Development and validation of a coupled neutron diffusion-thermal hydraulic calculation procedure for fast reactor applications

2020 ◽  
Vol 139 ◽  
pp. 107243
Author(s):  
Xuebei Zhang ◽  
Qin Zeng ◽  
Hongli Chen
Keyword(s):  
Fast Reactor ◽  
Calculation Procedure ◽  
Hydraulic Calculation ◽  
Neutron Diffusion ◽  
Development And Validation

Current Status and Development of Core Calculation Methodology for Sodium Cooled Fast Reactor in Mitsubishi

2010 ◽  
Author(s):  
Kazumi Ikeda ◽  
Hiroyuki Moriwaki ◽  
Wataru Nakazato
Keyword(s):  
Fast Reactor ◽  
Physical Phenomenon ◽  
Hydraulic Calculation ◽  
Current Status ◽  
Monte Carlo Code ◽  
Design Codes ◽  
Calculation Methodology ◽  
Calculation Results ◽  
Dimensional Cell ◽  
Calculation Code

This paper presents current status and development of nuclear calculation methodology for sodium cooled fast reactor in Mitsubishi and two core design codes; an original two dimensional cell/lattice calculation code, PIJHEX and a thermal hydraulic calculation code, MIX-MKII. Mitsubishi Atomic Power Industries, Inc. developed several core design codes for Monju and JOYO and recently this activity continues still further for the next FBR in Japan. It is explained in this paper that the PIJHEX is confirmed to be valid from the comparison of Monte Carlo code and the examination on physical phenomenon in calculation results. Besides, the approach toward design and development of code system in Mitsubishi and MIX-MKII are introduced briefly.

scholarly journals CFD Simulation of a Natural Circulation Helium Loop

2015 ◽  
Vol 23 (1) ◽  
pp. 31-36 ◽  
Author(s):  
Branislav Knížat ◽  
Peter Hlbočan ◽  
Marek Mlkvik
Keyword(s):  
Fast Reactor ◽  
Heat Exchangers ◽  
Cfd Simulation ◽  
Natural Circulation ◽  
Calculation Procedure ◽  
Pressure Losses

Abstract The paper deals with a flow in a closed helium loop serving for cooling of a fast reactor. The flow in pipeline branches of the system is simulated by methods of CFD. The purpose is to find exact values of pressure losses, so that heat exchangers could be successfully designed and so that the power available for a loop drive could be optimally utilized. General approach to the simulation is presented, as well as the calculation procedure and achieved results.

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

2021 ◽  
Author(s):  
Shiying Wei ◽  
Weimin Ma ◽  
Chenglong Wang ◽  
Ronghua Chen ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Evaluation Code ◽  
Transient Thermal ◽  
Development And Validation
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