The cover depicts the chemical and physical complexity of the various species and 

interfaces within a molten salt reactor. To advance new approaches to molten salt 

technology development, it is necessary to understand and predict the chemical and 

physical properties of molten salts under extreme environments; understand their ability to 

coordinate fissile materials, fertile materials, and fission products; and understand their 

interfacial reactions with the reactor materials. Modern x-ray and neutron scattering tools 

and spectroscopy and electrochemical methods can be coupled with advanced 

computational modeling tools using high performance computing to provide new insights 

and predictive understanding of the structure, dynamics, and properties of molten salts over 

a broad range of length and time scales needed for phenomenological understanding. The 

actual image is a snapshot from an ab initio molecular dynamics simulation of grapheneorganic electrolyte interactions. Image courtesy of Bobby G. Sumpter of ORNL. 
Molten Salt Chemistry Workshop 

Technology and Applied R&D 

Report for the US Department of Energy, Office 

of Nuclear Energy Workshop 

Molten Salt Chemistry Workshop 

Technology and Applied R&D Needs for Molten Salt Chemistry 

April 10–12, 2017 

Oak Ridge National Laboratory 

Co-chairs: 

David F. Williams, Oak Ridge National Laboratory 

Phillip F. Britt, Oak Ridge National Laboratory

Working Group Co-chairs 

Working Group 1: Physical Chemistry and Salt Properties 

Alexa Navrotsky, University of California–Davis 

Mark Williamson, Argonne National Laboratory 

Working Group 2: Analytical Chemistry 

Sam Bryan, Pacific Northwest National Laboratory 

Sheng Dai, Oak Ridge National Laboratory 

Working Group 3: Molten Salt Fission Product Chemistry and Solid Salt Radiolysis

Bill DelCul, Oak Ridge National Laboratory 

Tina Nenoff, Sandia National Laboratories 

Working Group 4: Materials Compatibility 

James Keiser, Oak Ridge National Laboratory 

Preet Singh, Georgia Institute of Technology 

Working Group 5: Computational Chemistry and Materials Science 

Bobby Sumpter, Oak Ridge National Laboratory 

Brian Wirth, University of Tennessee–Knoxville 

Charles Henager, Pacific Northwest National Laboratory 

Office of Nuclear Energy Leads 

John Herczeg, Deputy Assistant Secretary for Nuclear Technology Research and Development, 

DOE-NE 

Stephen Kung, Program Manager, DOE-NE 

ORNL Creative Services and ORNL Research Library 

Deborah Counce, LeJean Hardin, Cindy Johnson, Kathy Jones, James Kidder 

Molten Salt Chemistry Workshop 

Table of Contents iii 

Table of Contents 

Abbreviations, Acronyms, and Initialisms ............................................................................................... v

Executive Summary .................................................................................................................................. vii

1. Introduction ......................................................................................................................................... 1

1.1 Workshop Organization ................................................................................................................ 3

1.2 Future Research Directions ........................................................................................................... 3

References ............................................................................................................................................. 4

2. FRD 1: Understanding, Predicting, and Optimizing the Physical Properties of Molten 

Salts ...................................................................................................................................................... 5

2.1 Background and Current Status ..................................................................................................... 5

2.2 Technical Challenges and Research Directions ............................................................................ 7

2.2.1 Determining the Physical Properties of Molten Salts ....................................................... 7

2.2.2 Predicting the Chemical Speciation, Structure, and Dynamics of Molten Salts 

Solutions .............................................................................................................................. 15

References ........................................................................................................................................... 19

3. FRD 2: Understanding the Structure, Dynamics, and Chemical Properties of Molten 

Salts .................................................................................................................................................... 23

3.1 Background and Current Status ................................................................................................... 23

3.2 Technical Challenges and Research Directions .......................................................................... 28

3.2.1 Determining Molecular Structure by X-ray and Neutron Scattering and 

Spectroscopy ....................................................................................................................... 28

3.2.2 Correlating Electrochemistry and Spectroscopy for Online Monitoring and 

Predictive Modeling ............................................................................................................ 30

References ........................................................................................................................................... 33

4. FRD 3: Understanding Fission and Activation Product Chemistry and Radiation 

Chemistry .......................................................................................................................................... 37

4.1 Background and Current Status ................................................................................................... 37

4.2 Technical Challenges and Research Directions .......................................................................... 38

4.2.1 Understanding the Physical, Chemical, and Radiochemical Properties of the 

Fission and Activation Products in Molten Salts ................................................................. 38

4.2.2 Correlating Fission and Activation Product Behavior with Surrogates .......................... 40

4.2.3 Understanding the Physical and Chemical Impact of Short-lived Isotopes .................... 42

References ........................................................................................................................................... 43

5. FRD 4: Understanding Materials Compatibility and Interfacial Phenomena ............................ 45

5.1 Background and Current Status ................................................................................................... 45

5.2 Technical Challenges and Research Directions .......................................................................... 46

5.2.1 Advancing Spectroscopic and Scattering Investigation of MSR Interfaces .................... 46
Molten Salt Chemistry Workshop

iv Table of Contents 

5.2.2 Understanding Materials Compatibility in the Molten Salts Reactor 

Environment ........................................................................................................................ 47

5.2.3 Understanding Degradation Processes at the Material–Salt Interface ............................ 50

5.2.4 Understanding the Combined Effect of Chemistry and Radiation at the Interface ......... 51

5.2.5 Predicting Interfacial Interactions between Molten Salts and Structural Materials 

and/or Gases ........................................................................................................................ 53

References ........................................................................................................................................... 56

6. FRD 5: Guiding Next-Generation Materials for Molten Salt Reactors ....................................... 59

6.1 Background and Current Status ................................................................................................... 59

6.2 Technical Challenges and Research Directions .......................................................................... 59

6.2.1 Enabling Rapid Deployment of MSR Materials: Computational Challenges ................. 59

6.2.2 Guiding Experimental Exploration of Advanced “Super” Materials .............................. 63

6.2.3 Advancing Code Qualification of New Materials ........................................................... 64

References ........................................................................................................................................... 66

7. FRD 6: Creating a Virtual Reactor Simulation ............................................................................. 67

7.1 Background and Current Status ................................................................................................... 67

7.2 Technical Challenge and Research Directions ............................................................................ 67

7.2.1 Creating a Virtual Reactor Simulation ............................................................................ 67

References ........................................................................................................................................... 70

APPENDIX A: WORKSHOP RESOURCE DOCUMENT ................................................................ A-1

APPENDIX B: A DEVELOPER’S PERSPECTIVE ON THE NECESSITY OF SALT 

PROPERTY RESEARCH ............................................................................................................. B-1

APPENDIX C: WORKSHOP AGENDA ............................................................................................. C-1

APPENDIX D: WORKSHOP PARTICIPANTS ................................................................................. D-1
Molten Salt Chemistry Workshop 

Table of Contents v 

Abbreviations, Acronyms, and Initialisms 

AIMD ab initio molecular dynamics 

ARE Aircraft Reactor Experiment 

CALPHAD Computer Coupling of Phase Diagrams and Thermochemistry

CMD classical molecular dynamics 

DFT density functional theory 

DOE Department of Energy 

DSC differential scanning calorimetry or differential scanning calorimeter 

EMF electromotive force 

EXAFS extended XAFS 

FLiBe 7

Li2BeF4 

FRD Future Research Direction 

GAIN Gateway for Acceleration of Innovation in Nuclear

GI-XAFS grazing incidence x-ray absorption fine structure spectroscopy 

IR infrared 

JICS Joint Institute for Computational Sciences 

MAS magic-angle spinning 

MD molecular dynamics 

MOF metal-organic framework 

MSBR Molten Salt Breeder Reactor 

MSR molten salt reactor 

MSRE Molten Salt Reactor Experiment 

NE Office of Nuclear Energy 

NMR nuclear magnetic resonance 

ORNL Oak Ridge National Laboratory 

PDF pair distribution function 
Molten Salt Chemistry Workshop

vi Abbreviations, Acronyms, and Initialisms 

QM/MD quantum mechanics molecular dynamics 

R&D research and development 

SEMO semi-empirical molecular orbital theory 

TB-DFT tight binding DFT 

TGA thermogravimetric analysis 

UV-vis ultraviolet-visible 

UV-vis-IR ultraviolet-visible-infrared 

VR virtual reactor 

VUQ validation and uncertainty quantification 

WG working group 

XAFS x-ray absorption fine structure spectroscopy 

XAS x-ray absorption spectroscopy 

XRD x-ray diffraction 
Molten Salt Chemistry Workshop 

Executive Summary vii 

Executive Summary 

At the dawn of this century, nuclear experts in the Generation IV International Forum recognized the unique 

capabilities of molten salts in one of six next-generation reactor designs. Since that time, various national 

research and development programs, most notably in China, have focused resources on developing molten 

salt reactors (MSRs) of various designs. In the past decade in the United States, new designs and 

applications have matured to the point that a number of companies have dedicated resources, enthusiasm, 

and new approaches to MSR technology development. The US Department of Energy (DOE) and its 

national laboratories have supported this evolution by working with multiple companies and universities; 

this relationship matured in 2015 with the creation of the Gateway for Acceleration of Innovation in Nuclear 

(GAIN) initiative working in concert with the US Industry MSR Technical Working Group. 

In response to the sustained resurgence of the MSR concept, the DOE Office of Nuclear Energy 

organized the Molten Salt Chemistry Workshop at Oak Ridge National Laboratory on April 10–12, 2017, 

for the purpose of identifying innovative science-based, technology driven approaches to accelerate MSR 

development and deployment. In the United States and across the world, the MSR concept is recognized 

as one of the few game-changing technologies that can overcome existing barriers to the adoption of 

commercial nuclear power. Not only is the generation of electricity within a once-through uranium fuel 

cycle foreseen; in addition, proponents of MSR technology envision completely new opportunities to 

rationalize fuel cycles by securely using a variety of fissile and fertile resources to supply hightemperature, low-pressure industrial heat and to create a cost-competitive, safe, and more sustainable 

commercial nuclear option. 

This workshop was rooted in a multidisciplinary dialogue among a wide range of experts spanning 

chemistry, materials science, nuclear science, and engineering from academia, national laboratories, the 

US government, and commercial organizations. There was strong participation from the US Industry 

Technical Working Group, as well as participation from small and large companies with specific interests 

and expertise in molten salt technology. Since MSR technology has a chemical focus and chemical 

challenges, the participants were focused toward the chemical sciences—but a very strong contingent of 

materials, computational, and nuclear scientists/engineers also participated. These experts were organized 

into five working groups: (1) physical chemistry and salt properties, (2) analytical chemistry, (3) molten 

salt fission product chemistry and solid-state radiolysis, (4) materials compatibility, and (5) computational 

science and material science. The Introduction section of this document and the appendices provide 

details on how the workshop was conducted and who participated.

Fourteen R&D topics identified during the workshop were integrated into 6 Future Research Directions 

(FRDs) after the workshop. The input for and definition of each FRD contained contributions from 

multiple working groups and represent a collective effort from the workshop participants. 

The 21st century MSR mission is distinctly different from the original thermal-spectrum molten salt 

breeder reactor, and the next generation of MSR salts will be different. Thus a deeper understanding is 

needed of molten salt physical and chemical properties to optimize the salt composition for new 

applications. Choosing the right salt composition requires a knowledge of the salt’s capacity to dissolve 

fissile, fertile, and particular fission products and of the physical properties of the salt. Phase diagrams 

and thermophysical properties of molten salts are needed in the early stages of development to justify the 

design and to define the safety boundaries. The first FRD (Understanding, Predicting, and Optimizing the 

Physical Properties of Molten Salts) highlights the need to apply modern measurement techniques and 

modeling and simulation tools to accelerate the design, discovery, and characterization of salts optimized 

for various types of MSRs. Purification, characterization, and handling procedures for molten salts—as 

well as analytical standards for pedigree molten salts—need to be accessible to the scientific community 

to ensure the field can rapidly process and share reliable data. 
Molten Salt Chemistry Workshop

viii Executive Summary 

The structure and dynamics of molten salts play a key role in understanding phase transitions, viscosity, 

thermal conductivity, volatility, solubility of fission products, and corrosion. Identifying and 

understanding the structure, properties, and reactivity of these complex salt species is dependent upon 

coordination chemistry, the redox state and Lewis-acidity of the system. Doing so requires a combination 

of advanced measurement tools to identify the structures, and modeling capabilities to understand and 

interpret the structure, dynamics, and properties. The second FRD (Understanding the Structure, 

Dynamics, and Chemical Properties of Molten Salts) highlights how modern x-ray and neutron scattering 

and spectroscopic tools (which were not available during the studies in the 1960s and 1970s) and 

electrochemical methods can be coupled with advanced modeling capabilities to provide new insights into 

the structure, dynamics, and properties of salt species on the length and time scales needed for 

phenomenological understanding. This capability is needed for advancement in all other FRDs. 

The third FRD (Understanding Fission and Activation Product Chemistry and Radiation Chemistry) 

highlights the need to understand the rapid decay and chemical transmutation of fission and activation 

products and the unique phenomena related to radiation-induced chemistry. It is essential to understand which 

fission products are soluble, insoluble, partially soluble, and volatile, and how fission products change the 

corrosivity of the melt. Highly specialized approaches are needed to address the fate of insoluble fission 

products that evolve through complex decay chains, as well as specific radiation effects on chemical species. 

The fourth FRD (Understanding Materials Compatibility and Interfacial Phenomena) addresses the need 

to fundamentally understand material degradation (such as corrosion) and interfacial reactions (including 

the combined chemical and radiation effect) to develop new, more stable materials for MSR. This 

requires specialized experimental approaches (including a flow loop), instrumentation, and analysis to 

study interfacial reactions—for both out-of-pile and in-pile measurements—to follow trace levels of 

dissolved and dispersed species in solution and concentrations of surface species and surface morphology. 

Computational modeling is needed to gain a molecular-level understanding of the chemical reactivity of 

molten salts to provide insights into mitigating corrosion, fission product off-gassing, gas entrainment, 

and materials precipitation. 

The fifth FRD (Guiding Next-Generation Materials for Molten Salt Reactors) addresses the need to 

develop the next generation of materials that will enable MSR developers to reach the reactor 

performance targets. Structural materials proposed for MSR must endure extreme environments, 

including high fuences of neutrons, high operating temperatures, and corrosive environments. New 

approaches are needed to develop superior alloys and composites that are optimized for MSR which build 

off detailed physical, chemical, and materials studies and combine modern experimental approaches—

including high throughput screening—with state-of-the art computational materials science approaches. 

Advances in computational modeling can help guide materials development through an understanding of 

chemomechanical degradation mechanisms, radiation damage, and thermomechanical degradation of 

microstructure and property evolution. Timelines for materials development, testing, and qualification 

must be accelerated to incorporate new materials into MSR designs through innovative approaches using 

modern analytical and computational tools with new testing protocols. 

The sixth FRD (Creating a Virtual Reactor Simulation) focuses on developing the modeling and simulation 

tools necessary to understand the behavior of the reactor throughout its lifetime and provide dynamic reactor 

and irradiation capsule emulation for chemical and isotopic source terms. To accurately predict MSR 

conditions—including neutron transport, thermal hydraulics, isotopic transmutation, thermochemical 

properties of the fuel salt, and corrosion effects—a virtual reactor simulator needs to be developed which 

builds off the experimental and computational advances made in the previous FRD and incorporates a new 

understanding of the physical, chemical, and radiochemical phenomena found in MSRs. This simulation 

tool will not only allow understanding of the operational issues but also allow analysis of the fuel cycle, 

prediction of waste forms and quantities, and assessment of new reactor concepts, salts, and chemistries. 
Molten Salt Chemistry Workshop 

Introduction 1 

1. Introduction 

Since 1948, scientists have sought the creation of a new type of high-temperature fluid-fueled nuclear 

reactor that is not constrained by the limitations of solid fuels and conventional coolants. The molten salt 

reactor (MSR) emerged as the only viable concept for this class of reactor. In 1954, scientists at Oak 

Ridge National Laboratory (ORNL) demonstrated the short-duration but very-high-temperature 

capabilities of a molten-fluoride–fueled reactor in the Aircraft Reactor Experiment (ARE) [1]. The fuel 

salt was a mixture of NaF-ZrF4-UF4, and the moderator was BeO. The reactor operated successfully for 9 

days at an outlet temperature of 860 °C and power up to 2.5 MWt. At the end of the Aircraft Nuclear 

Propulsion program, molten salt technology was repurposed with the objective of developing a thermal 

spectrum breeder reactor based upon the thorium fuel cycle. From 1965 through 1969, ORNL operated 

the 8 MW Molten Salt Reactor Experiment (MSRE), which demonstrated many of the key technologies 

needed to justify the pursuit of a commercial MSR. The MSRE fuel salt consisted of 7

LiF-BeF2-ZrF4-UF4

and it flowed through a graphite moderator at atmospheric pressure at approximately 650 °C. (7

Li was 

used to prevent the undesired production of tritium from 6

Li by neutron-induced fission.) The operation 

demonstrated the use of both 235U and 233U as the fissile driver in the fuel salt, as well as minor refueling 

adjustments with 239Pu. The coolant salt in the secondary loop was 7

Li2BeF4 (FLiBe). Experience with 

these test reactors and the associated research and development (R&D) showed the MSR concept was 

viable; but it also highlighted the challenges that remained in the domains of salt chemistry, fission 

product chemistry, materials and corrosion chemistry, radiation chemistry, analytical chemistry, and fuel 

processing. Figure 1 charts the evolution of MSR concepts. 

(a) High-temperature 

thermal propulsion short 

duration engine (Aircraft 

Reactor Experiment) 

(b) Extended multi-functional test 

reactor (Molten Salt Reactor 

Experiment) 

(c) Thermal breeder reactor design 

(Molten Salt Breeder Reactor) 

Figure 1. The 20th century evolution of the molten salt reactor concept. | Images courtesy of Oak Ridge 

National Laboratory

With the cessation of the US commitment to thorium-breeder reactor designs in 1976, a different focus 

eventually took hold for the MSR concept. It is now envisioned as a new way to rationalize the overall 

fuel cycle without conventional aqueous reprocessing, with novel fueling and refueling schemes that do 

not entail remote solid-fuel refabrication from recycled actinides; and as a potential pathway to safer, 

more economical nuclear electricity generation and supply of high-temperature, low-pressure industrial 

heat. This vision was recognized in the international commitment to the MSR concept within the 

Generation Four International Forum. Strong efforts in MSR development are sustained in many 

European and Asian countries (especially China) (see World Nuclear Association). Most of the interest in 
Molten Salt Chemistry Workshop

2 Introduction 

the US private sector revolves around the use of low-enrichment, high-actinide-content fuel salts, but a 

wide variety of concepts are being considered using fluoride and chloride salts (see Appendix A). MSRs 

under consideration today are either solid-fueled salt-cooled designs or liquid-salt–fueled designs, as 

shown in Figure 2. Solid-fueled salt-cooled reactors contain solid ceramic fuel in prisms, plates, or 

pebbles. Molten salt coolants flow over the solid fuel, and the heat is removed in the primary heat 

exchanger by a secondary coolant; it transports the heat to a power generating system, such as a gas or 

steam turbine. Molten salts have excellent heat transfer properties that rival those of water and permit 

them to operate at much higher temperatures than light water reactors and at near-atmospheric pressures 

in the primary loop. 

Figure 2. In an MSR, the choice of salt determines the mission of the reactor. It imparts versatility for 

multiple applications such as electricity production, generation of process heat, or burnup of actinide 

waste (see Appendix A). | Image courtesy of Oak Ridge National Laboratory

Liquid-salt-fueled reactors—as demonstrated during the operation of the ARE and MSRE and envisioned 

in the Molten Salt Breeder Reactor (MSBR) design—contain all fissile and fertile constituents (and many 

fission products) dissolved within a homogeneous fluid fuel that circulates in the primary loop. The vast 

majority of fission energy is deposited uniformly throughout the fluid fuel, which serves as both fuel 

matrix and primary coolant. Heat is shuttled to the balance-of-plant by a nonradioactive low-pressure 

secondary molten salt via two heat exchangers (Figure 2). Fluid-fueled-reactors have the significant virtue 

that they do not require solid fuel fabrication/refabrication/qualification; but they impose the requirement 

to control and manage all of the fission products that are released into the primary loop as either 

dissolved, insoluble, or gaseous species. 
Molten Salt Chemistry Workshop 

Introduction 3 

As interest continues to grow in MSRs, new applications involve new materials and chemistry 

requirements. Although some R&D requirements for salt chemistry and materials are the same as those 

remaining at the end of the MSBR development program in 1976 (e.g., see Engel et al. 1979 [2]), other 

requirements are driven by the use of new salts (e.g., chlorides), the realization of new conditions, and the 

pursuit of new application domains. To advance the potential viability of MSRs, additional fundamental 

and applied R&D is needed for salt chemistry, fission product chemistry, materials and corrosion 

chemistry, radiation chemistry, analytical chemistry, and fuel processing. 

1.1 Workshop Organization 

In recognition of the renewed commercial interest in MSRs, the US Department of Energy (DOE) Office 

of Nuclear Energy (NE), in coordination with MSR developers, plans to reestablish core competencies 

and R&D capabilities in molten salt chemistry at DOE national laboratories and US universities. This 

effort will provide a technical basis for supporting the goal of US industries to commercialize MSR 

technologies and will support the MSR research community in training a broad base of next-generation 

expertise. 

To identify potential innovative science-based, technology-driven approaches to accelerate the 

deployment of MSR technologies, and to engage the broader scientific community to advance the 

knowledge and technology base of molten salt chemistry, DOE sponsored a 3 day workshop, Technology 

and Applied R&D Needs for Molten Salt Chemistry. It was held at ORNL on April 10–12, 2017. In 

addition to molten salt experts, the workshop included experts in traditional disciplines of chemistry 

(physical, analytical, fission product, and materials chemistry) and nontraditional disciplines 

(computational chemistry and materials science) from national laboratories, universities, and industry. A 

total of 72 experts attended the workshop—15 from universities, 35 from national laboratories, 15 from 

industry, and 7 from government agencies. (Appendix B is list of the workshop participants.) 

In preparation for the meeting, a Resource Document (Appendix A) was drafted to orient and educate all 

participants on the scientific and technology advances underpinning MSR technologies. Emphasis was 

placed on identifying fundamental knowledge gaps. The agenda of the workshop is provided in 

Appendix C. The workshop began with a half-day plenary session of invited experts from DOE, industry, 

universities, and national laboratories to help define where the science and technology stands currently 

and where it needs to go. Following the plenary session, the workshop participants separated into five 

working groups: Physical Chemistry and Salt Properties, Analytical Chemistry, Molten Salt Fission 

Product Chemistry and Solid Salt Radiolysis, Materials Compatibility, and Computational Chemistry and 

Materials Science. The panels were charged with identifying (1) research topics that address both shortterm technology barriers and long-term (5–10 years) grand challenges that might enable revolutionary 

changes in MSR technology and (2) a set of Future Research Directions (FRDs) that might accelerate 

MSR technology development and deployment. At the midpoint of the workshop, each group reported to 

all of the workshop participants on its progress toward identifying FRDs. The FRDs were discussed and 

overlaps were identified. At the end of the workshop, each working group again reported its 

recommendations for FRDs to all the workshop participants. Each working group prepared a report that 

captured the research ideas and opportunities discussed in its sessions and the recommended FRDs. These 

results were integrated and are summarized below and described in detail in the full report. 

1.2 Future Research Directions

The working groups initially identified 14 FRDs; however, these contained significant overlap and thus 

were combined to formulate 6 FRDs. The following FRDs are described in detail in the chapters that 

follow. 
Molten Salt Chemistry Workshop

4 Introduction 

 Understanding, Predicting, and Optimizing the Physical Properties of Molten Salts 

 Understanding the Structure, Dynamics, and Chemical Properties of Molten Salts 

 Understanding Fission and Activation Product Chemistry and Radiation Chemistry 

 Understanding Materials Compatibility and Interfacial Phenomena

 Guiding Next-Generation Materials for Molten Salt Reactors 

 Creating a Virtual Reactor Simulation 

References 

1. E. S. Bettis, W. B. Cottrell, E. R. Mann, J. L. Meen, and G. D. Whitman. “The Aircraft Reactor 

Experiment—Operation.” Nucl. Sci. Eng. 2, 841–853, 1957. 

2. J. R. Engel, H. F. Bauman, J. F. Dearing, W. R. Grimes, and H. E. McCoy, Jr. Development Status 

and Potential Program for Development of Proliferation-Resistant Molten-Salt Reactors. ORNL/TM6415, Oak Ridge National Laboratory, 1979 
Molten Salt Chemistry Workshop 

FRD 1 5 

2. FRD 1: Understanding, Predicting, and Optimizing the Physical 

Properties of Molten Salts 

2.1 Background and Current Status 

The concept of a homogeneous, molten salt–cooled reactor was first envisioned in 1948 and its 

development is documented in many reviews [1]. The MSRE, the most significant MSR demonstration 

activity, operated at ORNL from 1965 until 1969. The results from that experiment and subsequent 

research have provided a wealth of material properties data and performance information that are 

thoroughly documented in the open literature. The MSRE was originally fueled with 235U (and later 233U) 

in a molten salt solution composed of 7LiF-BeF2-ZrF4-UF4 of the approximate composition 65-29-5-1 

mol%. To appreciate the derivation of this fuel composition, one must recognize the numerous criteria 

and constraints behind it, which were described in detail by Grimes [2]. At the top level, neutronics 

(parasitic neutron absorption and moderating properties) limit the choice of nuclides that can be 

considered; those limits combine with the limits associated with the mission of the reactor (breeder, 

burner, or converter) and the neutron energies (fast or slow) in the system. 

Figure 3 highlights the neutronic constraints that dictate the choice of elements for thermal and epithermal 

reactors (i.e., low neutron capture cross sections) and the chemical constraints that exert a controlling 

influence on all reactor applications. [3] The key criterion of the molten salt in a fluid-fueled reactor is 

that it must dissolve more than the critical concentration of fissile material (235U, 233U, or 239Pu), dissolve 

a high concentration of fertile material (232Th or 238U), and dissolve all fission products without loss of 

useful properties or subsequent chemical reaction. It also must be stable toward reactor radiation and high 

operating temperature (500–750 °C), survive fission of uranium (or other fissionable materials), and 

possess a low vapor pressure over the entire performance envelope. The fluid must also possess heat 

transfer and hydrodynamic properties allowing it to serve as a heat-exchange fluid without excessive 

power requirements for circulation, and it must be relatively nonreactive with the metal reactor, heat 

exchangers, and moderator material (typically graphite). Finally, to achieve a low fuel cycle cost, the fuel 

should be inexpensive and economical to purify via efficient schemes designed to recover unburned 

fission and bred fissile materials and to remove fission products and neutron poisons. 

The neutronic requirements for coolant salts are distinctly different from those for fuel salts (see 

Figure 2). For a primary coolant salt, low-neutron-absorbing-salts are necessary to obtain negative void 

and temperature coefficients [4]; whereas secondary coolant salts have no formal neutronic requirements, 

as the radiation intensities are significantly less, and the consequences of uranium fission and fission 

products are absent. In both cases, the coolant salt must be stable and compatible with the metal 

construction of the heat exchangers, have good heat transfer and hydrodynamic properties, have a low 

vapor pressure over a wide operating range (400 to over 900 °C), and not undergo a violent reaction with 

the fuel or steam in the event of a leak. 

The requirement that fuel salts be compatible with the container material places a significant constraint 

upon the choice of salt constituents. It means that the cationic constituents of the salt must consist of 

relatively active, electropositive elements (e.g., fluoride compound free energies of formation 

>95 kcal/molF), whereas the container material must consist of elements that are relatively noble in 

comparison (e.g., fluoride compound free energies of formation <75 kcal/molF). [5] This requirement 

eliminates the middle of the periodic chart for consideration as salt constituents and the periphery of the 

periodic chart for consideration as alloy constituents (only a small region in the center is suitable for 

metallic alloys; see Figure 3). Nonmetallic materials (e.g., ceramics and composite materials) can also be 

compatible with a high-temperature salt—for example, graphite is known to be fully compatible with 

molten halides to very high temperatures. Although nonmetallic materials have been used and may find 
Molten Salt Chemistry Workshop

6 FRD 1 

service as core internals, there is currently no joining and fabrication technology for these materials to 

support their deployment as the container material for many of the complex components that will be in 

contact with the salt. 

Figure 3. Early MSR program approach to screening and justifying candidate MSR salts (where t(abs) is 

total neutron absorption cross section). | Image courtesy of Oak Ridge National Laboratory

From a neutronics point of view, non-halide fluid-fueled systems (containing H, B, C, N, O, and Bi), such 

as oxides, carbonates, and nitrates, seem like reasonable choices. But they are either not chemically 

compatible with potential container materials or cannot dissolve enough fissile and fertile material at 

practical temperatures to serve as fluid fuels. Oxides typically have high melting points and are corrosive; 

carbonates also have a fairly high melting point and require a CO2 over-pressure to stabilize them; and 

nitrates are problematic because of thermal decomposition and their stability in a high radiation field. 

Thus, combinations of 7

Li+, Be2+, Al3+, Na+, Mg2+, and Zr4+ cations with F- and 37Cl- stand out for use in 

fuel salts and primary coolants. (Note: 37Cl-

 is preferred because the major chlorine isotope [35Cl] 

produces 36Cl as an activation product, which is an unwanted long-lived energetic beta emitter.) For 

secondary coolant salts, there are no formal neutronic requirements; therefore, isotopic enrichment is not 

needed and so combinations of Li+

, Na+, K+, Rb+, Mg2+, Be2+ and Zr4+ cations with F-, Cl- or BF4

-

 are 

viable [6]. There are limited thermodynamic data for candidate molten fluoride salt systems and even 

fewer data for candidate chloride salts systems (as shown in Figure 4). 

Reactor salt options need to be optimized in the context of applications that address current needs, new 

capabilities, and material developments that have emerged since the MSRE. Over the past half-century, 

experimental, computational, and analytical methodologies have shown enormous advances and can be 

used to explore structure, properties, and reactivities of molten salts in considerably greater detail. In this 

light, this section focuses on the need for state-of-the-art investigations to understand, predict, and 

optimize the physical properties of molten salt systems that could be applied to MSRs and to obtain the 

data necessary to support reactor design/development. 
Molten Salt Chemistry Workshop 

FRD 1 7 

Figure 4. Available thermodynamic data (marked by an x) for binary and ternary fluoride salt mixtures for 

MSRs developed at the Institute for Transuranium Elements of the Joint Research Centre of the European 

Commission, Karlsruhe, Germany. | Capelli, E. Thermodynamic Characterization of Salt Components for Molten Salt 

Reactor Fuel, Fig. 1.2, PhD Thesis, University of Delft, 2016, http://samofar.eu/wp-content/uploads/2016/11/2016_ 

Capelli_Elisa_PhD-thesis.pdf

2.2 Technical Challenges and Research Directions 

2.2.1 Determining the Physical Properties of Molten Salts 

Preparation and handling of high-purity salts 

Although ORNL established processes and practices for purifying and handling fluoride salts for R&D 

and the MSRE, [7] there is no national standard for testing and reporting the purity of the salts. Different 

purification processes and procedures are needed for other salts (e.g., carbochlorination for chloride salts). 

[8] Purification methods are needed not only to eliminate cationic metallic impurities and corrosive 

moisture, oxides, and hydroxides in the hydroscopic halide raw materials, but also to eliminate corrosive 

anionic contaminants such as sulfates, sulfides, and carbonates. It is necessary to positively identify salt 

impurities and quantify the degree of purity achieved. In some cases, this may require going beyond 

previous practices, especially for new salt systems. Such purification protocols and tabulated data sets are 

readily available in the literature for organic reagents and solvents, but complementary data and 

standardized and validated methods do not exist for halide salts intended for nuclear service. The 

development of protocols amenable to the production of high-purity salt will also require scaling to 

production levels and transferring proven methods and defined purity requirements to commercial 

vendors to facilitate more widespread availability of pure, quality-assured salts to accelerate research. 

There are many challenges to the accurate measurement of the high-temperature thermophysical 

properties of MSR salts. First, a wide range of salt and fuel compositions must be considered, including 

fission products and transuranic elements produced during power production. Each component requires 

extensive purification to remove impurities—such as oxides, sulfides, metals and water—which demands 

the use and engineering of specialized containers that not only are compatible with the salt but also will 

rigorously maintain salt purity. For highly accurate determination of salt physical properties, 

contaminants (e.g., atmospheric moisture) must be rigorously excluded following purification and during 

property measurement. To these ends, purification procedures need to be developed, externally validated, 

and published in the open literature; and a single source of pedigreed salt should be established for the 

production of analytical standards for the research community. Meeting those needs would advance our 

knowledge and confidence in the physical properties of the molten salts, since the community would be 

working with materials of similar quality. A primary contribution to the overall scientific and technical 

program would be a supply of pedigreed salt from which institutions could obtain small quantities to 

serve as standards to validate the purity of larger salt quantities. 

LiF NaF KF RbF CsF BeF2 CaF2 LaF3 ZrF4 ThF4 UF4 PuF3 UF3

x x x x x x x x x x LiF

x x x x x x x x x NaF

x x x x x KF

x x x RbF

x x CsF

x x x x BeF2

x x CaF2

x LaF3

ZrF4

x x ThF4

x UF4

PuF3

UF3

Binary

LiF NaF KF RbF CsF BeF2 CaF2 LaF3 ThF4 UF4 PuF3

x x x x x x x LiF-NaF

x x x LiF-KF

LiF-RbF

x x LiF-CsF

x x x LiF-BeF2

x LiF-CaF2

LiF-LaF3

x x LiF-ThF4

x LiF-UF4

x x x NaF-BeF2

x NaF-ThF4

x BeF2-ThF4

Ternary
Molten Salt Chemistry Workshop

8 FRD 1 

For large-scale experiments, analytical-grade salts available from commercial suppliers are of insufficient 

purity to accommodate meaningful investigations of molten salt behavior and the reliable characterization of 

physical properties essential for predictive modeling efforts and fuel qualification. Both phase behavior and 

corrosion responses are especially sensitive to small amounts of impurities. Previous work on the MSRE at 

ORNL revealed that minimal salt loop corrosion was achievable when care was taken to afford a high-purity 

salt that was preserved under air-free and moisture-free environments. Modest corrosion rates (0.013–0.044 

mm/year) have been obtained for purified salts in a closed loop, [9] whereas salts that were used as received 

and remained open to atmosphere corroded containments more than four orders of magnitude more rapidly 

(up to 167 mm/year). [10] This vast difference in corrosion response highlights the need to develop a series of 

best practices in the production, purification, characterization, and handling of molten salts, as well as a 

means by which the pedigree and reliability of the reported data can be assessed by the scientific community. 

