During hypothetical loss-of-coolant accidents (LOCA), zirconium-based nuclear fuel claddings can be exposed to high temperatures in the βZr phase domain and absorb substantial amounts of hydrogen (up about 3000 weight ppm) oxygen 1 %). This paper provides novel data combined effects contents on mechanical behavior (prior-)βZr phase, as a function temperature, upon cooling from temperature range...

Some experimental condit ions which generate t h e d i s loca t ion-enhanced Snoek peak i n Fe-C a l l o y s a r e presented. The r e s u l t s i n d i e a t e t h a t the movement of non-screw d is loca t ion segments i s involved i n t h e pelaxat ion mechanism. The r e s u l t s a r e cons i s ten t with t h e movement of non-screw d i s loca t ions i n a medium with a v i s c o s i t y prop...

Fuel rod analysis program transient (FRAPTRAN) code was used to study the fuel rod performance during a postulated large break loss of coolant accident (LBLOCA) in Maanshan nuclear power plant (NPP). Previous transient results from thermal hydraulic code, TRACE, with the same LBLOCA scenario, were used as input boundary conditions for FRAPTRAN. The simulation results showed that the peak claddi...

This paper describes a multi-scale modeling approach to compute the large-strain deformation of zirconium alloy cladding, which is of interest to the study of fuel rod behaviors in transient and accident conditions. A Visco-Plastic Self-Consistent (VPSC) polycrystal plasticity model representing polycrystals by weighted orientations of single crystals has been implemented as the constitutive la...

A quenching experiment is performed to investigate the heat transfer characteristics and cooling performance of CrAl-coated Zircaloy (Zr) cladding in a water flow. The Zr one accident tolerant fuels for light reactors. uncoated also used this comparison. This simulates reflood fuel rod during loss coolant (LOCA) nuclear power plant. test conditions were determined represent peak temperature, su...