Results 71 to 80 of about 340 (168)

Mechanism study of brittle-to-ductile transition temperature in W-0.2 wt% Al2O3 alloy

open access: yesJournal of Materials Research and Technology
W-0.2 wt% Al2O3 alloy plates were fabricated using composite powders prepared by the wet chemical method, followed by cold isostatic pressing, vacuum sintering, and hot-rolling.
Ziwen Zhao   +9 more
doaj   +1 more source

Effect of Neutron Fluence Rate on Irradiation Embrittlement of Low-copper RPV Steel

open access: yesYuanzineng kexue jishu
The reactor pressure vessels (RPVs) of pressurized water reactor (PWR) nuclear power plants currently in operation and under construction in China are all made of low-alloy ferritic steels obtained by quenching and tempering treatments (which meet the ...
WU Yazhen, LI Guoyun, WANG Haidong, HUANG Juan, ZHANG Haisheng, SUN Kai, LEI Yang, ZHU Lini
doaj   +1 more source

Effects of Y2Zr2O7 particles and rolling deformation process on the mechanical properties and heat load behavior of W–Y2Zr2O7 plate

open access: yesJournal of Materials Research and Technology
This paper addresses the significant requirements in China's fusion engineering and investigates the preparation technology and performance characterization of high-performance tungsten-based materials used as PFMs.
Gang Yao   +6 more
doaj   +1 more source

DBTT measurement by use of small punch(SP) test

open access: yesDBTT measurement by use of small punch(SP) test
微小試験片による延性脆性遷移温度(DBTT)の評価法について、スモールパンチ(SP)試験を中心に検討を行った。熱時効、冷間加工、中性子照射を行った21/4Cr-1Mo鋼及び鉄基モデル合金を用い、SP試験によるDBTT評価の課題を検討した結果、SPによるDBTTとシャルピー試験によるDBTTは全般的には良い相関が得られた。しかしながら、粒界脆化型のDBTTの上昇に関しては、SP試験片の破壊破面観察等を組み合わせることにより、脆化のより的確な判断を行っていく必要があることが判った。
openaire  

Influence of Stabilizing Elements on Ductile-Brittle Transition Temperature (DBTT) of 18Cr Ferritic Stainless Steels

open access: yesISIJ International, 2022
Yoshino, Masataka   +3 more
openaire   +2 more sources

Radiation damage in tungsten under high-energy He-ion irradiation

open access: yesCommunications Materials
Tungsten (W) is a primary candidate for plasma-facing components (PFCs) in fusion reactors and hybrid fission-fusion reactors. However, W suffers from severe microstructural damage under Helium (He) ion irradiation in operation conditions.
Yinghang Liu   +10 more
doaj   +1 more source

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