The current standard for quality assurance of nuclear fuels and material is Nuclear Quality Assurance-1, 

established by the American Society of Mechanical Engineers. There is a need to establish a quality 

assurance hierarchy for molten salt preparation and characterization with commensurate measurement, 

documentation, and reporting criteria explicitly documented. Such an approach will improve confidence 

in reported results and allow the research community to meaningfully contribute to the collection of highquality data required for fuel validation or development of predictive models. For the purposes of 

ensuring instrument calibration and good quality control practices, a single source of salt should be 

established for the production of analytical standards for purified, pedigreed salt. Additionally, 

experimental documentation and representative metrics of data quality need to be accessible for external 

reproducibility, validation, and verification. An example of such a system is the Cambridge 

Crystallographic Database, which serves as a free repository for single crystal data sets. Crystal data can 

be independently analyzed before they are uploaded, enabling the creation of a “checkcif” file that 

articulates any suspicious metrics in the proposed structure. These documents are peer-reviewed before 

publication of the crystal structure and preserved for subsequent inspection as needed. A similar review, 

documentation, and archival infrastructure is necessary to enable the reliable determination of salt 

properties and to facilitate future molten salt characterization and MSR development. 

Defining phase diagrams of molten salts 

The primary tool for understanding fuel and coolant salts is the phase diagram—as it defines the 

temperature-composition space for single-phase liquids; the order and identity of equilibrium phases that 

crystallize as the melt freezes; and the solubility of constituents that can grow, such as transuranics, 

fission products, corrosion and activation products, and impurities. A variety of phase transitions occur in 

molten salts: in a few instances, the melt and frozen solid have an invariant composition throughout the 

freezing process (i.e., the line-compound composition); in most instances, the transition is more complex 

and the frozen solid is different from the melt composition in a variety of ways; and in rare instances, two 

immiscible liquids can form or a melt may cool to a metastable phase that remains liquid below the 

indicated melting temperature of the mixture. [11]

A large number of binary and ternary phase diagrams have been measured for chloride and fluoride salt 

systems, [12] and a small number of higher-order systems have been characterized over the years (see 

Figure 5). The methods used to determine the phase diagrams include differential scanning calorimetry, 

electrochemical methods, and solidification along a temperature gradient [13–16]. Since a number of the 

species are volatile, their vapor-liquid and vapor-solid equilibria need to be known, along with additional 

thermodynamic data, to provide a complete understanding of the phase behavior.
FRD 1 9 

(a) Calculated LiF-BeF2 pseudo-binary phase diagram with 

fixed concentration of UF4 (2.55 mol%) and ThF4 (19.95 mol%). (b) The liquidus projection of the LiF-BeF2-Th

(c) Calculated entropy of mixing: BeF2-AnF4 (An=Th, U) 

1000 K and 1400 K. (d) Optimized BeF2-UF4 phase diagr

Figure 5. Representations for integrating phase diagram understanding. |Reprinted from J. Nucl. Mater. 449, E. Capelli, O. Ben

“Thermodynamic assessment of the LiF–NaF–BeF2–ThF4–UF4 system.” 111–121, 2014, with permission from Elsevier.

ideal 

BeF2-UF4, 1000K, 1400K 
Molten Salt Chemistry Workshop

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Previous methods of determining phase diagrams of molten salt mixtures used a relatively large number 

of samples and were highly labor intensive. The rigorous control and measurement of impurities—

including dissolved metals, oxygen-containing impurities (e.g., oxides, hydroxides, sulfates, carbonates, 

oxyhalides), water, and other types of impurities—were not always reported. Therefore, it is difficult to 

ascertain the accuracy of some of the existing phase diagrams in the literature. For some systems, there 

are discrepancies among various studies that have yet to be reconciled. These flaws reinforce the need for 

a source of high-purity salts and for the publication of complete analytical approaches and experimental 

practices used to obtain data sets. 

The phase diagram is also a manifestation of the thermodynamics of the system, as equilibrium is 

determined by the Gibb’s free energy of the system. Thermodynamic properties of each of the 

components (including their heats and entropies of fusion) and mixing properties within phases of varying 

compositions define the Gibb’s free energy. If the thermodynamic properties are known, the phase 

diagram can be calculated, using various established formalisms [17]. Conversely, thermodynamic 

properties can be extracted from the phase diagram. 

The thermodynamic properties of a system are determined by the atomic-level structure and inter-atomic 

interactions. Modern spectroscopic techniques (as described in Section 3) can provide insights into the 

structure and dynamics of molten salts and can be used to validate computational models. Ab-initio 

calculations can be used to guide the prediction of thermodynamic properties and identify the appropriate 

forms of equations for calculating phase equilibria. Comprehensive databases, which are just beginning to 

be developed from data produced decades ago, can be used for data mining and for compiling and 

reconciling data to gain insights into thermodynamics properties of molten salts and develop more 

complex thermodynamic models. An improved way to query and visualize phase diagram and 

thermodynamic databases is recommended, which would support more strategic thinking about the 

development of new chemical systems and help in developing more comprehensive thermodynamic 

models. 

Computational methods and databases can accelerate the collection and analysis of thermodynamic data 

and phase diagrams. As a starting point, we need to assess prior studies and identify missing data, 

compositions, and thermodynamic accuracy gaps in the experimental data for binary, ternary, and higher 

order multi-component systems. Recommended research directions include the following:

1. Use computational studies to define where additional experimental data are most needed. 

Experimental data should be taken in regions where the activity coefficients vary sharply, where there 

is evidence for short-order and complex-ion formation, from computational and experimental studies 

of salt structure and speciation, and for new systems under consideration. 

2. Develop advanced methods of high-throughput data collection, with well-quantified uncertainties, 

that are applicable to air- and water-sensitive high-temperature salts containing radionuclides and 

other toxic elements. For example, high-throughput phase diagram determination might entail 

spectroscopic interrogation of an array of simultaneously heated samples, or of a sample of given 

composition exposed to a temperature gradient. Approaches for obtaining meaningful data for salts 

containing volatile fission and activation products are also needed; experiments performed under 

modest pressures may overcome some of these challenges, but such an approach also requires 

validation against known phase diagrams. 

3. Develop high-purity calibration standards of known pedigree for use in analytical studies using 

multiple experimental methods for thermodynamic data collection. It is critical to validate newly 

developed methods on well-characterized simple systems, as well as to include quality control 

samples during data collection in compliance with good analytical practices. 
Molten Salt Chemistry Workshop 

FRD 1 11 

4. Identify and investigate potential metastable phases induced by ionizing radiation and compositional 

changes due to fission, activation, and decay. 

5. Validate density functional theory (DFT) and ab initio molecular dynamics (AIMD) calculations 

performed on known salts with experimental data and spectroscopic characterization, followed by 

modeling of thermodynamic properties of systems that are new or difficult to measure 

experimentally, (e.g., containing higher actinide activation products). 

6. Develop appropriate thermodynamic models for use as a predictive tool, enabling investigation 

beyond conditions that can be experimentally measured (e.g., short-lived fission products or highly 

activated sample). Database development requires selecting appropriate theoretical models for each 

specific system, then computationally optimizing the input thermodynamic parameters to match 

available experimental data points. One of the challenges is development and application of the 

appropriate thermodynamic model for each specific system. The question of appropriate entropies of 

mixing is a particular challenge because of the complex speciation, and the possibility of short- to 

mid-range molecular order in the salt. 

7. Develop tools to query salt property databases and visualize information in a way that is accessible to 

a variety of users and most useful in design and safety analysis. Recent developments in modern 

computing, data visualization, and user-interface tools need to be translated to those ends. The ability 

to trace the results of a query to the initial data sources, accompanied by experimental details, can 

take advantage of modern database and content management tools. Uncertainty quantification, error 

propagation, and sensitivity analysis should be integral to these models. Such analysis will also 

indicate where more experimental data are needed. The user interface will expand the community, 

which will also help with prioritization of data needs. 

Thermodynamic and thermophysical property determination 

In addition to phase behavior, other important physical properties need to be measured for the individual 

salts and mixtures of salts, including melting point, density, viscosity, heat capacity, thermal conductivity, 

vapor pressure, heat of fusion, expansivity, compressibility, surface tension, and gas and fission product 

solubilities. [18] Some physical properties (such as solubility, vapor pressure, and viscosity in the 

neighborhood of glass-forming domains) are sensitive to slight compositional changes; and the properties 

change rapidly, further underscoring the need for established purification protocols, high-purity salts, and 

rigorous analytical practices (Section 2.2.1). Given the number of measurements that need to be made, 

high-throughput experimental techniques that can be miniaturized and are amenable to operation in a 

glovebox environment could greatly accelerate property measurements and technology development. 

However, the need for high-throughput measurements must be balanced against and coordinated with the 

need for high-precision reference measurements. This subsection provides examples of some state-of-theart reference measurement techniques that may be miniaturized for high-throughput measurements. This 

discussion is not comprehensive but provides some examples of measurement techniques to inspire the 

reader. It must be stressed that most of these property measurement capabilities for extreme temperatures 

and controlled atmosphere are not off-the-shelf capabilities. They are typically customized systems with 

specialized procedures that are not standardized. 

Calorimetry 

Calorimetric methodology and sensitivity have been improved over the past three decades. A calorimeter 

suitable for high-temperature solution calorimetry is now commercially available and is being used in 

several laboratories at both universities and national labs. Heats of formation and transformation can 

typically be measured with an accuracy of ±1–3 kJ/mol for sample sizes of 10–100 mg. Heats of mixing, 

heats of fusion of binary and ternary fluorides and chlorides, heats of formation of crystalline ternary 
Molten Salt Chemistry Workshop

12 FRD 1 

compounds, and heats of dissolution of fission products and other impurities in molten salts as a function 

of both molten salt composition and solute concentration can all be measured by calorimetric methods. 

However, the precision and accuracy of such measurements will depend critically on the purity of the 

starting materials. 

Differential scanning calorimeters (DSCs) can provide heat capacities, melting points and heats of 

fusion/crystallization. Their accuracy in heat capacity measurements is of the order of 1–2% at room 

temperature to 500 °C and 2–5% at 500–1200 °C. Heats of phase transition and fusion can be determined 

at an accuracy of 1–5%, depending critically on appropriate calibration standards. Temperature accuracy 

below 1000 °C is generally on the order of ±1 °C. There are opportunities to adapt these methods to 

radionuclides and air- and water-free environments. Sample containment—especially because of sample 

creep along container walls—and materials compatibility pose challenges. New high-throughput methods 

on multiple samples, perhaps combining calorimetric and spectroscopic techniques, should be developed. 

Viscosity Measurements 

A number of different methods (e.g., rotational, capillary) have been used for molten salt viscosity 

measurement. [19] The miniature rolling-ball viscometer (Figure 6) uses an induction coil for rapid heatup and measurement of the rolling ball velocity (μ) as a function of inclination angle (θ). With this 

method, it is possible to determine the viscosity of molten salts in a range of 1–5 mPas at temperatures of 

up to 1400 °C. However, a set-up needs to be designed to use less than 10 mL of flux and use 

interchangeable tubes with rolling balls. 

Figure 6. Miniature rolling-ball viscometer description. | Image courtesy of Prof. Uday Pal, Boston University]

Thermal Diffusivity and Thermal Conductivity 

Optical techniques based on both emission spectroscopy and modulated heating have the potential to 

measure radiative and thermophysical properties that govern temperature in a molten salt. To 

determine thermophysical properties, such as thermal diffusivity or thermal conductivity, thermal lens 

spectroscopy can be implemented. [20] This technique uses an acousto-optical modulator to provide 

modulated laser heating of the liquid at a small focal point. A probe laser beam is passed through the focal 

region and is affected (broadened or deflected) by the change in refractive index caused by the heating 

beam. This change produces in turn a change in the detected laser intensity, and the resulting signal can 

be fit to heat transfer models to determine the thermal conductivity and thermal diffusivity of the fluid. 
Molten Salt Chemistry Workshop 

FRD 1 13 

The thermal lens technique has the added benefit of being able to provide an independent measurement of 

the absorption coefficient of a fluid by a rearrangement of the lasers and detectors. [21] 

Molar Volume and Electrical Conductivity Measurement 

Combined density and electrical conductivity (ionic and electronic) can be measured using coaxial 

electrodes that measure AC impedance spectra and DC currents as a function of electrode depth in the salt 

(Figure 7). Using the cell geometry and the measurement data, salt height at temperature, molar volume, 

and ionic and electronic conductivities can be determined. Furthermore, the partial ionic conductivities 

and the chemical diffusivities of species can be extracted from separate transference number 

measurements. [22] 

Figure 7. Electrical method for measuring molar volume, ionic and electrical conductivity. | Reprinted from 

S. L. Schiefelbein, N. A. Fried, K. G. Rhoades, and D. R. Sadoway. A high-accuracy, calibration-free technique for 

measuring the electrical conductivity of liquids. Review of Scientific Instruments, 69(9), 3308–3313, 1998, 

(http://dx.doi.org/10.1063/1.1149095), with the permission of AIP Publishing.

Surface Tension and Contact Angle 

A precise bubble pressure technique can be designed to measure surface tension by immersing an 

appropriate metal tube in the selected molten flux, blowing gas through the tube to acquire a surface 

tension value between the gas and the liquid salt, and examining the peak pressure exerted on the gas by 

the liquid (Figure 8). In many cases, contact angle is the key piece of information for understanding 

interfacial behavior (foaming, intrusion into pores, and wetting surfaces). Additional techniques that can 

yield both contact angle and surface tension are the sessile drop technique and the Langmuir-Wilhelmy 

plate method. The sessile drop method uses a very high–shutter speed camera and an optical setup to take 

an image of a molten flux piece resting on an appropriate substrate. This allows measurement of a contact 

angle at a three-phase interface (Figure 8). Using this contact angle, the solid-liquid surface tension can be 
Molten Salt Chemistry Workshop

14 FRD 1 

estimated. Each of these methods has advantages and disadvantages that will need to be considered based 

upon the system being investigated, the type of information desired, and the required accuracy. [23] 

Figure 8. Interfacial measurement methods. (a) Bubble-pressure method for measuring surface tension. (b) 

Sessile-drop method for measuring contact angle and surface tension. | Image courtesy of Prof. Uday Pal, Boston 

University

Volatility and Melting Point 

The melting point and volatilization rate of the salt as a function of temperature can be measured using a 

simultaneous DSC and thermogravimetric analysis (DSC/TGA) balance. To ensure salt purity is 

preserved, the instrument needs to be designed for operation in an inert-atmosphere glovebox. Although 

the recorded weight change of the TGA can be attributed to salt volatilization, the weight change is 

dependent upon salt purity (Section 2.2.1). Volatile species can be determined, both for understanding salt 

properties and for validating salt purity, by quadrupole mass spectrometer analysis or by condensing 

volatilized species on a water-cooled substrate with subsequent analysis by x-ray fluorescence, x-ray 

diffraction (XRD), or other techniques. [24] For measuring low vapor pressures, the traditional 

transpiration method is reliable and accurate and can be used as a reference method in this domain. It does 

not require significant deployment of sophisticated instruments within an inert envelope. [25] 

Solubility 

Small, thin capillary tubes containing the salt can be rapidly heated and quenched. The tubes with the salt 

can be placed in contact with the compound for which the solubility needs to be measured. Both the 

solubility and the diffusivity of the dissolving compound can be estimated by measuring the diffusion 

profile of the quenched sample (Figure 9). 

EMF Technique for Chemical Activity and Phase Diagrams 

The electromotive force (EMF) generated by electrochemical cells can be used to measure the partial 

molar Gibbs energies, activity, and activity coefficients, and to determine the phase diagram. For 

example, the phase diagram of Ba-Bi system was determined by the EMF technique using a CaF2-BaF2

binary electrolyte at 450–800 °C. [26] This technique is potentially powerful, as it is possible to define 

both the thermodynamic properties of the system and the phase diagram that will delineate liquid-state 

solubility, solidus temperature, and phase fraction/composition.
Molten Salt Chemistry Workshop 

FRD 1 15 

Figure 9. Rapid-quenched capillary setup for specie solubility and diffusivity. | S. Su, T. Villalon Jr., U. Pal, and 

A. Powell. “Techniques for measuring solubility and electrical conductivity,” pp. 465–475 in Advances in Molten Slags, 

Fluxes, and Salts: Proceedings of The 10th International Conference on Molten Slags, Fluxes and Salts (MOLTEN16), Eds. 

R. G. Reddy, P. Chaubal, P. C. Pistorius, and U. Pal. TMS (The Minerals, Metals & Materials Society), 2016. Used with 

permission of Springer.

2.2.2 Predicting the Chemical Speciation, Structure, and Dynamics of Molten Salts 

Solutions 

Computational salt chemistry is a major scientific challenge because of the complex nature of the salts. 

Although there has been prior work showing the value of molecular modeling with interatomic potentials 

to predict molten salt properties, [27] only a few studies of molten salts have used a fully first principles 

approach capable of explicitly treating polarization effects and chemical reactivity. [28] Looking forward, 

optimized approaches are needed which employ first-principles MD simulations of a molten salt system 

that are validated against existing experimental data. To develop those approaches, the basic 

thermodynamic and kinetic properties of molar volume, thermal expansion, bulk modulus, and 

diffusivity, for a molten salt at multiple temperatures, need to be computed. Simulations for pure molten 

chloride/fluoride salts at multiple temperatures are needed to obtain temperature-dependent 

thermodynamic, kinetic, and transport properties; and speciation information is clearly needed for the 

prediction of salt phase diagrams and other physical properties. In addition, knowledge of the constituents 

is needed to understand interfacial behavior leading to MSR material degradation, such as corrosion. 

Another complexity in computational salt chemistry is that the atomic composition of the salt changes as 

a function of time as a result of fission product formation and the emission of α, β, and γ-rays, which can 

lead to radiolysis in solid salts. In the molten salt, our understanding of the transient chemistry associated 

with fission and decay is limited, but what has been observed over longer time periods is that much of the 

energy that is deposited in the salt works to restore the salt to an equilibrium composition; and few 

structural changes are observed in the salt. Compositional and redox changes arise as results of 

transmutation and fission (loss of fissile species, grow-in of soluble fission products). Chemical evolution 

of the insoluble fission products, at all time scales, remains an open question Computational electronic 

structure and MD methods can be used to extend the limited experimental data sets to cover a broad range 

of parameter space. However, these predictions must have the appropriate accuracy required for upscaling 
Molten Salt Chemistry Workshop

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across multiple time and length scales. For example, nucleation is very sensitive to the rate constants 

used. An error of 0.3 kcal/mol in a reaction step at the MSR operating temperature of ~ 600 °C can lead to 

an overall error in the equilibrium prediction by a factor of about 850 over 40 steps. [29] 

Furthermore, the importance of experimental validation of computationally-predicted results cannot be 

overemphasized. In previous research studies, computational methods were applied that incorrectly 

predicted molten salt structure and were unable to predict the intermediate structure that was later 

demonstrated experimentally. [30] The power of predictive approaches is that they provide the ability to 

investigate systems under time scales and conditions that are not experimentally accessible; therefore 

demonstration of performance accuracy under experimentally testable conditions is an important 

prerequisite for computational development. Techniques suitable for physical characterization are 

discussed in detail earlier, and spectroscopic approaches capable of providing structural details at the 

atomistic and intermediate length scales are discussed in Section 3. 

Successful development and validation of this computational research approach will enable a new 

capability that goes beyond the traditional approach to optimizing molten salt composition with respect to 

fuel and fission product solubility, thermophysical properties, and in-reactor salt degradation. It will also 

minimize corrosion/degradation of materials of construction for the reactor, pump, and heat exchanger 

components. 

Building Accurate Multi-component Models for Prediction of Phase Diagrams 

The phase equilibria and thermochemistry of fluoride and chloride molten salts can be well represented 

by existing modeling frameworks (Figure 10). While free energies and solution parameters for a number 

of the proposed components are available, numerous critical elements are missing in the datasets. Highthroughput first-principles techniques and experimental methods need to be developed to efficiently 

provide phase equilibria and thermochemistry information. The models will then need to be expanded to 

include any additional components, and validated experimentally. Solution models for viscosity will also 

need to be developed, possibly based on the demonstrated successful correlations in oxide liquids. 

Importantly, parameters such as constituent solubility, oxidation state, and coordination environment need 

to be experimentally determined to allow theoretical models to accurately reproduce and predict the 

physicochemical behavior of the melts. 

Applying Molecular Dynamics Simulations to Predict Thermophysical and Transport 

Properties 

It is important to be able to accurately predict and understand liquid structure and dynamics both in the 

bulk and at salt/material or salt/vapor interfaces. MD simulation can help explain transport properties, 

including viscosity and the origin of different length-scale features in the x-ray structure function, S(q), as 

well as real-space radial distribution functions. It can also be particularly useful in understanding the 

effects of structural voids and defects formed under radiolytic conditions, as well as determining their 

effects on material transport and thermal conductivity. This is particularly important for understanding 

possible structural heterogeneities in the molten salt system. MD simulation can be used to understand 

wetting and permeation at degraded interfaces where surface energies are important, and to describe salts 

under confinement within graphitic or other structural materials when thermodynamic models are 

insufficient. AIMD will be needed for processes in the condensed phase in which reactivity may be 

important. 
Molten Salt Chemistry Workshop 

FRD 1 17 

Figure 10. Illustrative phase diagram of invariant points and liquidus projections for a fluoride fuel salt with a 

minor amount of Pu and low concentrations of Cr, Ni, Mo. (Diagrams and data such as this will be needed 

for MSR complex salt chemistries.) | Image computed from existing molten salt databases; courtesy of Prof. Ted 

Besmann, University of South Carolina

A simplistic view of a MSR salt fuel is a set of cations, which are changing their identity and their 

oxidation states (through nuclear transmutation and fission), embedded in a sea of fluoride or chloride 

anions. The interactions of the cations with the anions is thus quite complex, especially considering 

dynamic oxidation states. The current approach is to fit a classical potential model to a set of results from 

first-principles electronic structure calculations; if other liquid state data are available, such as densities or 

diffusivities, they can also can be used to improve parameterization of the model. This approach can work 

well if enough configurations can be sampled and there are no changes in the oxidation state or the 

valency (number of “bonds” to a center). However, modeling more complex systems will require many 

different force field interactions, with a concomitant increase in complexity. Thus, to treat the atom-atom 

interactions, different approaches are needed that are not necessarily biased by the input force field and 

can account for changing oxidation states and the presence of open shells due to unpaired electrons. One 

approach would be to employ fast electronic structure methods for the atom-atom interactions, for 

example, semi-empirical molecular orbital theory (SEMO) [31] and tight binding DFT (TB-DFT). Such 

methods require parameterization, but they can be tuned for the atom and oxidation state in a 

straightforward manner by comparison to high-level first-principles electronic structure calculations at the 

correlated molecular orbital theory or DFT levels. [32] Even though these approaches have the advantage 

of being able to deal with changes in electronic structure, they have some disadvantages: (1) the time 

scales that can be simulated are short; and (2) calculations of liquid structural properties, such as the x-ray 

structure function S(q), and dynamical properties, such as diffusion constants and viscosity, are not 

guaranteed to be superior to those derived from classical simulations. Therefore, experimental validation 

is an irreplaceable aspect of atomistic modeling. 
Molten Salt Chemistry Workshop

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Computational approaches also need to be developed to bound uncertainties in the results obtained from 

simulations of molten salt systems to provide useful information for system models. It is critical to 

provide error bars from the computations either through careful analysis of each term in the modeling, as 

has been done for first-principles electronic structure of thermodynamic properties, or through proper 

sampling. Current approaches for the latter case are usually based on Monte Carlo sampling, which is 

computationally very expensive. New mathematical approaches are needed for more efficient uncertainty 

analysis. 

Ultimately, quantum mechanics, classical dynamics, coarse-grained models, continuum models, and 

thermodynamic models must be integrated to achieve predictive insights ranging from chemical reactivity 

at the atomistic level to bulk thermophysical and transport properties. Computational studies will require 

large simulation sizes on the orders of tens of millions of atoms to cover the required spatial scale. Such 

simulations will require access to DOE’s highest-performance computers alongside optimized exascale 

software. For example, new and proposed heterogeneous computer architectures, such as graphics 

processing units, may require improvements in the software to attain optimal performance. 

Impact of Advanced Modeling 

The ability to model physicochemical properties of molten salts and predict optimal compositions, 

enabling a “materials by design” approach, would dramatically increase the rate of scientific advance. 

Identification of promising melt compositions could be performed predictively, allowing for experimental 

validation of promising compositions, rather than through brute-force approaches requiring large 

investments of time, money, and material. Furthermore, predictive modeling allows for the interrogation 

of conditions beyond what can be investigated by traditional experimental approaches, such as prolonged 

exposure (years to decades) to elevated temperatures and extreme radiation fields. 

A core capability will be the ability to design molten salts from a combination of simulation and 

experimental results with the appropriate chemical and physical properties that will provide optimal 

operation in a MSR. This will include the choice of an appropriate combination of anions and cations with 

the best set of properties to provide reactors that have long lifetimes and can be safely operated with 

maximal efficiency. An advantage of simulations is their ability to model complex mixtures with many 

cations in different compositions, covering a much broader range of parameter space than can be done 

experimentally in a shorter time. 

Addressing these scientific challenges will enable scientists and engineers to improve the modeling and 

simulation of MSRs, including the underlying physicochemical phenomena captured in the transport 

properties used in the continuum theories of momentum, mass, and energy transfer used to design nuclear 

reactors. This will allow predictions of effective diffusivities of fission products in complex molten salts, 

thermal conductivity, and viscosity, which are parameters used in the traditional modeling and simulation 

approach for reactor design. It will enable prediction of the nature and fate of volatile products, including 

release rates. The data generated can be used to extrapolate limited experimental data into broader areas 

of parameter space for use in reactor design, including new and improved materials. They will be able to 

predict the nucleation and growth of insoluble products within a flowing molten salt and be able to bound 

the uncertainty in our knowledge of molten salts. 

The ability to predict the chemical and thermophysical properties of molten salts over a broad range of 

temperature and composition will have significant impact on fields beyond MSR. A near-term goal 

should include the development of a global thermodynamic model of MSR salts that is valid under a 

broad range of conditions. This will be of great value to the reactor design community, and these 

approaches can be applied to the design of other reactor systems, as well as complex industrial processes. 

The tools and approaches developed for the design of molten salts will help enable the prediction of 
Molten Salt Chemistry Workshop 

FRD 1 19 

chemical properties under ionizing radiation conditions for other materials, in particular because of the 

necessary validation of computational data against an extensive experimental data set. The development 

of new MD approaches is required for simulations of complex systems over long times; and new 

approaches for the development of force fields and fast electronic structure methods will be broadly used 

by the simulation community for other systems, especially for materials and ionic liquids. [33] Finally, 

the ability to reliably predict the solubility of fission products in salts will lead to new concepts of 

solvation, which will be broadly used by the MSR community and others investigating the diverse 

industrial applications of molten salts.

References 

1. M. W. Rosenthal, P. R. Kasten, and R. B. Briggs. “Molten-salt reactors—History, status, and 

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24. U. Pal, S. Su, and T. Villalon. “Molten flux design for solid oxide membrane-based electrolysis of 

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28. A. Bengtson, H. O Nam, S. Saha, R. Sakidja, and D. Morgan. “First-principles molecular dynamics 

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29. S. M. Kathmann, G. K. Schenter, B. C. Garrett, B. Chen, and J. I. Siepmann. “Thermodynamics and 

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22 FRD 1 

structure in solid and liquid CuBr probed by multiple-edge x-ray absorption spectroscopy.” Phys. 

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Molten Salt Chemistry Workshop 

FRD 2 23 

3. FRD 2: Understanding the Structure, Dynamics, and Chemical 

Properties of Molten Salts 

3.1 Background and Current Status 

The structural and dynamical properties of molten salt systems have been investigated at varying levels of 

detail over the past five decades, resulting in established techniques amenable to the characterization and 

interrogation of MSR-relevant salts. Methods that have great utility include ultraviolet-visible-infrared 

(UV-vis-IR) spectroscopy, x-ray absorption spectroscopy (XAS), Raman spectroscopy, and nuclear 

magnetic resonance (NMR) spectroscopy. These methods provide information on (1) oxidation states of 

transition metals, lanthanides, and actinides and (2) local structure and bonding of species in solution. 

Complementary methods that apply to crystallized salts and salt mixtures in both the solid and liquid 

states include XRD, neutron diffraction, and x-ray absorption fine structure spectroscopy (XAFS), with 

current instrumentation and sampling methods permitting studies at temperatures of up to 1000 C or 

even higher in some cases. Application of these methods yields direct structural information, including 

inter-atomic distances and coordination numbers. Additional information about cation-anion interactions 

can be gained from NMR spectroscopy of ions like 7

Li+, 19F-, 11B (in BF4

-

), 23Na+, 27Al3+, selected 

lanthanides (e.g., 139La3+) and actinides (e.g., 235U3+,4+) cations, and other relevant NMR active ions. 

An idealized perspective of the structure of a 

molten salt is conceptualized as a set of cations 

embedded in a sea of halide anions, typified by the 

binary salt NaCl. Short-range structural properties 

are attributable to competition between coulombic 

interactions and Pauli repulsion, inducing strong 

ordering effects and determining the closest 

interatomic distances, respectively. A corollary 

resulting from the influence of these contrasting 

phenomena is that the innermost coordination shell 

will always be completely made up of oppositely 

charged species, affording an intermediate-scale 

ordering that extends to several solvation shells. [1] 

Such conclusions were validated by neutron 

scattering measurements performed on molten 

NaCl for three different isotopic compositions of 

Cl-

, shown in Figure 11. (Note the difference in the 

structure factor profile as a function of isotope.) 

The key finding here is that no evidence exists for 

covalency or effects associated with polarization 

(consistent with the fact that molten NaCl does not 

exhibit a Raman spectrum). The coordination 

number for unlike ions (first coordination shell) 

was determined to be 5.8  0.1 and the coordination number for like atoms (second shell) was found to be 

13.0  0.5; both quantities are consistent with values for their solid-state counterparts, i.e., in this simplest 

case, the melt is close to isostructural with the solid. 

However, introduction of any complexity beyond the aforementioned idealized system, such as 

substitution of a divalent cation for a monovalent Na+

, results in structural effects that remain difficult to 

understand and fully rationalize. While the identity of the inner sphere coordinating atoms remains as 

anticipated, variation of the inner-sphere cation-anion bond lengths is observed, and different outer sphere 

Figure 11. Total structure factor, F(Q), data for liquid 

Na35Cl (curve A), NaCl (Curve B), and Na37Cl 

(Curve C) at 875 C. The points are the 

experimental data and the curves are derived from 

the partial structure factors | Reproduced from F. G. 

Edwards, J. E. Enderby, R. A. Howe, and D. A. Page. “The 

structure of molten sodium chloride.” J. Phys. C: Solid State 

Phys. 8(21), 3483–3490, 1975, http://iopscience.iop.org/ 

article/10.1088/0022-3719/8/21/018/pdf 
Molten Salt Chemistry Workshop

24 FRD 2 

cationic species arise from the interplay of multiple atomic interactions. Polarization-induced intermediate 

structure can also occur, resulting in edge-sharing of the first-neighbor coordination shell. Such 

phenomena are further influenced by the energy of the outer shell of the atoms, with formation dependent 

upon the concentration of the salt and the number of cations present in the outer sphere. For example, the 

structure of a ZnCl2 melt was characterized by neutron diffraction and XRD, revealing a tetrahedral firstshell coordination environment for Zn2+, where corner sharing of adjacent octahedrals resulted in 

intermediate ordering. [2] For atomistic simulations to reasonably represent this system, unrealistically 

large dispersion interactions were required among the cations, [3] which were reconciled by the 

introduction of a component representing polarization of the ionic electron cloud. [4] 

It has also been demonstrated that for composite salts, the melt does not rearrange into a simple 

combination of component parts. For example, XAFS spectroscopy was used to investigate the melting of 

Rb2ZnCl4 [5] in comparison with the liquid structure of ZnCl2 and RbCl. While molten ZnCl2 displayed 

tetrahedrally-coordinated corner-sharing chlorines, forming the extended network discussed above, RbCl 

displayed significant distortion of the chloride shell around Rb, indicative of significant movement of Rb+

and Cl-

 in the molten state. [6] However, in molten Rb2ZnCl4, isolated [ZnCl4]

2- units were identified with 

a chlorine coordination of 7.6 around Rb+

, in contrast to a chlorine coordination of 4.8 in molten RbCl.

Thus, the salt structure formed by their combination does not afford the same metal coordination 

environment for Rb+

 and Zn2+ as in RbCl and ZnCl2, respectively. 

In the above example, the room-temperature solid state structure of Rb2ZnCl4 was similar to that in the 

molten form; however, this is not universally true. In other instances, atomic rearrangement occurs, so 

that the molten structure differs dramatically from that of the solid precursor. For example, Crozier and 

colleagues investigated the liquid structures of [Bu4N][MnBr3] and [Bu4N]2[MnBr4], identifying a 

bromine coordination number of 3.1 around manganese in the liquid state, compared with a coordination 

number of 6 in the solid-state structure. [7] Molten PbCl2 has also shown a change in local coordination 

environment upon melting, with a decrease in chlorine ligation from nine to six. [8] 

Studies of lanthanide halides have also revealed unanticipated structural changes. Upon melting, LaCl3

displays a first shell coordination number suggesting an equal mixture of [LaCl7]

4- and [LaCl8]5-. A 

decrease in the La3+-Cl-

 bond length has also been observed, contrary to what would be anticipated at 

elevated temperatures. [9] Insight into the latter comes from work performed on an AgBr salt. It revealed 

a similar first-shell bond length contraction upon melting, while radial distribution functions of atomic 

positions revealed greater asymmetry toward longer distances and indicated the average bond length to be 

increasing, as is expected from thermal expansion. [10] Such behavior suggests in both instances the 

metal is more appropriately conceptualized as forming MClx molecules with the anion than as isolated 

Mx+ dissolved in a Cl-

 solvent. Proceeding down the periodic group revealed significant structural 

differences in YCl3 compared with LaCl3. The Y3+ – Y3+ distance was 0.25 Å closer than the analogous 

La3+ – La3+, which was attributed to different linkages in the different salt. Whereas LaCl3 displays 

corner-sharing six-fold complexes, YCl3 is proposed to have edge-sharing complexes. [11] Thus, 

structural properties can vary dramatically based on the species of the cation, which has been previously 

demonstrated to significantly influence chemical properties, such as solubility of actinide complexes. [12] 

While molten salts are regarded as purely ionic solutions, the formation of MClx species upon melting, as 

discussed above, suggests covalent interactions are also a significant, albeit unanticipated, possibility. 

