Uncertainty calculations with openMOC for lattice reactor physics

M. Hashemi, René Reifarth, R. Macian-Juan, Augusto Hernandez Solis

    Research outputpeer-review


    OpenMOC is a deterministic neutron transport code using the method of characteristics (MOC) for full-core 3D reactor physics analysis developed at Massachusetts Institute of Technology. OpenMOC relies on user input for macroscopic cross-sections data and currently does not perform self-shielding of the multi-group macroscopic cross-sections and any burnup calculations. This work aims to add perturbation methods to OpenMOC concerning cross-section uncertainties and lattice perturbations. Typical measurement uncertainties of the thermal neutron flux in commercial LWRs are about 3 to 5% at the positions of measurement (fuel assembly guide tubes). In this work, a total 100 different input files were prepared for OpenMOC based on covariance information from the ENDF/B-VII. The final objective is to estimate the impact that covariance data have in the multi-group flux obtained by OpenMOC at a PWR lattice configuration domain. In the end, this project has successfully demonstrated how the total Monte Carlo method can in principle be used in conjunction with OpenMOC.
    Original languageEnglish
    Title of host publicationProceedings International Conference on Physics of Reactors 2022 (PHYSOR 2022)
    PublisherAmerican Nuclear Society
    Number of pages9
    StatePublished - 23 Jun 2022
    Event2022 - PHYSOR - International Conference on Physics of Reactors: Making Virtual a Reality: Advancements in Reactor Physics to Leap Forward Reactor Operation and Deployment - Sheraton Pittsburgh Hotel, Pittsburgh, PA
    Duration: 15 May 202220 May 2022


    Conference2022 - PHYSOR - International Conference on Physics of Reactors
    Country/TerritoryUnited States
    CityPittsburgh, PA

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