Toward an open-source neutronics code for circulating-fuel reactors

Julien De Troullioud De Lanversin, Alexander Glaser, Malte Göttsche

Research output: Chapter in Book/Report/Conference proceedingConference contribution

1 Scopus citations


In circulating fuel reactors, such as the Molten Salt Reactor, the fuel circulates throughout the reactor instead of being immobile as in solid fuel reactors. The vast majority of nuclear simulation codes are primarily designed to simulate solid fuel reactors. Hence, many features unique to circulating fuel reactors, such as fuel injection and removal, cannot be properly modeled with these codes. The work presented here focuses on developing a numerical simulation package that can effectively and accurately model these reactors. This package consists of the coupling of the Monte Carlo particle transport code OpenMC with a modified version of ORIGEN-S, and uses a novel algorithm that calculates the optimal fuel injection and removal schemes for such reactors to achieve certain conditions such as a stable reactivity. We demonstrate our code's accuracy by benchmarking the coupling module with the MCODE coupling code, and by simulating the operation of the ORNL Denatured Molten Salt Reactor using the coupling and fuel injection modules. The resulting fuel injection scheme is in agreement with the original study of that reactor while offering a much finer resolution for the injection scheme over time. This work is part of a broader project to develop an open-source neutronics code for circulating fuel reactors that will couple OpenMC with an inhouse open-source depletion module.

Original languageEnglish (US)
Title of host publicationAdvanced and Next Generation Reactors, Fusion Technology; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues
PublisherAmerican Society of Mechanical Engineers (ASME)
ISBN (Print)9784888982566
StatePublished - 2017
Event2017 25th International Conference on Nuclear Engineering, ICONE 2017 - Shanghai, China
Duration: Jul 2 2017Jul 6 2017

Publication series

NameInternational Conference on Nuclear Engineering, Proceedings, ICONE


Conference2017 25th International Conference on Nuclear Engineering, ICONE 2017

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering


Dive into the research topics of 'Toward an open-source neutronics code for circulating-fuel reactors'. Together they form a unique fingerprint.

Cite this