• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2139-2146
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• 2023

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 ℃ and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ~268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m² for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.

• 한국전산유체공학회지
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• 제22권1호
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• pp.26-36
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• 2017

A numerical study has been performed to evaluate thermal-hydraulic performance of a finger type cooling module with multiple-jet impingement in a divertor of nuclear fusion reactor. To analyze conjugate heat transfer in both solid and fluid domains, numerical analysis of the flow using three-dimensional Reynolds-averaged Navier-Stokes equations has been performed with shear stress transport turbulence model. The computational domain for the cooling module consisted of a single fluid domain and three solid domains; tile, thimble, and cartridge. The numerical results for the temperature variation on the tile were validated in comparison with experimental data under the same conditions. A parametric study was performed with four geometric parameters, i.e., angles between x-axis and centerlines of hole 1, 2, 3 and 4. The results indicate that the heat transfer rate was increased by 2.7% and 0.7% by the angle ${\theta}_1$ and angle ${\theta}_2$, respectively, and that the pressure drop was decreased by up to 1.8% by the angle ${\theta}_3$.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2307-2316
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• 2024

A melting failure of W monoblock components of the upper divertor outer target in EAST occurred during the plasma campaigns in 2019. The failure characters and microstructure evolution of the failed W monoblock have been well investigated on one string (W436 string). Near the strike point region where heat flux density is highest, macroscopic cracks and severe surface damage such as dimensional change, melting and solidification are visible in several W monoblocks. At the same time, debonding, melting and migration of Cu/CuCrZr cooling tube components introduced fatal damage to the structure and function. The heat-induced microstructure evolution in the rest part has been examined via hardness tests and metallography. From the heat flux surface to the cooling tube, hardness increased gradually and the recrystallized grains could be found in the region with the highest temperature, while recrystallization grains also appear in some W monoblocks near the cooling tube area. The detailed microstructure has been investigated by metallography and EBSD. Such cases in EAST provide experiences on the extreme condition of accidental loss of coolant or higher discharge power in future devices.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.727-728
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.510.1-510
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• 2001

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2630-2637
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• 2020

The structure design of divertor Multi-Functional Maintenance Platform (MFMP) actuated by hydraulic system for China Fusion Engineering Test Reactor (CFETR) was introduced in this paper. The model of MFMP was established according to maintenance requirements. In this paper, Newton-Euler method and the improved virtual work principle were used, the equivalent driving force of each actuator was obtained through the equivalent Jacobian inverse matrix derived from velocity relationship among the components. The accuracy of the model was verified by ADAMS simulation. The stability control of the heavy-duty components driven by hydraulic cylinders based on Newton-Euler method and improved virtual work principle was established.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2251-2260
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• 2021

The motion control of the divertor maintenance system of the China Fusion Engineering Test Reactor (CFETR) was studied in this paper, in which CFETR Multi-Functional Maintenance Platform (MFMP) was simplified as a parallel robot for the convenience of theoretical analysis. In order to design the motion controller of parallel robot, the kinematics analysis of parallel robot was carried out. After that, the dynamic modeling of the hydraulic system was built. As the large variation of heavy payload on MFMP and highly nonlinearity of the system, A Fuzzy-PID controller was built for self-tuning PID controller parameters by using Fuzzy system to achieve better performance. In order to test the feasibility of the Fuzzy-PID controller, the simulation model of the system was built in Simulink. The results have showed that Fuzzy-PID controller can significantly reduce the angular error of the moving platform and provide the stable motion for transferring the divertor.

• Interaction and multiscale mechanics
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• 제4권3호
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• pp.207-217
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• 2011

International efforts have focused recently on the development of tungsten surfaces that can intercept energetic ionized and neutral atoms, and heat fluxes in the divertor region of magnetic fusion confinement devices. The combination of transient heating and local swelling due to implanted helium and hydrogen atoms has been experimentally shown to lead to severe surface and sub-surface damage. We present here a computational model to determine the relationship between the thermo-mechanical loading conditions, and the onset of damage and failure of tungsten surfaces. The model is based on thermo-elasticity, coupled with a grain boundary damage mode that includes contact cohesive elements for grain boundary sliding and fracture. This mechanics model is also coupled with a transient heat conduction model for temperature distributions following rapid thermal pulses. Results of the computational model are compared to experiments on tungsten bombarded with energetic helium and deuterium particle fluxes.

• Nuclear Engineering and Technology
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• 제29권4호
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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.

• 진공이야기
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• 제1권2호
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• pp.9-18
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• 2014

In this article, we present a brief explanation of simulation for magnetic confined fusion plasma. Devices for nuclear fusion experiment become large, complex, and expensive these days, so the simulation can be a valuable tool for understanding and expecting the fusion plasma physics. Research areas presented here are plasma equilibrium and instability, turbulence study, heating and current driving, boundary and divertor area plasma physics, and integrated operation scenario study. Traditionally, many foreign codes have been used because those are verified and stable, however our own MHD and gyrokinetic codes with better performance are under developing recently. While researchers have devoted their effort to make and use a simulation code in individual areas, many ones also endeavor to integrate the simulation codes in different areas for thorough understanding of fusion plasma physics.