Monte Carlo and nodal neutron physics calculations of the IAEA MTR benchmark using Serpent/DYN3D code system

Marat Margulis, Erez Gilad

Research output: Contribution to journalArticlepeer-review

6 Scopus citations

Abstract

As part of recent efforts to utilize NPPs computational methodologies to safety analysis of research reactors, the Serpent and DYN3D codes were extensively compared with a variety of static and burnup calculations as defined in the IAEA benchmark for 10 MW MTR pool-type reactor. These calculations include unit cell calculations and few group constants generation, unit cell and full core k-eigenvalue and burnup calculations, and full core 3D flux and power distributions. The Serpent code capabilities as a lattice code for MTR plate-type fuel assemblies were evaluated and compared with EPRI-CELL and WIMS-D4 results and reference solutions for full 3D core models were compared with MCNP5 and OpenMC results. The DYN3D nodal diffusion code capabilities in modeling full 3D MTR cores were also evaluated using few group cross sections and assembly discontinuity factors obtained by Serpent unit cell calculations. The DYN3D results were compared with Serpent, MCNP5 and OpenMC.

Original languageEnglish
Pages (from-to)118-133
Number of pages16
JournalProgress in Nuclear Energy
Volume88
DOIs
StatePublished - 1 Apr 2016

Keywords

  • Burnup
  • DYN3D
  • Monte Carlo
  • Reactor physics
  • Research reactor
  • Serpent

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Safety, Risk, Reliability and Quality
  • Energy Engineering and Power Technology
  • Waste Management and Disposal

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