Further work was performed by Okamoto and colleagues in which YCl3 and YBr3 were dissolved in a 

LiCl-KCl eutectic with LiBr, respectively. Investigation of the extended XAFS spectrum of Y revealed a 

sharpening of the first major feature and a small decrease in bond length. These results are consistent with 

improved stabilization of a YX3 molecular complex upon further dilution with the alkali salt. [13] Copper 

and silver halides have also been shown to display significant covalent behavior. Work by Di Cicco and 

colleagues indicates nearly total covalent bonding observed in solid CuBr, which is similar to what is 
Molten Salt Chemistry Workshop 

FRD 2 25 

observed in the liquid state. A quasi-continuous phase change occurred during melting, which afforded a 

significant decrease in Cu-Br bond length in the liquid compared with the solid. [14] A similar structure 

for liquid CuI was also determined, supporting the same conclusion. [15] Studies of AgCl-AgI mixtures 

reveal silver coordination environments composed of two chloride and two iodide atoms at temperatures 

below the melting point; but upon melting, the chloride contribution decreased significantly while the 

iodide was invariant. This decrease is thought to have occurred as a result of the removal of chloride 

atoms from the first coordination shell as the bonding strength between silver and iodide increases. [16] 

Chemical bond lengths are representative of bond strengths, indicative in turn of the chemical reactivity 

of the molten salt. This is clearly demonstrated by Cho et al., who investigated the covalency of 

uranium(III) and uranium(IV) chlorides and observed that as the U-Cl distance became shorter, the 

interactions became stronger and the reactivity decreased. [17] As noted above, the molecular structure of 

the melt is known to influence the solubility and chemical reactivity of the constituents. Accordingly, 

detailed structural knowledge of molten salts is essential for understanding and predicting the chemical 

behavior in a MSR, not only to determine bulk physicochemical properties but also to understand the 

solubility and behavior of fission, activation, and corrosion products present in the melt. 

In addition to the majority constituents in the starting salt, it is necessary to investigate, identify, 

understand, and predict the multi-faceted behaviors of the accumulating fission products. There are 

numerous examples in the literature of spectroscopic and diffraction studies that are directly applicable to 

this research direction; and the behavior of the “sometimes soluble” niobium, tellurium, and insoluble 

noble metal fission products (elements between niobium and tellurium) are particularly important. For 

example, Toth et al. measured the UV-vis-IR absorption spectrum of telluride species in molten Li2BeF4. 

Their work showed that although Te2- was not detectable in Li2BeF4 by optical absorption spectroscopy, 

when the Te2- ion was mixed with metallic tellurium, a soluble poly-telluride species formed that could be 

detected in the absorption spectrum of the molten salt solution [18] (Figure 12a). This study provides 

insight into the solvation properties that could affect tellurium migration in an MSR coolant salt system. 

In light of the preference for using low-enrichment fissile constituents, very high actinide salt 

concentrations are needed, affording salts with very different chemical properties and coordination 

chemistry from dilute actinide-containing solutions. For example, the Raman spectrum of molten LiFNaF-ThF4 at 650 C shown in Figure 12b sheds light on the coordination behavior of Th4+ in molten 

fluorides. Eight-coordinated Th4+ was identified in melts with excess fluoride ions by comparing their 

spectra with that of crystalline K5ThF9. The existence of seven-coordinated Th4+, present in fluoride-iondeficient melts, was identified by shifts in the frequency of the 474 cm1

 band with changes in melt 

composition. This represents a case in which the coordination chemistry of an actinide cation is affected 

by solvent composition. 

High-temperature solid state and liquid NMR can provide insights into the local structure of species in the 

melt and interactions of the cation-anion. For example, in the high-temperature 19F NMR spectra for solid 

and molten YF3-LiF (Figure 13), there is a substantial down field shift in the 19F resonance with 

increasing LiF content. This large shift indicates that the nature of the interaction between F-

 anions and 

Y3+ cations is influenced by the amount of Y3+ in the melt. The higher the quantity of Y3+, the more 

bridging the F-

 ions become. The differences in line width for the solid and the liquid state are due to 

differences in relaxation times. Note also in Figure 13 that the 19F NMR resonance pattern of the solid 

shows a significant difference from that of the liquid with increasing YF3 content. One interpretation of 

this observation is that there are multiple F-

 site types in the solid that are randomized by the ionic 

mobility in the liquid state. 
Molten Salt Chemistry Workshop

26 FRD 2 

a

Ultraviolet-visible absorption 

spectrum of polytelluride ion in 

molten Li2BeF4. 

b

Raman spectra of (A) molten LiF-NaF-ThF4 (40-46-14 mol%) 

at 650 C and (B) polycrystalline K5ThF9 at 25 °C.

Figure 12. (a) Ultraviolet-visible and (b) Raman spectroscopy results reveal species identity in molten 

salts. | Images courtesy of Oak Ridge National Laboratory.

Figure 13. High temperature 19F NMR spectra of YF3–LiF as a function of composition. Black 

lines are for the molten state; dotted lines are for the solid state at temperatures 700–1100 °C 

(estimated). | Reproduced from Anne-Laure Rollet et al. ”In situ high temperature NMR and EXAFS 

measurements in rare-earth fluoride molten salts,” Comptes Rendus Chimie 2004 7(12), 1135–1140. 2004. 
Molten Salt Chemistry Workshop 

FRD 2 27 

Elsevier Masson SAS. All rights reserved. 

In similar measurements for the LaF3-LiF system, the corresponding 139La NMR spectra showed that the 

first coordination shell around the La3+ cations changed very little with composition. [19] The radial 

structure functions derived from analysis of the extended XAFS (EXAFS) spectra in Figure 14 support 

the 139La NMR results for the first coordination shell, but they give evidence of significant differences in 

coordination environment beyond the first shell when the solid and molten states are compared. 

The results in Figures 11 through 14 provide excellent examples of how the proposed measurement 

methods provide the information needed regarding oxidation states, coordination numbers, inter-ion 

spacing, local and long-range order, and solvation that will be used to elucidate the structure and 

dynamics of MSR-relevant molten salt systems. 

Figure 14. La LIII EXAFS spectra of molten LaF3-LIF (20-80 mole%): (a) solid phase (solid line), (b) melt at 780 

°C (open triangles), and (c) melt at 850 °C (open diamonds). | A.-L. Rollet, “Local structure analogy of 

lanthanide fluoride molten salts.” International Journal of Thermophysics, Vol. 26(4), 1115–1125, 2005. Used with 

permission of Springer.

Dissolved species in molten salts exist as highly mobile metal complexes, whereas insoluble and semisoluble elements may exist as suspended fine particles (metallic or salt precipitates) or as surface deposits 

on the wetted materials. In both instances, the insoluble species exist in dynamic equilibrium with the 

dissolved species and temperature, concentration, speciation, and other factors impact this equilibrium. At 

the concentrations required for soluble and semi-soluble elements to appear in the salt, the short-range 

structure of the salt (complexation between cations and anions) is sufficiently affected to impact a broad 

range of thermophysical properties, such as viscosity, diffusion and transport coefficients (thermal or 

concentration gradient induced migration), and chemical activity (the propensity of a particular phase to 
Molten Salt Chemistry Workshop

28 FRD 2 

be stable or instable). These changes in physicochemical behavior dramatically affect critical engineering 

factors, such as corrosion behavior and thermal conductivity. 

Design and development of a fully functional MSR that meets all the requirements for sustained 

operation, including materials compatibility, manageable radionuclide containment and disposition, and 

tenable economics will require detailed knowledge of the structure and chemical speciation of complex 

ions and other associated species in the molten salt system, as well as real-time monitoring of the species 

in solution. The net oxidation/reduction condition of the salt must be tracked and understood as a function 

of burnup, and the time evolution (dynamics) of molten salt structure and elemental speciation as a 

function of burnup must be fully elucidated on the local and global scales. The influence of these 

structural and dynamical behaviors on other thermophysical and transport properties must be measured 

and reconciled. 

3.2 Technical Challenges and Research Directions 

To understand how the structure and dynamics of molten salts impact their physical and chemical 

properties—such as viscosity, solubility, volatility, and thermal conductivity—it is necessary to determine 

the structure and speciation of the molten salt using advanced spectroscopic and scattering methods that 

provide information at the atomic/molecular scale. Where necessary, innovative experimental methods for 

microscopic and macroscopic property measurements will be employed, with emphasis on highthroughput techniques (i.e., ones applicable to radionuclides) to accommodate the efficient gathering of 

data for the large information base expected to be needed to underpin/support MSR design and 

development. Real-time analysis methods need to be developed for monitoring key chemical species in 

solution, allowing optimization of reactor performance and lifetime. This experimental work must be 

closely interfaced with first-principles computational and data analysis approaches to establish validated 

predictive models for system performance. 

3.2.1 Determining Molecular Structure by X-ray and Neutron Scattering and 

Spectroscopy 

In the nearly five decades following the decommissioning of ORNL’s MSRE, astounding gains have been 

made in the development of advanced x-ray and neutron scattering and spectroscopy techniques, which 

will enable precise investigation of the structure of molten salts and the coordinate environment of the 

actinides and fission products. Third-generation synchrotron light sources typically have brilliance in 

excess of 1018 photons s-1 with exquisite polarization, collimation, energy tunability, and low emittance, 

enabling real-time analysis with high-energy techniques that were not possible during the MSRE. 

Similarly, access to high-flux neutron scattering facilities now allows the investigation of structures, 

coordination environments, and buried structural features and dynamics at high temperatures; and 

additional information can be gained by leveraging isotopic contrast. 

Development of robust methodologies and high-temperature sample cells (up to 850 °C) is essential for 

the application of such techniques to the investigation of molten salt structure at operational temperatures 

to enhance the understanding of fundamental salt properties in the bulk. The application of XAFS (see the 

sidebar “XAFS of Molten Salts” on p. 29) to the molten salts relevant to a MSR has been limited, yet it 

provides an avenue for investigating the local atomic structures of melt constituents. It is also a potential 

means of identifying fission and corrosion product oxidation states and geometries, particularly in 

instances where dilute elemental concentrations prohibit the application of other techniques. 

Application of soft x-rays, suitable for XAFS of light elements (before scandium in the periodic table, 

with specific MSR-relevance including Na, Mg, and Al), may afford important structural information for 

the salts in the melt phase. A significant challenge is the design of sample holders capable of preserving 

an air- and moisture-free environment at MSR operational temperatures (up to and exceeding 850 C) 
Molten Salt Chemistry Workshop 

FRD 2 29 

while sufficiently enclosing the low-viscosity melt and permitting sufficient penetration of low-energy xrays. Another difficulty is the need for sample holders that are amenable to the analysis of lowconcentration components in the melt. Data for trace elements are customarily collected with fluorescence 

detectors, but few instances of such measurements exist in the molten salts literature, none of which 

involves elevated temperatures or carefully controlled atmospheres. [20] 

High energy x-ray or neutron scattering measurement affording the elemental pair distribution function 

(PDF) is another quantum-beam technique capable of affording otherwise unobtainable insights regarding 

molecular structure. The PDF is the Fourier transform of the total scattering pattern and affords the 

probability of finding an atom at a given distance. Investigation of the PDF allows interrogation of bond 

lengths and crystal structures and is a particularly powerful technique following the subtraction of 

background scattering (i.e., differential-PDF). [24] Recent work has demonstrated the capability of PDF 

studies to investigate the exchange kinetics of the first and second coordination environment of aqueous 

metal halides (see Figure 15), [25, 26] yet extension of such efforts to molten salts is non-trivial. NeutronPDF investigations afford the opportunity to introduce isotopic contrast (due to different scattering cross 

sections of the elements), allowing the elucidation of which molecular moieties contribute at a given 

spatial range. While the inherent opportunity for investigating the structure of materials devoid of longrange ordering is obvious, isotopic contrast nevertheless diminishes with increasing Z-number and is 
Molten Salt Chemistry Workshop

30 FRD 2 

typically insufficient for elements heavier than phosphorous. Therefore, the development of approaches 

by which maximum information can be gained from scattering data possessing minimal contrast remains 

an area requiring further development. Finally, it is important to mention that these techniques are just a 

subset of potential quantum-beam approaches to investigating the structure and dynamics of molten salts, 

and that the technique used is contingent upon the scientific question to be answered. 

Solid-state and liquid NMR can also be used to 

determine the local structure of species in the melt 

and the solid and the cation–anion interactions. 

Although the temperature range of magic-angle 

spinning (MAS) techniques is limited to about 

500 °C by the mechanical strength of the zirconia 

rotors, high-temperature MAS probes using a rotor 

encapsulated in a CO2 laser-absorbing sleeve are 

now commercially available. They can be used at 

temperatures above 2000 °C and enable containerless heating and aerodynamic levitation. [27] The 

measured chemical shift provides information on 

the local environment around the observed nucleus 

and depends on the nature and number of 

neighbors. As seen in Figure 13, the evolution of 

the high temperature 19F resonances in YF3 in 

varying amounts of LiF (measured at 20 K above 

the melting point of the binary mixture, 700–

1100 °C) signifies a progressive modification of the 

fluorine environment from free F-

 in LiF toward 

YFx

3-x. The high-temperature 19F and 7

Li NMR 

spectra also reveal that the coordination 

environment of YF3 in varying amounts of LiF is 

different between the melt and the fused state. [28] 

Since the fluoride is shielded less in the melt, the 

molten salt is more ionic and thus exhibits a 

different coordination environment. This is not 

expected to be the case with chlorides, in which the 

shielding is lower than in fluoride and complex ion 

formation has been observed. Coupling hightemperature NMR with complementary techniques such as EXAFS and Raman spectroscopy leads to a 

better description of the local structure of the salt at high temperatures and provides insights into the 

origin of the physical and chemical properties. 

3.2.2 Correlating Electrochemistry and Spectroscopy for Online Monitoring and 

Predictive Modeling 

One of the challenges in a MSR is that fission creates an oxidizing chemical environment that can lead to 

corrosion of the metal walls. For example, when fission of 235UF4 occurs, four fluoride ions are released 

from U4+ and two fission products are formed—such as 92Kr and 141Ba2+ or 140Xe and 94Sr2+—which 

subsequently decay to more stable isotopes. However, the net oxidation state of the products differ from 

those of the parent actinide, which leads to free fluoride in the system. To prevent corrosion, online 

monitoring of the redox properties of the melt is essential to enable real-time analysis of the state of the 

molten salt media. The approaches used should be capable of determining the total uranium as well as the 

total fraction of uranium in each oxidation state to maintain proper redox chemistry. A mixed-valance 

Figure 15. Differential-PDF measurements were used 

to investigate the first and second coordination 

environment of aqueous uranyl as a function of 

chloride concentration, revealing the formation of 

inner sphere chloride complexes and reorientation 

of second sphere solvent molecules upon chloride 

coordination. The extension of such techniques to 

investigating the speciation and structure of molten 

salt components is expected to afford new insights 

into physical phenomena such as fission product 

precipitation, corrosion, and gas entrainment. 

| Reprinted with permission from L. Solderholm, S. 

Skanthakumar, and R. E. Wilson. “Structural 

correspondence between uranyl chloride complexes in 

solution and their stability constants.” J. Phys. Chem. A

115(19), 4959–4967, 2011. Copyright 2011, American 

Chemical Society.
Molten Salt Chemistry Workshop 

FRD 2 31 

uranium fuel in a halide salt is the starting point, but the ratio of the oxidation states of U3+/U4+ will 

change over the course of the fuel lifetime. Direct monitoring of the fission product buildup is required to 

determine the optimum time for fission product removal/molten salt replenishing. Also, monitoring for 

trace impurities or corrosion products will enable feedback on the reactor health. 

Spectroscopic monitoring of the molten salts system can simultaneously provide information on the 

dissolved concentration and oxidation state of optically active species, and on the identity of the dissolved 

specie as a complex ion. The measurement of U [29–31], Np [32, 33], and Pu [34–36] by absorption 

spectroscopy in molten salt solutions has been extensively studied since as early as the 1960s. The 

absorption spectra of the various oxidation states of U, Pu, and Np in molten salt media are quite distinct 

(as they are in aqueous solution) and have been used as the basis for quantitative measurement. 

Measurement of U, Np, and Pu species in molten salt solutions has also been achieved by electrochemical 

methods (see review article by Zhang [37] and references therein). The strength of the electrochemical 

methods includes durability in strong radiation fields and high operating temperatures. The combination 

of electrochemical methods with spectroscopic measurements, in conjunction with chemometric modeling 

techniques, would prove a powerful analytical tool for real-time monitoring. For example, Figure 16a 

presents absorption spectra for the interconversion between U(III) and U(IV) by electrochemical 

modulation. [30] The spectroelectrochemical interconversion of Np(IV) to Np(III) is shown in 

Figure 16b. [32] 

Figure 16. (Left, a) UV-vis absorption spectra following the reduction of U(IV) to U(III) within an alkali 

chloride molten salt. Reprinted with permission from I. B. Polovov, V. A. Volkovich, J. M. charnock, et al. In situ 

spectroscopy and spectroelectrochemistry of uranium in high-temperature alkali chloride molten salts.” Inorganic 

Chemistry 47(17), 7474–7482, 2008. Copyright 2008, American Chemical Society. (Right, b) The 

spectroelectrochemical UV-vis measurement of the conversion of Np(VI) to Np(II) in molten LiCl-KCl 

eutectic media | Reprinted from I. B. Polovov, C. A. Sharrad, L. May, V. A. Volkovich, and B. D. Vasin. 

“Spectroelectrochemical study of neptunium in molten LiCl-KCl eutectic.” Z Naturforsch A 62(12), 745–748, 2007.

Spectroscopic measurements, [38] electrochemical measurements, [39] and combined spectroelectrochemical [39, 40] methods have been demonstrated in alkali-halide eutectic solutions containing a 

mixture of lanthanide chlorides. These techniques enable the quantitative analysis of mixtures of metal 

salts in molten salt media. [38] As shown in Figure 17, the use of multivariate regression techniques for 

the quantitative measurement of multiple species with overlapping/interfering absorption bands has been 

demonstrated. It would be advantageous to apply this approach to several key actinides (U, Np, and Pu). 

Furthermore, both Raman and UV-vis methods can be incorporated to expand the range of interrogation 

capabilities. 

U (III) 

U (IV) 
Molten Salt Chemistry Workshop

32 FRD 2 

 

Figure 17. UV-vis absorbance spectra of lanthanides within alkali chloride molten salt (left) and example 

chemometric modeling results to quantify neodymium within single-component and multicomponent 

molten salt solutions (right) where training sets data (blue + symbol) were used to build the models that 

were applied to validation sets (red triangles); a 1:1 correlation (black line) is bounded by the 95% 

confidence intervals (red dashed lines). |Reproduced from C. A. Schroll, A. M. Lines, W. R. Heineman, and S. A. 

Bryan. “Absorption spectroscopy for the quantitative prediction of lanthanide concentrations in the 3LiCl-2CsCl 

eutectic at 723 K.” Anal. Methods–UK 8(43), 7731–7738, 2016, with permission of the Royal Society of Chemistry.

In addition to redox state, the Lewis acid-base properties of the melt are important in determining the 

solubility of the actinides in the melt and the corrosivity toward structural materials. The solubility of 

actinides in a molten salt is fundamentally related to the local basicity of the anions surrounding the metal 

cation. Phenomenological approaches have been employed to investigate the effects of molten-salt 

compositions on the solubility properties of actinides. [41] No attempts have been made to correlate these 

observations with a microscopic rationalization of actinide species solubility in molten salts. In other 

words, a bridge between macroscopic solubility experiments and microscopic local structural properties is 

lacking. This information gap has prevented the development of predictive models of molten salts with 

good solubility properties. Research is needed to uncover the correlation between the solubility of 

actinides in molten salts and their fundamental spectral fingerprints. 

In the framework of Lewis acid-base theory, a metal ion behaves as a Lewis acid while the anionic 

network behaves as a Lewis base. The halide anions of a molten salt surrounding the metal ion donate the 

negative charge, and this donation reduces the positive charge on the metal ion. [42, 43] The greater the 

negative charge borne by the halide anions, the greater will be the electron donation. If the metal ion can 

signal (via optical spectroscopy) the extent to which its positive charge changes, then there is the 

possibility of using it as a probe for the quantitative measurement of the basicity of the corresponding 

molten salt medium. The basicity scale thus obtained has been called “optical basicity”; it is one of the 

most fundamental parameters that characterize a specific molten salt, in analogy to the way pH is used to 

describe properties of aqueous systems. Metal ions such as Tl+

, Pb2+, and Bi3+ have been employed as 

optical probes to measure the basicity of industrially interesting oxide glasses. As was pointed out earlier, 

actinide solutes are known to have characteristic UV-vis and near-IR absorption bands resulting from f-f 

and charge-transfer electronic transitions. Thus, a new online and accurate optical basicity scale is 

needed for actinide-containing molten salts that uses optical spectra of actinide ions as direct probes of 

the basicity of the melt. 
Molten Salt Chemistry Workshop 

FRD 2 33 

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Molten Salt Chemistry Workshop 

FRD 3 37 

4. FRD 3: Understanding Fission and Activation Product Chemistry 

and Radiation Chemistry 

4.1 Background and Current Status 

Understanding the fate and transport of fission and activation products in MSRs is key to monitoring, 

predicting, and controlling reactor performance and salt properties (Figure 18). Although the decay 

pathways of fission and activation products are well established and independent of the chemical 

environment, the chemical and physical properties of these elements and their interactions with their 

surroundings are different at each step of the decay chain; they are complicated by the fact these elements 

exhibit distinct chemical behaviors in different salt mixtures and behave differently in different shortdecay regimes. The general understanding is far from complete and remains a major scientific and 

technology challenge associated with the MSR system. It is essential to understand which fission 

products remain dissolved in the salt; which ones are volatile; and which ones precipitate on surfaces 

(and where), which ones are sometimes soluble, and how the fission products change the corrosivity of 

the melt. Furthermore, determination of fission product interactions and solubility is necessary for each 

unique salt system (e.g., fluoride salts vs. chloride salts vs. new salts). To achieve this level of 

understanding requires knowledge regarding the chemical nature of the fission and activation products—

i.e., the chemical species that exist under particular conditions of salt composition, temperature, and redox 

state—and how they may react or distribute on surfaces in a dynamic environment. 

Figure 18. Complex behavior of molten salt and insoluble and volatile fission products. The metal strip 

was in the MSRE pump bowl gas space; it shows evidence of mist in the gas space and noble metal 

fission products. | Images courtesy of Oak Ridge National Laboratory

Although fresh fuel salt mixtures have been studied, understanding of transport phenomena, effects of 

short-term decay, neutron activation, and interfacial interactions are necessary. Individual fuel atoms in 

the salt undergo fission (see Figure 19), splitting into a few of 40+ elements representing more than 400 

different isotopes, while neutron capture generates various transuranic and other activation products (such 

as 36Cl). Fission liberates large amounts of energy, causing various local disruptions such as ionization or 

material sputtering. The individual fission product atoms will be highly ionized and will quickly interact 

with nearby atoms, reaching one of several possible end states dependent on their chemistry and the 

atoms with which they interact. The interactions with the salt, structural materials, or other fission and 
Molten Salt Chemistry Workshop

38 FRD 3 

activation products, causes complex speciation. The process is repeated with many variant outcomes for 

each fission product. The end result can be macroscopically understood based on whether the fission 

products are soluble, insoluble, volatile, or corrosive. Each of these have implications for the reactor and 

the physical properties of the salt. On the mesoscale, interactions will be highly dependent on interfacial 

chemistries between fission products and the materials present. Understanding this and translating it to the 

macroscale effects, such as corrosion, is necessary. 

Figure 19. Yields of different elements from fission of 235U. | Image courtesy of Peter Lux. In “Composition of Spent 

Fuel,” Peter Lux Blog, http://www.plux.co.uk/. Licensed under a Creative Commons license Attribution-NonCommercialShareAlike 3.0 unported license

At some time after the fission event, the isotopes of the fission products reach chemical equilibrium and 

behave according to their elemental identities, eventually forming stable species. Soluble species are 

incorporated within the salt solution and often behave according to their elemental periodic group; 

insoluble species likely deposit on surfaces, such as the heat exchanger, or at the interface between wetted 

and non-wetted surfaces; volatile species are driven to the off-gas system; and corrosive species may 

degrade structural and container materials (Figure 18). To understand the impact of fission products on 

reactor operation, an understanding the basic physics and chemistry at the atomic, molecular, meso-scale, 

and continuum length scales, as well as phenomena that occur over a wide range of time scales is 

required. With full knowledge of the fission product decay pathways and their chemistry, we should be 

able to design systems that mitigate detrimental effects of fission products with regard to reactor 

performance and disposition in waste streams. 

4.2 Technical Challenges and Research Directions 

4.2.1 Understanding the Physical, Chemical, and Radiochemical Properties of the 

Fission and Activation Products in Molten Salts 

As previously discussed, there is a need to understand the chemical bonding and coordination 

environment (and associated energies) between the solutes (actinides, fission products, and activation 

products) and the molten salt solvent, since interactions could form out to a length scale that includes 
Molten Salt Chemistry Workshop 

FRD 3 39 

many atoms (in fact, determining the number of atoms that must be included is one of the underlying 

uncertainties.) Spectroscopic studies (described in Section 3) can provide insights into the coordination 

environment of the elements, but computational insights would be particularly useful for the more 

radioactive and short-lived species that are hard to study experimentally. Unfortunately, little theoretical 

work has been reported on the solvation of individual fission products. Although various computer codes 

can be used to obtain this information (e.g., the methods described for understanding the solvation of Gd3+

dissolved in LiCl and KCl), [1] the more atoms that are involved, the greater the computational expense. 

Therefore, appropriate assumptions for simplifying the models will need to be developed and validated. In 

moving from single atoms to complex fission product species, more solute-solvent configurations must be 

included, as well as an understanding of the internal bonds within the fission product complex. To solve 

the problem satisfactorily will require combining the results from computer codes that handle atomic- and 

molecular-scale interactions and properties all the way up to bulk material properties and energetics 

associated with changes in the physical states of transmuting isotopes. 

To understand the binding, solvation, and properties of fission and activation products, a representative 

set of fission products (several from each of the possible end states, e.g., soluble, insoluble, sometimes 

soluble, and volatile) that can be both modeled and measured in the laboratory (potentially though the use 

of nonradioactive surrogates) needs to be investigated. A comparison between experimental and computer 

models is extremely important, since only validated computational models can to trusted to provide 

insights into a broader range of fission products and yield meaningful results. This means that the systems 

that are studied must also possess unique, reliable experimental signatures that extend from single atoms 

to more complex molecular species. 

In addition to out-of-pile chemical studies, in the future, it will be necessary to develop experimental 

systems with in situ monitoring for the irradiation of salts to probe dynamic fission product chemistry 

effects and their impact on the physical properties of the molten salt (and reactor operation). With this 

understanding, we can begin to design systems to mitigate any negative effects associated with fission 

product transport, localization, or specific radiation/interfacial chemistry. For example, one importance 

result from in-pile studies in the MSRE was the discovery that the fission product tellurium contributed to 

surface cracking in the alloy Hastelloy N. Tellurium is unusual in that it has a high vapor pressure at the 

operating temperature of the MSRE, so it can survive as both a dissolved gas and (possibly) as a soluble, 

reactive ion. [2] To determine the cause of the observed surface cracking, a series of temperaturecontrolled tests were conducted both to confirm that tellurium was contributing and to subsequently learn 

how to mitigate its degradation effects. This study showed how a particular fission product’s behavior 

changed with salt redox potential, how it reacted at the interface, and how the negative effect could be 

mitigated in future designs. 

The tellurium study is an example of what needs to be done in greater depth for a range of “saltinsoluble” fission products that reside in the middle of the periodic chart. Although the understanding of 

tellurium is the most advanced finding for these fission products, it is not yet complete, especially for the 

new proposed salt compositions. An experiment design that accomplishes these goals with dynamic 

online measurement capability, as distinct from static sample analysis, will greatly accelerate the 

understanding of fission product behavior. 

Validating the ex situ surrogate testing requires high radiation fields, high-temperature operation, and 

realistic production and decay of fission products. In preparation for operation of the MSRE in the 1960s, 

a series of experiments were conducted at the Materials Test Reactor in Idaho and the Low Intensity Test 

Reactor and the Oak Ridge Research Reactor at ORNL (Figure 20). A similar set of experiments is 

needed for next-generation molten salts for the MSR to address the concerns discussed above. 
Molten Salt Chemistry Workshop

40 FRD 3 

Figure 20. Natural circulation loop used in the Oak Ridge Research Reactor. The experiment tested both 7LiF-BeF2-ZrF4 and 7LiF-BeF2-ZrF4-UF4 salts. | Image courtesy of Oak Ridge National Laboratory

4.2.2 Correlating Fission and Activation Product Behavior with Surrogates 

Before establishing the ability to install a molten salt test loop within a reactor with useful emulation of a 

range of MSR environments, the path for learning could proceed in a limited way with out-of-pile studies 

using both nonradioactive and radioactive surrogates to understand the behavior of fission and activation 

products. Whereas in situ studies will be very complex and will require the development of advanced 

analytical techniques, ex situ studies can be started immediately. These studies can be carried out with 

both nonradiological surrogates and the actual fission products. Ex situ studies can involve capsule testing 

of materials prior to thermal convection loop experiments to understand the influence of the fission 

products on reactor materials. The anticipated radiological fission products include elements such as I, 

Xe, Cs, Cd, In, Sn, Sb, Te, Zr, Nb, Mo, Tc, Ra, and Tl (see Figure 19). The stable isotopic forms for most 

of these elements are routinely used as surrogates in adsorbent and separations studies. Therefore, 

laboratory studies can be performed in a variety of research settings to gain insights into fission product 

behavior without requiring access to hazardous radioactive materials. 

There are numerous published examples of successful studies with surrogate/stable isotopes. For example, 

Nenoff et. al, have used stable iodine gas (I2 and organic iodides such as CH3I) to study the fundamental 

materials chemistry of iodine fission gas capture and storage materials. Structure property relationships 

between nanoporous materials such as zeolites and metal-organic frameworks (MOFs) and the iodine 

species have been detailed (Figure 21). [3] Subsequent studies indicate there is no degradation of the 
Molten Salt Chemistry Workshop 

FRD 3 41 

framework structures due to gamma irradiation. Furthermore, stable isotope studies of cesium and 

strontium ion capture from aqueous and seawater solutions were performed to confirm the stability and 

selectivity of a crystalline silicotitanate framework before commercialization and use at the Fukushima 

Daiichi nuclear reactor accident site. [4] 

Figure 21. Metal-organic frameworks for capture of fission products (left): A schematic of iodine capture 

by silver-containing mordenite (MOR). Whereas pre-reducing the silver MOR yields a mixture of -AgI 

nanoparticles and subnanometer -AgI, direct iodine uptake by silver-exchanged MOR produces 

exclusively subnanometer -AgI. | Reprinted with permission from K. W. Chapman, P. J. Chupas, and T. M.; 

Nenoff. “Radioactive iodine capture in silver-containing mordenites through nanoscale silver iodide formation.” 

J. Amer. Chem. Soc. 132 (26), 8897–8899, 2010. Copyright 2010, American Chemical Society. Right: Captured I2

crystallographically located inside the small pored ZIF-8 MOF | Reprinted with permission from D. F. Sava, M. A. 

Rodriguez, K. W. Chapman, P. J. Chupas, J. A. Greathouse, P. S. Crozier, and T. M. Nenoff. “Capture of volatile iodine, 

a gaseous fission product, by zeolitic imidazolate framework-8.” J. Amer. Chem. Soc. 133(32), 12398–12401, 2011. 

Copyright 2011, American Chemical Society.

Some fission and activation products are commonly substituted with surrogate elements for laboratory 

studies, for example, cesium for plutonium and europium for americium. Requirements include 

comparable atomic size, redox states, and chemical reactivity. For MSRs, the substitution of rhenium for 

technetium is particularly important for investigation of its solution chemistry and volatilization 

behaviors. [5] It is important to note that in one set of glass waste form degradation studies, rhenium 

transport was reported to behave quite differently from technetium transport under hydrothermal 

conditions. [6] Experiments reacting Re, ReO2, and TcO2 with Cl2 and HCl in molten chloride eutectics 

also revealed different chemical products, with a volatile TcOxCly proposed. [7] Therefore, control 

experiments need to be performed to ensure surrogate elements possess representative chemical and 

physical properties compared with the fission product generated under MSR operational conditions. 

Radiological studies, either with appropriate radiological isotopes or with the introduction of the 

appropriate decay energy, are needed to confirm the experimental results obtained with the stable or 

surrogate isotopes. For example, the electron beam from a TEM can be used to irradiate surrogate fission 

products [8] and the resulting reactions and redox responses can be studied. [9] 

As noted earlier, modeling the effects of radiological degradation of the salt or containment metals will 

provide a predictive understanding of the behavior of radiological fission products in the molten salt 

operational environment. It will also provide a computational means of near-term investigation of surrogate 

behavior compared with behavior of the actual radionuclide. These computational findings will rapidly 

provide information, which will necessarily be validated or revised upon completion of the more involved 

and complicated studies of radioactive materials following in the course of MSR development. [10] 
Molten Salt Chemistry Workshop

42 FRD 3 

4.2.3 Understanding the Physical and Chemical Impact of Short-lived Isotopes 

Short-lived fission products in the salt are of great concern because of the chemical differences between 

the original elements and the daughter products. The effects from short-lived fission products are complex 

and special experimental methods are required to probe them as they undergo many iterations and 

speciations until becoming stable isotopes. For instance, 137I (t1/2 = 24 sec) will be soluble in the salt; 

however, upon decay it becomes the entrained gas 137Xe (t1/2 = 3.81 min), which can further decay to 137Cs 

(t1/2 = 30 year) (Figure 22). If the system is flowing rapidly, the xenon will be removed from the salt (if 

not entrained) and will decay to cesium. Cesium from entrained 137Xe will redissolve into the halide salt 

and will stay soluble based on its free energy as a fluoride or chloride. Alternatively, cesium from the 

liberated 137Xe, removed during circulation, will result in cesium plating within the outgassing system. 

