Study on advection–dispersion behavior for simulation of 3H, 99Tc, and 90Sr transport in crushed sandstone of column experiments
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Yu-Hung Shih
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Liang-Cheng Chen
Abstract
In order to establish a universal and safety-compliant post-closure safety assessment technique, it is necessary to develop appropriate models to explain the migration behavior of radioactive materials within the rock system. Advection-dispersion experiments (ADE) have proven successful in designing transport models through a calibration/validation process of breakthrough curves (BTCs). In the present investigation, we employed a dynamic column device to examine the transport of Tritium (3H), Technetium-99 (99Tc), and Strontium-90 (90Sr) in crushed sandstone. Non-reactive transport experiments utilizing conservative tracers enabled us to determine the transport parameters, including retardation factors (R) and dispersivity (α). Our study focused on investigating the surface complexation model (SCM) using an additive approach coupled with the advection-dispersion equation to simulate the reactive transport behavior of 90Sr. The results affirm the robustness of our chosen thermodynamic database and modeling approach, emphasizing the criticality of accurately modeling and predicting reactive transport behavior in porous media.
Acknowledgments
The authors would like to thank the researchers at the Radioactive Waste Disposal Technology Research and Development Center, National Tsing Hua University for technical discussions.
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Research ethics: Not applicable.
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Author contributions: The authors have accepted responsibility for the entire content of this manuscript and approved its submission.
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Competing interests: The authors state no conflict of interest.
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Research funding: The Science and Technology Development Fund supported this project under contract number NSTC 112-3111-Y-042A-012.
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Data availability: Not applicable.
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Articles in the same Issue
- Frontmatter
- Numerical study on the effect of the PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code
- Analyses of the unavailability dynamics of emergency core cooling system
- Study on spent fuel heatup during spent fuel pool complete loss of coolant accident
- Numerical simulation analysis of high-temperature bent sodium heat pipes
- Influence of the twisting and nano fluids on performance of a triangular double tube heat exchanger
- Neutronic simulation of Traveling Wave Reactor (TWR) core in multi-cycles using Monte Carlo method
- Gain scheduled internal model control based on the dynamic sliding mode method for the water level of nuclear steam generators
- Verification and validation optimization method for signal quality bits in digital control system application software of nuclear power plant
- Investigation of Li–Be and B halides as blanket in future fusion molten salt reactor
- A study on porosity investigation of compacted bentonite in various densities by using micro-computed tomography images analysis
- CTAB modification bentonite for enhanced Re adsorption and diffusion suppression
- Study on advection–dispersion behavior for simulation of 3H, 99Tc, and 90Sr transport in crushed sandstone of column experiments
- Investigating advection–dispersion behavior for simulation of HTO and 238Pu transport in argillaceous shale with different varying degrees of weathering
- Study on analysing the potential benefits of utilizing nuclear waste for biodiesel production
- Calendar of events
Articles in the same Issue
- Frontmatter
- Numerical study on the effect of the PI-controller type on the quasi-steady reactor pressure in MAAP 5.04 code
- Analyses of the unavailability dynamics of emergency core cooling system
- Study on spent fuel heatup during spent fuel pool complete loss of coolant accident
- Numerical simulation analysis of high-temperature bent sodium heat pipes
- Influence of the twisting and nano fluids on performance of a triangular double tube heat exchanger
- Neutronic simulation of Traveling Wave Reactor (TWR) core in multi-cycles using Monte Carlo method
- Gain scheduled internal model control based on the dynamic sliding mode method for the water level of nuclear steam generators
- Verification and validation optimization method for signal quality bits in digital control system application software of nuclear power plant
- Investigation of Li–Be and B halides as blanket in future fusion molten salt reactor
- A study on porosity investigation of compacted bentonite in various densities by using micro-computed tomography images analysis
- CTAB modification bentonite for enhanced Re adsorption and diffusion suppression
- Study on advection–dispersion behavior for simulation of 3H, 99Tc, and 90Sr transport in crushed sandstone of column experiments
- Investigating advection–dispersion behavior for simulation of HTO and 238Pu transport in argillaceous shale with different varying degrees of weathering
- Study on analysing the potential benefits of utilizing nuclear waste for biodiesel production
- Calendar of events
