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Toward an Open-Source Neutronics Code for Circulating-Fuel Reactors

[+] Author Affiliations
Julien de Troullioud de Lanversin, Alexander Glaser, Malte Göttsche

Princeton University, Princeton, NJ

Paper No. ICONE25-66584, pp. V005T05A019; 8 pages
  • 2017 25th International Conference on Nuclear Engineering
  • Volume 5: Advanced and Next Generation Reactors, Fusion Technology; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues
  • Shanghai, China, July 2–6, 2017
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5783-0
  • Copyright © 2017 by ASME


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 in-house open-source depletion module.

Copyright © 2017 by ASME
Topics: Fuels



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