Another example is the 99Zr fission product (t1/2 = 2 sec), which will decay to 99Nb (t1/2 = 15 sec), then to 

99Mo (t1/2 = 2.75 days) and eventually to 99Tc (t1/2 = 2.1×105

 year), which will stay in the metal state and 

not be soluble in the molten salt; thus it will plate out on the reactor components or heat exchanger. 

Cadmium- 131,-132 (t1/2 = 68 msec, 97 msec) provides a third example; it decays to 131In (t1/2 = 280 

msec), 131Sn (t1/2 = 56 sec), 131Sb (t1/2 = 23 min), and 131Te (t1/2 = 25 min), and finally to 131I (t1/2 = 8 days). 

Iodine-131 decays to 131Xe gas, which is stable and must be removed from the melt to minimize neutron 

poisoning in the reactor. These phenomena are depicted in Figure 22. Although individual elements can 

be experimentally studied with surrogates, it is difficult to study the dynamics and changing chemistries 

of short-lived fission products. Thus, computational methods need to be developed in tandem with the 

best experimental methods to provide insights into the impacts of short-lived isotopes on the reactor 

chemistry and corrosion. 

Figure 22. Examples of short-lived fission products that change solubility behavior during decay. | Image 

courtesy of Oak Ridge National Laboratory

~1970 out-of-pile chemistry

alsoluble  gaseous  soluble deca

137I  137Xe  137Cs 2 5 sec. 4 ‐m in.

nsitional(soluble  insoluble  sometimes soluble  soluble  gaseousy exampl

131,132Cd  131In  131Sn  131Sb  131Te  131I  131Xe < 1 sec. < 1 sec. 1 min. 2 3 ‐m in. 2 5 ‐m in. 8 ‐d a ys

elemental volatility

alsoluble  sometimes soluble  insoluble deca

99Zr  99Nb  99M o  99Tc 2.1 sec. 1 5 ‐sec. 2.7 5 ‐da y

New capabilities in 

• radiation imaging

• multiphysics modeling

could be coordinated with 

need for neutronics analysis.

Reactor 

Core

Primary 

Loop

decay

and

stripping

fission,

decay,

activation

pump

heat‐exchanger

noble gasstripping

< 30 sec

residence

time

< 30 sec

residence

time
Molten Salt Chemistry Workshop 

FRD 3 43 

References 

1. S. Hazebroucq, G. S. Picard, and C. Adamo. “A theoretical investigation of gadolinium (III) solvation 

in molten salts.” J. Chem. Phys. 122, 224512, 2005; S. Hazebroucq, G. S. Picard, C. Adamo, T. 

Heine, S. Gemming, and G. Seifert. “Density-functional-based molecular-dynamics simulations of 

molten salts.” J. Chem. Phys. 123, 134510, 2005. 

2. E. L. Compere, S. S. Kirslis, E. G. Bohlmann, F. F. Blankenship, and W. R. Grimes. Fission Product 

Behavior in the Molten Salt Reactor Experiment, ORNL-4865, Oak Ridge National Laboratory, 1975. 

3. T. M. Nenoff, M. A. Rodriguez, N. R. Soelberg, and K. W. Chapman. “Silver-mordenite for 

radiologic gas capture from complex streams: Dual catalytic CH3I decomposition and I confinement.” 

Micro. Meso. Mater. 200, 297–303, 2014; K. W. Chapman, P. J. Chupas, and T. M. Nenoff.

“Radioactive iodine capture in silver-containing mordenites through nanoscale silver iodide 

formation.” J. Am. Chem. Soc. 132(26), 8897–8899, 2010; D. F. Sava, K. W. Chapman, M. A. 

Rodriguez, J. A. Greathouse, P. S. Crozier, H. Zhao, P. J.Chupas, and T. M. Nenoff. “Competitive I2 

sorption by Cu-BTC from humid gas streams.” Chem. Mater. 25(13), 2591–2596, 2013; D.F.S. 

Gallis, I. Ermanoski, J. A. Greathouse, K. W. Chapman, and T. M. Nenoff. “Iodine gas adsportion in 

nanoporous materials: A combined experiment–modeling study.” Ind. Eng. Chem. Res. 56(8), 2331–

2338, 2017; J. T. Hughes, D. F. Dava, T. M. Nenoff, and A. Navrotsky. “Thermochemical evidence 

for strong iodine chemisorption by ZIF-8.” J. Am. Chem. Soc. 135(44), 16256–16259, 2013. 

4. T. M. Nenoff and J. L. Krumhansl. “Cs+ removal from seawater by commercially available molecular 

sieves.” Solvent Extraction and Ion Exchange 30, 33–40, 2012; C. A. Abney, K. M. L. TaylorPashow, S. R. Russell, Y. Chen, R. Samantaray, J. V. Lockard, and W. Lin. “Topotactic 

transformations of metal-organic frameworks to highly porous and stable inorganic sorbents for 

efficient radionuclide sequestration.” Chem. Mater. 26, 5231–5243, 2014. 

5. M. H. Langowski, J. G. Darab, and P. A. Smith. Volatility Literature of Chlorine, Iodine, Cesium, 

Strontium, Technetium, and Rhenium: Technetium and Rhenium Volatility Testing, PNNL-11052, 

Pacific Northwest National Laboratory, 1996. 

6. D. A. McKewon, A. C. Buechele, W. W. Likens, D. K. Shuh, and I. L. Pegg. “Tc and Re behavior in 

borosilicate waste glass vapor hydration tests.” Environ. Sci Technol. 41(2), 431–436, 2007. 

7. V. A. Vokovich, I. May, J. M. Charnock, and B. Lewin. “Reactions and speciation of technetium and 

rhenium in chloride melts: A spectroscopy study.” Phys. Chem. Chem. Phys. 4, 5753–5760, 2002. 

8. M. Nyman, B. X. Gu, L. M. Wang, R. C. Ewing, and T. M. Tenoff. “Synthesis and characterization of 

a new microporous cesium silicotitanate (SNL-B) molecular sieve.” Micro. Meso. Mater. 40, 115–

125, 2000. 

9. C. L. Tracy, M. Lang, J. M. Pray, F. Zhang, D. Popov, C. Park, C. Trautmann, M. Bender, D. 

Severin, V. A. Skuratov, and R. C. Ewing. “Redox response of actinides materials to highly ionizing 

radiation.” Nature Comm. 6, 6133, 2015. 

10. A. Dunn, R. Dingreville, and L. Capolungo. “Multi-scale simulating radiation damage accumulation 

and subsequent hardening in neutron-irradiated α-Fe.” Model. Simul. Mater. Sci. Eng. 24(1), 015005, 

2015; R. Dingreville and S. Berbenni. “On the interaction of solutes with grain boundaries.” Acta 

Mater. 104, 237–249, 2016. 

Molten Salt Chemistry Workshop 

FRD 4 45 

5. FRD 4: Understanding Materials Compatibility and Interfacial 

Phenomena 

5.1 Background and Current Status 

The reemergence of MSR concepts for future energy production has prompted an evaluation of the state 

of knowledge and fit-for-use (or lack of such) of materials in contact with the coolant/fuel. Materials 

compatibility under such extreme conditions presents unique challenges, many of which have been 

defined and addressed, at least for fluoride salts, during the course of the MSRE and subsequent MSBR 

program. [1] However, concerns remain regarding the chemical compatibility and application of Ni-Mo 

alloys (specifically Hastelloy N) in molten fluoride salt because of the lack of long-term operational 

experience and the possibility of embrittlement by fission products. As an expanded performance 

envelope has been proposed in some recent MSR designs, including higher temperatures and use of 

chloride salts, it is apparent that alternative structural materials need to be evaluated. Chloride salts for 

MSR applications have a much smaller compatibility database, require higher temperatures to liquefy, 

and appear to display greater corrosion issues, although results show a great degree of variability. 

Candidate nickel-based alloys and/or advanced ceramics, such SiC-SiC composites, encompass greater 

alloy strength (without significant degradation of fracture toughness), radiation tolerance, and resistance 

to corrosion and oxidation than the alloys developed during the MSRE. 

Molten salt coolants and fuels present a number of extreme environmental demands on reactor structural 

and functional materials. These include 

 Resistance to corrosion and embrittlement by the fuel/coolant salt 

 Low rates of mass transfer burden due to thermal or dissimilar material gradients in the reactor 

system 

 High strength and fracture toughness with minimal/acceptable degradation over time due to radiation 

damage and high temperatures 

 Corrosion/oxidation resistance by the external reactor environment (whether air or a secondary 

coolant) 

 Fabricability, including joining and cladding (for certain applications) 

In general, materials degradation processes underlying these requirements are approached as either 

irradiation or chemical compatibility effects. Significantly less attention is paid to their combined 

influence or the origins of those effects at the atomistic scale. Similarly, although mechanical and 

chemical behaviors tend to be treated separately, their effects can be synergistically compounding; 

relevant examples include stress corrosion cracking in aqueous systems and oxidative embrittlement of 

nickel-based alloys. To some extent, the problems can be separated by using corrosion-resistant coatings 

on materials to provide adequate mechanical strength, although this practice introduces the additional 

issue of coating integrity. The manifestations of these effects can be different for alloys, graphite, and 

silicon carbide composites. 

Materials degradation (corrosion) occurs at an interface between the salt and a metal in a MSR. Thus, 

understanding the molecular-level chemistry of the fluid, solid, and gas interfaces is key to controlling the 

interfacial reactions that result in materials degradation. Interfacial behavior pertinent to MSRs is an 

under-investigated area—in particular at atomistic scales—because the boundary regions between phases 

or molecular regions comprise only a small fraction of all species in any sample, making their focused 

investigation particularly difficult. For MSRs, a molecular-level understanding is needed of the salt/alloy, 
Molten Salt Chemistry Workshop

46 FRD 4 

salt/graphite, salt/gas, and salt/salt interfaces; and an understanding of their molecular structure and 

chemical reactivity will provide important insights into mitigating phenomena such as corrosion, fission 

product off-gassing, gas entrainment, and material precipitation.

5.2 Technical Challenges and Research Directions 

5.2.1 Advancing Spectroscopic and Scattering Investigation of MSR Interfaces 

The ability to characterize the molecular-level chemistry that takes place at the liquid/solid interface is 

key to understanding interfacial reactions such as corrosion. A number of surface-sensitive techniques 

have been developed in the 20th century to address the challenges of characterizing an interface, 

including infrared (IR) adsorption spectroscopy (attenuated total reflectance, reflection-adsorption 

infrared spectroscopy, transmission IR), Raman spectroscopy, sum frequency generation, UV-vis 

spectroscopy, and x-ray and neutron-based techniques, such as reflectometry. [2] Second harmonic 

generation, which was used to investigate molecular monolayers adsorbed to surfaces, [3] has been 

extended to study the air-water interface [4] and more recently the interfaces of spherical particles [5] and 

colloidal systems. [6–8] Recent developments in quantum-beam approaches have also enabled the 

investigation of interfacial structure and reactivity. Perhaps of greatest note are x-ray and neutron 

reflectivity, which provide a molecular density profile of molecular ordering as a function of distance 

from the interface. [9] For example, a review of the water/mineral interface reveals perturbations of the 

liquid structure near the solid surface boundary. [10] This liquid structure is expected to have a crucial 

influence on physical interfacial behaviors—such as diffusion, adsorption, and dielectric properties—that 

directly affect phenomena of immediate pertinence, such as metal precipitation or alloy corrosion. 

Numerous grazing incidence techniques have also been developed for the investigation of various 

interfaces. For example, grazing incidence XAFS (GI-XAFS) has been used to investigate the speciation 

of Er3+ at the electrolyte-vapor interface; it revealed simple hydration of Er3+ in the bulk phase to form 

[Er(OH2)8]

3+, but formation of a neutral [(H2O)6-7ErCl3] at the surface (Figure 23). [11] This work was 

complementary to an investigation of the same system using x-ray reflectivity. [12] Grazing incidence xray diffraction provides information regarding in-plane structure and ordering at interfaces, and x-ray 

fluorescence near total reflection provides element-specific information regarding the binding and 

coverage of target metal ions. 

There is a need to characterize the molecular-level chemistry at the molten salt/solid interface. The use of 

approaches like those described to investigate interfaces within a molten salt would enable the 

identification of metal species and oxidation states in the salt phase. There is also great potential to 

answer unresolved questions regarding corrosion mechanisms in MSRs, in particular via the development 

of hyphenated techniques in which manipulation of interfacial behavior is achieved electrochemically. 

Corrosion studies are commonly performed by applying a voltage to a metal surface or creating a 

chemically reactive environment and observing how metal components react. [13–15] Electrochemical 

impedance spectroscopy is routinely used to explore corrosion of a metal surface, because it provides 

information regarding solution phase resistance, charge transfer properties of the surface, and diffusion 

characteristics of the solution phase. Traditional corrosion analysis is performed using post-mortem 

methods—such as transmission electron microscopy with electron energy loss spectroscopy or energy 

dispersive spectroscopy, x-ray diffraction, and optical spectroscopy—in which a sample is analyzed after 

retrieval from a reaction vessel. While informative, this approach is limited by challenges associated with 

discerning the chemistry after it has occurred. In contrast, the spectroscopic approaches discussed above 

allow the direct interrogation of the interface itself and, in some instances, are even amenable to in situ 

reaction monitoring. 
Molten Salt Chemistry Workshop 

FRD 4 47 

Figure 23. GI-XAFS results reveal unanticipated molecular complex formation at the water/gas interface. 

The extension of similar techniques to molten salts could be used to answer fundamental questions 

regarding corrosion mechanisms, particularly if successfully paired with electrochemical control. | Reprinted 

with permission from M. K. Bera, G. Luo, M. L. Schlossman, S. Lee, and M. R. Antonio. “Erbium(III) coordination at the 

surface of an aqueous electrolyte.” J. Phys. Chem. B 119(28), 8734–8745, 2015. Copyright 2015, American Chemical 

Society. 

Development of multi-faceted in situ and in operando approaches to observe interfacial physical 

phenomena is needed to advance MSR development. The advancement of corrosion science is greatly 

enhanced by the ability to collect information as the chemistry occurs. For example, a combination of 

electrochemical methods with advanced x-ray and neutron scattering techniques would allow the 

exploration of corrosion in operando, providing process-relevant information that cannot be observed 

during traditional post-mortem analysis. 

5.2.2 Understanding Materials Compatibility in the Molten Salts Reactor Environment 

Results from the Aircraft Nuclear Propulsion program in the 1950s and the MSRE in the 1960s clearly 

showed the benefits of the Ni-Mo alloy system for use with molten fluoride salts. Most existing hightemperature nickel-based superalloys were designed for operation in oxidizing external atmospheres (e.g., 

air in the MSRE radiator and secondary coolant piping/tanks, and supercritical steam in the MSBR 

secondary heat exchanger). Therefore, their corrosion resistance is derived from the formation of a 

passivating layer of chromia or alumina scale. The situation with regard to salt-side corrosion is 

fundamentally different because of the dissolution of passivating oxide layers in fluoride and chloride 

salts. Corrosion in these salts is instead controlled by the thermodynamic stability of the bare alloy 

surface in the halide salt environment. Degradation occurs by dissolution of the containment material into 

the salt, driven by the difference in free energy of formation between the salt constituents and the most 

susceptible transition metal corrosion product (the more negative the free energy of reaction, the greater 

extent to which the reaction will occur). In the simplest systems, these reactions are governed by 

elemental solubility considerations and the kinetics of the rate-controlling step of the dissolution process. 

Those dependencies are reflected in the fundamental rate equation for solvation of solids in liquids [16]:
Molten Salt Chemistry Workshop

48 FRD 4 

Ji = ks,i(Co,i – C) , (5.1) 

where Ji is the rate of mass loss of element i due to dissolution, Co,i is the solubility of component i in the 

salt, C is the actual concentration of i in the bulk of the liquid, and ks,i is the effective solution rate 

constant. (Each constituent of the condensed phase is separately governed by Eq. (5.1) but, for 

simplification, the subscripts are not used hereinafter.) Because J is proportional to (Co – C), high 

solubilities do not necessarily translate into high dissolution rates (and vice-versa), because the driving 

force for dissolution depends on the actual concentration of the solute in the salt as well as Co. [17, 18] 

Elimination of concentration gradients by convective or mechanical mixing thus forms the basis for the 

difference between isothermal and/or stagnant salt systems and nonisothermal dynamic ones (such as in 

most reactor schemes) with regard to material stability. This can be particularly important for heat 

exchangers and systems with large temperature differentials in which deposition can block flow passages 

and lead to thermal excursions or excessive localized radioactivity. Furthermore, preferential dissolution 

of one constituent into the salt can destabilize the structure of the solid or its interface with the liquid, in 

some cases, leading to porous zones. 

Under nonisothermal conditions normally encountered in reactor systems, [19] C should be uniform 

within a flowing liquid system. Consequently, J is only dependent on time (t) through ks because (Co – C) 

is effectively constant. Common practice then allows J to be described in a generalized kinetic form: 

J = m/tn . (5.2) 

The time dependence of the dissolution rate process (n) can be determined by fitting measured mass loss 

(m) using Eqs. (5.1) and (5.2). Therefore, direct measurement of mass loss from corrosion affords a 

straightforward method of determining a key kinetic parameter for dissolution, yielding information 

regarding rate-controlling steps (surface reaction, liquid-phase diffusion through boundary layer, and 

solid-state diffusion) [16, 17, 20, 21] for corrosion mechanisms under nonisothermal conditions. 

Collection of these data at various temperatures affords the representation of different positions in a 

molten salt loop. Such kinetic information can guide and validate appropriate modeling and lifetime 

prediction approaches for a particular solid-salt system. 

The flux description described by Eq. (5.1) applies equally well to deposition when C > Co. If the 

dissolution and deposition behaviors are controlled by the respective forms of Eq. (5.1), measurements of 

mass change as a function of loop position (x) should yield data resembling the schematic depiction 

shown in Figure 24. Using the appropriate analysis methods and simplifying assumptions, measurements 

of the mass—or, for alloys and compounds, composition change—as a function of loop position (and 

therefore temperature) can provide further data regarding kinetics and the rate controlling mechanism of 

mass transport associated with a given material–liquid salt system. Initial results could be obtained in ex 

situ studies to guide future experimental investigations. 

The free energy driving force for dissolution of the solid can be controlled by manipulating the redox 

potential of the salt. While more reducing conditions were found to suppress corrosion, other reactions 

between the salt and solid, such as alloying, become more likely. [22] Therefore, corrosion control by 

manipulation of the redox potential is bound by the necessity of identifying conditions that minimize both 

oxidative and reductive degradation reactions; limiting other deleterious reactions; and, for fuel salts, 

maintaining the necessary U3+/U4+ ratio. The ability to effectively control the salt redox properties in real 

time requires implementation of online monitoring (discussed in Section 3). 
Molten Salt Chemistry Workshop 

FRD 4 49 

(a) Mass transfer schematic 

(b) Idealized profile of mass change per unit 

surface area for a liquid circuit. Tmax, Tmin are 

maximum, minimum loop temperatures, 

respectively; x is loop position.

Figure 24. Steps and balance points in temperature-gradient mass transfer. | Image courtesy of Oak Ridge 

National Laboratory

Impurities in the as-produced salts, such as HF, HCl, H2O; residual metal oxides; and/or polyvalent metal 

ions can also cause or affect degradation. They can change the driving forces (free energies) for 

compound formation/alloying of the solid, provide new reaction pathways for dissolution and deposition, 

and affect kinetic factors governing mass transfer of the principal products. In closed systems with 

purified salts, these impurity effects should be limited and insignificant (after an initial short time period) 

compared with the persistent, continuous corrosion mechanisms. 

As discussed in Section 4, fission products that develop in the salt during reactor operation can also 

impact materials performance. Tellurium, for example, significantly impacts the mechanical performance 

of Hastelloy N in fluoride salts by promoting cracking via leaching of chromium from the alloy. 

However, the cracking can be reduced by controlling the redox potential of the salt through variation of 

the U3+/U4+ ratio, as shown in Figure 25. [23] This approach will also affect the dissolution rate of the 

alloy constituents. A proven alternative approach was using metal additions to form metal alloys, for 

example, adding niobium reduced the propensity for this type of embrittlement. [23] 

A rigorous predicative capability for corrosion and environmental effects in the molten salt operating 

environment is currently lacking. Although this gap is not unique with respect to MSR systems, the 

technology would certainly profit from research to advance the fundamental understanding of corrosion 

processes at the atomistic level, as well as provide the necessary underpinnings and experimental data for 

accurate modeling and simulation of materials behavior. Such developments would facilitate the 

identification and remediation of design issues while quantifying lifetime prediction for MSR concepts. A 

focus on predictive capabilities would provide the framework to conduct a coordinated experimental 

program in corrosion, mass transfer, physical and mechanical properties, and radiation effects. These 

would directly impact and inform the modeling effort and provide the means for validation of models and 

the simulations built on them. Such efforts need to address 

 Reaction pathways and kinetics 

 Prediction of failure mode and lifetime 
Molten Salt Chemistry Workshop

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 Materials, component, and structural design rules 

 Mechanistic underpinnings for input into modeling and simulation modules 

 Coupling of radiation and mechanical effects to corrosion and mass transfer phenomena 

Figure 25. Plot showing the effect of the salt oxidation potential (as measured by U4+/U3+) on the 

propensity for tellurium embrittlement of Hastelloy N. | Image courtesy of Oak Ridge National Laboratory

5.2.3 Understanding Degradation Processes at the Material–Salt Interface 

To understand the degradation of salt-facing structural materials, it is critical that the corrosion 

mechanisms at the material–salt interface be well understood. However, to date, studies have mostly 

relied on macroscopic ex situ measurements in static conditions, leaving a knowledge gap in our 

understanding of the nature of the material–salt interface and the kinetic parameters that determine its 

evolution with time. This knowledge gap needs to be addressed by designing, constructing and 

establishing techniques and methodologies for in situ characterization of the material–salt interface (see 

Section 5.2.1). In situ spectroscopy/diffraction analysis of material surfaces in the molten salt under both 

static and ex situ flow loop testing will greatly advance the understanding of relevant corrosion 

mechanisms. 

To aid in the understanding of interface behavior under realistic conditions, flow loops should be 

constructed for the integration of thermal gradients, fretting, and erosion into the mechanism studies (see 

Figure 20). In-core management and monitoring of corrosion is an area of research requiring much more 

investment, and flow loop testing is the ideal first step in investigating potential in situ monitoring 

methods and testing new materials. Flow loops will also allow for the effects of joining different 

structural materials together to enable a clearer understanding of the overall integrity of the system. Use 

of in situ experimental techniques to study the evolution of the material–molten salt interfacial chemistry 

is critical to understanding and predicting the degradation behavior and lifetime of the material of interest. 
Molten Salt Chemistry Workshop 

FRD 4 51 

Characterizing any surface deposits that are formed in the reactor as a result of exposure to molten salts 

and fission products, and identifying and quantifying the elements that are preferentially leached out of 

the material, will lead to an understanding of the chemical interactions and reaction kinetics between the 

molten salt and the material. Further, in situ material surface analysis can be combined with in situ salt 

studies using techniques such as microfluidic sampling and inductively coupled plasma optical emission 

spectroscopy. 

To accurately model material degradation in molten salt, accurate inputs are required; and without a solid 

base of experimental data, a modeling effort would be limited in usefulness. For this purpose, a 

coordinated effort is needed to gain fundamental experimental data on simple model systems, in which 

parameters can be accurately controlled and questions or hypotheses can be answered, coupled to more 

complex experiments in a test loop. Greater emphasis must be placed on the use of controlled, sciencebased experiments to find thermodynamic parameters, diffusion rates, and kinetic dissolution parameters 

that control corrosion and degradation of materials at the molten salt interface. Targeted experiments 

may include capsule testing, flow loops, calorimetry, or in situ techniques designed to measure the 

dissolution of species in salt. 

Using data specifically acquired to generate thermodynamic databases, thermodynamic models can be 

constructed. While thermodynamic modeling is not necessarily predictive, it is an important step toward 

more complex predictive models that account for corrosion, materials aging, and defect kinetics. 

CALPHAD (Computer Coupling of Phase Diagrams and Thermochemistry) [24] codes have proven to be 

a useful tool for making thermodynamic predictions and have already been used for calculating phase 

equilibria in salts. However, the calculation of phase equilibria relevant to alloy–salt interactions requires 

fundamental thermodynamic parameters, such as Gibbs free energies and solubilities, which can only be 

obtained from experimental data (see Section 2).

Beyond thermodynamic modeling, true predictive capability will require the development of multiphysics 

models that incorporate corrosion kinetics with radiation and salt chemistry. The purpose of these models 

is to use the mechanistic understanding of materials behavior to forecast the behavior of new conceptual 

materials in these extreme environments. This will allow for the development of materials with targeted 

properties. 

Computational methods need to be developed and coupled to experimental data to predict corrosion due to 

thermal gradients, fission products, and dynamic chemistries in the MSR. Models need to be tightly coupled 

to fluid flow, heat transport, and mass transport; and the production and dispersion of soluble gases, such as 

xenon or tritium (from neutron-induced fission of 6

Li), need to be understood. The exposure of the salt to an 

inert atmosphere in the pump section can be modeled as a boundary condition allowing the entrainment of 

gases (helium and/or argon) into the coolant. Understanding the distribution of tritium, when present, is 

important for determining the optimal placement of tritium-getters and determining places in the coolant 

loop where tritium might naturally deposit or evolve. To capture these effects, models need to be developed 

to determine a sink of tritium through off-gas or other chemical reactions. 

Understanding the interfacial chemistry will help predict material behavior and lifetime and accelerate the 

deployment of MSRs. As an added benefit, the development of these in situ techniques for molten salts 

will also help with real-time monitoring of the quality of the salts in the reactor and time-dependent 

transfer and disposition of materials for accountability. 

5.2.4 Understanding the Combined Effect of Chemistry and Radiation at the Interface 

The generation of fission products and their subsequent decay is well documented and understood and 

supports accurate predictions of their production in molten salt systems. This enables the prediction of 
Molten Salt Chemistry Workshop

52 FRD 4 

fission product concentrations with high accuracy. However, it does not explain the evolution of dissolved 

fission product concentrations completely because of the dynamic effects associated with short-lived 

fission products and the complex solubility behavior of many transition metals. Although previous studies 

have never revealed any new chemical species due specifically to radiation-induced-reactions in molten 

coolant or fuel salt, the possibility exists for new chemistry and compounds to be produced by the highly 

energetic fissioning salt environment. Some of these could have detrimental effects (especially if the 

reactor operates for 30 years). 

The most dramatic impact of ionizing irradiation occurs during the self-irradiation (due to fission and 

activation products) of solid salts at lower temperatures, with the production of metal centers and reactive 

halogens (F2), and must be considered for the case of spent salt storage and capsule irradiation (see 

Appendix A, Section 2.3.3). Based on the composition of the salt, the radiolysis effects of each solid salt will 

be unique. In the case of mixed halides—including mixed halides involving fission products such as iodofluorides and iodo-chlorides—interhalogens are very reactive, with different chemical interactions from those 

of pure halogen salts. In particular, boron and iodine are hard to remove, and they each have long-term effects 

in molten salts. High-intensity gamma-radiolysis studies of solid salt will need to be conducted as a precursor 

activity to in-pile salt irradiations to define the necessary safety envelopes for the experiment. 

The interactions at the gas–salt interface in the MSRE pump bowl appeared to be much more complex 

than in the salt solution or at the container-salt interface (see Figure 18); and molten salt mist was found 

in the gas space, leading to deposition of fission products. No corrosion acceleration—akin to that shown 

in Figure 26 for zirconium in water reactors—was seen during intense irradiation of coolant salt to high 

fluence and high power density. So far, only the tellurium embrittlement effect was obvious for fluoride 

fuel salt in the MSRE, as shown in Figure 27. 

Figure 26. Irradiation enhancement of corrosion rates.  B. F. Kammenzind, J. A. Gruber, R. Bajaj, and J. D. Smee. 

“Neutron irradiation effects on the corrosion of Zircaloy-4 in a PWR environment.” Paper B-T-3965, 18th International 

Symposium on Zirconium in the Nuclear Industry, Hilton Head, South Carolina, May 15–19, 2016. Used by permission.
Molten Salt Chemistry Workshop 

FRD 4 53 

 (a) (b) 

Figure 27. Cross-section analysis of INOR-8 (heat 5065) sample tested at 25 °C (a) after being exposed to 

the MSRE core for 19,136 h above 500 °C and irradiated to a thermal fluence of 1.51021 neutrons/cm2

and (b) after being exposed to static unenriched fuel salt for 19,136 h above 500 °C. | Image courtesy of 

Oak Ridge National Laboratory

5.2.5 Predicting Interfacial Interactions between Molten Salts and Structural Materials 

and/or Gases 

There is a need for predictive computational models to describe the molecular, mesoscale, and macroscale 

interfacial processes that occur in an MSR over a wide range of time scales (from nanoseconds to years) 

(see Figure 28). There are two primary types of interfaces relevant to MSRs: (1) molten salt fluid–solid 

and (2) cover gas–molten salt. Both interface types need to be understood in terms of fundamental 

interactions and interfacial transport processes. Relevant phenomena include 

 Noble metal (fission products) plating on both the structural materials and the moderator 

 Deposition of fissile materials on the structural interface and subsequent migration into the bulk solid 

phase 

 Nucleation of particles and transport and collection at the gas–liquid interface that adversely affect 

the mass transport of volatile fission products into the gas phase 

 Gas-phase entrainment into the molten salt (in the form of microbubbles from the cover gas) either by 

design or unintentionally 

 Entrainment of molten salts in the gas (mist formation) 

 Corrosion of structural components 

The underlying timescales span from nanoseconds, for gas diffusion, to months to years for corrosion. 

Current modeling approaches can treat fluids and solids on timescales that typically are from picoseconds 

to a microsecond at the atomistic/quantum level of theory, and microseconds to seconds at the mesoscale 

level of theory. Extending the time scale assailable by predictive simulations can provide essential 

missing knowledge of interfacial transport phenomena, in many instances available by no other practical 

means, thereby aiding the design approaches for commercial MSRs. 
Molten Salt Chemistry Workshop

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Figure 28. Illustration of the length and time scales (and inherent feedback) involved in the multiscale 

processes responsible for microstructural changes in irradiated materials. | Reprinted from B. D. Wirth, X. Hu, A. 

Kohnert, and D. Xu. “Modeling defect cluster evolution in irradiated structural materials: Focus on comparing to highresolution experimental characterization studies.” J. Mater. Res. 30, 1440–1455, 2015. Used by permission.

There are several critical needs that must be addressed for interfacial interactions between molten salts 

and structural materials and/or gases that pose significant challenges to the current computational-based 

approaches: 

 Prediction of long time processes (particularly regarding corrosion) by scaling up insights from 

molecular interactions 

 Prediction of interfacial transport processes 

 Prediction of surface layer formation leading to degradation of reactor performance with fission 

processes (e.g., metal plating) 

 Prediction of surface layer formation leading to degradation of heat exchangers 

 The ability to couple chemical and physical phenomena with microstructure and composition 

One of the challenges is that interfacial regions are inherently inhomogeneous, making transport 

phenomena near such regions difficult to quantify. The transport of reactive chemical species across 

interfaces and/or species reactivity at the interface are particularly challenging. Nevertheless, a concerted 

approach coupling information from simulations at the quantum level of theory (first principles MD), 
Molten Salt Chemistry Workshop 

FRD 4 55 

classical MD with accurate/informed force fields, and mesoscale simulations will provide far more insight 

into the physicochemical behavior of interfaces than is currently available. Polarization effects and 

chemical changes dominate phenomena at interfaces; therefore, the combination of methods is a necessity 

despite the challenges of passing information from one level of theory to another in either direction. This 

multiscale simulation approach will provide new understanding of interfacial transport phenomena in 

view of the challenges of experimental methods for probing molten salt interfaces under realistic 

conditions of reactor operation. Hence the research directions should follow from the following three 

platforms. 

1. Predict dynamics at molten salt–materials and molten salt–gas interfaces: First principles 

quantum mechanical MD (QM/MD) simulations are needed for modeling redox states of salt 

mixtures, fission and activation products, structural metals, and other impurities near the interfacial 

region to predict transfer of species to and from the salt across the interface. The results of this type of 

investigation provide information on the mobility of interfacial species; residence times in the 

interfacial region; and changes in interfacial structure—including the formation of complexes and 

nucleation of particles—that lead to precipitation and coating. With these data in hand, classical 

mechanical simplifications can be constructed to allow for longer time-scale and larger length-scale 

coarse-grained simulations. As with all modeling, comparison with experimental data is indispensable 

for validating calculated results and for imparting confidence in the computational results. 

2. Develop approaches for coupling chemical and physical phenomena with microstructure and 

composition: Various simulations of molten salts have been performed to study the structure and 

dynamics of bulk molten salt mixtures using classical MD (CMD) simulations. Equipped with new 

information from interfacial QM/MD, classical models can be built considering redox states, 

interfacial speciation, and polarization to carry out simulations that allow for species to cross the 

interfacial region at representative length and time scales. Of interest are species involved in 

corrosion of the solid–liquid interfaces, migration of fission products into structural material defects, 

and escape of volatiles into the gas phase in contact with the molten salt phase. In this effort, trends of 

mesoscopic mass transport can begin to be assessed and proposed for the next level of coarse-graining 

models aimed at stimulating rate-based processes 

3. Predict diffusion and chemical transformations in confined spaces near interfaces: Mesoscale 

models and transport simulations based on classical conservation laws of convective-diffusivereactive processes in nonequilibrium, when calibrated by subscale models (typically CMD), allow for 

further insight into trends of species transport and reactivity in interfacial regions on time scales of up 

to seconds and length scales of tens of microns. For example, phase field methods coupled with 

accurate thermodynamic data sets and kinetic data can provide detailed transport simulations with 

chemical reactivity, including nucleation and new phase formation. Together with cluster dynamics 

methods, mesoscale approaches can also address radiation damage processes within this framework. 

