Engineering topics

Francois, Juan-Luis

Publications and source records attributed to Francois, Juan-Luis.

3 records

DOE OSTI2017

Fast Reactor Core Analysis Using AZNHEX Neutron Diffusion Code

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 In this work, a sodium-cooled fast reactor core in steady state, with control rods fully withdrawn, is simulated.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.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

AZTRAN and AZKIND verification with a BWR study case

The AZTLAN Platform project is a national initiative led by the National Institute of Nuclear Research which gather the main public institutions of higher education in Mexico: the National Polytechnic Institute, the National Autonomous University of Mexico and the Autonomous Metropolitan University. It consists in the development of a modeling platform for the analysis and design of nuclear reactors. This project aims to modernize, improve and integrate the neutronic, thermohydraulics and thermomechanical codes, developed in the national institutions, in a platform integrated and developed by Mexican experts. The development of AZTLAN Platform is divided technically into four working groups: neutronics, thermohydraulics, coupling and uncertainty analysis, and users group. The present work is the result of the activities of the Users Group focused on the follow-up, testing and use of the different codes (modules) being developed under the AZTLAN project; with the objective of providing feedback and to suggest development priorities to the team of developers, as well as to promote collaboration and synergy between the participating institutions. In particular, in this work, the AZTRAN and the AZKIND modules were verified and tested against the results of well-known codes of the nuclear industry for boiling water reactor (BWR) models. AZTRAN is a 3D transport code which solves numerically the multi-group time independent neutron transport equation in Cartesian geometry, based on the discrete ordinates (SN) method and the RTN-0 nodal scheme. AZKIND is a 3D diffusion code that solves numerically the time dependent neutron diffusion equations in Cartesian geometry, based on a Galerkin Finite Element Method using the RTN-0 polynomial nodal approximation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Verification of AZTRAN v1.2 code with a PWR-MOX benchmark problem

The AZTLAN platform project is an effort focused on developing a modeling Mexican platform to design and analyze nuclear reactors. It is a national initiative lead by the National Institute of Nuclear Research (ININ), in collaboration with the main Mexico's public universities: the National Polytechnic Institute (IPN), the National Autonomous University of Mexico (UNAM) and the Autonomous Metropolitan University (UAM). The AZTRAN code is a module of AZTLAN that solves the 3D neutron transport equation using the discrete ordinates (SN) method and the nodal scheme RTN-0 on steady state and Cartesian geometry. AZTRAN works with cross sections generated externally by other code. It provides a solution to the eigenvalue problem and determine the neutron distribution for a specific medium at desired conditions. On the other hand, it is important to mention that the Working Group 4 (users group) of AZTLAN platform is responsible of verify and validate the codes development, in order to give feedback to the group of developers: neutronics and thermal-hydraulics. In this work, the AZTRAN v1.2 performance was studied for a PWR-MOX lattice, using as example the Yamamoto's benchmark for LWR fuels, comparing AZTRAN v1.2 results again those obtained with the well-known neutronic code SERPENT-2 (v1.27), and some results from an article presented at PHYSOR-2002. Results from AZTRAN for a BWR fuel lattice is presented in a companion paper in this meeting.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