Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion
Abstract
Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with other system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advancedmore »
- Authors:
-
[1];
[1];
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Publication Date:
- Research Org.:
- Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
- Sponsoring Org.:
- USDOE
- OSTI Identifier:
- 1454407
- Grant/Contract Number:
- AC05-00OR22725
- Resource Type:
- Accepted Manuscript
- Journal Name:
- Nuclear Technology
- Additional Journal Information:
- Journal Volume: 203; Journal Issue: 1; Journal ID: ISSN 0029-5450
- Publisher:
- Taylor & Francis - formerly American Nuclear Society (ANS)
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 70 PLASMA PHYSICS AND FUSION TECHNOLOGY; Tritium; modeling; simulation
Citation Formats
Rader, Jordan D., Greenwood, Michael Scott, and Humrickhouse, Paul W. Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion. United States: N. p., 2018.
Web. doi:10.1080/00295450.2018.1431505.
Rader, Jordan D., Greenwood, Michael Scott, & Humrickhouse, Paul W. Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion. United States. https://doi.org/10.1080/00295450.2018.1431505
Rader, Jordan D., Greenwood, Michael Scott, and Humrickhouse, Paul W. Tue .
"Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion". United States. https://doi.org/10.1080/00295450.2018.1431505. https://www.osti.gov/servlets/purl/1454407.
@article{osti_1454407,
title = {Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion},
author = {Rader, Jordan D. and Greenwood, Michael Scott and Humrickhouse, Paul W.},
abstractNote = {Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with other system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.},
doi = {10.1080/00295450.2018.1431505},
journal = {Nuclear Technology},
number = 1,
volume = 203,
place = {United States},
year = {Tue Mar 20 00:00:00 EDT 2018},
month = {Tue Mar 20 00:00:00 EDT 2018}
}
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Works referencing / citing this record:
Demonstration of the Advanced Dynamic System Modeling Tool TRANSFORM in a Molten Salt Reactor Application via a Model of the Molten Salt Demonstration Reactor
journal, July 2019
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Modified Point-Kinetics Model for Neutron Precursors and Fission Product Behavior for Fluid-Fueled Molten Salt Reactors
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