Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

doi:10.1080/00295450.2018.1431505

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

Journal Article · Tue Mar 20 00:00:00 EDT 2018 · Nuclear Technology

DOI:https://doi.org/10.1080/00295450.2018.1431505· OSTI ID:1454407

^[1]; ^[1]; Humrickhouse, Paul W. ^[2]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Idaho National Lab. (INL), Idaho Falls, ID (United States)

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.

View Accepted Manuscript (DOE)

Research Organization:: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Sponsoring Organization:: USDOE

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1454407

Journal Information:: Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 1 Vol. 203; ISSN 0029-5450

Publisher:: Taylor & Francis - formerly American Nuclear Society (ANS)Copyright Statement

Country of Publication:: United States

Language:: English

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Cited By (2)

Demonstration of the Advanced Dynamic System Modeling Tool TRANSFORM in a Molten Salt Reactor Application via a Model of the Molten Salt Demonstration Reactor Greenwood, M. Scott; Betzler, Benjamin R.; Qualls, A. Lou Nuclear Technology, Vol. 206, Issue 3 https://doi.org/10.1080/00295450.2019.1627124	journal	July 2019
Modified Point-Kinetics Model for Neutron Precursors and Fission Product Behavior for Fluid-Fueled Molten Salt Reactors Greenwood, M. Scott; Betzler, Ben Nuclear Science and Engineering, Vol. 193, Issue 4 https://doi.org/10.1080/00295639.2018.1531619	journal	November 2018

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70 PLASMA PHYSICS AND FUSION TECHNOLOGY
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Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

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