Experimental and reactor results can be used to improve the confidence of macroscopic simulations at the 

continuum level of theory commonly used to develop nuclear reactor simulation codes. Ultimately, this is 

the simulation tool used to design, build, and test prototype reactors. Therefore, tools are needed for the 

key steps toward connecting the research needed to bridge the knowledge gap and to connect directly to 

practical simulation tools used in the licensing space of nuclear reactors. These tools need to be made 

broadly available to the community with the necessary computational resources, software, and data sets 

(see Section 7 discussion on the Virtual Reactor Simulation). 

Progress in these research directions will lay down a scientific foundation for building models and 

performing simulations of interfacial transport phenomena of pure electrolytes at high temperatures 

(800 C and potentially higher) in the presence of solid and gaseous interfaces. This class of systems 

applies not only to MSRs but also to metallurgy, nonaqueous materials synthesis, electrochemical 
Molten Salt Chemistry Workshop

56 FRD 4 

processing, and solar energy capture and conversion to solar-based fuels. Therefore, the simulation tools 

developed in this effort will apply broadly to various crosscutting applications of relevance to the energy 

and materials sectors. Specifically, these tools will enable an improved understanding of impurity 

interactions at interfaces, of interfacial selectivity and binding, of the volatility of dissolved species, and 

of the effects of confined phases. Numerous crosscutting applications will benefit from the modeling and 

simulation framework developed, in addition to MSR research. 

The recommended computational research directions, supported and validated by experimental results, 

will pave the way to understanding and predicting a class of interfacial phenomena that are serious 

concerns in designing commercial MSRs with long operating lives (tens of years) that do not require 

reprocessing of the salt. For example, this work will enable the prediction of the corrosion of structural 

components by the salt at the inside surfaces and at the heat exchanger, enabling the design of new 

materials and processes to minimize corrosion and the prediction of material lifetimes under reactor 

conditions. The advances will enable the design of reactor materials and alloy compositions. This will 

accelerate the qualification of novel materials for MSR applications. Such work will improve the 

operando removal of fission products, xenon, krypton, iodide, and tritium. In addition, it will lead to an 

improved understanding of 

 The loss of tritium through diffusion in the vessel walls 

 Plating of surfaces by fissile products and noble metals 

 The role of confinement effects on interfacial transport of the entrained gas phase 

 Permeation of fission products and/or fissile species in structural materials with defects and/or grain 

boundaries 

References

1. J. H. DeVan and R. B. Evans III. Corrosion Behavior of Reactor Materials in Fluoride Salt Mixtures, 

ORNL/TM-328, Oak Ridge National Laboratory, 1962; J. H. DeVan, Effect of Alloying Additions on 

Corrosion Behavior of Nickel-Molybdenum Alloys in Fused Fluoride Mixtures, master’s thesis, 

University of Tennessee, August 1960 (see also reference 20). 

2. F. Zaera. “Probing liquid/solid interfaces at the molecular level.” Chem. Rev. 112, 2920–2986, 2012. 

3. T. F. Heinz, C. K. Chen, D. Ricard, and Y. R. Shen. “Spectroscopy of molecular monolayers by 

resonant second-harmonic generation.” Phys. Rev. Lett. 48(7), 478–481, 1982. 

4. K. B. Eisenthal. “Equilibrium and dynamic processes at interfaces by second harmonic and sum 

frequency generation.” Annual Rev. Phys. Chem. 43(1), 627–661, 1992. 

5. J. I. Dadap, J. Shan, and T. F. Heinz. “Theory of optical second-harmonic generation from a sphere of 

centrosymmetric material: Small-particle limit.” J. Optical Soc. Am. B 21 (7), 1328–1347, 2004. 

6. S. W. Chan, R. Barille, J. M. Nunzi, K. H. Tam, Y. H. Leung, W. K. Chan, and A. B. Djurišić. 

“Second harmonic generation in zinc oxide nanorods.” App. Phys. B 84(1–2), 351–355, 2006. 

7. J. Zeng, H. M. Eckenrode, S. M. Dounce, and H.-L. Dai. “Time-resolved molecular transport across 

living cell membranes.” Biophys. J. 104(1), 139–145, 2013. 

8. W. Fan, S. Zhang, N. C. Panoiu, A. Abdenour, S. Krishna, R. M. Osgood, K. J. Malloy, and S. R. J. 

Brueck. “Second harmonic generation from a nanopatterned isotropic nonlinear material.” Nano Lett. 

6(5), 1027–1030, 2006. 

9. P. S. Pershan and M. L. Schlossman. Liquid Surfaces and Interfaces: Synchrotron X-ray Methods. 

1st ed., Cambridge University Press, 2012, p 334. 
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10. P. Fenter and N. C. Sturchio. “Mineral–water interfacial structures revealed by synchrotron x-ray 

scattering.” Prog. Surf. Sci. 77(5–8), 171–258, 2004. 

11. M. K. Bera, G. Luo, M. L. Schlossman, L. Soderholm, S. Lee, and M. R. Antonio. “Erbium(III) 

coordination at the surface of an aqueous electrolyte.” Phys. Chem. B 119(28), 8734–8745, 2015. 

12. G. Luo, W. Bu, M. Mihaylov, I. Kuzmenko, M. L. Schlossman, and L. Soderholm. “X-ray reflectivity 

reveals a nonmonotonic ion-density profile perpendicular to the surface of ErCl3 aqueous solutions.” 

J. Phys. Chem. C 117(37), 19082–19090, 2013. 

13. G. Gao, F. H. Stott, J. L. Dawson, and D. M. Farrell. “Electrochemical monitoring of hightemperature molten-salt corrosion.” Oxid. Met. 33, 79–94, 1990. 

14. D. Ludwig, L. Olson, K. Sridharan, M. Anderson, and T. Allen. “High temperature electrochemistry 

of molten fluoride salt for measurement of dissolved chromium.” Corros. Eng. Sci. Technol. 46(4), 

360–364, 2011. 

15. A. G. Fernández, A. Rey, I. Lasanta, S. Mato, M. P. Brady, and F. J. Pérez. “Corrosion of aluminaforming austenitic steel in molten nitrate salts by gravimetric analysis and impedance spectroscopy.” 

Mater. Corros. 65, 267–275, 2014. 

16. L. F. Epstein. “Static and dynamic corrosion and mass transfer in liquid metal systems. Liquid metals 

technology.” Chem. Eng. Prog. Symp. Ser. 53(20), 67–81, 1957. 

17. R. B. Evans, III, and P. Nelson, Jr. Corrosion in Polythermal Systems, I. Mass Transfer Limited by 

Surface and Interface Resistances as Compared with Sodium-Inconel Behavior, vol. 1, ORNL-4575, 

Oak Ridge National Laboratory, 1971. 

18. J. H. DeVan. Effect of Alloying Additions on Corrosion Behavior of Nickel-Molybdenum Alloys in 

Fused Fluoride Mixtures, ORNL/TM-2021, Oak Ridge National Laboratory, 1969. 

19. W. N. Gill, R. P. Vanek, R. V. Jelink, and C. S. Grove, Jr. “Mass transfer in liquid -lithium systems.” 

A.I.Ch.E. J. 6, 139–44, 1960. 

20. C. E. Sessions and J. H. DeVan. “Thermal convection loop tests of Nb-1% Zr alloy in lithium at 1200 

and 1300 °C.” Nucl. Appl. and Technol. 9, 250–259, 1970. 

21. J. J. Keyes, Jr. Some Calculations of Diffusion Controlled Thermal Gradient Mass Transfer, CF 

report 57-7-115, Oak Ridge National Laboratory, 1957. 

22. D. F. Williams. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature 

Reactor (AHTR), ORNL/TM-2006/12, Oak Ridge National Laboratory, 2006. 

http://info.ornl.gov/sites/publications/Files/Pub57476.pdf

23. J. R. Keiser. Status of Tellurium-Hastelloy N Studies in Molten Fluoride Salts, ORNL-TM-6002, Oak 

Ridge National Laboratory, 1977. 

24. P. J. A. Spencer. “A brief history of CALPHAD.” Calphad 32(1), 1–8, 2008.

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6. FRD 5: Guiding Next-Generation Materials for Molten Salt Reactors 

6.1 Background and Current Status 

Structural materials (ranging from graphite to advanced metallic alloys) proposed for use in MSRs will be 

exposed to extreme environments, including high fluences of neutrons, high operating temperatures, and 

corrosive environments. Challenges associated with the high temperatures will primarily be the thermal 

stability of such materials, such as creep strength and fracture toughness; and irradiation effects could 

lead to structural evolutions such as irradiation-induced swelling, creep, or embrittlement. These effects 

could be exacerbated by the potential dissolution of solutes into or out of the structural materials from 

molten salt/structural material interactions. The compositions of the melt and structural material are 

expected to be the dominating factors influencing degradative processes. However, salt composition is 

also anticipated to change as a function of time because of the formation of transuranics, fission products, 

and corrosion products and the introduction of contaminants. These will create a highly complex and 

dynamic environment that will affect the structural materials. Despite the performance already 

demonstrated by candidate materials, the development of advanced materials that can mitigate 

degradation remains a key element for the safe, successful operation of structural components in MSRs. 

The stability of nickel-based alloys, as well as of steels, graphite, and ceramics, under various reactor 

irradiation conditions (real or simulated) has been the subject of study for decades; and the various modes 

of materials degradation are fairly well defined for different regimes of dose and dose rate. [1] 

Nevertheless, advances in the fundamental understanding of radiation damage and incremental 

improvements in the radiation tolerance of materials have been made, [2] and efforts to discover and 

design new materials with revolutionary stability under irradiation are being actively pursued. [3] 

Although many of these studies are not specific to MSRs, several have focused on concepts that require 

higher temperatures; but there have not been as many studies on radiation effects. Furthermore, there has 

been relatively little effort invested in designing new nickel-based alloys for radiation resistance because 

much of the attention of the community over the past 25 years has been on other materials systems (partly 

because of the observed poor phase stability of 1970s-era nickel-based superalloys after moderate 

irradiation doses at elevated temperatures).

Neutron irradiation effects on alloys in the temperature range of interest for MSRs are mostly cavity 

swelling, phase instabilities associated with radiation-induced segregation, irradiation creep, and helium 

embrittlement of grain boundaries. [1, 4] Neutron irradiation effects on graphite involve an initial volume 

shrinkage that is followed by volume expansion. [5] Unlike for metals, irradiation-induced creep is 

beneficial in the case of graphite because it can partially relieve some of the large stresses associated with 

high volumetric swelling. 

6.2 Technical Challenges and Research Directions 

6.2.1 Enabling Rapid Deployment of MSR Materials: Computational Challenges

MSRs represent a unique materials challenge with limited compatible materials availability and open 

questions regarding material lifetimes and durability. New improved methods are needed which combine 

experimental characterization data with predictive modeling capabilities to develop new materials, 

including alloys and composites with the required radiation and chemical stability that are optimized or 

qualified for MSR use. In addition, the time for new materials development needs to be greatly accelerated 

to meet required timelines for testing and code qualification purposes. Addressing these needs requires 

that an innovative combination of computational materials science and high-throughput combinatorial 

material testing be developed. A comprehensive set of thermochemical models and data are needed to 

elucidate how elemental content, reduction potential, and temperature interact to control chemical 
Molten Salt Chemistry Workshop

60 FRD 5 

potential, thermophysical properties, and viscosity and predict such properties at the required level of 

accuracy. Additional studies and computational methods are needed to increase knowledge of the 

microstructure-property-corrosion correlation to improve the ability to successfully develop needed 

material solutions for MSRs. Finally, new efficient synthesis and processing methods are needed to 

prepare materials compatible with MSR operation. 

Methods are needed to predict material behavior in complex molten salt environments. MSRcompatible materials can be designed to exhibit optimal behavior in complex molten salt environments if 

researchers can understand and predict the interactions at the interface between the material and the salt, 

as discussed in Section 5. This represents a significant challenge because the salt chemistry and properties 

change with time, temperature, and burnup as fission products and tritium are produced. It is imperative 

that these computational models be dynamic and comprehensive in treating these interface reactions. 

Achieving this capability will require close coupling between model development, material testing in salt 

systems, and fundamental investigations of salt interfaces affording an atomistic understanding of the 

resulting physicochemical phenomena. Such modeling capabilities are at the crossroads of traditional 

computational material science, computational chemistry, and experimental materials science and 

experimental physical chemistry. A challenge will be to understand the MSR salt chemistry evolution in 

parallel to material testing so that the right experimental conditions are included. 

There is a need to understand and predict the microstructural evolution of materials in dynamic 

radiation and chemical environments. MSRs create a unique, dynamic chemical and radiation 

environment where any given structural material (e.g., graphite or Hastelloy N) will be undergoing 

simultaneous radiation damage and chemical attack while the MSR salts are undergoing radiolysis. This 

situation represents a significant challenge in that bulk radiation damage of structural materials is coupled 

to interfacial salt–material interactions. The effort to address these issues can draw on multiscale 

computational methods that are being developed for the study of materials operating under irradiation 

conditions or of plasma surface interactions. Radiation damage evolves microstructures in known ways, 

which include solute segregation that will compete with salt–interface reactions to produce a unique 

interfacial reaction zone. Ideally, this interface segregation can be optimized to provide a protected or 

stable interfacial region during MSR operation. Fission product uptake and diffusion, along with tritium 

uptake and helium production in the structural material, represent a further challenge in material design. 

MSR materials must be designed to mitigate high-temperature helium embrittlement and fission product 

corrosion. It will be necessary to predict many of these phenomena—such as helium-bubble nucleation at 

grain boundaries—over multiple time and length scales. For many materials, MSR operational 

temperatures provide relief from radiation hardening and embrittlement but push the material into a 

regime in which helium embrittlement can become an issue as helium is transported to grain boundaries 

and retained. This requires the ability to couple detailed helium trap models with helium transport and 

mechanics models to design materials that are resistant to helium production. 

To identify and design structural materials with MSR compatibility, it is necessary to develop novel 

multi-resolution (in terms of physics, length scales and time scales) modeling capabilities (see Figure 

29). This is a complex challenge with no suitable methods yet developed that can address these multiple 

time and length scales. Ab initio and classical methods will be required for structural material–salt 

interactions and for radiation damage resistance. The latter can be partly addressed using classical 

approaches coupled with cluster dynamics, spatial-dependent rate theory, kinetic Monte Carlo, and phase 

field or other mesoscale methods. However, as yet there are no suitable methods that can address 

radiation damage, helium trapping and migration, and tritium production and uptake for complex alloy 

systems. In any event, some starting point is required, such as Hastelloy N, for example, for alloy design. 

Chemical reaction models will need to be integrated with damage and deformation models to capture the 

range of corrosion, radiation damage, and thermal degradation that are anticipated. 
Molten Salt Chemistry Workshop 

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The study of structural material–salt interactions will require the development of improved ab initio 

methods (in terms of speed) or improved classical potentials for alloy/salt systems to describe the 

complex corrosion and microstructural evolution that will occur at the material–salt interface. Rapid 

deployment approaches will rely on close coupling of computational methods and rapid characterization 

of materials exposed to salt test loops. However, the lack of knowledge regarding concurrent radiation 

damage cannot be overcome using this approach. Thus, a MSR test reactor system is required. 

The paradigm envisioned in Figure 29 will enable the elucidation of the fundamental chemomechanical 

degradation mechanisms leading to corrosion at molten salt–structural materials interfaces, in addition to 

bulk radiation damage and thermomechanical degradation [6–9]. The ability to adequately simulate the 

response of such microstructures, to elucidate the interdependent compositional and microstructural 

evolution, and to gauge the associated change in thermomechanical properties in the extreme 

environments experienced in MSRs is crucial and necessary. Such a capability is essential not only for 

preventive and remedial actions to ensure safety and for performance evaluation over the lifetimes of 

MSR components, but also for the development of new generations of high-performance, corrosionresistant MSR materials. 

Compatibility with salt systems will require thermodynamic data for MSR salt fuels, as discussed in 

Section 2. The complex elemental content of a thermochemical solution model, together with those 

necessary for the panoply of crystalline phases that can precipitate, is unprecedented. Research is 

necessary to develop methods that efficiently provide accurate input data for such models and 

accommodate means of statistical validation. Determining the most advantageous salt composition will 

require the development of relationships for use in optimization algorithms. 

Advanced computational methods are needed to model and predict corrosion as a function of 

microstructure evolution. Any framework designed to improve the modeling and prediction of the 

corrosion behavior in MSR materials must be based on the concurrent coupling between composition 

evolution, microstructure evolution, localized deformation, and damage accumulation. Atomic-

/molecular-scale ab initio–based input will be necessary to calibrate this type of model and capture the 

adequate fundamental mechanisms influencing corrosion in MSR salts. These models can then be 

upscaled to phase field models, for example, to predict phase evolution and corrosion at the microscale 

and macroscale. The complexity envisioned for these models, however, is unprecedented and will require 

new methods of coarse-graining materials or upscaling to comprehensive mesoscale approaches. 

Models are needed to predict materials mechanical degradation processes under MSR conditions. The 

mechanical properties of the materials in an MSR, such as creep strength and fracture toughness, will 

gradually degrade over time because of the severe environmental exposure. Although adequate models 

exist for such degradation mechanisms, the MSR environment is unique. For example, models for highheat-flux materials for advanced fusion reactors are generally inadequate when plasma surface 

interactions must be coupled to bulk radiation damage models of hardening. Likewise, MSR corrosion 

models must be coupled with similar hardening models using new methods. Reactor transients, such as 

startup and shutdown, exacerbate this problem by introducing nonequilibrium states into the degradation 

mechanism space. This adds additional complexity by forcing the models to consider time-dependent 

stresses instead of steady state stresses. As a starting point, radiation damage models of hardening and 

swelling should be investigated. 
Molten Salt Chemistry Workshop

62 FRD 5 

Figure 29. Example of multi-resolution modeling capabilities for microstructure evolution and corrosion. |

(From left) W. Barrows, R. Dingreville, and D. Spearot. “Traction–separation relationship for hydrogen induced grain 

boundary embrittlement in nickel via molecular dynamics simulations.” Mater. Sci. Eng., A 650(5), 354–364, January 2016; 

F. Soisson. “Kinetic Monte Carlo simulations of radiation induced segregation and precipitation.” J. Nucl. Mater. 349, 235–

250, 2006; X. Q. Ma , S. Q. Shi, C. H. Woo, and L. Q. Chen. “Simulation of -hydride precipitation in bi-crystalline zirconium 

under uniformly applied load.” Mater. Sci. Eng., A 334, 6–10, 2002; A. Musienko and G. Cailletaud. “Simulation of interand transgranular crack propogation in polycrystalline aggregates due to stress corrosion cracking.” Acta Mater. 57, 

3840–3855, 2009; A. Y. Dunn. Radiation Damage Accumulation and Associated Mechanical Hardening in Thin Films and 

Bulk Materials. PhD Thesis, Georgia Institute of Technology. 2017.

Predictive multi-resolution modeling capabilities, addressing the degradation of structural materials 

subjected to extreme environments experienced in MSRs, will not only advance our current understanding 

of corrosion mechanisms due to molten salts but also create a new paradigm in the computational 

materials science and computational chemistry community for studying corrosion phenomena for a wide 

range of applications beyond MSRs. Indeed, since corrosion of materials is one of the main life-limiting 

processes for materials in a variety of settings, and is even more critical for MSRs, the development of 

improved corrosion models will greatly benefit predictive capabilities in a variety of fields. This research 

will greatly facilitate alloy development by advancing innovative ways of coupling advanced 

characterization methods and data sets with advanced computational models of salt/material interactions 

and bulk materials degradation models. The advanced chemistry approaches developed will have wideranging impacts in a variety of material-centric fields.
Molten Salt Chemistry Workshop 

FRD 5 63 

The ability to predict material properties under extreme conditions is essential to understanding how 

materials degrade and fail. The ability to accurately predict material lifetimes provides needed safety 

information as well as a foundation for the design of improved materials. The rational design of materials 

using these models will allow new and accelerated approaches for material deployment in extreme 

environments, such as MSRs, by increasing materials testing throughput. Tailored and optimized 

materials will enable MSRs and other technologies required for future energy production. These methods 

could impact materials design efforts for other advanced nuclear systems or other high-temperature 

material systems.

6.2.2 Guiding Experimental Exploration of Advanced “Super” Materials 

Although much work has focused on the quantification of current materials for use in MSRs, there exists 

a need for the development of new materials to meet the increased demands of advanced reactor concepts. 

The development of INOR-8 (Hastelloy N) after the ARE and its deployment in the MSRE serves as an 

example of a successful development of a new material for use in molten salt environments. Although 

reactor vendors are currently pursuing designs that incorporate alloys from the current ASME Boiler 

Pressure Vessel Code, there is a tremendous opportunity to take advantage of innovative advanced 

materials or “super materials.” Although code-qualifying a new alloy or material can take over a decade 

and cost tens of millions of dollars, the benefits in terms of operational life, maintenance costs, and safety 

could be significant. Nevertheless, most companies are not willing to investigate non-code materials 

because of the time required and the cost. 

Current materials, including Hastelloy N, cannot effectively address the higher temperatures and more 

aggressive radiation and chemical environments presented by some advanced MSR concepts. Proposed 

reactors with core temperatures above 900 °C and radiation damage over 1000 displacements per atom 

(dpa) represent pathways to energy independence through the use of tailored thermochemical processes 

(e.g., thermochemical H2 cycles), increased power-cycle thermal efficiency, and added utility from 

enabling flexible fissile fuel options. To achieve these goals requires developing new classes of super 

materials that can withstand the more aggressive conditions. Materials must be sufficiently 

thermodynamically stable in the coolant/fuel (i.e., molten salt), accommodate imposed mechanical loads, 

and be immune to significant dimensional and property changes from radiation damage over the design 

lifetime of the reactor system. Performance requirements must be balanced with practical considerations 

such as availability, affordability, fabricability, and manufacturability and acceptance by regulatory 

bodies. 

High temperatures are typically associated with loss of strength, creep, and higher corrosion rates. High 

dpa from exposure to neutron flux is associated with hardening, void swelling, and helium embrittlement. 

In MSRs, the challenge is compounded by temperature gradient, mass transfer, corrosion, and reactions 

with fission products, i.e., tellurium. The materials design space for MSRs includes the use of metals, 

ceramics, coatings, claddings, multilayer and nanostructured materials and, as necessary, composites. 

This variety of materials needs to be explored comprehensively to discover the super materials needed for 

maximum MSR performance. There is also a need for selective design of materials to ensure lifetime 

requirements are met. These requirements include maintaining material strength at temperatures in excess 

of 750 °C without suffering cracks or embrittlement, resisting corrosion, and also maintaining size and 

shape within the design specifications. 

The development cycle for new structural materials can be shortened through the application of modern 

materials science computational tools, as discussed earlier. Computational studies can be performed to 

down-select favorable chemical compositions from thermodynamic and kinetic stabilities; predict 

properties; and simulate component lifetimes through known/verified mechanisms of time-dependent 

mechanical, corrosion, and radiation behavior before trial materials are manufactured and properties 
Molten Salt Chemistry Workshop

64 FRD 5 

screened (see the sidebar “Optimizing Superalloys” on p. 65). This approach is an improvement over the 

traditional development path of fabricating trial materials from a large chemistry matrix for screening 

tests, and it has been successfully employed in a number of DOE–NE programs. Improvements in the 

accuracy of these tools and the development of new computational tools and databases to cover additional 

material behavioral responses would further shorten the development cycle. When the justification exists 

to actually start materials testing, there is also the possibility of much smaller alloy synthesis routes 

compared with the traditional large “heats” of years past. 

It also needs to be recognized that some of the components in an MSR will have different requirements at 

different locations, leading to more severe restrictions on material development. This issue could be 

addressed by developing innovative advanced manufacturing methods, such as additive manufacturing, 

that would allow tailoring of properties via the adjustment of compositions throughout the structural 

component. In addition, innovative methods of materials development can have a significant scientific 

impact on how material design is performed. This impact may lead to a deeper understanding of the 

underlying scientific principles of how materials behave in extreme environments and how they are best 

tailored for use in high-temperature, molten salts environments that are subjected to high radiation levels. 

Furthermore, the current limitations in understanding the joining of dissimilar materials, such as joining 

ceramics to metal seals, must be addressed; doing so would be an enabling technology. 

6.2.3 Advancing Code Qualification of New Materials 

A significant hurdle for the development of MSR structural materials is code qualification. The design of 

high-temperature coolant boundary structural components for deployment in nuclear reactor systems is 

required to follow the construction rules of the ASME Boiler and Pressure Vessel Code (Section III, 

Division 5). The qualification of new metallic materials requires multiple steps, including composition 

selection, processing and heat-treating steps, and finding a vendor willing to produce large quantities of 

the material. 

After development, ASME-based qualification of the metal includes long-time scale testing at operational 

temperatures (anywhere from 1/5 to 1/3 of the expected usage lifetime), preferred examples of the 

material applicability in other non-nuclear extreme environments, corrosion testing, and irradiationinduced changes to the physical/thermal/mechanical properties. Many of these steps are not easily 

accelerated, since the mechanisms controlling the changes may be not be the same at different conditions. 

Testing at operational temperatures could easily require 6–10 years of continuous testing for a component 

for which a 30 year usage lifetime is desired. In many cases, such a long period of testing is costprohibitive; therefore, materials used in other industries often are leveraged toward ASME code 

validation. 

Study of the changes caused by the irradiation environment in the reactor also is very time-intensive. In 

the current materials test reactors (High Flux Isotope Reactor and Advanced Test Reactor), lifetime 

damage accumulation can be sped up by a factor of 5–10. As a result, a 30 year lifetime component 

requires a minimum of 3–6 years of irradiation time alone (while the development and construction of the 

irradiation capsules and/or the post-irradiation examination of the properties and microstructure could 

expand the testing program to 7–10 years). Corrosion studies in the MSR environment will also require 

experiments to study the base metal–salt interaction at high temperature, as well as exposure to the salt in 

an irradiation environment, since the chemistry of the salt will be continuously changing as fission occurs. 

These experiments are technically challenging and require complex experimental setups in addition to 

long times to perform the tests. These issues highlight the need for innovative experimental test protocols 

and analytical methods to efficiently gather the required information for code qualification. 
Molten Salt Chemistry Workshop

FRD 5 65

The qualification of nonmetallic materials requires a similar development process to that for metallic 

materials. The primary difference is that qualification for nonstructural materials usually does not include 

long-term testing at operational temperatures (not required for ASME code qualification). But long-term 

studies of irradiation-induced changes and corrosion behavior are still critical knowledge requirements for 

in-reactor use.

Optimizing Superalloys

Nickel-based superalloys have an exceptional combination of high-temperature mechanical 

properties, primarily as a result of the formation of coherent nano-precipitates of the Ni3Al L12-

type in the nickel-based face-centered-cubic matrix. However, achieving similar microstructure and 

performance in other low-cost materials continues to be a challenge. Today it is feasible to use a 

modern accelerated approach to high-temperature structural alloy development based on atomiclevel control of interfaces. Predictive design of favorable properties in materials under extremes 

requires that the microstructural interfaces, such as grain and phase boundaries, be stable under 

high temperature and stress. While phase diagrams for stability of individual phases are often well 

known, an integrated understanding of the interfacial stability in the presence of individual or 

multiple atomic species is still needed. Microalloying elements that maximize stress and 

temperature resistance by computing the energetics of higher mobility interfaces are an aspect that 

needs to be verified and refined through experiments. Large-scale, high-throughput ab initio 

calculations can be the basis for simulating interfacial evolution at high stress and homologous 

temperature (ratio of temperature and absolute melting point). A suite of experimental 

characterization techniques—such as atom probe tomography, transmission electron microscopy, 

neutron diffraction, and mechanical characterization—can be used to help verify the computational 

predictions. Such models and experimental results can form a scientific framework for

understanding and controlling interfacial stability in superalloys, thereby dramatically accelerating 

the design of next-generation, advanced alloys optimized for molten salt reactors.
Molten Salt Chemistry Workshop

66 FRD 5 

References 

1. S. J. Zinkle and J. T. Busby. “Structural materials for fission and fusion energy.” Mater. Today 12, 

12–19, 2009; R. M. Boothby. “Radiation effects in nickel-based alloys,” in ed. R.J.M. Konings 

Comprehensive Nuclear Materials, vol. 4. Elsevier, Amsterdam, 2012. pp. 123–150; A. F. Rowcliffe, 

L. K. Mansur, D. T. Hoelzer, and R. K. Nanstad. “Perspectives on radiation effects in nickel-base 

alloys for applications in advanced reactors.” J. Nucl. Mater. 392, 341–352, 2009. 

2. S. J. Zinkle and L. L. Snead. “Designing radiation resistance in materials for fusion energy.” Annu. 

Rev. Mater. Res. 44, 241–267, 2014. 

3. J. B. Roberto and T. Diaz de la Rubia. Basic Research Needs for Advanced Nuclear Energy Systems. 

Office of Basic Energy Sciences, US Department of Energy, 2006; Y. W. Zhang, G. M. Stocks, 

K. Jin, C. Y. Lu, H. B. Bei, B. C. Sales, L. M. Wang, L. K. Beland, R. E. Stoller, G. D. Samolyuk, et 

al. “Influence of chemical disorder on energy dissipation and defect evolution in concentrated solid 

solution alloys.” Nature Comm. 6, 8736–8739, 2015; G. R. Odette, “Recent progress in developing

and qualifying nanostructured ferritic alloys for advanced fission and fusion applications.” JOM 66, 

2427–2441, 2014. 

4. T. R. Allen, M. Anderson, R. Ballinger, T. D. Burchell, E. D. Blandford, D. Clark, W. R. Corwin, 

G. Flanagan, C. W. Forsberg, P. Hosemann, et al. “Fluoride-salt cooled, High-temperature Reactor 

(FHR) Materials, Fuels and Components White Paper,” UCBTH-12-003, Department of Nuclear 

Engineering, University of California, Berkeley, California, 2013. 

5. L. L. Snead, T. D. Burchell, and Y. Katoh. “Swelling of nuclear graphite and high-quality carbon 

fiber composite under very high irradiation temperature,” J. Nucl. Mater. 381, 55–61, 2008; L.L. 

Snead. “Fusion energy applications,” in ed. T. D. Burchell, Carbon Materials for Advanced 

Technologies. Pergamon, New York, 1999, pp. 389–427. 

6. W. Barrows, R. Dingreville, and D. Spearot. “Traction-separation relationships for hydrogen induced 

grain boundary embrittlement in nickel via molecular dynamics simulations.” Mater. Sci. and Eng., A 

560(5), 354–364, 2016. 

7. F. Soisson. “Kinetic Monte Carlo simulations of radiation induced segregation and precipitation.” 

J. Nucl. Mater. 349(3), 235–250, 2006.

8. A. Musienko and G. Cailletaud. “Simulation of inter-and transgranular crack propagation in 

polycrystalline aggregates due to stress corrosion cracking.” Acta Mater. 57(13), 3840–3855, 2009. 

9. A. Dunn, R. Dingreville, E. Martínez, and L. Capolungo. “Synchronous parallel spatially resolved 

stochastic cluster dynamics.” Comput. Mater. Sci. 120, 43–52, 2016. 
Molten Salt Chemistry Workshop 

FRD 6 67 

7. FRD 6: Creating a Virtual Reactor Simulation 

7.1 Background and Current Status 

MSR technology development requires the use of modeling and simulation to understand the behavior of 

the reactor throughout its lifetime, including during normal operation, operational transients (such as 

startup and shutdown), and accident conditions. To accurately predict reactor conditions, an appropriately 

validated reactor system simulator is needed which incorporates the major physical phenomena that drive 

design and operational challenges at a suitable level of fidelity and accuracy. A virtual reactor (VR) 

simulator needs to include the reactor core and primary heat exchanger, plus related support systems, such 

as off-gas and online salt processing and refueling. The VR simulator should describe phenomena 

including neutron transport, thermal hydraulics, isotopic transmutation of the fuel, thermochemical 

properties of the fuel salt, and resulting corrosion effects, while also possessing sufficient fidelity to 

address material transport due to corrosion and thermal gradient redeposition. The VR would also define 

the source term for the MSR. This effort represents a significant scientific challenge because of the 

coupling of transport codes with thermodynamic models and material–salt chemistries, but it can build off 

the foundational studies outlined in the previous sections of this report. 

7.2 Technical Challenge and Research Directions 

7.2.1 Creating a Virtual Reactor Simulation 

The challenges associated with the deployment of MSR technology are summarized by Briant and 

Weinberg [1]: 

Two very different schools of reactor design have emerged since the first reactors were 

built. One approach, exemplified by solid fuel reactors, holds that a reactor is basically a 

mechanical plant; the ultimate rationalization is to be sought in simplifying the heat 

transfer machinery. The other approach, exemplified by liquid fuel reactors, holds that a 

reactor is basically a chemical plant; the ultimate rationalization is to be sought in 

simplifying the handling and reprocessing of fuel. 

The existing assumptions and simplifications in current modeling and simulation tools are specialized for 

solid-fuel systems; adapting these tools for liquid fuel systems requires significant effort, particularly in 

the coupling to the molten salt chemical analysis, as previously discussed. The major scientific challenge 

is to develop an accurate simulation of a MSR, including flowing fuel, neutronics, transport, chemistry, 

and in situ continuous processing and refueling. The model would need to be dynamic, which represents 

a grand challenge because of the complex material properties and transport. The simulator needs to 

include the primary loop, i.e., the reactor core and the primary heat exchanger, and related support 

systems such as off-gas and online salt processing (if needed) and refueling. The phenomena described by 

the simulator must include neutron transport, thermal hydraulics, isotopic transmutation of the fuel, 

thermochemical properties of the fuel salt, and corrosion effects. The simulator must also be validated 

with well-quantified uncertainties, since the nuclear industry and regulating authorities will need such a 

tool to ensure the efficiency and safety of this new technology. 

