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Development and validation of reactor nuclear design code CORCA-3D

  • An, Ping (Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China) ;
  • Ma, Yongqiang (Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China) ;
  • Xiao, Peng (Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China) ;
  • Guo, Fengchen (Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China) ;
  • Lu, Wei (Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China) ;
  • Chai, Xiaoming (Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China)
  • Received : 2019.04.12
  • Accepted : 2019.05.16
  • Published : 2019.10.25

Abstract

The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.

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