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Real-time wall conditioning and edge-localized mode (ELM) suppression through real-time boron powder injection into high-confinement mode discharges of the Experimental Advanced Superconducting Tokamak (EAST) have been investigated. The boron powder injection effectively conditioned the plasma-facing components, as shown by the reductions in low-Z and high-Z impurities and the global recycling level in the whole inner vessel, and suppressed the strong magnetohydrodynamic (MHD) activity in the plasma core. The improved wall conditions also resulted in an earlier transition from L- to H-mode. Boron powder injection also effectively suppressed ELM in the lower single null configuration with BT in the favorable direction while leaving the plasma density and stored energy largely unaffected.

Neutron irradiation damage is an important issue in fusion reactor applications. The present work evaluates the neutron-induced displacement damage cross section of the EUROFER97 reduced activation alloy based on the state-of-the-art neutron reaction data libraries JEFF-3.3 and ENDF/B-VIII.0. The issues related to nuclear data and their processing have been considered for evaluating reliable damage cross section for EUROFER97. In addition to the recently published improvements for EUROFER97 compositions, the JEFF-3.3-based damage cross section of 56Fe is significantly increased by treating the questionable recoil energy distributions for continuum inelastic neutron scattering. The final evaluation combines the revised JEFF-3.3 and ENDF/B-VIII.0 calculations and the KIT evaluation in 2015. The athermal recombination-corrected one is also calculated by using the above evaluation and the neutron energy-dependent efficiency deduced from the two KIT evaluations. The current evaluations are higher than the KIT evaluations by 8 % for a D-T fusion neutron and by 5–6 % for an ITER first wall neutron spectrum. The conversion factors from (fast) neutron fluences to DPA levels are included for simple applications.

Functional filler modified fiber reinforced polymer composites are promising structural/radiation shielding materials in application fields of nuclear technology. In this work, MCNP simulating models of epoxy resin matrix composites containing shielding micro-fillers (tungsten oxide and boron carbide) and continuous fiber for gamma radiation shielding were established by Lattice and Universe features in MCNP. The effects of filler size and uniformity of dispersion, types of filler and fiber, stacking order of composite layers on liner attenuation coefficient (μ) and mass attenuation coefficient (μm) were investigated using MCNP simulation. The corresponding shielding mechanisms were discussed. Furthermore, typical kinds of composites were prepared by hot-press process and their μ were tested. The numerical results show that μ decreases more significantly with large filler size and uneven dispersion in low-energy gamma rays. For fibers, μ is proportional to density independent of element type of fiber. Besides, stacking order has negligible effect on μ. All MCNP results are in good agreements with the calculation data from XCOM and the experimental values, which demonstrates the validity of the proposed modeling method. Thus, this simulation model can be used to design and evaluate radiation shielding composites with complex component and microstructure.

Heterophase interfaces have been considered as effective sinks for radiation induced defects and impurities in nuclear materials. However, the quantitative correlation of interface structure and the ability of interface sinking radiation damage is still largely elusive. In this study, we found that the interface angle plays important role in determining interface’s sink ability to radiation damage. Their correlations were investigated by performing He ion irradiation on the Nb film with column grain structure grown on single crystal MgO(100) substrate and we found that the sink efficiency of Nb/MgO interface is related to the interface angle. Considering the dependence of a He atom’s diffusion direction on interface angle, one-dimensional linear differential equation model was corrected by a structure factor to understand the mechanisms for the influence of interface angle on the sink efficiency of Nb/MgO interfaces, and the sink efficiencies of Nb/MgO interface with different interface angles were well predicted by the corrected model. This study could provide novel insights and mechanisms for further understanding the role of interface structure in enhancing radiation tolerance.

To solve the power exhaust handling challenge, meanwhile maintaining acceptable divertor target erosion, the snowflake divertor (SFD) is proposed on HL-2M tokamak. In this work, simulations are carried out by using scrape-off layer plasma simulation code SOLPS-ITER to understand the advantages of SFD on plasma detachment, with emphasis on effects of the magnetic configuration on the neutral and impurity particles transport. Two magnetic configurations, i.e. the standard divertor (SD) and SFD, are chosen for comparison. By increasing the upstream plasma density, the simulation reveals the outer target of SFD reaches plasma detachment at much lower density than that of SD. The reason lays to that SFD can enhance the accumulation of neutral particles and carbon impurity in divertor region, thus increase divertor power radiation and divertor impurity screening capabilities. The target shape combined with the magnetic configuration also plays important role in neutral and impurity accumulation and transport. Further analysis of the roles of deuterium and lower charge-state carbon on the power radiation in divertor region is presented. Finally, the performance of SFD on the divertor plasma with the increasing input power is also simulated and discussed, which is informative for the future fusion device.

To excite stronger magnetic fields, high temperature superconducting (HTS) insert coils are planned for the next generation central solenoid magnet for China fusion engineering test reactor (CFETR). The development of high strength and high toughness jacket has become one of the challenges in the application of high-field cable-in-conduit conductors (CICC). 0.2 % proof stress of jacket should be over 1500 MPa at 4.2 K for HTS conductors, thus 316LN and JK2LB jackets developed by ITER do not meet the requirements. Nitronic 50 (N50) super-austenitic stainless steel material has great optimization potential for the development of jacket. Based on the modified N50 material, China prepared the CICC jacket and has done some R&D work on the N50 jacket. The entire process of ReBCO CICC preparation, including extrusion, bending and straightening, was simulated using the N50 stainless steel jacket. The N50 circle-in-square jackets prepared in China showed a yield strength greater than 1600 MPa and a fracture toughness KIC better than 110 MPa·m1/2 after cold work at 4.2 K. This paper will focus on the preparation, cold work processing and performance test of the N50 jacket. This study will present experimental data and discuss the feasibility of modified N50 as a high-magnetic field jacket material for next-generation fusion reactors.

Tungsten is considered as the most promising material for plasma facing components (PFCs) in the magnetic confinement fusion devices, due to its high melting temperature, high thermal conductivity, low swelling, low tritium retention and low sputtering yield. However, the brittleness, poor machinability and low strength at high-temperatures of tungsten limits its application. Focusing on this issue, various W alloys with enhanced mechanical properties have been developed over recent decades. Among them, the W-ZrC alloys exhibit high strength, high ductility, low ductile-to-brittle transition temperature, good high-temperature stability, excellent resistance to thermal shock and low H/He plasma etching, making it one of the most promising candidate plasma facing materials (PFMs) for the future fusion devices. Therefore, the R&D experience of W-ZrC materials for PFCs, including the design idea, optimization of composition, fabrication technology (from powder metallurgy processing to hot working, from small specimen to large-sized bulk material), microstructure and performance (mechanical properties, thermal conductivity, resistance to thermal shocks and to H/He irradiation, H isotope retention behaviors) were reviewed in this paper. Key words: Plasma-facing materials, Tungsten, ZrC, Microstructure, Mechanical property