• Nuclear Engineering and Technology
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• v.55 no.6
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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.

• Journal of computational fluids engineering
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• v.22 no.1
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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 ${\theta}_3$.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.727-728
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.510.1-510
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• 2001

• Nuclear Engineering and Technology
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• v.52 no.11
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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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• v.53 no.7
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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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• v.4 no.3
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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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• v.29 no.4
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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.

• Vacuum Magazine
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• v.1 no.2
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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.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.02a
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• pp.472-472
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• 2010

The tomography has played a key role in tokamak plasma diagnostics for image reconstruction. The Phillips-Tikhonov (P-T) regularization method was attempted in this work to reconstruct cross-sectional phantom images of the plasma by minimizing the gradient between adjacent pixel data. Recent studies about the comparison of the several tomographic reconstruction methods showed that the P-T method produced more accurate results. We have studied existing Laplacian matrix used in Phillips-Tikhonov regularization method and developed modified Laplacian matrix (Modified L). The comparison of the reconstruction result by the modified L and existing L showed that modified L produced more accurate result. The difference was significantly pronounced when a portion of plasma was reconstructed. These results can be utilized in the Edge Plasma diagnostics; especially in divertor diagnostics on tokamak a large impact is expected. In addition, accurate reconstruction results from received data in only one direction were confirmed through phantom test by using P-T method with modified L. These results can be applied to the tangentially viewing pin-hole camera diagnostics on tokamak.