About

Raluca O. Scarlat is an Associate Professor and Vice Chair of Undergraduate Matters at the University of California, Berkeley. She holds a Ph.D. in Nuclear Engineering from UC Berkeley, earned in 2012, and a B.S. in Chemical Engineering from Cornell University, obtained in 2006. Her research focuses on the chemical and thermophysical characterization of high-temperature molten salts and other inorganic fluids, as well as heat and mass transport related to energy systems. Her expertise includes electrochemistry, corrosion, thermodynamics, nuclear reactor safety analysis, licensing and design, and engineering ethics.

Research topics

• Computer Science
• Materials science
• Metallurgy
• Inorganic chemistry
• Chemistry
• Physics
• Mechanical engineering
• Chemical physics
• Physical chemistry
• Organic chemistry
• Nuclear physics
• Engineering
• Computational chemistry

Selected publications

• Operando spectroelectrochemical identification of peroxide intermediate in molten carbonate CO2-to-carbon electroreduction

  Nature Communications · 2026-04-21

  articleOpen accessSenior author

  Abstract Electrolysis of CO 2 in molten salts promises efficient carbon capture, but the underlying reaction mechanisms remain incompletely understood, as research thus far has been limited by a lack of tools for operando investigations. Here, we use a high-temperature operando Raman spectroelectrochemical system to look for signatures of reaction intermediates and study the evolution of carbon structures over electrolysis time. The analysis reveals the existence of O 2 2− concurrently with the deposition of carbon on Au, W, Inconel, and Ni electrode materials, pointing to a common reaction mechanism with O 2 2− as an intermediate. Secondly, the G peak of the as-deposited carbon experiences a noticeable blue-shift as the material is cooled down and purified, suggesting either a growth in crystallite size even after the electrolysis is stopped or lithium deintercalation. Elucidating the cathodic carbon deposition mechanism could help create greater value-added products and increase the economic viability of carbon capture.

  PublisherOA PDFDOI

• Toward the performance assessment of advanced nuclear waste forms: temperature dependence of lanthanide borosilicate glass dissolution

  npj Materials Degradation · 2026-02-18

  articleOpen access

  Abstract Lanthanide borosilicate (LaBS) glasses are among the most promising waste forms for the immobilization of high-level radioactive waste generated from advanced nuclear fuel cycles. However, the temperature dependence of their dissolution kinetics remains poorly understood and constrained, limiting the integration of these materials into established performance assessment models. Here, we investigate the dissolution behavior of the legacy AmCm2-19 LaBS glass and the benchmark alkali aluminoborosilicate ISG-1 in deionized water between 50 °C and 250 °C using ASTM C1285 (Product Consistency Test-B) protocols. For AmCm2-19 LaBS glass, normalized elemental release rates for boron and silicon increase with temperature before plateauing near 150 °C, consistent with solubility-limited behavior. From data obtained at 50 °C and 100 °C, Arrhenius analysis yields activation energies of E a (B) = 24.8 ± 0.3 kJ mol⁻¹ and E a (Si) = 14.4 ± 0.2 kJ mol⁻¹, similar or slightly lower than those previously reported for two other compositions of LaBS glasses. No secondary phases or alteration layers were detected by SEM-EDX or pXRD. These results establish one of the first temperature-dependent kinetic datasets for LaBS glass dissolution, providing quantitative parameters to inform mechanistic corrosion models and predictive simulations of glass degradation in geological disposal environments.

  PublisherOA PDFDOI

• Effect of cold forging on the microstructure and corrosion behavior of type 316L stainless steel in molten FLiNaK salt

  Journal of Nuclear Materials · 2025-01-09 · 5 citations

  articleOpen access
  PublisherOA PDFDOI

• Solid structure of Li₂BeF₄ (FLiBe) from room temperature to melting studied by neutron and X-ray diffraction

