ORIGINAL RESEARCH article
Front. Nucl. Eng.
Sec. Nuclear Reactor Design
This article is part of the Research TopicEnsuring Accuracy: Verification and Validation of High-Fidelity Tools for Advanced Nuclear Reactor SimulationsView all articles
Multi-Step Verification of the Copenhagen Atomics Molten Salt Experiment Radioactive Inventory at the Paul Scherrer Institute using OpenMC, Serpent and EQL0D
Provisionally accepted
Lorenzo Chierici1,2*
Alessio Rossi2- 1Copenhagen Atomics, Copenhagen, Denmark
- 2Paul Scherrer Institut PSI, Villigen, Switzerland
Select one of your emails
You have multiple emails registered with Frontiers:
Notify me on publication
Please enter your email address:
If you already have an account, please login
You don't have a Frontiers account ? You can register here
The Molten Salt Experiment (MSE) is a 30-day 1 MW (thermal) molten salt criticality experiment designed by Copenhagen Atomics, that will be conducted at the Paul Scherrer Institute (PSI) Swiss research center. A detailed analysis of the radioactive inventory available during the operation and after shutdown has been carried out, for future estimates of potential source term release in case of undesired accidents, using the Monte Carlo transport and depletion code OpenMC. To benchmark the radioactive inventory obtained through the OpenMC burn-up calculation (version 0.15.2), code-to-code comparison with the Serpent Monte Carlo and depletion code (v2.2.1), as well as with Serpent coupled to the EQL0D depletion module developed at PSI, were performed. All of the utilized codes support neutron transport on CAD-based geometries and allow continuous material reprocessing during depletion. This feature is essential for modeling the MSE concept, where gaseous fission products are continuously extracted from the circulating fuel salt. For the present verification study, three models of increasing geometric and physical detail were developed: (1) a nested hollow sphere model in constructive solid geometry (CSG), with the scope of aligning basic settings beyond geometry, (2) a geometrically simplified CAD model of the Copenhagen Atomics Onion Core, mainly utilized for testing the integration of CAD models within the above depletion codes and finally (3) the same simplified CAD model, but with the implementation of reactor fluid flows from the inner active core to the outer inactive region, mainly to take into account the effect of out-of-core decay. The results were compared to benchmark the original OpenMC model. Particular attention was given to the end-of-cycle (EOC) isotopic inventory, with detailed comparisons for key radionuclides and total radio-toxicity.
Keywords: Monte Carlo neutron transport, Burnup calculations, Molten salt reactor, OpenMC, serpent, EQL0D, CAD, CSG
Received: 31 Jul 2025; Accepted: 27 Oct 2025.
Copyright: © 2025 Chierici and Rossi. This is an open-access article distributed under the terms of the Creative Commons Attribution License (CC BY). The use, distribution or reproduction in other forums is permitted, provided the original author(s) or licensor are credited and that the original publication in this journal is cited, in accordance with accepted academic practice. No use, distribution or reproduction is permitted which does not comply with these terms.
* Correspondence: Lorenzo Chierici, l.chierici89@gmail.com
Disclaimer: All claims expressed in this article are solely those of the authors and do not necessarily represent those of their affiliated organizations, or those of the publisher, the editors and the reviewers. Any product that may be evaluated in this article or claim that may be made by its manufacturer is not guaranteed or endorsed by the publisher.