The neutronic response of the reactor depends on salt properties such as density and viscosity, which will 

evolve with depletion of the fuel. Changes in fuel composition may also change the preferred speciation 

of elements, the liquidus temperature of the melt, and surface-salt interactions. Additionally, tritium 

produced by ternary fissions or reaction with salt constituents will easily permeate hot metal alloys and 

needs to be taken into account. All these phenomena are coupled together in a MSR system and need to 

be accurately described by the simulator. 
Molten Salt Chemistry Workshop

68 FRD 6 

A transformational advancement of a VR is a salt thermochemical transport modeling and simulation 

capability. While this is important for solid-fueled MSR designs, a more extensive contribution will be 

made for liquid-fueled systems. The salt thermochemical transport will focus on describing the redox 

reactions in the flowing salt and reactivity leading to corrosion, in addition to radiolysis and chemical 

speciation in the liquid and gas phases. Online monitoring can provide input and validation of these 

computational predictions. The monitoring will be augmented by considerations of reactions with 

impurities, and of the nucleation and growth of particulates, which have ramifications for performance, 

safety, and off-gas operation. 

The molten salt VR must be able to determine the fluence on the core structural material, the neutron 

reflector, and the vessel wall, which can be used along with the temperature values and salt composition 

to determine the lifetimes of these components. Additionally, these data will enable analysis of tritium 

production and the migration through the primary loop. It can also be used to optimize alternative salts 

that would reduce tritium production while maintaining good nuclear and chemical properties. The VR 

must also be able to couple to power plant system models to investigate transient and safety analysis of 

these reactors. 

The development of a molten salt VR can leverage existing neutronic and thermal hydraulic capabilities, 

but it will need to be extended to capture the complexities of MSRs. A specific challenge that will be 

addressed is the “delayed neutron precursor drift” phenomenon, which results in the birth of delayed 

neutrons (i.e., the decay of precursors) in a different location from the fission that produces the precursor. 

In solid-fuel designs, the fission products are stagnant; but with a liquid-fuel design, they are carried away 

with the fluid fuel flow. To capture this effect, delayed neutron precursor groups need to be explicitly 

modeled even during steady state operation. The standard delayed neutron precursor models can be 

adapted by adding in a convection term. Since the change in the core flow speed changes the effective 

delayed neutron fraction in the core, this feature is critical for modeling power transients during accidents 

such as a pump trip. In addition to the delayed neutron precursors, there are other elements that require 

special treatment because of the flowing fuel, such as xenon, which is a neutron poison. Also, the 

redistribution of heat coming from nuclear decay must be properly dispersed throughout the system. 

Several key advancements are needed to capture the thermal hydraulic performance of MSRs. The natural 

first extension is to add the capability for direct fluid heating as the dominant heat source. Another key 

feature is the ability to transport nuclides throughout the system. This capability needs to be extended to 

capture multiple species at the same time, which will lead to a coupled set of species transport equations 

to account for the interaction between species. An additional complication is the thermophysical 

properties variation with concentration of species; since fission products will be generated pointwise, and 

convected nonuniformly, the implementations of such variations must be considered. In addition, models 

or correlations for dynamic phase diagrams for estimating thermophysical properties and melting point 

will be required. 

The addition of mass transport and thermochemistry requires information from both the heat and the 

momentum fields of the bulk fluid. Additionally, the reactor physics component should provide a 

radiation environment that will continually create fission events and transmute the components of the salt 

as they are irradiated. It is necessary to account for these effects in considering the thermochemical 

reactions in the reactor. The mass transport and thermochemistry will also impact the reactor physics and 

the thermalhydraulic solution. The reactor physics model will need the detailed distribution of salt 

components throughout the core. For short-lived fission products, there may be a non-trivial gradient in 

the concentration during standard reactor operation that must be accounted for. Likewise, the properties of 

the molten salt—density, viscosity, thermal conductivity, and specific heat—will change as the salt 

changes. Both feedback mechanisms must be accounted for. Figure 30 shows the three major components 

of a VR and the data that are needed between the various components. 
Molten Salt Chemistry Workshop 

FRD 6 69 

Figure 30. A schematic showing the three major components of a virtual reactor simulation and the data 

that are needed between the various components. | Image courtesy of Oak Ridge National Laboratory

The virtual simulator code will need validation and uncertainty quantification (VUQ). The VUQ will 

require sets of separate effects and integral effect tests to quantify uncertainty for each phenomenon and 

each of the implemented multiphysics components. Data are available from international molten-salt loop 

tests that can be used for validation. Additionally, the MSR VR should be used to simulate the ORNL 

MSRE operation and demonstrate the capability developed to measured data from an operating reactor. 

The MSRE operation represents the only data available in the world for a well-characterized operating 

MSR and will be crucial for system-level validation. 

Two sets of critical experiments were performed at MSRE. The first was performed using 235U fuel 

loading. [2] The data measured include the critical position as well as the measurement of various 

reactivity coefficients, including temperature, enrichment, differential control rod worth, and the effects 

of fuel circulation. The second was performed using 233U fuel. [3] Very similar measurements were 

performed with the different driver fuel. Additionally, the dynamic behavior of the MSRE was studied in 

detail for these different fuels, [4-6] and its fuel flow behavior characterized. [7] 

It is recognized that an MSR is a tightly coupled dynamic multiphysics system consisting of neutronics, 

thermal hydraulics, fission product generation and transport, and the chemistry of constituents. The 

thermophysical properties will be changing as the fuel moves through the system. The solubility of 

solutes will change as new fission products are added. Developing a VR simulator will require 

specification of figure of merit or safety parameters and identification and ranking of phenomena with 

respect to the figure of merit. Then each physics will have its own constitutive relationships that may 

come from experiments or first principles. We anticipate that advanced 3-dimensional meshing 

algorithms will need to be developed to adequately handle these demands, with perhaps a phase field 

model handling the thermochemistry coupled to other transport models. 

Overall, the capabilities of a virtual MSR simulator with a tightly coupled model with fuel flow, salt 

chemistry, surface corrosion, neutron transport, and thermal hydraulics would enable advanced design 

and analysis of complex fluid-fueled systems. This would allow for the understanding of operational 

issues including fuel cycle analysis, fuel loading analysis, prediction of waste forms and quantities, and 
Molten Salt Chemistry Workshop

70 FRD 6 

prediction of chemical processes required to reduce chemistry issues. The virtual model would also define 

the source term for the MSR. In addition, this capability will allow for the assessment of new molten salts 

and new reactor designs. 

Another application of this simulation capability would be in the design of irradiation test stands and the 

analysis of results from irradiation experiments that will be needed for the development of MSR 

technology. Nearly all of the phenomena that need to be simulated in a reactor can be probed to some 

extent in irradiation tests of increasing complexity, and this simulation capability can be of great value in 

designing the irradiation experiments. 

References

1. R. C. Briant and A. Weinberg. “Molten fluorides as power reactor fuels.” Nucl. Sci. Engr. 2, 797–

803, 1957. 

2. B. E. Prince, S. J. Ball, J. R. Engel, P. N. Haubenreich, and T. W. Kerlin. Zero-Power Physics 

Experiments on the Molten Salt Reactor Experiment, ORNL-4233, Oak Ridge National Laboratory, 

February 1968. 

3. J. R. Engel and B. E. Prince. Zero-Power Experiments with 233U in the MSRE, ORNL-TM-3963, Oak 

Ridge National Laboratory, December 1972. 

4. T. W. Kerlin, S. J. Ball, R. C. Steffy, and M. R. Buckner. “Experiences with dynamic testing methods 

at the Molten-Salt Reactor Experiment.” Nucle.Appl. Tech. 10, 2, February 1971. 

5. T. W. Kerlin and S. J. Ball. Experimental Dynamic Analysis of the MSRE, ORNL-TM-1647, Oak 

Ridge National Laboratory, October 1966. 

6. R. C. Steffy. Experimental Dynamic Analysis of the MSRE with 233U Fuel, ORNL-TM-2997, Oak 

Ridge National Laboratory, April 1970. 

7. R. J. Kedl. Fluid Dynamic Studies of the MSRE Core, ORNL-TM-3229, Oak Ridge National 

Laboratory, November 1970. 
Molten Salt Chemistry Workshop 

Appendix A A-1 

APPENDIX A: WORKSHOP RESOURCE DOCUMENT 

Publicly available at https://info.ornl.gov/sites/publications/Files/Pub73652.pdf and included in 

the following pages. 

A-3
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makes any warranty, express or implied, or assumes any legal liability 

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A-4
ORNL/LTR-2017/135 

TECHNOLOGY AND APPLIED R&D NEEDS

FOR MOLTEN SALT CHEMISTRY

Resource Document for the

Molten Salt Chemistry Workshop: 

Innovative Approaches to Accelerate Molten Salt Reactor

Technologies Deployment

Prepared by David F. Williams

Oak Ridge National Laboratory

Date Published: April 2017 

Prepared by 

OAK RIDGE NATIONAL LABORATORY

Oak Ridge, TN 37831-6283 

managed by 

UT-BATTELLE, LLC

for the

US DEPARTMENT OF ENERGY

under contract DE-AC05-00OR22725 

A-5

iii

CONTENTS 

CONTENTS .......................................................................................................................... iii

FIGURES ................................................................................................................................v

TABLES .................................................................................................................................v

ACRONYMS ....................................................................................................................... vii

1. INTRODUCTION...........................................................................................................1

2. MOLTEN SALT REACTOR CHEMISTRIES ..............................................................3

2.1 PHYSICAL CHEMISTRY CHALLENGES .........................................................4

2.1.1 Reasoning Framework for Salt Choices .....................................................4

2.1.2 Phase Diagrams and Solubility Behavior ...................................................6

2.1.3 Thermophysical and Transport Properties .................................................7

2.1.4 Modeling and Simulation ...........................................................................9

2.2 ANALYTICAL CHEMISTRY CHALLENGES (including measurement 

instruments and online sensors) .............................................................................9

2.3 IRRADIATION AND FISSION PRODUCT CHEMISTRY CHALLENGES ...11

2.3.1 Molten Salt Fission Product Chemistry ....................................................11

2.3.2 Barren-Salt Activation Product Chemistry: Tritium and Sulfur ..............14

2.3.3 Solid Salt Irradiation Behavior .................................................................15

2.4 MATERIALS COMPATIBILITY CHALLENGES ............................................18

2.4.1 Corrosion ..................................................................................................18

2.4.2 Irradiation (Neutron) Damage Limits .......................................................21

3. TECHNOLOGY AND APPLIED R&D NEEDS.........................................................23

4. REFERENCES..............................................................................................................25

APPENDIX A. THUMBNAIL SUMMARIES OF REACTOR CONCEPTS |UNDER 

CONSIDERATION BY MSR TECHNICAL WORKING GROUP ......................... A-1

APPENDIX B. 1969 MOLTEN SALT REACTOR EXPERIMENT VIDEO ...................B-1

APPENDIX C. MOLTEN SALT REACTOR EXCERPTS FROM “AN ACCOUNT OF 

ORNL’S 13 RESEARCH REACTORS” BY MURRAY ROSENTHAL ..................C-1

A-7

v 

FIGURES

Figure 1. Functional categories of molten salts for reactor applications. ....................................... 3

Figure 2. Early MSR program approach to screening and justifying candidate MSR salts. .......... 4

Figure 3. Effect of cationic radius on solubility of tri-fluorides in molten fluorides at 550°C 

[6]. ........................................................................................................................................... 6

Figure 4. Trivalent solubility effects in plutonium and minor actinides fueled fluoride salts 

[6]. ........................................................................................................................................... 6

Figure 5. Three-dimensional model of LiF-BeF2-ZrF4 phase diagram [11]. .................................. 8

Figure 6. Electroanalytical results from measurement on MSRE fuel [22]. ................................. 10

Figure 7. MSRE-era electroanalytical and spectrophotometric apparatuses [22]......................... 11

Figure 8. Fission product classes identified during MSRE operation highlighted as groups 

in the periodic chart (below) and the fission-yield curve (above). ........................................ 13

Figure 9. Exemplar short-lived fission products that change solubility behavior during 

decay in a fuel circuit (typically < 1 minute salt recirculation residence time). ................... 14

Figure 10. MSRE salt phase segregation on cooling [52]. ........................................................... 17

Figure 11. Temperature-gradient mass transfer [53]. ................................................................... 20

Figure 12. Persistent salt corrosion as a temperature-gradient-driven redox cycle [54]. ............. 20

Figure 13. Dual challenges of higher temperatures and displacements for MSR in-core 

structural materials [61]. ........................................................................................................ 22

TABLES

Table 1. Qualitative assessment of salt coolant properties database and predictive methods 

[12] .......................................................................................................................................... 9

Table C-1. ARE parameters ........................................................................................................ C-7

Table C-2. MSRE parameters ..................................................................................................... C-9

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vii 

ACRONYMS

ANP Aircraft Nuclear Propulsion Program

ARE Aircraft Reactor Experiment (fuel salt composition: 53.1-40.7-6.2 mol% NaFZrF4-UF4) 

MSBR Molten Salt Breeder Reactor (fuel salt composition: 71.7-16-12-0.3 mol% 7LiFBeF2-ThF4-UF4

MSR Molten Salt Reactor

MSRE Molten Salt Reactor Experiment (fuel salt composition: 65-29.1-5-0.9 mol% 7

LiFBeF2-ZrF4-UF4) 

ORNL Oak Ridge National Laboratory

PMA plutonium and minor actinides 

R&D research and development

A-11

1 

1. INTRODUCTION

The Department of Energy (DOE) Office of Nuclear Energy is sponsoring the Molten Salt 

Chemistry Workshop (Innovative Approaches to Accelerate Molten Salt Reactor [MSR] 

Technologies Deployment) to be held at Oak Ridge National Laboratory (ORNL) April 10–12, 

2017. This workshop is intended to identify science-based, technology-driven research 

opportunities to facilitate and accelerate MSR technologies development. 

Highlighted areas will include physical chemistry, irradiation and fission product chemistry, 

analytical chemistry (including instruments and in-line sensors), and materials compatibility. 

The goals of the workshop are to identify (1) research topics that address both short-term 

technology showstoppers and long-term grand challenges that may enable revolutionary changes 

in molten salt technologies and (2) a set of priority research directions that may accelerate 

molten salt technology development and deployment. The workshop discussion will be focused 

on high-risk/high-payoff research in molten salt chemistry and salt chemical properties and 

reactions that affect the application of molten salt for advanced nuclear systems.

This resource document was prepared by the author in consultation with subject matter experts 

from other national laboratories. In particular, the author acknowledges review and suggestions 

by Bill Del Cul and Sheng Dai (ORNL), Guy Frederickson (Idaho National Laboratory), and 

Mark Williamson and Sam Sham (Argonne National Laboratory). A draft of this report was 

circulated on March 20 to those registered for the workshop for comment (comment deadline 

was March 29 for revision) and all comments were addressed. It was released in final form on 

April 3. 

This document provides information from the scientific research community that underpins MSR 

technologies. Emphasis is placed on identifying fundamental knowledge gaps and potential 

solutions. 

A-13

3 

2. MOLTEN SALT REACTOR CHEMISTRIES

New molten salt nuclear reactor concepts have recently garnered significant attention in the 

United States and other countries. For this workshop, the following systems are emphasized: 

• Fast-spectrum chloride fuel salts containing NaCl, sometimes other chloride diluents, and

high concentrations of (U/Pu)Cl3

• Fluoride fuel salts (primarily for thermal spectrum applications)

– Similar to Molten Salt Reactor Experiment (MSRE)/Molten Salt Breeder Reactor (MSBR)

salts containing 7

LiF and/or NaF and BeF2 combined with low concentrations (< 5 mol%)

of UF4/PuF3 and sometimes containing larger concentrations of ThF4 (~10 mol %)

– Containing 7LiF and/or NaF and high concentrations of UF4/ThF4 (near eutectic: 25–

30 mol% actinide)

• Primary and secondary coolant salts

– Fluoride salt primary and secondary coolant candidates (e.g., LiF/NaF-BeF2, FLiNaK,

NaF-ZrF4) and sometimes other types of low-pressure secondary coolants (chlorides,

nitrates, liquid metals)

Seven of the eight concepts under consideration by the MSR Technical Working Group in 

conjunction with the Gateway for Innovation in Nuclear initiative are fluid-fueled reactors, and 

one is for a solid-fueled salt-cooled reactor. Figure 1 provides a functional diagram for these 

systems and Appendix A provides a thumbnail description of each one. 

Figure 1. Functional categories of molten salts for reactor applications.

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4 

2.1 PHYSICAL CHEMISTRY CHALLENGES

2.1.1 Reasoning Framework for Salt Choices

Salt systems selected for investigation must be justified to ensure that they are viable based on 

our current understanding of their physics and chemistry, they are good research and 

development (R&D) targets, and there are no better choices. Sometimes lost in the wealth of 

recent literature is the underpinning of reasoning that motivated and sustained the early MSR

programs. This traditional approach defined requirements for the application at hand to 

downselect salt systems based on their nuclear and chemical behavior, as outlined in Figure 2 

[1–4]. 

Figure 2. Early MSR program approach to screening and justifying candidate MSR salts.

At the top level, neutronics (parasitic neutron absorption and moderating properties) limit the 

choices of nuclides that can be considered, in combination with limits associated with the 

mission of the reactor (breeder, burner, converter) and neutron energies in the system. Figure 2 

highlights the neutronic constraints that strongly limit the choice of constituents for thermal and 

epithermal reactors and the chemical constraints that exert a controlling influence on all reactor 

applications as a result of the following requirements [1]: 

The fuel must dissolve more than the critical concentration of fissionable material at 

temperatures safely below the heat exchanger exit temperature. It must consist of elements 

of low capture cross sections for neutrons of the energy spectrum of the design chosen. 

The mixture must be thermally stable and must possess low vapor pressure over the 

temperature range proposed for operation. The fuel mixture must possess heat transfer and 

hydrodynamic properties adequate for its service as heat exchange fluid without excessive 

power required for circulation. It must be relatively non-corrosive to some (otherwise 

suitable) material, presumably a metal, of construction. The fuel must be stable to reactor 

radiation, must be able to survive fission of uranium or other fissionable material and must 

tolerate fission product accumulation without deterioration of its useful properties [1]. 

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5 

Early references explain why non-halide fluid-fueled systems (systems containing H, B, C, N, O, 

Bi) were rejected for large-scale high-temperature, low-pressure applications based on one of the 

chemical criteria listed in the previous paragraph [1–4]. 

The requirement that fuel salt be compatible with a container material places the most wideranging constraint on the choice of salt constituents because it requires the cationic constituents 

of the salt to be composed of relatively active, electropositive elements (fluoride compound free 

energies of formation > 95 kcal/mol-F), whereas the container material must be composed of 

elements that are relatively noble in comparison (fluoride compound free energies of formation < 

75kcal/mol-F) [3]. This requirement eliminates the middle of the periodic chart for consideration 

as salt constituents and the periphery of the periodic chart for consideration as alloy constituents

(only a small region in the center is suitable for metallic alloys). Nonmetallic materials (e.g., 

ceramics, composite materials) can also be compatible with high-temperature salt—for example, 

graphite is known to be fully compatible with molten halides to very high temperatures. 

Although nonmetallic materials have been used and may find service as core internals, there is 

currently no joining and fabrication technology to support and deploy these as the container 

material for many of the complex components that are in contact with salt. 

The neutronic requirements for a primary coolant are distinctly different from those for a fuel 

salt, and initial studies indicate that low-neutron-absorbing salts are necessary to obtain negative 

void and temperature coefficients [5]. Secondary coolant salts have no formal neutronic 

requirements, but consideration is typically given to the ability to recover/remediate leakage of 

secondary coolant into primary coolant or fuel salt. The chemical requirements for primary and 

secondary coolant salts are similar to those for fuel salts.

A-17
6 

2.1.2 Phase Diagrams and Solubility Behavior

Knowledge of phase diagrams is an essential requirement. Maintaining fissile and fertile material in 

solution is a fundamental requirement for solution reactors, so knowing the phase diagram behavior of 

actinides is of primary importance. For some MSR concepts, very high burnups and accumulation of 

fission products are foreseen, and research must account for how this increasing concentration of 

fission/activation products affects the solubility of fissile and fertile materials and the melting point of the 

salt. Trivalent elements often play a key role in determining melting point increases and solubility limits 

as actinides and fission products accumulate in a fuel salt, as shown in Figures 3 and 4 [6]. 

Figure 3. Effect of cationic radius on solubility of tri-fluorides in molten fluorides

at 550°C [6].

Figure 4. Trivalent solubility effects in plutonium and minor actinides fueled fluoride salts [6].

A-18
7 

Sometimes unrecognized is the interplay of oxidation-reduction (“redox”) equilibria and phase 

diagram behavior. Given the recent interest in salts with high actinide concentrations (up to and 

exceeding 30 mol%), the distribution of species between trivalent and tetravalent states 

according to the redox condition of the salt will need to be accounted for. For uraniumcontaining salts to be noncorrosive toward container materials, a small fraction of uranium must 

be trivalent for fluoride salts (UF3) and tetravalent (UCl4) in chlorides. For fluorides, there is a 

demonstrated benefit to maintaining salts at more reducing conditions to reduce corrosion and 

tellurium-attack [7–8]. Yet these very same conditions tend to increase the trivalent uranium 

fluoride content of the salt and could reduce the trivalent solubility margins (trivalents often 

exhibit an approximate “group solubility”). The complementary perspective is that trivalent 

actinide chlorides are the majority species in chlorides, and the tetravalent solubility associated 

with UCl4 and tetravalent fission products will need to be accounted for, as well as the native 

volatility of UCl4. Increases in the trivalent content of the fuel salts will occur as a result of

accumulation of activation and rare earths fission products and will likely increase the melting 

point of the salt because of a group trivalent solubility effect. 

The preceding discussion emphasizes minimizing oxidative corrosion by maintaining the salt in a 

reduced state. However, salts kept at very reduced conditions can promote other types of 

interactions with container materials and reactor internals, such as alloying, carburization, and 

wetting interactions that can cause unwanted intrusion of salt into the pores of core internals 

(e.g., graphite pores). This type of interaction between the salt and its container material imposes 

a limit on how reducing the salt can be maintained. 

Bounds on avoiding both oxidative corrosion and unwanted reductive interactions define a 

suitable redox window. Bounds on the redox window for the MSRE and MSBR fluoride salts 

were defined (~0.5%–15% UF3/UF4) [9], but those for the desirable redox windows for other 

fluorides and chlorides have not yet been established [10]. 

2.1.3 Thermophysical and Transport Properties

A significant number of molten salt properties have been measured and correlated and this 

knowledge should be harvested and used for present needs. However, not all properties are 

known or can be accurately estimated, especially those of new salt compositions proposed for 

reactor application. A few of the salts under consideration (e.g., 2LiF-BeF2) have a deep 

properties database (Figure 5); but for many, we are pushing the boundaries of our knowledge. 

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8 

Figure 5. Three-dimensional model of LiF-BeF2-ZrF4 phase diagram. Source: Reprinted from J. Nucl. 

Mater. 27(2), R. E. Thoma, H. Insley, H. A. Friedman, and G. M. Herbert, “Equilibrium phase diagram of 

the litium fluoride-beryllium fluoridezirconium fluoride system,” 166–180, 1968, with permission from 

Elsevier. 

For some properties (density, heat capacity), simple methods exist for correlating or obtaining an 

approximate estimate of the true value, but for others (melting point, vapor pressure, viscosity, 

and thermal conductivity), measurements are usually needed to proceed with confidence.

Phase transition thermodynamics govern melting point and species volatility (vapor pressure) 

and reflect the behavior of a number of complex solution species. A priori predictions of these 

phase transitions are difficult without some data to determine what species and interactions exist. 

Interaction parameters based on lower order datasets (e.g., binary mixtures) can sometimes lead 

to reasonable predictions for higher order datasets (e.g., ternary mixtures).

Viscosity and thermal conductivity are defined by constitutive equations for momentum and heat 

transfer and depend upon the structure of the molten salt in a complex manner. For thermal 

conductivity, trustworthy measurements are scarce because it is difficult to measure this property 

A-20
9 

at elevated temperatures in a manner that factors out the parallel heat transfer modes of thermal 

radiation and convection. The transmissive and absorptive properties of the salt for thermal 

radiation will be needed for detailed heat transfer analyses. With the exception of thermal 

conductivity, all salt properties have demonstrated measurement methods that can be refined to 

obtain high accuracy and reproducibility. A summary of these points is provided in Table 1, 

which is from a previous study. 

Table 1. Qualitative assessment of salt coolant properties database and predictive methods [12]

2.1.4 Modeling and Simulation

There is a significant opportunity to extend the knowledge base and strategically fill database 

gaps by using modeling and simulation to improve our ability to understand and correlate salt 

properties of all types. The understanding of property trends with compositional changes can be 

of great value in guiding any measurement campaign.

Phase Transitions/Diagrams

It is typically acknowledged that ab initio prediction of phase transitions/equilibria is more 

difficult than the prediction of stationary thermodynamic properties. A thermodynamic 

framework that makes use of experimental data on simpler systems is the current method in use 

[13]. However, recent progress in predicting phase transitions in molecular crystals suggests a 

similar approach might be developed for ionic crystals [14–15].

Thermophysical and Transport Properties (density, heat capacity, viscosity, thermal 

conductivity) 

Atomistic models, and molecular dynamic simulations in particular, have recently been applied 

with success to correlating molten salt physical properties [16–21]. The trend of property 

changes with composition is sometimes revealed more accurately than the absolute value of the 

property itself; but in some cases, remarkably accurate property estimates can be derived from 

reliable atomistic interaction parameters. 

2.2 ANALYTICAL CHEMISTRY CHALLENGES (INCLUDING MEASUREMENT 

INSTRUMENTS AND ONLINE SENSORS)

Traditional destructive chemical and isotopic analysis methods were successfully applied to 

irradiated and unirradiated molten salt samples during the Aircraft Nuclear Propulsion (ANP), 

MSRE, and MSBR efforts. Although improvements in some archival sample analyses are likely 

A-21
10 

needed, the primary R&D need is for short-turnaround (“nearly instantaneous”) feedback from 

nondestructive online analytical methods. Ongoing surveillance of the salt chemistry is required 

to ensure the reactor is operating within safe bounds with regard to corrosion and other unwanted 

chemical interactions; in this regard, measurement of the U3+/U4+ ratio and level of dissolved 

alloy constituents (e.g., Cr, Fe, Ni) is the first priority. Electroanalytical and spectrophotometric 

methods were tested in the lab during the MSRE and MSBR efforts on unirradiated and 

irradiated MSRE salts, and their viability and value was firmly established. Refinement of these 

techniques for application (and durability) in the complex MSR solution and its accompanying 

high-radiation field was recognized as a significant R&D challenge. Examples of the results and 

apparatus used at the end of the MSRE/MSBR program are depicted in Figures 6 and 7. 

Figure 6. Electroanalytical results from measurement on MSRE fuel [22].

A-22
11 

Figure 7. MSRE-era electroanalytical and spectrophotometric apparatuses [22].

Spectrophotometry provides fundamental and specific information about species in solution and 

has the potential to determine absolute values of most actinide constituents. At the end of the 

MSBR program, higher spatial- and energy-resolution gamma radiation detection methods were 

being tested. With the advent of much more capable radiation detection and imaging methods in 

the intervening years, it is likely that both assay and surveillance by radiation detection methods 

will find new applications. Since the end of the MSBR program, development of molten salt 

electroanalytical and spectrophotometric methods and apparatuses has continued [23–29].

2.3 IRRADIATION AND FISSION PRODUCT CHEMISTRY CHALLENGES

2.3.1 Molten Salt Fission Product Chemistry

Noble gas behavior was a focus of early fission product studies because these neutron poisons 

controlled breeding gain and could produce strong reactivity effects in thermal reactors. The 

solubility of noble gases in fluoride salts was found to be very low; therefore, the near-complete 

removal of these elements was achieved by use of low-porosity moderator graphite, non-wetting 

salts, and straightforward methods to strip the noble gas from the salt. 

A-23
12 

For the other fission products, it is important to understand where they go and how they behave 

to provide a radiation map to support dose control, maintenance, and accidentmitigation/response, as well as to design features into the plant and components to account for 

their behavior and manage their disposition. The detailed chemical behavior of many fission 

products remained uncertain at the end of the MSBR program (1976), and not much more is 

known today. What we know about their behavior in the MSRE salt is summarized in Figure 8 

[30], and we believe other fluoride and chloride salts follow similar patterns of behavior. There 

is also a body of evidence developed for fission product behavior during pyroprocessing of solid 

fuels using chlorides and fluorides, but the redox conditions applied and salt compositions used 

are significantly different from those under consideration as fluid fuels. The short-lived isotope 

effects described below are largely absent from the pyroprocessing experience.

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13 

Mass Number of Fission Product Decay Chain

Figure 8. Fission product classes identified during MSRE operation highlighted as groups in the 

periodic chart (below) and the fission-yield curve (above). (Upper graphic source: “Fission product 

yield,” Wikipedia, https://en.wikipedia.org/wiki/Fission_product_yield.) 

gas soluble insoluble soluble-gas-soluble

Cumulative Fission Product Yield (%)

A-25
14 

Four classes of fission products were identified based on the MSRE experience: gaseous species, 

soluble species, insoluble species, and sometimes-soluble species. The noble gases comprise the 

gaseous species. The IA (alkali), IIA (alkaline earth), IIIA&B (rare earths and Al-group), IVB 

(active refractories), and VIIB (halogens) comprise the soluble species. The transition metals 

between Nb and Te comprise the insoluble “noble metal group.” The Nb, Te, and Zn periodic 

columns comprise the sometimes-soluble species—their behavior changing with the redox state 

of the salt.

How these insoluble and sometimes-soluble species transition from birth at an atomic level to an 

insoluble phase or deposit is not completely understood, and several hypotheses still compete to 

explain what occurs [31]. Although a major fraction of the fission products end up as salt-soluble 

species (including iodine), the decay behavior of the fission product chain is quite complex in 

many cases and can result in partition of short-lived species before the longer-lived isotope 

emerges. This situation occurs for many of the noble metal fission products and for the key 

isotopes of 137Cs, 99Mo, and 131I, as shown in Figure 9. 

Figure 9. Exemplar short-lived fission products that change solubility behavior during decay in a 

fuel circuit (typically < 1 minute salt recirculation residence time). 

2.3.2 Barren-Salt Activation Product Chemistry: Tritium and Sulfur

Tritium is produced in MSRs by both ternary fission and a number of light-element neutron 

reactions. With the exception of its production from the irradiation of lithium, the amounts 

produced are similar to those produced in high-temperature reactors and light water reactors; and 

its mobility through alloys at high temperatures is a problem that is shared with gas-cooled hightemperature reactors. It is the use of lithium (and even lithium highly enriched in 7

Li) in the fuel 

A-26
15 

salt or primary coolant that is attended by large amounts of tritium production (rivaling or 

exceeding the amounts produced in heavy-water reactors) [32–34]. 

Although tritium is not particularly long lived and has only weak beta-particle emissions, 

because of its mobility, it has the potential to create large amounts of low-level radioactive waste 

if it reaches water in the environment or in the balance of plant. It must also be controlled to 

levels safe for workers and the public. There is a wealth of information on the technology used to 

control hydrogen and tritium transport, and many types of solutions can be envisioned, including 

various barriers and trapping approaches. Because of the fluxing property of halide salts, it is 

very unlikely that any type of film on salt-wetted alloy surfaces could serve as a barrier to 

tritium.

During the operation of the MSRE, a significant fraction of the tritium produced permeated the 

primary systems and was not controlled or trapped by the off-gas systems [35]. As part of the 

post-MSRE studies, a fluoroborate coolant-salt-trapping modality was identified for the MSBR 

fluoroborate secondary coolant [35–36]. Modeling and experimental studies on the behavior of 

tritium in MSR systems have continued since the MSBR-era studies [37–39]. It was also 

proposed that an oxygen-containing secondary salt, such as a nitrate mixture, would be able to 

chemically trap tritium, as was demonstrated with the oxygen moiety in fluoroborates. The 

ability to manipulate the chemistry in other fluoride or chloride salts (such as the behavior of 

trace oxygen levels as dissolved or undissolved components) to provide tritium trapping is not 

known. 

In chloride fuel salts not heavily enriched in 37Cl, the activation of 35Cl will produce significant 

quantities of 36S by (n,p) activation. Taube predicted sulfur levels in the thousands of ppm at 

steady state in a molten chloride fast reactor composed of unenriched chlorides, and it is clear 

that this production has both a redox effect and potentially a chemical corrosive/reactive effect 

(akin to that associated with its periodic chart neighbor tellurium in group 6A) [40–44].

2.3.3 Solid Salt Irradiation Behavior

The behavior of solid salt during interim or long-term storage is an important issue because solid 

halide salts can sustain considerable radiation damage and chemical change upon irradiation by 

the alpha, beta, and gamma radiation produced by the decay of dissolved fission and activation 

products. The production of metal centers and reactive halogens (e.g., F2) within the salt was

observed in discharged irradiation capsules and the stored MSRE fuel salt [45–46]. The 

importance of distinguishing solid-salt irradiation effects from molten-salt irradiation effects was 

clearly demonstrated during the testing that led up to the design and operation of the MSRE [47].

Testing of fuel and coolant fluoride salts by in-reactor irradiation confirmed that molten salt is 

largely immune to irradiation damage—any primary radiolytic event is rapidly reversed by 

recombination. No materials damage or chemical effects of beta, gamma, neutron, or alpha 

radiation (or fission fragments) were observed in molten fluorides even when very high 

(> 1 kw/cc) in-reactor fission power densities were applied. This testing also revealed that if the 

MSRE salt was allowed to freeze and cool to temperatures below 200°C, significant radiolysis 

and chemical change occurred very quickly, driven by the intense radiation associated with the 

A-27
16 

short-lived decay of fission products. Thus any study of irradiated salts must account for this 

factor, and the domain where radiolysis occurs must be understood for any new salts. 

Molten chloride salts have not seen the intense in-pile power densities proposed for molten salt 

fast reactor concepts, so their immunity to radiation damage under these types of conditions will 

need to be confirmed. Our understanding of radiolytic effects in chloride salts builds from 

experience in using them as reprocessing media to dissolve solid nuclear fuels and perform 

separations. Molten alkali chloride mixtures used for pyroprocessing exhibited good chemical 

and radiation stability at the radiation levels associated with processing fuel with short decay 

periods (3–6 months). Experiments are yet to be conducted to confirm the stability of molten 

chlorides under the more intense conditions of irradiation as a fuel salt. Out-of-reactor studies 

have also been conducted on solid fluoride and chloride salts using alpha, beta, gamma, and xray irradiation [48–51].

