Preliminary Study of Transient Simulations in the MSRE Primary Loop with Modelica/TRANSFORM.
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| Title: | Preliminary Study of Transient Simulations in the MSRE Primary Loop with Modelica/TRANSFORM. |
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| Authors: | Mao, Chenrui1 (AUTHOR), Guo, Jian1,2 (AUTHOR), Zou, Yang1,2 (AUTHOR) zouyang@sinap.ac.cn, Yan, Rui1,2 (AUTHOR) |
| Source: | Energies (19961073). Jan2026, Vol. 19 Issue 1, p13. 17p. |
| Subjects: | Molten salt reactors, Transients (Dynamics), Neutrons, Modeling languages (Computer science), Delayed neutrons, Thermal hydraulics, Xenon |
| Abstract: | Compared to conventional solid-fueled reactors, the liquid fuel transport in molten salt reactors (MSRs) leads to a strong coupling between thermal-hydraulics and neutronics. To enable system-level analysis of MSR, this study focuses on the main loop of the Molten Salt Reactor Experiment (MSRE). A system model is developed using the open-source, multiphysics modeling platform Modelica/TRANSFORM. The model is validated against ORNL experimental data under various conditions, including zero-power pump start/stop, natural circulation. In addition, the xenon transport behavior is compared with predictions from a two-region analytical model. Results indicate that the number of discretized core nodes significantly influences the estimation of delayed neutron precursor (DNP) losses due to fuel circulation. The applicability of the ANSI/ANS-5.1 decay heat model, originally developed for light water reactors, is confirmed to be conservative when applied to MSRE conditions. Finally, natural circulation behavior with decay heat transport is further analyzed. [ABSTRACT FROM AUTHOR] |
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| Database: | Engineering Source |
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| Abstract: | Compared to conventional solid-fueled reactors, the liquid fuel transport in molten salt reactors (MSRs) leads to a strong coupling between thermal-hydraulics and neutronics. To enable system-level analysis of MSR, this study focuses on the main loop of the Molten Salt Reactor Experiment (MSRE). A system model is developed using the open-source, multiphysics modeling platform Modelica/TRANSFORM. The model is validated against ORNL experimental data under various conditions, including zero-power pump start/stop, natural circulation. In addition, the xenon transport behavior is compared with predictions from a two-region analytical model. Results indicate that the number of discretized core nodes significantly influences the estimation of delayed neutron precursor (DNP) losses due to fuel circulation. The applicability of the ANSI/ANS-5.1 decay heat model, originally developed for light water reactors, is confirmed to be conservative when applied to MSRE conditions. Finally, natural circulation behavior with decay heat transport is further analyzed. [ABSTRACT FROM AUTHOR] |
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| ISSN: | 19961073 |
| DOI: | 10.3390/en19010013 |
