Development of a point-kinetics model in OpenFOAM, integration in GeN-Foam, and validation against FFTF experimental data

Radman, Stefan; Fiorina, Carlo; Song, Ping; Pautz, Andreas

doi:10.1016/j.anucene.2021.108891

Cited by 11 publications

(1 citation statement)

References 25 publications

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“…The resulting tool, called GeN-Foam, is a multiphysics solver for steady-state and transient nuclear reactor analysis [11]. It couples together a multiscale fine/coarse mesh subsolver for thermal hydraulics (one and two phases) [16,17], a subsolver for neutronics (point kinetics [18], diffusion [19], SP3 [20], and SN [21,22]), and a displacement-based subsolver for thermal-mechanics to allow deforming the neutronics mesh for an explicit modeling of the expansion reactivity feedback. The developed solver can easily be maintained and modified thanks to object-oriented programming.…”

Section: The Gen-foam Multiphysics Solvermentioning

confidence: 99%

Investigation and Validation of Unstructured Mesh Methodologies for Modeling Experimental Reactors

et al. 2022

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This paper summarizes a methodology developed at École Polytechnique Fédérale de Lausanne for the neutronic modeling of the CROCUS experimental reactor and proposes solutions to the challenges one may face while modeling a research reactor with a complex geometry. Indeed, the double-lattice configuration of CROCUS makes it difficult to use codes for neutron diffusion and transport relying on a structured mesh description. For this reason, and based on the available in-house competences, we decided to make use of the neutronic capabilities of the GeN-Foam multiphysics solver, which takes advantage of general finite volume methodologies on unstructured meshes to provide sufficient flexibility for the study of unconventional reactor designs. In this work, GeN-Foam is used to build a first SP3 model of CROCUS based on an unstructured mesh to have an explicit modeling of the double lattice and the water gap between the two lattices. Form functions are then used to reconstruct the intra-pin fission rates for validation against measured distributions. We also discuss the limitations of the SP3 approximation of neutron transport in regions with steep neutron flux gradients and the planned future developments.

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