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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.
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]
Copyright of Energies (19961073) is the property of MDPI and its content may not be copied or emailed to multiple sites without the copyright holder's express written permission. Additionally, content may not be used with any artificial intelligence tools or machine learning technologies. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)
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  Data: Preliminary Study of Transient Simulations in the MSRE Primary Loop with Modelica/TRANSFORM.
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  Data: <searchLink fieldCode="AR" term="%22Mao%2C+Chenrui%22">Mao, Chenrui</searchLink><relatesTo>1</relatesTo> (AUTHOR)<br /><searchLink fieldCode="AR" term="%22Guo%2C+Jian%22">Guo, Jian</searchLink><relatesTo>1,2</relatesTo> (AUTHOR)<br /><searchLink fieldCode="AR" term="%22Zou%2C+Yang%22">Zou, Yang</searchLink><relatesTo>1,2</relatesTo> (AUTHOR)<i> zouyang@sinap.ac.cn</i><br /><searchLink fieldCode="AR" term="%22Yan%2C+Rui%22">Yan, Rui</searchLink><relatesTo>1,2</relatesTo> (AUTHOR)
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  Data: <searchLink fieldCode="JN" term="%22Energies+%2819961073%29%22">Energies (19961073)</searchLink>. Jan2026, Vol. 19 Issue 1, p13. 17p.
– Name: Subject
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  Data: <searchLink fieldCode="DE" term="%22Molten+salt+reactors%22">Molten salt reactors</searchLink><br /><searchLink fieldCode="DE" term="%22Transients+%28Dynamics%29%22">Transients (Dynamics)</searchLink><br /><searchLink fieldCode="DE" term="%22Neutrons%22">Neutrons</searchLink><br /><searchLink fieldCode="DE" term="%22Modeling+languages+%28Computer+science%29%22">Modeling languages (Computer science)</searchLink><br /><searchLink fieldCode="DE" term="%22Delayed+neutrons%22">Delayed neutrons</searchLink><br /><searchLink fieldCode="DE" term="%22Thermal+hydraulics%22">Thermal hydraulics</searchLink><br /><searchLink fieldCode="DE" term="%22Xenon%22">Xenon</searchLink>
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  Data: 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]
– Name: AbstractSuppliedCopyright
  Label:
  Group: Ab
  Data: <i>Copyright of Energies (19961073) is the property of MDPI and its content may not be copied or emailed to multiple sites without the copyright holder's express written permission. Additionally, content may not be used with any artificial intelligence tools or machine learning technologies. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract.</i> (Copyright applies to all Abstracts.)
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        Value: 10.3390/en19010013
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      – Code: eng
        Text: English
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        PageCount: 17
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      – SubjectFull: Molten salt reactors
        Type: general
      – SubjectFull: Transients (Dynamics)
        Type: general
      – SubjectFull: Neutrons
        Type: general
      – SubjectFull: Modeling languages (Computer science)
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      – SubjectFull: Delayed neutrons
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      – SubjectFull: Thermal hydraulics
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      – SubjectFull: Xenon
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      – TitleFull: Preliminary Study of Transient Simulations in the MSRE Primary Loop with Modelica/TRANSFORM.
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            NameFull: Mao, Chenrui
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            NameFull: Guo, Jian
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            NameFull: Zou, Yang
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            NameFull: Yan, Rui
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              Text: Jan2026
              Type: published
              Y: 2026
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