Zobrazeno 1 - 7
of 7
pro vyhledávání: '"Yungan Zhao"'
Publikováno v:
Annals of Nuclear Energy. 110:1098-1106
SiC-based ceramic matrix composites (CMCs) are attractive materials for Very High Temperature Gas Cooled Reactor and advanced reactor. The present work is focused on the Helium permeability measurements and Young’s modulus calculation of SiSiC samp
Publikováno v:
Fusion Engineering and Design. 125:378-383
The present work represents an experimental investigation on the wetting behaviour of lead bismuth eutectic (LBE) on corroded T91. Firstly, the T91 samples were exposed to LBE at saturated oxygen concentration for 400 h, 1200 h and 2000 h respectivel
Autor:
Yeyun Wang, Yu Yu, Jun-Chi Cai, Tengfei Ma, Xiao-Wei Su, Weiqian Zhuo, Yungan Zhao, Fenglei Niu, Yingqiu Hu
Publikováno v:
Nuclear Engineering and Design. 298:14-24
Thermal mixing and stratification, which are common phenomena in the passive containment cooling system, may occur in the containments of the small modular reactors as well during the loss of coolant accidents or the main steam line break accidents.
Publikováno v:
Nuclear Science and Techniques. 27
The underground nuclear power plant (NPP) makes full use of land resources, reduces costs, makes better use of its passive safety, and avoids radioactivity release into the atmosphere in serious nuclear accidents. In this paper, for obtaining compreh
Publikováno v:
Nuclear Science and Techniques. 27
Thermal mixing and stratification phenomena may occur during the loss of a coolant accident or main steam line break accident in the containment of a Passive Containment Cooling System, or in the suppression pools in BWR. However, the present study p
Publikováno v:
Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Radiation Protection and Nuclear Technology Applications.
Permeability of helium gas of Silicon carbide ceramic composites material, which is one of the most important properties in application of SiC composite for advanced reactors, is studied by using a simple, low-cost test system. The test system can no
Publikováno v:
Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Radiation Protection and Nuclear Technology Applications.
Considering the core outlet temperature of high temperature gas-cooled reactor (HTGR) is above 1000 °C and in the accident conditions, a new type of heat-resistant silicon carbide (SiC) ceramic matrix composites needs to be developed. Analysis of th