Abstract

Highly uniform oxide dispersion-strengthened materials W–1 wt % Nd2O3 and W–1 wt % CeO2 were successfully fabricated via a novel wet chemical method followed by hydrogen reduction. The powders were consolidated by spark plasma sintering at 1700 °C to suppress grain growth. The samples were characterized by performing field emission scanning electron microscopy and transmission electron microscopy analyses, Vickers microhardness measurements, thermal conductivity, and tensile testing. The oxide particles were dispersed at the tungsten grain boundaries and within the grains. The thermal conductivity of the samples at room temperature exceeded 140 W/m·K. The tensile tests indicated that W–1 wt % CeO2 exhibited a ductile–brittle transition temperature between 500 °C and 550 °C, which was a lower range than that for W–1 wt % Nd2O3. Surface topography and Vickers microhardness analyses were conducted before and after irradiations with 50 eV He ions at a fluence of 1 × 1022 m−2 for 1 h in the large-powder material irradiation experiment system. The grain boundaries of the irradiated area became more evident than that of the unirradiated area for both samples. Irradiation hardening was recognized for the W–1 wt % Nd2O3 and W–1 wt % CeO2 samples.

Highlights

  • The development of fusion technology for electrical power production is one of the major challenges faced in the 21st century [1]

  • The synthesized powder precursors were analyzed via field emission scanning electron microscopy (FE-SEM)

  • Reduction of the precursor in different followed by hydrogen insize the(i) formation of atopowder consisting of particles with particles with a cubic shape, as shown in

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Summary

Introduction

The development of fusion technology for electrical power production is one of the major challenges faced in the 21st century [1]. In the International Thermonuclear Experimental Reactor, the divertor material in the deuterium discharge phase is purely tungsten because of this material’s low tritium retention [3]. Problems such as coarse grain, inherent high ductile–brittle transition temperature (DBTT), poor ductility, low fracture toughness, recrystallization brittleness, and radiation-induced brittleness inhibit the material from meeting the harsh wall loading requirements of future fusion reactors [4,5]. Novel tungsten materials with improved ductility and stability against high temperatures and neutron radiation must be developed. Y2 O3 , La2 O3 , ZrC, TiC, and other similar dispersoids

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