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Full core 3D simulation of the BEAVRS benchmark with OpenMOC

  • Published:2019-12 Issue: Volume:134 Page:299-304
  • ISSN:0306-4549
  • Container-title:Annals of Nuclear Energy
  • language:en
  • Short-container-title:Annals of Nuclear Energy

Author:

Gunow Geoffrey,Forget BenoitORCID,Smith Kord

Funder

Center for Exascale Simulation of Advanced Reactors

Idaho National Laboratory

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference31 articles.

1. Reactor Agnostic Multi-group Cross Section Generation for Fine-mesh Deterministic Neutron Transport Simulations;Boyd,2017

2. The OpenMOC method of characteristics neutral particle transport code;Boyd;Ann. Nucl. Energy,2014

3. An analysis of condensation errors in multi-group cross section generation for fine-mesh neutron transport calculations;Boyd;Ann. Nucl. Energy,2018

4. CASMO-4, A Fuel Assembly Burnup Program, User’s Manual. Studsvik0SOA-9501;Edenius,1995

5. A linear source approximation scheme for the method of characteristics;Ferrer;Nucl. Sci. Eng.,2015

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