There is another dimension to irradiation damage in solid salts that has not received much 

attention. In the solid state, a multicomponent halide is a mixture of different crystal sizes and 

types. The distribution of crystal sizes and types depends on the cooling history of the melt. A 

solid boule of the unirradiated MSRE salt contained four major phases, as shown in Figure 10. 

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17 

 

Figure 10. MSRE salt phase segregation on cooling [52]. 

A-29
18 

The purpose of the study [52] depicted in Figure 10 was to bound how much uranium 

segregation could occur during freezing. As irradiated salt is cooled to a solid, not only does the 

fissile component distribute according to phase equilibrium and crystallization dynamics but the 

fission products also distribute themselves according to their phase equilibria. For the MSRE and 

MSBR salts, fission-product phase distributions received some study. In general, it was found (as 

expected) that soluble fission products (e.g., Rb, Cs, Sr, Ba, Zr) within the same chemical/space 

group (e.g., IA, IIA, IVA) already present in the salt (e.g., LiF, BeF2, ZrF4) would be substituted 

into the phases of the major salt constituents. The rare earths were an exception. This class of 

soluble fission product was not incorporated into a phase associated with the unirradiated fuel 

salt.

The impact of this nonuniform distribution of fission products throughout the solid salt is that 

different regions (and even micro-domains) will receive different radiation exposures. Since each 

crystal also exhibits a distinct radiolysis behavior, taking account of this heterogeneity is 

important, as is accurate thermal modeling, because the radiolysis is strongly influenced by 

temperature. Some simplification occurs for the effect of long-energy-track radiation, such as 

gamma rays; but the fraction of gamma energy deposited in the salt must also be accounted for 

from both internal and external sources (e.g., neighboring tanks/sources).

2.4 MATERIALS COMPATIBILITY CHALLENGES

The identification of suitable core and container materials for molten salt service must take into 

account a number of requirements and involves a number of trade-offs. The core and container 

material must retain adequate strength, toughness, and creep resistance for the neutron exposure 

that is accumulated; endure contact with molten salt over a range of temperatures without 

excessive corrosion; be stable with regard to the external atmosphere surrounding the exterior of 

the container (and tolerate whatever atmospheric in-leakage that forms the basis for safety 

evaluation); and be able to be fabricated into the complex shapes of the primary loop (and 

sometimes the secondary loop) of the reactor. For service in the secondary loop, the material 

must also exhibit compatibility with a separate working fluid (e.g., steam, CO2, helium for a 

power conversion system, or another high-temperature heat transfer agent), but the requirements 

for neutron exposure durability are almost completely absent. Although durability under neutron 

exposure is a very important property, it is not unique to molten salt systems. Because the 

corrosion phenomenon in molten salts is distinctive, it is highlighted first. 

2.4.1 Corrosion

There are a number of excellent reviews of the extensive corrosion studies on fluoride salts at 

ORNL that led to the development of Hastelloy-N, its highly successful deployment in the 

MSRE, and its optimization for the MSBR program [53]. For the reader less familiar with this 

history, the following section summarizes some key points. There are parallels with the 

chemistry in chloride melts to the principles outlined below for fluorides.

Unlike the case for more conventional oxidizing media, the products of oxidation of metals by 

fluoride melts tend to be completely soluble in the corroding media [54]; hence, passivation is 

precluded, and corrosion depends directly on the thermodynamic driving force of the corrosion 

reactions [55]. Design of a chemically stable system using molten fluoride salts therefore 

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demands the selection of salt constituents that are not appreciably reduced by available structural 

metals and the development of containers whose components are in near thermodynamic 

equilibrium with the salt medium.

Examination of the free energies of formation for the various alloy components in Inconel or 

Hastelloy-N shows that chromium is the most active of the metal components. Therefore, any 

oxidizing reactions with nickel-base alloys are be expected to show selective attack on the

chromium constituent. Such oxidation and selective attack follow from reactions such as the 

following [56] for the fuel salt. 

Impurities dissolved in the melt attack chromium directly producing dissolved CrF2: 

Cr + NiF2 → CrF2 + Ni , 

Cr + 2HF → CrF2 + H2 . 

Oxide films on the metal are converted to corrosive dissolved species: 

2NiO + ZrF4 → 2NiF2 + ZrO2 . 

These reactions are followed by reaction of NiF2 with Cr cited above, and reduction of excess 

UF4 to UF3: 

Cr + 2UF4 ↔ 2UF3 + CrF2 . 

Of course, in the case of a coolant salt with no fuel component, this reaction with uranium would 

not be a factor. 

The phenomenon described above falls into the class of “temperature-gradient mass transfer” 

that occurs without a protective film of any sort and has some features in common with liquid 

metal corrosion. When dissimilar materials contain molten salt, an activity gradient corrosion can 

occur in the absence of any temperature-gradient driving force (or it can act in combination with 

a thermal gradient). The generalized concept of temperature-gradient mass transfer is depicted in 

Figure 11. 

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20 

Figure 11. Temperature-gradient mass transfer [53].

It was the thermally driven UF4/UF3 redox couple that was believed to be the primary source of 

persistent corrosion in the fuel salt loop after impurity-driven corrosion had been played out, and 

is depicted in Figure 12. 

Figure 12. Persistent salt corrosion as a temperature-gradient-driven redox cycle [54].

Redox processes that are responsible for attack by molten fluoride mixtures on high-temperature 

alloys result in selective oxidation of the least noble metal—typically chromium. Removal of 

chromium from the alloy occurs primarily in the regions of the highest temperature and results in 

the formation of discrete voids in the alloy [57]. These voids are not, in general, confined to the 

grain boundaries in the metal but are relatively uniformly distributed throughout the alloy surface 

that is in contact with the melt. The rate of corrosion has been measured and was found to be 

controlled by the rate at which chromium diffuses to the surfaces undergoing attack [58].

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With regard to our current understanding of the corrosion process, there are some important 

gaps. The following questions can be posed: 

• In the absence of a redox-sensitive soluble species (as was largely the case for the MSRE 

2LiF-BeF2 coolant), what is the mechanism of corrosion in a polythermal loop? 

• Is it better to have a substantial redox buffer (which will have some redox-temperature

coefficient) in a coolant salt and drive it to a reducing condition, or is it better to have no (or 

very low levels of) redox-sensitive species in solution and maintain it in a reducing state by 

other means? 

• If a redox-sensitive species is a major constituent of the salt (as is the case for a uraniumcontaining fuel salt), is it more advantageous to have high concentrations and maintain the salt 

in a very reducing conduction (and thus have a very large buffer capacity), or is it better to 

have a lower concentration of redox-sensitive species? 

– Work in the 1950s during the ANP program indicated that higher uranium concentrations 

in the salt accelerated the persistent polythermal corrosion mechanism. 

• What is the effect of different salt compositions on polythermal redox-driven corrosion? 

– Work in the late 1950s and 1960s indicated that very basic1 uranium fuel salts (e.g., low 

concentrations of UF4 dissolved in FLiNaK, a.k.a. “Fuelinak”) had much larger thermal 

coefficients of corrosion than other uranium-containing salts that had fewer uncomplexed 

fluoride anions (e.g., neutral 2LiF-BeF2 or acidic NaF-ZrF4 salts). 

– The very high uranium- and plutonium-content salts proposed for some new MSR 

concepts are significantly different from those tested in the past. 

• Are iron-based alloys an option for halide salts if the salt is maintained under very reducing 

conditions [59]? What are the salt redox window and compositional window for iron-based 

alloys? What is the response of iron alloys to design-basis corrosive inputs compared with 

nickel-based alloys?

2.4.2 Irradiation (Neutron) Damage Limits

The irradiation damage limits associated with the use of conventional nickel superalloys and 

stainless and ferritic/martensitic steels have been studied extensively [60–62], although 

incremental improvements in alloy performance and our understanding of it continue [63].

These conventional materials do not exhibit the long lifetimes typically desired for nuclear 

components; this is an especially dramatic shortfall for fast-reactor applications. However, the 

thermal neutron effects remain an important constraint limiting higher temperatures and 

exposures for nickel-based superalloys. Figure 13 depicts the dual materials challenges posed for 

MSR systems as higher temperatures and atomic displacements are sought.

 

1 Basic in the sense of the Lewis acid-base theory—possessing an abundance of electron donors (uncomplexed 

fluoride anions in salts that are predominantly alkali halides such as FLiNaK).

A-33
22 

Figure 13. Dual challenges of higher temperatures and displacements

for MSR in-core structural materials [61]. 

The situation is not as challenging for container alloys or primary loop components outside the 

core because they see considerably less intense radiation fields, but existing materials were still 

seen as having some shortfalls at the end of the MSBR program. Whether these shorter material 

lifetimes are suitable is an economic decision and depends on the particular system design. 

A-34
23 

3. TECHNOLOGY AND APPLIED R&D NEEDS

In the most general terms, the technology and applied R&D needs highlighted in the previous 

sections can be summarized as follows: 

1. Create a systematic computational approach to salt selection that balances and 

encompasses all neutronic and chemical factors.

2. Predict phase diagrams, solubilities, vapor pressures, and all relevant thermophysical and 

transport properties. 

3. Develop rugged and reliable in-line chemical and isotopic sensors in conjunction with the 

development of out-of-pile and in-pile test stands. 

4. Create a spatially and temporally accurate predictive model of fission/activation product 

production coupled with the ability to accurately track the fate and transport of this 

activity.

5. Develop a discriminating model of corrosion and materials compatibility for a variety of 

salt redox conditions and salt compositions that permits the most favorable operating 

envelope to be identified and the most appropriate salts and materials (or groups of 

materials) to be discovered, developed, and designed. 

6. Discover new materials and ways to use existing materials that can extend the hightemperature and neutron flux/fluence boundaries that currently exist. 

7. Conceive and formulate a fine-grained predictive understanding of radiolytic processes in 

solid salts to permit in-pile testing, interim salt storage, and disposition of salt-based 

wastes.

A-35

25 

4. REFERENCES

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system,” Journal of Nuclear Materials 462, 43–53 (2015).

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Chemical Reviews 116, 5567−5613 (2016).

15. S. Hirata et al., “Ab initio molecular crystal structures, spectra, and phase diagrams,” Acc. Chem. Res.

47, 2721−2730 (2014).

16. D. Corradini et al., “Coordination numbers and physical properties in molten salts and their

mixtures,” Faraday Discuss. 190, 471 (2016).

17. Y. Ishii et al., “Thermal conductivity of molten alkali metal fluorides (LiF, NaF, KF) and their

mixtures,” The Journal of Physical Chemistry B 118, 3385−3391 (2014).

A-37
26 

18. A. E. Gheribi et al., “Prediction of the thermophysical properties of molten salt fast reactor fuel from

first-principles,” Molecular Physics 112, 1306–1312 (2014).

19. F. Lantelme and H. Groult, Molten Salts Chemistry: From Lab to Applications (2013).

20. P. A. Madden and M. Salanne, “Thorium molten salts: Theory and practice,” Thorium Energy for the

World, pp. 111–116 (2016). http://link.springer.com/chapter/10.1007/978-3-319-26542-1_15

21. L. C. Dewan et al., “Molecular dynamics simulation of the thermodynamic and transport properties of

the molten salt fast reactor fuel LiF–ThF4,” Journal of Nuclear Materials 434 322–327 (2013).

22. M. W. Rosenthal, The Development Status of Molten-Salt Breeder Reactors, ORNL-4812, Oak Ridge

National Laboratory, August 1972. http://web.ornl.gov/info/reports/1972/3445604501750.pdf

23. H. A. Laitinen, “History of electroanalytical chemistry in molten-salts,” ACS Symposium Series 390,

417–428 (1989).

24. D. K. Corrigan et al., “Advances in electroanalysis, sensing and monitoring in molten salts,” Faraday

Discuss. 190, 351–366 (2016).

25. M. M. Tylka et al., “Method development for quantitative analysis of actinides in molten salt,”

Journal of the Electrochemical Society 162(9), H625–633 (2015).

26. B. Kelleher et al., “Observed redox potential range of Li2BeF4 using a dynamic reference electrode,”

Nuclear Technology 195, 239 (2016).

27. G. Duran-Klie et al., “Dynamic reference electrode development for redox potential measurements in

fluoride molten salt at high temperature,” Electrochimica Acta 195, 19–26 (2016).

28. J. Zhang, “Electrochemistry of actinides and fission products in molten salts—Data review,” Journal

of Nuclear Materials 447, 271–284 (2014).

29. K. Song et al., “Spectroscopic studies of lanthanides ion in high-temperature molten salt,” Applied

Spectroscopy Reviews 50(8), 654–669 (2015).

30. E. L. Compere et al., Fission Product Behavior in the MSRE, ORNL-4865, Oak Ridge National

Laboratory, 1975. http://web.ornl.gov/info/reports/1975/3445605698493.pdf

31. R. J. Kedl, The Migration of a Class of Fission Products in the MSRE, ORNL/TM-3884, Oak Ridge

National Laboratory, 1973. https://www.osti.gov/scitech/biblio/4471292-migration-class-fissionproducts-noble-metals-molten-salt-reactor-experiment

32. R. B. Briggs, “Tritium in molten salt reactors,” Reactor Technology 14(4), 335 (1972).

33. R. B. Briggs, “Tritium in molten salt reactors—Reply,” Reactor Technology 15(1), 79 (1972).

34. R. B. Briggs, Calculation of the Tritium Distribution in the MSRE, CF-70-7-13, Oak Ridge National

Laboratory, 1970, and Additional Calculations of the Distribution of Tritium in the MSRE, ORNL/CF

71-7-8, Oak Ridge National Laboratory, 1971.

35. G. T. Mays, Distribution and Behavior of Tritium in the Coolant-Salt Technology Facility,

ORNL/TM-5759, Oak Ridge National Laboratory, 1977.

http://moltensalt.org/references/static/downloads/pdf/ORNL-TM-5759.pdf

36. L. Maya, “Identification of polyborate and fluoropolyborate ions in solution by Raman spectroscopy,”

Inorganic Chemistry 15(9), 2180 (1976).

37. R. B. Briggs and C. W. Nestor, A Method for Calculating the Steady-State Distribution of Tritium in

a Molten-Salt Breeder Reactor Plant, ORNL/TM-4804 Oak Ridge National Laboratory, 1975.

http://web.ornl.gov/info/reports/1975/3445605505619.pdf

A-38
27 

38. J. D. Stempien et al., “An integrated model of tritium transport and corrosion in fluoride salt-cooled

high-temperature reactors (FHRs)—Part I: Theory and benchmarking,” Nuclear Engineering and

Design 310, 258–272 (2016).

39. D. Zhang, “Permeation of hydrogen in Hastelloy C-276 at high temperature,” Fusion Science and

Technology 67, 681 (2015).

40. E. Ianovici and M. Taube, Chemical State of Sulphur Obtained by the 35Cl (n, p) 35S Reaction during

in-pile Irradiation, Swiss Institute for Reactor Technology (Wurenlingen) Report, EIR-267 (1974).

http://moltensalt.org/references/static/downloads/pdf/EIR-267.pdf

41. M. Taube and J. Ligou, “Molten plutonium chlorides fast breeder reactor cooled by molten uranium

chloride,” Annals of Nuclear Science and Engineering 1, 277 (1974).

42. E. Ianovici and M. Taube, “Chemical behavior of radiosulphur obtained by 35Cl (n, p) 35S during inpile irradiation,” Journal of Inorganic and Nuclear Chemistry 37, 2562 (1977).

43. M. Furrer, “Chemical behaviour of radiosulphur obtained by 35Cl (n, p) 35S during in-core irradiation

of alkali-chloride melts,” Journal of Inorganic and Nuclear Chemistry 39, 1086 (1977).

44. M. Taube, Fast Reactors Using Molten Chloride Salts as Fuel: Final Report (1972–1977), Swiss

Institute for Reactor Technology (Wurenlingen) Report, EIR-332 (1978).

http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/13/648/13648304.pdf

45. D. F. Williams et al., A Descriptive Model of the MSRE after Shutdown: Review of FY 1995 Progress,

ORNL/TM-13142, Oak Ridge National Laboratory, 1996.

https://www.osti.gov/scitech/servlets/purl/230260

46. D. F. Williams, A. S. Icenhour, L. D. Trowbridge, G. D. Del Cul, L. M. Toth, “Radiolysis studies in

support of the remediation of the Molten Salt Reactor Experiment,” Transactions of the American

Nuclear Society 81, 89–90 (1999).

47. W. R. Grimes, Radiation Chemistry of MSR System, ORNL/TM-500, Oak Ridge National Laboratory,

1963. http://web.ornl.gov/info/reports/1963/3445605493999.pdf

48. L. M. Toth and L. K. Felker, “Fluorine generation by gamma radiolysis of a fluoride salt mixture,”

Radiat. Eff. Def. Solids 112, 201–210 (1990).

49. A. B. Lidiard, “The radiolysis of alkali halides—The nucleation and growth of aggregates,”

Zeitschrift fur Physikalische Chemie 206, 219–248 (1998).

50. L. Tandon, Radiolysis of Salts and Long-Term Storage: Issues for Both Pure and Impure PuO2 

Materials in Plutonium Storage Containers, LA-13725-MS, Los Alamos National Laboratory, 2000.

http://www.iaea.org/inis/collection/NCLCollectionStore/_Public/32/016/32016719.pdf

51. D. K. Veirs et al., Evidence of Corrosive Gas Formed by Radiolysis of Chloride Salts in PlutoniumBearing Materials, SRNL-STI-2010-00051, Savannah River National Laboratory, 2010.

http://sti.srs.gov/fulltext/SRNL-STI-2010-00051.pdf

52. H. A. Friedman and R. E. Thoma, “MSRE fuel fractionation experiments,” in Reactor Chemistry

Division Annual Progress Report for Period Ending January 31, 1965, W. R. Grimes et al., eds.,

ORNL-3789, Oak Ridge National Laboratory, April 1965.

http://web.ornl.gov/info/reports/1965/3445605716008.pdf

53. J. W. Koger, “Chromium depletion and void formation in Fe-Ni-Cr alloys during molten salt

corrosion and related processes,” in Advances in Corrosion Science and Technology, vol. 4, M. G.

Fontana and R. W. Staehle, eds. (1974).

A-39
28 

54. W. D. Manly et al., “Metallurgical problems in molten fluoride systems,” Progress in Nuclear

Energy, Series IV 2, 164–179 (1960).

55. J. H. DeVan and R. B. Evans III, Corrosion Behavior of Reactor Materials in Fluoride Salt Mixtures,

ORNL/TM-328, Oak Ridge National Laboratory, September 1962.

http://web.ornl.gov/info/reports/1962/3445605700621.pdf

56. W. R. Grimes, Chemical Research and Development for the Molten-Salt Breeder Reactor,

ORNL/TM-1853, Oak Ridge National Laboratory, 1967.

http://web.ornl.gov/info/reports/1967/3445605995384.pdf

57. L. S. Richardson, D. E. Vreeland, and W. D. Manly, Corrosion by Molten Fluorides, ORNL-1491,

Oak Ridge National Laboratory, March 1953.

http://web.ornl.gov/info/reports/1953/3445603532247.pdf see also

http://web.ornl.gov/info/reports/1969/3445605136769.pdf

58. J. H. Devan, Effect of Alloying Additions on Corrosion Behavior of Nickel-Molybdenum Alloys in

Fused Fluoride Mixtures, University of Tennessee Thesis, August 1960, see also ORNL/TM-2021,

Oak Ridge National Laboratory, 1969. https://www.osti.gov/scitech/servlets/purl/4789147

59. J. R. Keiser et al., The Corrosion of Type 316 Stainless Steel to Li2BeF4, ORNL/TM-5782, Oak Ridge

National Laboratory, April 1977. https://www.osti.gov/scitech/servlets/purl/4789147

60. H. E. McCoy, Status of Materials Development for Molten Salt Reactors, ORNL/TM-5920, Oak

Ridge National Laboratory, January 1978. https://www.osti.gov/scitech/servlets/purl/5195742

61. S. J. Zinkle and G. S. Was, “Materials challenges in nuclear energy,” Acta Materialia 61 (2013) 735–

758.

62. S. J. Zinkle, K. A. Terrani, and L. L. Snead, “Motivation for utilizing new high-performance

advanced materials in nuclear energy systems,” Current Opinion in Solid State and Materials Science

20, 401–410 (2016).

63. L. Jiang et al., “M2C andM6C carbide precipitation in Ni-Mo-Cr based superalloys containing

silicon,” Materials and Design 112, 300–308 (2016).

A-40
APPENDICES

A. Molten Salt Reactor Technical Working Group Descriptions of Reactor Systems under

Consideration

B. 1969 Molten Salt Reactor Experiment Video

C. Molten Salt Reactor Excerpts from “An Account of ORNL’s 13 Research Reactors” by

Murray Rosenthal

A-41

APPENDIX A. 

THUMBNAIL SUMMARIES OF REACTOR CONCEPTS

UNDER CONSIDERATION BY MSR 

TECHNICAL WORKING GROUP

A-43

A-3

Elysium Molten Salt Reactor Concept

• Rated electric output 1,000 MWe

• Chloride fuel salt

• Fast spectrum neutron flux

• Core outlet temperature of >600°C

• Core inlet temperature of ~500°C

• Employs drain valve

www.elysiumindustries.com/

FLiBe Energy

• Liquid Fluoride Thorium Reactor (LFTR) – Tho

• Graphite moderated, thermal spectrum, LiF-Be

• Freeze valve and drain tank based decay hea

• Described in EPRI technical report – 3002005

− https://www.epri.com/#/pages/product/000

− https://aris.iaea.org/PDF/LFTR.pdf

http://flibe-energy.com/

Kairos Power

• Solid-fueled fluoride-salt-cooled concept with demonstration

targeted by 2030

• Uses ceramic TRISO coated fuel particles in a pebble fuel

element design.

• Proposes nuclear air combined cycle for flexible, responsive

power generation

− co-fired with natural gas or hydrogen for peaking

power

• PB-FHR described at

https://aris.iaea.org/PDF/Mk1%20PB-FHR%20ARIS%20final.pdf

http://kairospower.com/

Moltex Energy

• UK company – Stable Salt Reactor (SSR) con

• NaCl-(Actinide)Cl3 fuel salt (LEU, <15% 235U)

− Redox stabilized with Zr/ZrCl couple

• NaF-ZrF4 coolant salt

− ZrF2/ZrF4 redox buffer

• Natural circulation of fuel salt within vented fue

• Steel cladding similar to that used in Dounreay 

https://aris.iaea.org/PDF/SSR.pdf

http://www.moltexenergy.com/

Fig. A-1. Thumbnail descriptions of Molten Salt Reactor Technical Working Group reactor concep

A-45
A-4

TerraPower

• Molten Chloride Fast Reactor (MCFR) uses chloride-based fuel

salt

• Employs reflector

− Reduce dose to containment

− Improve neutron economy

− Reduce radiation level within shielding

− Integral heat exchangers

• Integral heat exchangers

http://terrapower.com/news/terrapowers-continuing-innovation

Terrestrial Energy USA

• Integral Molten Salt Reactor concept

− Denatured U/Pu fuel cycle and LE

containing constituents such as LiF

− Graphite moderator

− Sealed primary unit with integrate

exchangers, and shutdown rods

• Seven year reactor vessel, moderator, and 

− Twin reactor module configuration 

swapped every seven years

https://aris.iaea.org/PDF/IMSR40

https://aris.iaea.org/Publications/SMR-B

Thorcon Power

• Fuel salt comprised of NaF and BeF2 with 12mol% actinide

− 80% of actinide is thorium, 20% uranium (~20% 235U)

• Fuel salt and moderator replaced every four years in twin reactor

modular configuration

https://aris.iaea.org/PDF/ARISThorCon9.pdf

http://thorconpower.com/

Transatomic

• 5% enrichment LEU fuel salt, such as LiF-

• Zirconium hydride in movable rods

http://www.transatomicpower.com/wp-content/up

White-Paper-v2.1.pdf

http://www.transatomicpower.c

Fig. A-2. Thumbnail descriptions of Molten Salt Reactor Technical Working Group reactor concepts (con

A-46
APPENDIX B. 

1969 MOLTEN SALT REACTOR

EXPERIMENT VIDEO

A-47

B-3

Oak Ridge National Laboratory. 1969. “The Molten Salt Reactor Experiment.” Video. 

Available: https://www.youtube.com/watch?v=tyDbq5HRs0o. 

A-49

APPENDIX C. MOLTEN SALT REACTOR EXCERPTS FROM “AN 

ACCOUNT OF ORNL’S 13 RESEARCH REACTORS” BY 

MURRAY ROSENTHAL

(Complete document at http://info.ornl.gov/sites/publications/files/Pub20808.pdf) 

A-51

C-3

AIRCRAFT NUCLEAR PROPULSION PROGRAM

In 1946, the Air Force initiated a program to develop a nuclear-powered airplane and contracted 

with the Fairchild Engine and Airplane Corporation to evaluate its feasibility. Fairchild set up a 

project in an unused building near the K-25 gaseous diffusion plant (now the East Tennessee 

Technology Park) in Oak Ridge, Tennessee. Oak Ridge National Laboratory (ORNL) provided 

assistance.

There was considerable skepticism, and before getting more deeply involved, the Atomic Energy 

Commission (AEC) asked staff at the Massachusetts Institute of Technology to perform an 

independent evaluation. A summer project at Lexington, Massachusetts, concluded after initial 

doubts that developing nuclear-powered aircraft would cost as much as $1 billion and take 15 y 

but was feasible.

So under Cold War pressure, the AEC directed ORNL to establish the Aircraft Nuclear 

Propulsion (ANP) Program. It grew to be the Laboratory’s largest activity for a number of years. 

General Electric (GE) eventually replaced Fairchild and moved its program to Evendale, Ohio. 

But about 180 people remained in Oak Ridge and joined ORNL’s program, which by 1951 filled 

a big building at the Oak Ridge Y-12 Plant (now the Y-12 National Security Complex). 

Two approaches to nuclear propulsion were considered. In the direct cycle, air for the jet engine 

would pass through the reactor and be directly heated to 1,700°F. In the indirect cycle, a molten 

metal would flow through a reactor and then through a heat exchanger where it would heat the 

air. GE undertook to develop the direct cycle and ORNL undertook development of the indirect 

cycle.

Aircraft Reactor Experiment Development

ORNL proposed building a small reactor to guide its development program. Funded by the AEC, this 

reactor was called the Aircraft Reactor Experiment (ARE). To permit high temperature, BeO was chosen 

as the moderator. The core was to be a stack of hexagonal blocks with vertical passages for the fuel 

elements and molten sodium coolant. Because of the time needed for production of the BeO blocks, they 

were ordered before the design was completed. The blocks stacked up to form the core are shown in 

Figure C-1. The fuel elements themselves were to be long, thin stainless-steel tubes containing UO2. 

Details are given in Table C-1. 

A-53
C-4

Figure C-1. Stacked BeO moderator blocks for the ARE.

An extensive research and development program was undertaken in chemistry, materials, hightemperature sensors, and other areas, and components such as pumps and valves were tested in 

loops. Protecting the aircraft’s crew and components and the public from radiation was critical. 

Consequently, ORNL launched a broad program on the penetration of gamma rays and neutrons, 

including measurements of the shielding properties of a variety of materials and the development 

of calculation methods. That led to reactors built just for shielding studies, about which more 

will be said later.

As the ARE design progressed, however, two serious questions arose. One came from physics 

calculations that showed that the cross section of xenon-135 at elevated temperatures could cause 

a positive temperature coefficient in a high-power solid-fuel reactor. The other was doubts about 

the reliability of “spaghetti-like” fuel elements operating at the temperatures that would be 

necessary.

An alternate idea arose in which a uranium compound would be dissolved in a molten salt and 

solid fuel elements avoided. The salt would flow through the core, where it would be heated by 

fissions, and then pass through a heat exchanger. The xenon gas was expected to evolve from the 

salt and be removed, but even if it did not, much of it would be outside the core in the external 

system. So the fuel rods were replaced by a molten salt.

After considering other salts, the choice of carrier salt was a “relatively” low melting 

temperature (950°F) mixture of sodium and zirconium fluorides into which UF4 was dissolved. 

The fluoride salts when solid look somewhat like green table salt, but as seen in Fig. C-2, when 

molten they resemble water. Alkali fluorides are among the most stable of compounds, which 

means they do not corrode stainless steels and similar alloys. They also have very low vapor 

pressures, and when molten, they are not damaged by radiation. 

A-54
C-5

Figure C-2. ORNL photograph of molten salt, featured on the cover of the British Science Journal 

in June 1969. Used by permission; (Science Journal was incorporated into New Scientist, 

www.newscientist.com, in 1971). 

The salt would flow through the core in the passages that were designed to hold the solid 

elements, but in a single pass the velocity would have been too low for turbulent flow. This was 

dealt with by having the fuel make multiple passes at higher velocity. The core with the multiplepass tubes is seen in Figure C-3. The salt emerging from the core was pumped through a finnedtube heat exchanger, where it was cooled by helium, which was in turn cooled by water, as 

shown schematically in Figure C-4. 

A-55
C-6

Figure C-3. Elevation view of the ARE core. (ORNL Drawing 16336). 

Figure C-4. ARE flow diagram. 

Heat generated in the moderator and reflector was removed in a sodium system that was similar 

to the salt system. For safety, the sodium system had two independent circuits, and the fuel salt 

system contained a standby pump and heat exchanger to ensure continued operation. Sump-type 

centrifugal pumps with overhung impellers were used, and the sumps served as expansion tanks. 

A-56
C-7

The reactor was designed to operate with the fuel salt at around 1500°F and at a power of 1 to 3 

MW of heat.

Table C-1. ARE parameters

Power (heat) 2.5 MW

Fuel salt 53% NaF, 41% ZrF4, 6% UF4

(93% enriched)

Fuel melting point 990°F

Fuel inlet temperature 1225°F

Fuel outlet temperature 1580°F (1620° at peak)

Fuel flow rate 45 gal/min

Moderator BeO hexagonal blocks, 3.6 in. 

across × 6 in. high with 1.25 in. 

coolant passages

Coolant Helium to water

Container Inconel

Critical November 3, 1954

Shut down November 12, 1954

The salt and sodium pipes and all major components had double walls to provide a space in 

which helium was circulated to maintain even temperatures and to provide leak monitoring. 

Control and safety rods in the core and fission chambers in the reflector were also cooled by the 

circulating helium.

Electrical heaters around the core and the piping raised the temperature before molten salt was 

added. All salt and sodium-containing systems were made from Inconel, a high-temperature 

nickel- base alloy.

A mill-type building was constructed in Melton Valley near the homogeneous reactors to house 

the ARE. All of the reactor systems were located below grade in concrete-lined pits. 

ARE Operation

Operation of the ARE began in early November 1954. First the system was heated with the 

electrical heaters. Then it was filled with uranium-free carrier salt. And finally, fuel salt as 

2NaF·UF4 was added in increments from a heated pot that emptied into the pump bowl. 

The fuel concentrate had a higher melting temperature than the carrier salt and tended to freeze 

in the fuel addition system. Heating the transfer line and getting the fuel concentrate to flow was 

the biggest operating problem of the reactor, and much of three days was spent in clearing plugs.

Overheating of the line with a welding torch caused a small hole which, however, did not prevent 

the loading from being completed. 

Physics experiments were run at low power for several days, and then the power was raised in 

steps to full power over 30 hours. The ARE ran very smoothly. However, small amounts of 

radioactive gas leaked into the reactor compartment from the hole in the loading line and set off 

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alarms. That was handled by hooking up a construction compressor that exhausted air from the 

compartment and blew it into an uninhabited area about 1000 ft away. 

The temperature coefficient of reactivity was negative as expected, and the power level could be 

raised or lowered by changing the speed of the blower that cooled the heat exchanger. The total 

power extracted in the fuel salt and sodium reached 2.5 MW with a steady-state fuel temperature 

of 1580°F, but the temperature exceeded 1620°F in transients. At steady conditions, the 

temperature rise across the core was 355°F. 

The last experiment conducted was a measurement of xenon retention during a 25-hour run at 

full power. It indicated that no xenon remained in the salt. After all scheduled tests were 

completed, the reactor was cycled between full and low power 21 times within 12 hours. The 

operating plan was to run the ARE for 100 hours; however, when that target was reached, only 9 

days of operation had been at power. Running longer was considered, but the experiment was 

shut down on schedule because of concern about weakness in the overheated fill line. That 

turned out to be a wise decision, because the line opened completely 5 days later and released 

radioactive gas into the reactor compartment.

The Air Force was pleased with the performance of the ARE and brought Pratt and Whitney 

Aircraft Company aboard to develop the indirect cycle power plant. ORNL began the design of a 

compact 60 MW reactor. And in spite of growing skepticism about success and the recognition 

that missiles might substitute for bombers, industrial and political support kept the national 

program going. But it was killed in March 1961 soon after John Kennedy took office. 

Thus ORNL’s ANP program came to an end, but in its 12-year run, it greatly expanded 

knowledge of the chemistry and technology of molten salts and made advances in materials, 

shield design, and other areas that enlarged the Laboratory’s ability to undertake new projects. 

MOLTEN-SALT REACTOR EXPERIMENT

With the ARE having shown the feasibility of molten-salt fuel, ORNL persuaded the AEC to 

fund a study of molten-salt power reactors. Two concepts were evaluated, both graphite 

moderated and based on the U-233/thorium fuel cycle. In one, the uranium and thorium were in 

the same salt. In the other, a thorium salt that formed a fertile blanket was kept separate from the 

fuel salt by a graphite barrier. The single-fluid concept was simpler, but a net breeding gain 

appeared to be possible in a two-fluid reactor.

The results of the study were encouraging. However, the AEC was still funding the 

homogeneous reactor program as well as a liquid-bismuth reactor program at Brookhaven 

National Laboratory, and it had more fluid-fuel reactors than it wanted. So a task force was 

created to evaluate all three programs. It concluded that the molten-salt reactor “had the highest 

probability of achieving technical feasibility.” This led to approval of a program at ORNL to 

develop molten-salt power reactors and to the demise of Brookhaven’s program and, as 

mentioned earlier, the homogeneous reactor program.

