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Title: Thermal analysis of projected molten salt compositions during FFTF and EBR-II used nuclear fuel processing

Abstract

This work studies the change in liquidus temperature of the salt used in the Mk-IV electrorefiner at the Idaho National Laboratory during processing campaigns of sodium-bonded driver fuels from Experimental Breeder Reactor II and the Fast Flux Test Facility reactor. Modeling and Simulation Tool for Electrochemical Recycling Systems (MASTERS), an INL proprietary pyroprocessing flowsheet simulation tool, was used to simulate the processing campaigns and determine the resulting composition changes of the Mark-IV electrorefiner salt. Surrogate salt samples simulating the Mk-IV electrorefiner salt during the fuel processing campaigns were prepared, and the thermal properties were measured via differential scanning calorimetry. Results from this study indicate that after processing approximately 2,150 kg of uranium metal from used fuel, the liquidus temperature of the molten salt exceeds 520°C, which is a significant increase from 352°C (the melting temperature of eutectic LiCl-KCl). It was also discovered that the liquidus temperature is strongly dependent on the amount of NaCl accumulating in the Mk-IV electrorefiner salt. Thus, the assessment of the liquidus temperature during the subsequent fuel processing campaigns can be simplified by considering the ternary LiCl-KCl-NaCl system only.

Authors:
ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1];  [1]; ORCiD logo [1];  [2]
  1. Idaho National Laboratory (INL), Idaho Falls, ID (United States)
  2. Univ. of Idaho, Moscow, ID (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1508609
Report Number(s):
INL/JOU-18-51475-Rev000
Journal ID: ISSN 0022-3115
Grant/Contract Number:  
AC07-05ID14517
Resource Type:
Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 520; Journal Issue: C; Journal ID: ISSN 0022-3115
Publisher:
Elsevier
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CHEMISTRY; liquidus temperature; uranium electrorefining; Mk-IV electrorefiner; molten salt; pyroprocessing

Citation Formats

Karlsson, Toni Y., Fredrickson, Guy L., Yoo, Tae-Sic, Vaden, DeeEarl, Patterson, Michael N., and Utgikar, Vivek. Thermal analysis of projected molten salt compositions during FFTF and EBR-II used nuclear fuel processing. United States: N. p., 2019. Web. doi:10.1016/j.jnucmat.2019.04.016.
Karlsson, Toni Y., Fredrickson, Guy L., Yoo, Tae-Sic, Vaden, DeeEarl, Patterson, Michael N., & Utgikar, Vivek. Thermal analysis of projected molten salt compositions during FFTF and EBR-II used nuclear fuel processing. United States. doi:10.1016/j.jnucmat.2019.04.016.
Karlsson, Toni Y., Fredrickson, Guy L., Yoo, Tae-Sic, Vaden, DeeEarl, Patterson, Michael N., and Utgikar, Vivek. Wed . "Thermal analysis of projected molten salt compositions during FFTF and EBR-II used nuclear fuel processing". United States. doi:10.1016/j.jnucmat.2019.04.016.
@article{osti_1508609,
title = {Thermal analysis of projected molten salt compositions during FFTF and EBR-II used nuclear fuel processing},
author = {Karlsson, Toni Y. and Fredrickson, Guy L. and Yoo, Tae-Sic and Vaden, DeeEarl and Patterson, Michael N. and Utgikar, Vivek},
abstractNote = {This work studies the change in liquidus temperature of the salt used in the Mk-IV electrorefiner at the Idaho National Laboratory during processing campaigns of sodium-bonded driver fuels from Experimental Breeder Reactor II and the Fast Flux Test Facility reactor. Modeling and Simulation Tool for Electrochemical Recycling Systems (MASTERS), an INL proprietary pyroprocessing flowsheet simulation tool, was used to simulate the processing campaigns and determine the resulting composition changes of the Mark-IV electrorefiner salt. Surrogate salt samples simulating the Mk-IV electrorefiner salt during the fuel processing campaigns were prepared, and the thermal properties were measured via differential scanning calorimetry. Results from this study indicate that after processing approximately 2,150 kg of uranium metal from used fuel, the liquidus temperature of the molten salt exceeds 520°C, which is a significant increase from 352°C (the melting temperature of eutectic LiCl-KCl). It was also discovered that the liquidus temperature is strongly dependent on the amount of NaCl accumulating in the Mk-IV electrorefiner salt. Thus, the assessment of the liquidus temperature during the subsequent fuel processing campaigns can be simplified by considering the ternary LiCl-KCl-NaCl system only.},
doi = {10.1016/j.jnucmat.2019.04.016},
journal = {Journal of Nuclear Materials},
number = C,
volume = 520,
place = {United States},
year = {2019},
month = {4}
}

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This content will become publicly available on April 10, 2020
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