APOLLO3®: Overview of the new code capabilities for reactor physics analysis

P Mosca, L Bourhrara, A Calloo… - Nuclear Science and …, 2024 - Taylor & Francis
APOLLO3® is a French deterministic reactor code for lattice and core calculations. It has
been developed at CEA (Commissariat à l'Energie Atomique) with the financial and …

Rattlesnake: A MOOSE-based multiphysics multischeme radiation transport application

Y Wang, S Schunert, J Ortensi, V Laboure… - Nuclear …, 2021 - Taylor & Francis
Advanced reactor concepts span the spectrum from heat pipe–cooled microreactors,
through thermal and fast molten-salt reactors, to gas-and salt-cooled pebble bed reactors …

Development of a reduced order model for fuel burnup analysis

C Castagna, M Aufiero, S Lorenzi, G Lomonaco… - Energies, 2020 - mdpi.com
Fuel burnup analysis requires a high computational cost for full core calculations, due to the
amount of the information processed for the total reaction rates in many burnup regions …

Overview of SERMA's Graphical User Interfaces for Lattice Transport Calculations

D Tomatis, F Bidault, A Bruneton, Z Stankovski - Energies, 2022 - mdpi.com
This article presents an overview of the graphical user interfaces (GUIs) developed at
CEA/SERMA (Service d'Études des Réacteurs et de Mathématiques Appliquées) in Saclay …

[HTML][HTML] APOLLO3 homogenization techniques for transport core calculations—application to the ASTRID CFV core

JF Vidal, P Archier, B Faure, V Jouault, JM Palau… - Nuclear Engineering …, 2017 - Elsevier
This paper presents a comparison of homogenization techniques implemented in the
APOLLO3 platform for transport core calculations: standard scalar flux weighting and new …

Overview of multiphysics R&D activities at the CEA/IRESNE institute

C Vaglio-Gaudard, I Ramière, N Seiler - Annals of Nuclear Energy, 2023 - Elsevier
The major mission of the IRESNE (Research Institute on Nuclear Systems for low carbon
Energy production) institute of the DES (Energy Division) at CEA (French Alternative …

Research and development of the neutron-photon coupled transport method in the fast reactor code system MOSASAUR

C Zhao, K Hu, L Wang, B Zhang, L Lou, D Lu… - Progress in Nuclear …, 2024 - Elsevier
In the high-fidelity reactor physics design of fast reactors, the influence of photon effect
needs to be explicitly simulated with the neutron-photon coupled transport method. The …

A New Calculation Strategy for Molten Salt Reactor Neutronic–Thermal-Hydraulic Analysis Implemented with APOLLO3® and TRUST/TrioCFD

N Greiner, F Madiot, Y Gorsse, C Patricot… - Nuclear Science and …, 2023 - Taylor & Francis
Molten salt nuclear reactors (MSRs) constitute a promising technology to produce safe,
reliable, abundant low-carbon energy. To design MSR systems and perform safety analyses …

ANT-MOC: Scalable Neutral Particle Transport Using 3D Method of Characteristics on Multi-GPU Systems

S Li, Z Wang, L Bu, J Wang, Z Xin, S Li… - Proceedings of the …, 2023 - dl.acm.org
The Method Of Characteristic (MOC) to solve the Neutron Transport Equation (NTE) is the
core of full-core simulation for reactors. High resolution is enabled by discretizing the NTE …

A neutron transport characteristics method for 3D axially extruded geometries coupled with a fine group self-shielding environment

S Santandrea, D Sciannandrone… - Nuclear Science and …, 2017 - Taylor & Francis
In this paper we describe some recent developments in the Method of Characteristics (MOC)
for three-dimensional (3D) extruded geometries in the nuclear reactor analysis code …