  Journal of Applied Crystallography · 2025-03-13

  articleOpen accessSenior author

  Molten fluoride salts such as Li 2 BeF 4 (FLiBe) are used in molten salt reactors, fluoride-salt-cooled high-temperature reactors and fusion reactors as a fuel solvent, coolant and/or tritium breeding medium. In engineered systems that use molten salt, solid-state material will be present during melting and freezing scenarios, and therefore the temperature-dependent properties of the solid and solid/liquid phase transition merit investigation. To observe the behavior of the solid state of Li 2 BeF 4 from room temperature to melting, this work used neutron and X-ray diffraction to measure the changes in the lattice parameters and volume of the crystalline unit cell and compared the results with prior low-temperature data for solid Li 2 BeF 4 . From neutron diffraction data it is also possible to identify anisotropy: centimetre-scaled crystals align preferentially with the a axes parallel to the direction of freezing front propagation, and the c axes expand 54% more than the a axes. This work provides the lattice constants as a function of temperature, quantifies the thermal expansion, and determines the equation describing the change in density for solid Li 2 BeF 4 from room temperature to 459°C to be ρ solid (kg m −3 ) = 2182 (3) − 0.115 (2) T (°C) and the volume expansion upon melting to be less than 5%. This density changes depending on molecular weight and enrichment.

  PublisherOA PDFDOI

• Emissivity of post-corrosion stainless steel 316 exposed to FLiBe molten salt in a non-isothermal flow loop

  Journal of Nuclear Materials · 2025-07-05 · 1 citations

  articleCorresponding
  PublisherDOI

• A Case for Nuclear Chemical Engineering in the Era of Fission and Fusion Reactors that Employ Molten Salts

  Nuclear Technology · 2025-06-30 · 1 citations

  articleOpen accessSenior authorCorresponding
  PublisherDOI

• Evaluating Hydrogen Bonding in Molten Halides: Implications for 2LiF-BeF ₂ (FLiBe) Tritium Breeding Blankets

  ECS Meeting Abstracts · 2025-11-24

  articleSenior author

  For a deuterium-tritium (D-T) fusion facility like Commonwealth Fusion Systems’ ARC or Xcimer Energy’s HYLIFE-III to operate at scale, considering that the current global tritium inventory is estimated at less than 27 kilograms 1 compared to estimated start-up requirements alone (~10kg) 1 , it is clear that the viability of D-T fusion depends on the development of technologies that close the tritium fuel cycle. For instance, assuming cycle energies for the HYLIFE-III system, one day’s worth of 3 GJ fusion events at a 0.75Hz repetition rate 2 would burn approximately 343.5 grams of tritium before including additional requirements to cover operational inefficiencies, radioactive decay (t 1/2 = 12.32 years), or build a reserve inventory. Therefore, upcoming fusion facilities are planned to include "tritium breeding blankets", which generate the isotope through neutron capture reactions on lithium-6. The molten salt 2LiF-BeF 2 (or FLiBe) is an attractive candidate for use as a liquid breeder. It combines a favorable neutron economy from the inclusion of beryllium as a neutron multiplier with compatibility as a heat transfer fluid to transport fusion energy to power generation at a relatively low corrosivity. However, a major research gap between today and the eventual scale-up of nuclear fusion facilities is the wide spread in the reported solubilities and diffusivities of tritium atoms (ions or otherwise) in FLiBe 3 , hindering the development of predictive engineering models for tritium retention, release, and accountancy. To address this, we will examine the behavior of hydrogen species in molten halides to build a mechanistic understanding of how a selected behavior depends on an equilibrium for or perturbation of a specific salt chemistry . We will review the body of literature examining hydrogen species (H - , H o , and H + of any isotope) behavior in the molten halides to assess how the cationic and anionic environments about hydrogen affect the solubility mechanisms and transport properties. Finally, we will extend this understanding to analyze the possible bond characteristics in Tritium-FLiBe systems incorporating the most recent electrochemical measurements of hydrogen speciation in FLiBe collected within the SALT group. We also reflect on strategies for communicating these complex chemical concepts to experts in other fields besides molten salts and ionic liquids. For example, words like species , behavior , or salt chemistry communicate a group of concepts to a reader which we cannot assume has the context to convert this terminology to actionable understandings to be implemented within system design of molten salt-facing technology. The behavior of hydrogen in molten FLiBe is not just a chemical curiosity; it directly impacts tritium safety, accountability, and supply. For these difficult to study systems (be it from the HF hazards, high temperatures, or strict beryllium safety protocols), it is important we make these chemical concepts accessible to the broader nuclear engineering community and extract as much value as possible from every measurement. To bridge the research and communication gaps, we propose a method to identify the excess chemical potential wells likely relevant to tritium-FLiBe chemistry like charge-dipole bonding between HF (d) and F - , BeF 4 2- , or larger anions formed from tetrahedral crosslinking 4 or covalent bonding character between elemental hydrogen and electropositive metals present in the melt. Using this method, we will define specific activity coefficients which relate input parameters (the exact LiF to BeF 2 ratio, redox control additives, and tritium production rates) to the observed activities of T 2 and TF in the melt from hypothetical tritium extraction rates for permeator or sparger systems. By demonstrating how the currently unpredictable data from tritium-FLiBe systems emerges from the chemistry at play in molten salts, we will build a case for molten salt chemical analysis as a core capability for fusion facilities through our exploration of charge-dipole and covalent bonding in our data and the body of literature on hydrogen-containing halide melts. (1) Abdou, M.; Riva, M.; Ying, A.; Day, C.; Loarte, A.; Baylor, L. R.; Humrickhouse, P.; Fuerst, T. F.; Cho, S. Physics and Technology Considerations for the Deuterium–Tritium Fuel Cycle and Conditions for Tritium Fuel Self Sufficiency. Nucl. Fusion 2020 , 61 (1), 013001. https://doi.org/10.1088/1741-4326/abbf35. (2) Ogando, F.; Tobin, M. T.; Meier, W. R.; Farga-Niñoles, G.; Marian, J.; Reyes, S.; Sanz, J.; Galloway, C. Preliminary Nuclear Analysis of HYLIFE-III: A Thick-Liquid-Wall Chamber for Inertial Fusion Energy. Fusion Eng. Des. 2024 , 202 , 114333. https://doi.org/10.1016/j.fusengdes.2024.114333. (3) Humrickhouse, P. W.; Fuerst, T. F. Tritium Transport Phenomena in Molten-Salt Reactors ; INL/EXT-20-59927-Rev000; Idaho National Lab. (INL), Idaho Falls, ID (United States), 2020. https://www.osti.gov/biblio/1777267 (accessed 2025-02-05). (4) Baes, C. F. A Polymer Model for BeF2 and SiO2 Melts. J. Solid State Chem. 1970 , 1 (2), 159–169. https://doi.org/10.1016/0022-4596(70)90008-3.

