Deterministic neutron transport and burnup code coupling assessment

Author:

Vivancos Arturo,Miró RafaelORCID,Barrachina TeresaORCID,Bernal Álvaro,Verdú Gumersindo

Funder

European Regional Development Fund

European Commission

Agencia Estatal de Investigación

Ministerio de Ciencia e Innovación

European Social Fund Plus

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference48 articles.

1. Aldama, D. (2014). International Atomic Energy Agency INDC International Nuclear Data Committee Documentation for WIMSD-formatted libraries based on ENDF/B-VII.1 evaluated nuclear data files with extended actinide burn-up chains and cross section data up to 2000 K for fuel materials. Retrieved from http://www-nds.iaea.org/publications.

2. Simulation of the CROCUS REACTOR in the Framework of the H2020 CORTEX (CORE monitoring Techniques and EXperimental validation and demonstration) PROJECT;Barrachina;European Research Reactor Conference,2022

3. Solution of a system of differential equations occurring in the theory of radioactive transformations;Bateman;Proc. Cambridge Philos. Soc.,1910

4. Bernal, Á. (2018). Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses.

5. 2010;Cacuci,2010

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