Fast Reactor Core Analysis Using AZNHEX Neutron Diffusion Code

Galicia-Aragon, Juan; Francois, Juan-Luis; Bastida-Ortiz, Guillermo-Elias; Del-Valle-Gallegos, Edmundo; Gomez-Torres, Armando

DOE OSTI · 23050358

Fast Reactor Core Analysis Using AZNHEX Neutron Diffusion Code

Galicia-Aragon, Juan·Francois, Juan-Luis·Bastida-Ortiz, Guillermo-Elias·Del-Valle-Gallegos, Edmundo·Gomez-Torres, Armando

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Abstract

The AZTLAN platform project emerged as an initiative to rank Mexico at a competitive level in software aspects for the analysis of nuclear reactors, in an effort to achieve autonomy in this area. The AZTLAN platform is conformed by a thermo-hydraulic code (AZTHECA), a neutron transport code (AZTRAN) and two neutron diffusion codes (AZKIND, AZNHEX). AZNHEX is a 3D diffusion code that solves numerically the time dependent neutron diffusion equations in hexagonal z-geometry using the RTN0 nodal finite element method. With this code, it is possible to simulate and analyze light water reactors, such as the Russian VVER, and variety of fast reactors. In this work, a sodium-cooled fast reactor core in steady state, with control rods fully withdrawn, is simulated. The comparison and validation of the values obtained with AZNHEX was done with the results obtained with MCNP6 Monte Carlo code.

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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS 72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS COMPUTER CODES COMPUTERIZED SIMULATION CONTROL ELEMENTS FAST REACTORS FINITE ELEMENT METHOD HYDRAULICS MONTE CARLO METHOD NEUTRON DIFFUSION EQUATION NEUTRON TRANSPORT NEUTRONS REACTOR CORES SODIUM COOLED REACTORS STEADY-STATE CONDITIONS TIME DEPENDENCE VALIDATION WATER COOLED REACTORS WATER MODERATED REACTORS

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BibTeX RIS

Galicia-Aragon, Juan, Francois, Juan-Luis, Bastida-Ortiz, Guillermo-Elias, Del-Valle-Gallegos, Edmundo, Gomez-Torres, Armando. 2017-07-01. Fast Reactor Core Analysis Using AZNHEX Neutron Diffusion Code. https://www.osti.gov/biblio/23050358

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