The molten-salt development program proceeded well, and by the end of 1959, the Laboratory 

felt justified in proposing a small reactor to investigate the technologies needed for civilian 

power. To keep the reactor simple and inexpensive, it had a single region like a converter, but the 

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salt did not contain thorium and in that sense was similar to the core of a breeder. (Later a 

concept was developed in which a virtual two-region reactor was created with a single salt by 

reducing the moderation in the outer region.) The proposed reactor, called the Molten-salt 

Reactor Experiment, was approved by the AEC. Design began in 1960, and construction started 

early in 1962. Details are given in Table C-2, and a sketch of the reactor is shown in Figure C-5. 

Table C-2. MSRE parameters

Power (heat) 7.4 MW

Fuel salt 65% Li7F, 29.1% BeF4,

5% ZrF4, 0.9 % UF4

Fuel melting point 813°F

Fuel inlet temperature 1,175°F

Fuel outlet temperature 1,225°F

Fuel flow rate 400 gal/min

Coolant salt 66% Li7F, 34% BeF4

Moderator Graphite

Container Hastelloy-N (68% Ni,

17% Mo, 7% Cr, 5% Fe)

Critical on 235U June 1, 1965

Critical on 233U October 2, 1968

Shut down December 1969

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C-10

Figure C-5. Sketch of the MSRE. (ORNL-LR-DWG 61097R). 

The fuel salt chosen for the MSRE was a mixture of the fluorides of lithium-7, beryllium, and 

zirconium selected to have good physical and nuclear properties. The melting point was 840°F. 

Fluoride salts do not wet graphite, so the MSRE was designed with the bare-graphite core shown 

in Figure C-6. It was 54 in. in diameter and made up of 2 in. square bars of specially produced 

low-porosity graphite. Flat passages for the salt were machined into the surfaces of the bars. 

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Figure C-6. Graphite core of the MSRE.

The molten salt flowed from the core to a heat exchanger where it gave up heat to another 

fluoride salt that carried it to an air-cooled radiator. As shown in Figure C-7, the salts were 

stored below the reactor in heated vessels, from which they could be forced up into the various 

systems by gas pressure. “Freeze valves” formed by air-cooled flattened sections of pipe retained 

the salts in the core. Figure C-8 is a photograph of the core, fuel pump, and heat exchanger. The 

radiator with its doors open is seen in operation in Figure C-9. 

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Figure C-7. Schematic of the MSRE. (ORNL-LR-DWG 63-1209R). 

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Figure C-8. MSRE core, fuel pump, and heat exchanger in the primary system containment tank.

(ORNL Photo 67051-64). 

Figure C-9. MSRE air-cooled radiator with its door open.

The salt pumps were at the high points in the system and like those on the ARE, were sump type 

and performed several functions. The sumps served as expansion tanks and were the free 

surfaces for the salt systems. A spray of salt into the gas space above the sump allowed noble 

gases to escape and be conducted to charcoal beds. In addition, a “sampler-enricher” tube leading 

from above provided access to the salt in the sump. With it, capsules of frozen fuel-salt 

concentrate could be lowered into the sump on a cable for small fuel additions, or thimbles could 

A-63
C-14

be lowered to collect salt samples.

A simple processing facility was installed in a cell near the drain tanks. It contained a tank in 

which salt could be sparged with hydrogen fluoride to remove water and oxygen or with fluorine 

to remove uranium. 

The reactor and all of the components that contacted salt were made of INOR-8, a nickel-base 

alloy developed by ORNL and the International Nickel Company for use in molten-salt reactors. 

It was later commercialized as Hastelloy-N and is still in use today. All salt piping and vessels 

were electrically heated.

Careful provisions drawing on HRT experience were made for maintenance, in particular the 

design and layout of components so that they could be removed from above with long-handled 

tools. A 16 in. thick shield surrounding the reactor reduced activation of components by 

neutrons. Effective procedures were devised to control the spread of radioactivity. 

Construction of the MSRE was completed in 1964, and critical experiments were run in mid1965. To bring the reactor critical, part of the enriched uranium was added to carrier salt in the 

drain tank in the form of a UF4-LiF eutectic. Then the final additions were made 88 grams at a 

time using the sampler-enricher.

Approach to full power began early in 1966, but within a few hours, plugs developed at several 

points in the fuel off-gas system. Apparently oil vapors that leaked past a gasket in the pump 

were polymerized by heat and radiation and deposited in small passages. Only a few grams of oil 

were involved, but three months was required to investigate and remedy the problem. 

After full-power operation was resumed, the reactor ran well with only a few small problems 

until the blades in one of the main blowers that cooled the radiator broke up. Redesigned blowers 

were procured and installed, but other problems, including recurrences of the off-gas plugs, kept 

interfering with operation until mid-December when an uninterrupted run of 30 days at full 

power was achieved. During it the reactor was seen to be stable and load-following as a result of 

the strong negative temperature coefficient.

The 30-day operation was followed by progressively longer runs; one lasted for 6 months, during 

which the behavior of fission products was investigated. Xenon-135 poisoning was found to be 

well below what it would have been if all of the xenon had remained in the core. Between runs, 

surveillance specimens were removed from the core for examination, tests made of containment 

and instrumentation, and maintenance operations performed.

Even though numerous problems were encountered during the early months of power operations, 

they were always overcome in spite of high radiation fields, and the capability for sustained 

operation was demonstrated. So at the end of the 6-month run, the fuel salt was transferred to the 

processing tank and the uranium removed from it by treatment with fluorine. The emerging UF6

was passed though a bed of hot sodium fluoride pellets that trapped volatile impurities, then was 

routed to a series of canisters in the high bay and absorbed on NaF pellets. The fuel-salt 

fluorination took only 47 hours, and the UF6 that it produced was clean enough that the canisters 

could be handled without shielding. 

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Next, some U-233 that had been made in a production reactor was added to the carrier salt, 

making the MSRE the first reactor ever to operate on U-233. Since Glenn Seaborg, chairman of 

the AEC, had led the team that first created U-233, he was invited to come to Oak Ridge and 

start up the reactor. He is seen in Figure C-10 raising the control rod. Although U-233 has a 

smaller delayed-neutron fraction than U-235, the reactor was quite stable and the dynamic 

behavior close to predictions. Later a small amount of plutonium was added to show the fissilefuel flexibility of molten-salt systems.

Figure C-10. Glenn Seaborg at the MSRE control panel.

The MSRE experience had increased confidence in the performance and practicality of moltensalt systems, and the Laboratory was anxious to move ahead with further development that 

would lead to demonstration of a breeder reactor. Since the budget was quite limited, the reactor 

was shut down at the end of 1969 and its funds devoted to development, particularly of a 

processing system that would be needed for breeding in a single-fluid reactor. Two problems that 

turned up during operation of the MSRE were also investigated. One was the confirmation that 

tritium produced in the fuel salt by neutron reactions with lithium, while small in amount, had 

diffused through the heat- exchanger tubes and would need to be intercepted in future reactors. 

The second was the discovery that although general corrosion was less than expected, fissionproduct tellurium had caused fine cracks on Hastelloy-N surfaces.

In spite of the technical success of the MSRE, the AEC was strongly committed to sodiumcooled fast breeder reactors and in 1973 gave orders to shut down the molten-salt program. A 

low-budget extension was soon granted in which solutions of the tellurium and tritium problems 

were demonstrated, but the program was stopped completely 2 years later. So ORNL’s moltensalt adventure came to an end after 20 years.

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MOLTEN-SALT REACTORS GENERAL REFERENCES

E. S. Bettis, R. W. Schroeder, G. A. Cristy, H. W. Savage, R. G. Affel, and L. F. Hemphill. 1957. “The 

Aircraft Reactor Experiment—Design and Construction.” Nuclear Science and Engineering 2: 

804–25. 

E. S. Bettis, W. B. Cottrell, E. R. Mann, J. L. Meem, and G. D. Whitman. 1957. “The Aircraft Reactor 

Experiment—Operation.” Nuclear Science and Engineering 2: 841–43.

D. W. Cardwell, and P. N. Haubenreich. 1971. Indexed Abstracts of Selected References on Molten-Salt

Reactor Technology. Rep. No. ORNL-TM-3595, Oak Ridge National Laboratory, Oak Ridge, 

TN.

W. K. Ergen, A. D. Callihan, C. B. Mills, and D. Scott. 1957. “The Aircraft Reactor Experiment—

Physics.” Nuclear Science and Engineering 2: 826–40.

H. G. MacPherson, ed. 1958. “Part II: Molten-Salt Reactors.” In Fluid Fuel Reactors, 565–697. Reading, 

MA: Addison-Wesley, 1958. 

H. G. MacPherson. 1985. “The Molten Salt Reactor Adventure.” Nuclear Science and Engineering 90: 

374–80. 

“Molten-Salt Reactor Experiment.” 1966. Oak Ridge National Laboratory, Oak Ridge, TN. Pamphlet. 

“A Review of Molten Salt Reactor Technology.” 1970. Nuclear Applications and Technology 8 (2): 102–

219. Special issue comprising a set of ten papers including M. W. Rosenthal, P. R. Kasten, and R.

B. Briggs, “Molten-Salt Reactors: History, Status, and Potential”; Paul N. Haubenreich and J. R.

Engel, “Experience with the Molten-Salt Reactor Experiment”; and W. R. Grimes, “Molten-Salt

Reactor Chemistry.”

M. W. Rosenthal, P. N. Haubenreich, and R. B. Briggs. 1972. The Development Status of Molten-Salt

Breeder Reactors. Rep. No. ORNL-4812, Oak Ridge National Laboratory, Oak Ridge, TN. 

Molten-Salt Reactor Program Semiannual Progress Report for Period Ending [August 1958–August 

1976]. Rep. Nos. ORNL-2474–ORNL-5132. Oak Ridge National Laboratory, Oak Ridge, TN. 

http://web.ornl.gov/info/reports/1976/3445604452553.pdf., (last in a series of semiannual 

progress reports that began in 1957; they cover the MSRE and all aspects of the related R&D 

program. 

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Molten Salt Chemistry Workshop 

Appendix B B-1 

APPENDIX B: A DEVELOPER’S PERSPECTIVE ON THE NECESSITY OF SALT 

PROPERTY RESEARCH 

During the initial stages of the design process, it is vital to establish the capability to rapidly evaluate the 

base performance of different salt systems in order to select the optimal composition for a given 

application. This requires both extensive experimental data for the salt system in question and correlations 

and predictive tools that can give insight into the fundamental thermophysical salt properties for systems 

that have yet to be explored or have been explored only briefly. 

Figure 1 outlines the salt properties data required for input to the framework of a MSR design project. 

Although this discussion centers on the design phase of an MSR, the same information will be required 

for the safety analysis. The following thermophysical properties must be known as a function of 

temperature, both in the normal operating reactor temperature range and at higher temperatures to 

account for criticality excursions: 

 Melting point (phase diagram, and associated sensitivity with composition) 

 Heat capacity 

 Thermal conductivity 

 Viscosity 

 Density 

 Vapor pressure 

 Gibbs free energy of formation (and the associated redox potential)

Figure 1. Organization schematic in a molten salt reactor design project. 

These properties must be known for the initial salt composition (at the first startup for the reactor), as well 

as during reactor operation, when the salt composition will contain increasing concentrations of fission 

Chemistry

Team

Melting Point (sets Cold Leg Temperature)

Density (Pumps)

Vapor Pressure (Pumps)

Viscosity

Heat Capacity

Thermal Conductivity

Emissivity Thermo‐

hydraulics

Team

Materials

Team

Salt Potential

Composition

Temperature

Melting Point

Density

Composition

Physics Enrichment

Team

Burnup Composition

Salt Cleanup

Design Team

Composition

Melting Point

Vapor Pressure
Molten Salt Chemistry Workshop

B-2 Appendix B 

products as a function of burnup (time and power at which the reactor operates), burnout of the fissile 

species (for example, 235U → 2 fission products), and an increase in bred species (for example, 238U →

239Pu). 

It will be necessary for national laboratories, universities, and industry to measure these properties where 

gaps exist in historically published data. It is likely that each reactor design will require its own unique 

data set, based upon its starting chemistry, temperature range, and burnup profile. 

The following summarizes the rationale for the properties requirements for each of the reactor design 

groups shown in Figure 1. 

Physics

This evaluation begins with neutronics, shown as “physics” in Figure 1. If a fuel salt cannot achieve and 

maintain criticality, it should not be explored as an option. The multiplication factor (reactivity) of any 

given MSR, whether using a thermal or a fast neutron spectrum, depends upon the volumetric density of 

fissile isotopes. This is dependent upon the composition of the salt (e.g., percent uranium halide in a 

halide salt mix), the enrichment of the fissile isotope (e.g., 235U% within the total uranium), and the 

density (molar volume) of the salt mixture (where the salt is a mixture of an actinide salt and other 

nonfissile “carrier” salt species, e.g., NaCl, NaF, LiF, or others). The density of the salt mixture is 

temperature-dependent, i.e., the salt becomes less dense with increasing temperature. The resulting 

negative temperature-coefficient of reactivity is the basis of an MSR’s reactivity control, as well as an 

important passive safety feature. Reactor designers need accurate density data to accurately design the 

reactor size and to ensure that reactivity is adequately controlled. 

The salt density of a mixture is not always an additive combination of the sum of the pure components. 

For some salts, the total volume of the salt mixture exceeds the sum of the individual component 

volumes. This molar volume anomaly is not completely understood but is thought to be due to complex 

ion formation and a resulting open structuring of the salt structure. This stresses the need for experimental 

measurements of density, since theoretical calculations are not necessarily predictive of the density of salt 

mixtures. Determinations of the density of the salt in and out of a reactor (with and without fission 

products) will be necessary to understand reactor control requirements during startup and shutdown. 

Thermohydraulics 

Once it has been shown that criticality can be achieved and maintained, it must then be determined that 

the heat transfer properties of the given system allow for heat to be effectively transported so that power 

may be produced. Heat exchanger design requires knowledge of salt density, heat capacity, viscosity, and 

thermal conductivity as a function of salt temperature and composition. Thermal radiation transport 

coefficients (e.g., absorption, transmission, and fluorescence) and other radiation transport effects may 

need to be measured and understood to predict radiative heat transport. Fluid dynamics and heat transfer 

properties will determine the size and geometry of the heat exchanger given a designed power output. The 

heat exchanger must be designed so as to prevent salt freeze-up at the outlet (cold leg) of the heat 

exchanger. Optimal heat exchanger designs will minimize fuel salt volume and therefore reduce the cost 

of the reactor. 

Materials 

Corrosion of reactor structural alloys is strongly affected by the electrochemical potential of the salt 

mixture. It is necessary to keep the reactor structural materials at sufficiently cathodic (reductive) 

potentials to provide thermodynamic immunity toward corrosion of structural alloys. Specific chemical 

effects of fission products that grow in during reactor operation and potentially attack structural alloys 
Molten Salt Chemistry Workshop 

Appendix B B-3 

need to be investigated. For example, tellurium, a fission product, was seen in the MSRE to diffuse into 

structural grain boundaries and produce cracks by leaching chromium from the Hastelloy N alloy The 

cracking could be reduced by controlling the redox potential of the salt. Materials performance as a 

function of salt composition, electrochemical potential, and temperature must therefore be studied. 

Auxiliary systems (cleanup) 

Fission products will have to be managed at some level in all MSR designs. Off-gas hold-up and disposal 

systems will need to be designed for gaseous fission products. Some mechanism to remove or manage 

insoluble fission products will need to be in place. Some MSR designs may require a soluble fission 

product removal system. The composition of the fuel/burnup salt, electrochemical potentials, and vapor 

pressures must be known for the successful deployment of these systems. 

Down-selection of candidate salts will be accomplished by each MSR design team according to its own 

selection criteria for salt properties. Salts with high melting points, or high neutron capture cross sections, 

will likely be eliminated from further consideration; and testing of other properties of these salt systems 

may not be useful. Fluid-fueled systems require a uranium or plutonium driver as a constituent of the 

primary salt—thorium can be included as a fertile constituent in some designs 

A quality control program must be instituted to guarantee the quality of the data to the level required. 

Access to pure salts for laboratory studies 

There is currently no supply chain for many actinide salts. Most researchers will find it necessary to 

synthesize their own salts using a variety of synthesis pathways. Although ORNL established processes 

and practices for purifying and handling fluoride salts for research and for and the MSRE (J. H. Shaffer in 

ORNL-4616; Bamberger in Advances in Molten Salt Chemistry 3 (1975): “Experimental techniques in 

molten fluoride chemistry”), there is no national standard for testing and reporting the purity of the salt. 

Different purification processes and procedures will be needed for other (e.g., chloride) salts. Although 

much, but not all, work can be done with surrogate salt systems, the properties and chemistry of the 

actinides will also need to be studied directly. 

To enable progress in MSR salt properties research, a central repository is strongly encouraged for both 

actinide and surrogate salts. This would provide universities, national labs, and industry with consistent 

salts that would allow better cross-comparison between research groups. Site licensing controls exist for 

natural uranium salts, limiting the amount of uranium salt that can be stockpiled at a given research 

location. A central repository would allow researchers access to adequate quantities of salt, while keeping 

their inventories below regulatory limits. Additionally, handing plutonium salts requires extremely 

rigorous and specialized controls for any isotopic composition. Special consideration could be given to 

doing work with non-fissile plutonium as a way to avoid the nuclear safety and security burden associated 

with the effort. Disposal pathways must be identified as a part of any site license, and central repository 

disposal services would be most helpful. 

Salt purification 

Most salts are hydroscopic and contain significant amounts of water. Thus, best practices for dewatering 

salts should be studied, published, and standardized. Best practice procedures for establishing clean salt 

may involve some or all of the following: handing the salt in an inert atmosphere, heating the salt, and/or 

purging the salt with a purifying gas (e.g., HF, HCl or CCl4), perhaps in the presence of hydrogen gas. 

Once a salt is purified, standard methods of handling and assaying the purity of the salt economically and 

routinely should be developed and standardized. 
Molten Salt Chemistry Workshop

B-4 Appendix B 

Glovebox-compatible advanced instrumentation 

All handling and analysis of salts outside of closed or inert-atmosphere-blanketed containers needs to be 

avoided, and inert-atmosphere gloveboxes should be used. The dimensions of the analysis equipment 

must be such that it can be located within the volume of the glovebox. Electronics must be located outside 

the glovebox, since those experienced in the field are aware that standard electronics tend to fail (arcing) 

in the inert atmosphere. Easy operation and maintenance by the operator in gloves must be designed into 

the instruments. 

The following characterization methods are not meant to be a complete list of candidate methods. These 

serve as known examples of methods currently in use or being developed for future use: 

Melting point determinations

Melting points can be determined in a variety of ways, most commonly with differential scanning 

calorimetry (DSC). Standardization of operating parameters for DSC operation needs to be established. 

For example, recommendations should be made for temperature ramp rate, number of freeze/thaw cycles, 

purity of cover gas, purge time of the cover gas, and sample crucible material. Other melting point 

measurement methods may be used, and the operating parameters for these methods should be established 

as well. 

Density and viscosity 

There are a variety of methods for measuring the density and viscosity of a molten salt. The best 

method(s) and procedures need to be determined by the molten salt community. The Archimedes-bob Ptmethod is the reference method most commonly employed for density, and various capillary and 

rotational methods have been used for viscosity measurement. 

Electrochemical potential 

Salt electrochemical potential can be calculated with quality Gibbs free energy data, which are 

determined from experimental thermochemical data. Electrochemical potential can also be measured 

directly in an electrochemical cell that uses a stable reference electrode. 

If a given salt system has been shown to effectively produce and transport heat, the feasibility of the 

design will then rest on the safety and operational performance of the reactor system, as stated earlier. 

The following are some additional questions that can inform this safety and operations evaluation: 

 Do the properties of the salt change as a result of the presence of fission products, transuranics, and 

radiation, to the point where heat production and transport are significantly affected? 

 What are the solubility limits of species that will be present over the course of operations? Is there a 

risk that some of these species could precipitate out in ways that would be unfavorable? 

 What is the free energy of formation of the species contained in the salt in relation to the reactor’s 

structural materials, and can it inform the extent of corrosion that will occur within the system? 

 What is the vapor pressure of the species that will be present over the course of operations, to 

sufficiently inform considerations for liquid-gas interfaces of all systems containing a fuel salt and its 

byproducts? 
Molten Salt Chemistry Workshop 

Appendix C C-1 

APPENDIX C: WORKSHOP AGENDA 

MOLTEN SALT CHEMISTRY WORKSHOP: Defining Technology and Applied R&D Needs for 

Molten Salt Chemistry 

Building 5100, Joint Institute for Computational Sciences 

Oak Ridge National Laboratory 

Chairs: David F. Williams and Phillip F. Britt (Oak Ridge National Laboratory) 

Event contact: David F. Williams, 865-574-8853 (office), 865-255-4737 (mobile); 

williamsdf2@ornl.gov

Monday, August 10, 2017

Time Event Lead 

7:15–8:00 a.m. ORNL Visitor Check-in and Workshop Registration at Bldg. 5100 Lobby 

8:00–8:30 a.m. 

Welcome and Workshop Charge (Joint 

Institute for Computational Sciences [JICS] 

Lecture Hall) – working breakfast 

Alan Icenhour, Associate Laboratory 

Director, Nuclear Science and Engineering 

Directorate, ORNL 

John Herczeg, Deputy Assistant Secretary, 

Nuclear Technology Research and 

Development, DOE-NE 

8:30–9:00 a.m. Technical Working Group Perspective on 

Workshop 

Nick Smith, MSR Technical Working Group 

Representative 

9:00–9:40 a.m. Challenges in MSR Chemistry Vic Maroni, Argonne National Laboratory 

Distinguished Fellow 

9:40–10:20 a.m. Group Photo and Coffee Break All 

10:20–11:00 a.m. Irradiation Effects in MSR Systems Steve Zinkle, Nuclear Engineering, University 

of Tennessee–Governor’s Chair 

11:00–11:40 a.m. Materials Compatibility in MSR Systems James Keiser, Materials Science and 

Technology Division, ORNL 

11:40 a.m.–12:00 p.m. Separation Options and Waste Forms Dave Williams (ORNL) 

12:00–1:00 p.m. 

Working Lunch: (JICS Lecture Hall) 

 12:00–12:20—serving line in Lobby 

o Bring lunch to Lecture Hall 

 12:30–1:00 p.m. 

o Working Group (WG) Charge 

Dave Williams and Phil Britt, Workshop Cochairs (ORNL) 

1:00–5:30 p.m. 

Parallel Working Group Sessions 

 WG#1: Physical Chemistry and Salt 

Properties 

 WG#2: Analytical Chemistry (includes 

measurement instruments and online 

sensors) 

 WG#3: Molten Salt Fission Product 

Chemistry and Solid Salt Radiolysis 

 WG#4: Materials Compatibility 

 WG#5: Computational Chemistry and 

Materials Science 

WG Co-Chairs 

WG#1: Alex 

Navrotsky, Mark 

Williamson 

WG#2: Sam Bryan, 

Sheng Dai 

WG#3: Bill DelCul, 

Tina Nenoff 

WG#4: James Keiser, 

Preet Singh 

WG#5: Bobby 

Sumpter, Brian 

Wirth 

Locations 

1. Rm 130 

2. Rm 262 

(Boardroom) 

3. Rm 128 (Lecture 

Hall) 

4. Bldg 5700, Rm 

E104*(Monday 

only) 

5. Roams among all 

groups 
Molten Salt Chemistry Workshop

C-2 Appendix C 

Monday, August 10, 2017 (continued)

Time Event Lead 

5:45–7:00 p.m. Onsite Working Dinner — WG Status Dinner for All, WG status from chairs (JICS 

Lobby) 

7:00–9:00 p.m. Parallel Working Group Sessions (cont.) WG Members led by chairs / designated 

location above 

9:00 p.m. Adjourn Day 1 All 

Tuesday, August 11, 2017

Time Event Lead 

8:00–10:15 a.m. 

Working Group Sessions – Working Breakfast 

Breakfast served in JICS Lobby

 WG#1: Physical Chemistry and Salt 

Properties 

 WG#2: Analytical Chemistry 

 WG#3: Molten Salt Fission Product 

Chemistry and Solid Salt Radiolysis 

 WG#4: Materials Compatibility 

 WG#5: Computational Chemistry and 

Materials Science 

WG Co-Chairs 

WG#1: Alex 

Navrotsky, Mark 

Williamson 

WG#2: Sam Bryan, 

Sheng Dai 

WG#3: Bill DelCul, 

Tina Nenoff 

WG#4: James Keiser, 

Preet Singh 

WG#5: Bobby 

Sumpter, Brian Wirth 

Locations 

1. Rm 130 

2. Rm 262 

(Boardroom) 

3. Rm 128 (Lecture 

Hall) 

4. Rm 134 (Learning 

room) 

5. Bldg. 4100, 

Rm 203 

10:15–10:40 a.m. Break and Reconvene in JICS Auditorium All 

10:40–11:00 a.m. 

Report 

 WG#1: Physical Chemistry, Salt 

Properties 

WG#1: Alex Navrotsky, Mark Williamson 

11:00–11:20 a.m. Report 

 WG#2: Analytical Chemistry WG#2: Sam Bryan, Sheng Dai 

11:20–11:40 a.m. 

Report 

 WG#3: Molten Salt Fission Product 

Chemistry and Solid Salt Radiolysis 

WG#3: Bill DelCul, Tina Nenoff 

11:40 a.m.–12:00 p.m. Report 

 WG#4: Materials Compatibility WG#4: James Keiser, Preet Singh 

12:00–1:00 p.m. 

Working Lunch – Report 

(Serving line in JICS Lobby, return to 

Lecture Hall)

 WG#5: Computational Chemistry and 

Materials Science 

WG#5: Bobby Sumpter, Brian Wirth 

1:00–5:30 p.m. WG Sessions (cont.) WG members/designated location as above 

5:30–6:30 p.m. Onsite Working Dinner – WG Status Dinner for all, WG chair status (JICS Lobby) 

6:30–9:00 p.m. WG Preparation for Final Workshop Plenary All 

9:00 p.m. Adjourn Day 2 All 

 
Molten Salt Chemistry Workshop 

Appendix C C-3 

Wednesday, August 12, 2017

Time Event Lead 

8:00–8:30 a.m. 

Report 

 WG#1: Physical Chemistry, Salt 

Properties 

WG#1: Alex Navrotsky, Mark Williamson 

8:30–9:00 a.m. Report 

 WG#2: Analytical Chemistry WG#2: Sam Bryan, Sheng Dai 

9:00–9:30 a.m. 

Report 

 WG#3: Molten Salt Fission Product 

Chemistry and Solid Salt Radiolysis 

WG#3: Bill DelCul, Tina Nenoff 

9:30–10:00 a.m. Report 

 WG#4: Materials Compatibility WG#4: James Keiser, Preet Singh 

10:00–10:30 a.m. Break All 

10:30–11:00 a.m. 

Report and Discussion 

 WG#5: Computational Chemistry and 

Materials Science 

WG#5: Bobby Sumpter, Brian Wirth 

11:00–11:30 a.m. Final Discussion and Closing Remarks John Herczeg, DOE-NE 

Dave Williams, ORNL 

11:30 a.m. Adjourn Day 3* All 

Report Team Agenda (same as previous breakout rooms) 

1:30–5:00 p.m. Writing Team Tuesday Breakout Rooms 

*You are welcome to lunch at the ORNL Cafeteria on Wednesday. A workshop lunch is not provided on Wednesday. 

Molten Salt Chemistry Workshop 

Appendix D D-1 

APPENDIX D: WORKSHOP PARTICIPANTS 

Registered participants for April 10–12, 2017, Molten Salt Chemistry Workshop at ORNL 

Name Organization Role 

Abney, Carter Oak Ridge National Laboratory Analytical Chemistry (scribe)

Besmann, Ted University of South Carolina Computational Chemistry/Materials 

Betzler, Ben Oak Ridge National Laboratory Computational Chemistry/Materials (scribe)

Bresee, James Department of Energy, Office of Nuclear 

Energy

Observer

Britt, Phil Oak Ridge National Laboratory Meeting co-chair

Bryan, Sam Pacific Northwest National Laboratory Analytical Chemistry

Calderoni, Pattrick Idaho National Laboratory Analytical Chemistry

Chidambaram, 

Dev 

University of Nevada, Reno Materials Compatibility

Chvala, Ondrej University of Tennessee Computational Chemistry/Materials

Cooper, Thurman Terrestrial Energy USA Ltd Physical Chemistry and Salt Properties
Molten Salt Chemistry Workshop

D-2 Appendix D 

Name Organization Role 

Czerwinski, Ken TerraPower Physical Chemistry and Salt Properties

Dai, Sheng Oak Ridge National Laboratory Analytical Chemistry (WG co-chair)

DelCul, Bill Oak Ridge National Laboratory Molten Salt Fission Product Chemistry (WG 

co-chair)

Dingreville, Remi Sandia National Laboratories Computational Chemistry/Materials

Dixon, David The University of Alabama Computational Chemistry/Materials

Dunstan, Don Energy Storage Solutions Materials Compatibility

Fredrickson, Guy Idaho National Laboratory Molten Salt Fission Product Chemistry

Hammelman, 

James

US Nuclear Regulatory Commission Observer

Hanson, Michael Elysium Industries Physical Chemistry and Salt Properties

Henager, Charles Pacific Northwest National Laboratory Computational Chemistry/Materials (WG 

co-chair)

Herczeg, John Department of Energy, Office of Nuclear 

Energy

Speaker/Observer

Horton, Linda Department of Energy, Basic Energy 

Sciences

Observer

Icenhour, Alan Oak Ridge National Laboratory Speaker/Observer

Keiser, James Oak Ridge National Laboratory Speaker/ Materials Compatibility (WG 

co-chair)

Kim, Hojong The Pennsylvania State University Molten Salt Fission Product Chemistry

Kruizenga, Alan Sandia National Laboratories Materials Compatibility

Kung, Stephen Department of Energy, Office of Nuclear 

Energy

Observer

Longford, Des PHDS Co. Analytical Chemistry

Margulis, Claudio University of Iowa Computational Chemistry/Materials

Maroni, Vic Argonne National Laboratory Speaker/Physical Chemistry and Salt Properties 

(scribe)

Marsden, Kenneth Idaho National Laboratory Molten Salt Fission Product Chemistry

Massie, Mark Transatomic Power Corporation Observer/ Industry Representative

Mayes, Richard Oak Ridge National Laboratory Molten Salt Fission Product Chemistry (scribe)

Mays, Gary Oak Ridge National Laboratory Physical Chemistry and Salt Properties

McDuffee, Joel Oak Ridge National Laboratory Molten Salt Fission Product Chemistry

McLaughlin, David Westinghouse Electric Co. Ltd Physical Chemistry and Salt Properties

McNeese, James Lawrence Livermore National Laboratory Molten Salt Fission Product Chemistry

Meeks, Noah Southern Company Services Observer/ Industry Representative

Memmott, 

Matthew

Brigham Young University Physical Chemistry and Salt Properties

Motsegood, Perry Argonne National Laboratory Analytical Chemistry

Muralidharan, 

Govindarajan 

Oak Ridge National Laboratory Materials Compatibility

Navrotsky, 

Alexandra 

University of California–Davis Physical Chemistry and Salt Properties (WG 

co-chair)
Molten Salt Chemistry Workshop 

Appendix D D-3 

Name Organization Role 

Nenoff, Tina Sandia National Laboratories Molten Salt Fission Product Chemistry (WG 

co-chair)

Pal, Uday Boston University Physical Chemistry and Salt Properties

Pheil, Ed Elysium Industries Physical Chemistry and Salt Properties

Qualls, Lou Oak Ridge National Laboratory Observer

Raiman, Stephen Oak Ridge National Laboratory Materials Compatibility (scribe)

Rappleye, Devin Lawrence Livermore National Laboratory Analytical Chemistry

Robertson, Sean Transatomic Power Corporation Observer/ Industry Representative

Rohatgi, Upendra Brookhaven National Laboratory Observer

Salay, Michael US Nuclear Regulatory Commission Observer

Scarlat, Raluca University of Wisconsin Physical Chemistry and Salt Properties

Shaka, A. J. University of California–Irvine Analytical Chemistry

Sham, Sam Argonne National Laboratory Materials Compatibility

Singh, Preet Georgia Institute of Technology Materials Compatibility (WG co-chair)

Smith, Nicholas Southern Company Services Speaker/ Materials Compatibility

Sorensen, Kirk Flibe Energy Physical Chemistry and Salt Properties

Stoddard, Michael Alpha Tech Research Corp Molten Salt Fission Product Chemistry 

Sumpter, Bobby Oak Ridge National Laboratory Computational Chemistry/Materials (WG 

co-chair) 

Sun, Xaidong University of Michigan Materials Compatibility 

Terrani, Kurt Oak Ridge National Laboratory Materials Compatibility 

Todd, Terry Idaho National Laboratory Molten Salt Fission Product Chemistry 

Tortorelli, Peter Oak Ridge National Laboratory (retired) Materials Compatibility 

Wells, Daniel Electric Power Research Institute Physical Chemistry and Salt Properties 

Williams Dave Oak Ridge National Laboratory Meeting co-chair 

Williams, Joseph US Nuclear Regulatory Commission Observer 

Williamson, Mark Argonne National Laboratory Physical Chemistry and Salt Properties (WG 

co-chair) 

Willit, James Argonne National Laboratory Molten Salt Fission Product Chemistry 

Wilson, Dane ThorCon Power Materials Compatibility 

Wirth, Brian University of Tennessee Computational Chemistry/Materials (WG 

co-chair) 

Zinkle, Steve University of Tennessee Speaker/Materials Compatibility 

Ziock, Klaus Oak Ridge National Laboratory Molten Salt Fission Product Chemistry 
Molten Salt Chemistry Workshop

D-4 Appendix D 

This report was prepared as an account of work sponsored by an agency of the United States 

Government. Neither the United States Government nor any agency thereof, not any of their 

employees, makes any warranty, express or implied, or assumes any legal liability or 

responsibility for the accuracy, completeness, or usefulness of any information, apparatus, 

product, or process disclosed, or represents that its use would not infringe privately owned 

rights. Reference herein to any specific commercial product, process, or service by trade name, 

trademark, manufacturer, or otherwise, does not necessarily constitute or imply its 

endorsement, recommendation, or favoring by the United States government or any agency 

thereof. The views and opinions of authors expressed herein do not necessarily state or reflect 

those of the United States Government or any agency thereof.