Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT) (방사성폐기물학회지)

Korean Radioactive Waste Society (한국방사성폐기물학회)
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Phase Behavior of the Ternary NaCl-PuCl3-Pu Molten Salt

  • Received : 2022.09.06
  • Accepted : 2023.03.07
  • Published : 2023.03.31
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Abstract

There is a gap in our understanding of the behavior of fused and molten fuel salts containing unavoidable contamination, such as those due to fabrication, handling, or storage. Therefore, this work used calorimetry to investigate the change in liquidus temperature of PuCl3, having an unknown purity and that had been in storage for several decades. Further research was performed by additions of NaCl, making several compositions within the binary system, and summarizing the resulting changes, if any, to the phase diagram. The melting temperature of the PuCl3 was determined to be 746.5℃, approximately 20℃ lower than literature reported values, most likely due to an excess of Pu metal in the PuCl3 either due to the presence of metallic plutonium remaining from incomplete chlorination or due to the solubility of Pu in PuCl3. From the melting temperature, it was determined that the PuCl3 contained between 5.9 to 6.2mol% Pu metal. Analysis of the NaCl-PuCl3 samples showed that using the Pu rich PuCl3 resulted in significant changes to the NaCl-PuCl3 phase diagram. Most notably an unreported phase transition occurring at approximately 406℃ and a new eutectic composition of 52.7mol% NaCl-38.7mol% PuCl3-2.5mol% Pu which melted at 449.3℃. Additionally, an increase in the liquidus temperatures was seen for NaCl rich compositions while lower liquidus temperatures were seen for PuCl3 rich compositions. It can therefore be concluded that changes will occur in the NaCl-PuCl3 binary system when using PuCl3 with excess Pu metal. However, melting temperature analysis can provide valuable insight into the composition of the PuCl3 and therefore the NaCl-PuCl3 system.

Keywords

Acknowledgement

This work was supported through the Idaho National Laboratory's (INL) Laboratory Directed Research and Development Program under Department of Energy's Idaho Operations Office Contract DE-AC07-05ID14517 and The Molten Salt Reactor Campaign, work package number AT22IN070502 "Thermochemical and Thermophysical Property Database Development-INL". The authors would like to acknowledge and thank the staff in the Fuel Manufacturing Facility at INL for coordinating this research.

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