Validation and application of the 3D neutron transport MPACT code within CASL VERA-CS

Brendan Kochunas, Thomas Downar, Dan Jabaay, Benjamin Collins, Shane Stimpson, Andrew Godfrey, Kang Seog Kim, Jess Gehin, Scott Palmtag, Fausto Franceschini

Research output: Chapter in Book/Report/Conference proceedingConference contributionpeer-review

2 Scopus citations

Abstract

MPACT is a new high-fidelity neutron transport code designed to provide an advanced pin-resolved transport capability for the VERA (Virtual Environment for Reactor Applications) which is the end-user reactor simulation tool being developed for the Consortium for the Advanced Simulation of Light Water Reactors (CASL). The transport methods currently implemented in MPACT are based on the 2-D/l-D method of characteristics (MOC) capabilities to provide a pin-resolved solution of the neutron flux and power throughout the reactor. The cross section resonance treatment utilizes the subgroup method and thermal-hydraulic feedback capability within MPACT includes a simplified T/H feedback model, as well as coupling to the subchannel code CTF and ongoing work to couple MPACT to Computational Fluid Dynamics (CFD). Other features include full core depletion and parallel execution enabling efficient scaling up to 0(105) processors. This capability has been validated for several benchmark problems and several cycles of operating Pressurized Water Reactors (PWR).

Original languageEnglish
Title of host publicationInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
PublisherAmerican Nuclear Society
Pages2931-2944
Number of pages14
ISBN (Electronic)9781510811843
StatePublished - 2015
Event16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015 - Chicago, United States
Duration: Aug 30 2015Sep 4 2015

Publication series

NameInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
Volume4

Conference

Conference16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015
Country/TerritoryUnited States
CityChicago
Period08/30/1509/4/15

Funding

ACKNOWLEDGEMENTS This research was supported by the Consortium for Advanced Simulation of Light Water Reactors (wvw.casl.gov'). an Energy Innovation Hub (http://www.energy.gov/hubs') for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy, and used resources of the Oak Ridge Leadership Computing Facility at the Oak Ridge National Laboratory.

Keywords

  • CASL
  • MOC
  • Neutronics
  • VERA-CS

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