing parameters are various, for example, the operating pressure, gas temperature, TBS volume, VV volume, and mass flow rate. Thus, a scoping study is an essential strategy when attempting to determine the proper design specification for a Korean TBS. In this paper, given the preliminary accident analysis results for the current HCCR TBS, a parametric study was performed. For this transient simulation, the Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used.</P>
Mechanical properties of ARAA steel after electron beam welding
Yoon, Jae Sung,Kim, Suk-Kwon,Lee, Eo Hwak,Jin, Hyung Gon,Lee, Dong Won,Cho, Seungyon Elsevier 2017 Fusion engineering and design Vol.124 No.-
<P><B>Abstract</B></P> <P>Korea has designed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) that includes a TBM shield, called a TBM set, that will be tested in ITER. The HCCR TBM is composed of four sub-modules and a back manifold. In addition, each sub-module is composed of a first wall (FW), a breeding box with a seven-layer breeding zone (BZ), and side walls with a cooling path. Korean RAFM steel was developed as a structural material for the HCCR TBM, and advanced reduced activation alloy (ARAA) was selected as the primary candidate from various program alloys. Welding technologies for fabrication of the HCCR TBM were developed using ARAA. Tensile, impact, bend tests were performed after post-weld heat treatment, and hardness, microstructure characteristics were determined before and after post-weld heat treatment to evaluate the welded specimen under the chosen welding conditions.</P> <P><B>Highlights</B></P> <P> <UL> <LI> E-beam welding in ARAA plate. </LI> <LI> Performed mechanical properties of ARAA for fabrication of KO HCCR TBM. </LI> <LI> Evaluation of the tensile, Charpy impact, bend tests after PWHT. </LI> <LI> Evaluation of the hardness and microstructure on the welded area before and after PWHT. </LI> </UL> </P>