  PublisherDOI

• Round Robin Measurements of Molten Salt Properties for LiF-NaF-KF (FLiNaK) and NaCl-KCl Mixtures

  Journal of Chemical & Engineering Data · 2025-10-30 · 1 citations

  articleSenior author

  The development, operation, and regulation of nuclear reactors that utilize molten salts as fuel or as heat transfer media require knowledge of the thermal properties of the salt systems and quantification of the corresponding uncertainties. Knowledge of molten salt properties is also necessary for applications in material synthesis, processing, separations, solar thermal power generation, and energy storage. A round robin was conducted with national laboratory and university participants from twenty-one laboratories in five countries to compare property measurements, to better understand uncertainties, and to identify possible best practices. Two salt mixtures, each from a common batch, were distributed to participants for evaluation: equimolar NaCl-KCl and 45.0LiF-13.7NaF-41.3KF mol % (FLiNaK). Measurements were performed to determine the major constituent composition, oxygen content, density, thermal expansivity, melting point, and thermal conductivity. Error analysis was performed on each measurement for uncertainty quantification for each type of property that was explored. The resulting discussion of the methodologies used in this work is meant to lay the groundwork for the development of standard methods and reference materials for future high-temperature property measurements on halide melts.

  PublisherDOI

• Safety Analysis on Fission Induced Thermophysical Property Alterations of MCFR Fuel Salt

  Transactions of the American Nuclear Society · 2025-01-01

  articleSenior author
  PublisherDOI

• Experience Building a Molten-Salt Forced-Circulation Loop at the MIT Reactor

  Transactions of the American Nuclear Society · 2025-01-01

  article
  PublisherDOI

Frequent coauthors

• Francesco Carotti

  Kairos Power (United States)

  18 shared

• Huali Wu

  Institut Européen des Membranes

  17 shared

• Lorenzo Vergari

  17 shared

• Mark Asta

  University of California, Berkeley

  12 shared

• H. T. Williams

  12 shared

• Charles Forsberg

  Massachusetts Institute of Technology

  9 shared

• David Sprouster

  8 shared

• Sara Mastromarino

  Kairos Power (United States)

  8 shared

Education

• Ph.D., Nuclear Engineering

  University of California, Berkeley

  2005

• M.S., Nuclear Engineering

  University of California, Berkeley

  2001

• B.S., Physics

  University of Bucharest

  1998