Molten Salt Reactors

Yousif Kelaita
February 18, 2015

Submitted as coursework for PH241, Stanford University, Winter 2015


Fig. 1: Schematic of a typical molten salt reactor. (Source: Wikimedia Commons)

Despite being introduced in the early 1960's, molten salt reactors (MSRs) are only now beginning to receive attention. The current viability of the technology is built upon numerous US government and military experiments that originated over 50 years ago. While the current leading example of the technology is the liquid fluoride thorium reactor, numerous other designs exist with their own advantages and disadvantages. Herein, a historical perspective of MSRs will be presented along with a brief overview of the principles of operation and the role that various molten salts play in such reactors.

Principles of Operation

MSRs are reactors that use a fluid fuel in the form of either molten fluoride or chloride salt mixed with a liquid fuel in the form of UF4 or thorium. Hence, the salt can be both the fuel and the coolant at the same time. A typical schematic is displayed in Fig. 1. While fuel choices are relatively limited, there are numerous fused salts to choose from that are prioritized based on safety and practicality. In general, fluoride salts are preferred for two reasons: (a) fluorine has only one stable isotope (F-19) and (b) it does not easily become radioactive under neutron bombardment.

On the other hand, chlorine has two stable isotopes (Cl-35 and Cl-37), as well as an unstable isotope between them that facilitates neutron absorption. Of the fluoride salts, there are a variety of options that can be chosen based on at least two different important characteristics: neutron capture and moderating efficiency. Low neutron capture leads to absorption of fewer neutrons, while higher moderating ratio leads to slower neutrons that can more readily sustain a nuclear chain reaction. Table 1 presents a range of various materials that can be used in nuclear reactors as either coolants or moderators, with characteristic neutron captures and moderating ratios listed. [1] As is clear, fluoride salts offer nominally low neutron captures while simultaneously working as moderators for sustained fission reactions.

Material Total neutron capture relative to graphite (per unit volume) Moderating ratio (Avg. 0.1 to 10 eV)
Heavy water 0.2 11449
Light water 75 246
Graphite 1 863
2LiF-BeF2 8 60
LiF-NaF-BeF2 (31-31-38) 20 22
LiF-ZrF4 (51-49) 9 29
LiF-NaF-ZrF4 (26-37-37) 20 13
KF-ZrF4 (58-42) 67 3
RbF-ZrF4 (58-42) 14 13
LiF-NaF-KF (46.5-11.5-42) 90 2
Table 1: Comparison of the neutron capture and moderating efficiency of several materials. [1]

So far, the most common salt used is FLiBe, one form of which is listed in the table as a mixture of LiF and BeF2. [2] Lithium and beryllium both act as effective moderators, and the two salts form a eutectic salt mixture that has a lower melting point. Lower melting point salts are advantageous because they simplify startup and reduce the risk of the salt freezing as it is cooled. In addition, beryllium also performs neutron doubling, where the nucleus will re-emit two neutrons after absorbing a single neutron.

In general, with the proper choice of fuel and fused salt to form the mixture, MSRs offer the following advantages in comparison to light water reactors: (a) operation at lower pressures, (b) no need for fuel rod manufacturing, and (c) operation at higher temperatures and thus higher efficiency for electricity production. Disadvantages include relative lack of wide-scale development and the need for vast regulatory changes. [3]

Brief History

The U.S. aircraft reactor experiment (ARE) that began in 1946 was the first extensive look into MSRs. The ARE was originally designed to be an engine for a nuclear-powered bomber. The ARE used a molten fluoride salt with uranium tetrafluoride fuel, beryllium oxide as a moderator, and sodium as a coolant. In the initial test in 1954, the reactor operated for 100 MW-hours over nine days. While the reactor was an important initial step in MSR design, it only ran for a few weeks and produced zero nuclear power. [4]

The more significant effort in MSR design occurred at Oak Ridge National Laboratory (ORNL) in the 1960s, which culminated in the Molten-Salt Reactor Experiment (MSRE). This was a 7.4 MW test reactor that relied on the now popular liquid fluoride thorium fuel salt, with U-235 and U-233 as the primary fuel. The secondary coolant was a FLiBe salt and the moderator was a graphite core. The reactor operated for about 1.5 years of full- power operation, proving the viability of the MSR design principle. For example, it confirmed that the salt was immune to radiation damage, the graphite did not attack the fuel salt, and corrosion of the metal used to build the reactor did not occur on any meaningful scale. Despite this, the MSR largely fell out of favor and research in the U.S. did not continue at an appreciable rate. While other countries would pick up the mantle, only recently has new interest in the technology once again picked up. [5]

Future Outlook

Numerous entities both private and public are continuing to advance the technology and application of MSRs by primarily focusing on the LFTR design. For example, Reactive IVS, a company in Denmark, is developing a molten salt waste-burner (MSW) that is designed to use nuclear waste from conventional nuclear reactors as a fuel. On larger scales, a consortium of countries including Japan, the U.S., and Russia is working on the Fuji MSR, which will be a 100 to 200 MW reactor upon completion. [6] Similarly, both China and India are working on MSRs for deployment within the decade. So while the technology certainly took its time coming to maturity, the future of the MSR is bright provided nuclear power continues to assert its prevalence.

© Yousif Kelaita. The author grants permission to copy, distribute and display this work in unaltered form, with attribution to the author, for noncommercial purposes only. All other rights, including commercial rights, are reserved to the author.


[1] D. T. Ingersoll, "Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)," Oak Ridge National Laboratory, ORNL/TM-2005/218, December 2005.

[2] D. F. Williams, L. M. Toth, and K. T. Clarno, "Assessment of Candidate Molten Salt Coolants for the Advanced High-Temperature Reactor (AHTR)", Oak Ridge National Laboratory, ORNL/TM-2006/12, March 2006.

[3] "The Use of Thorium in Nuclear Power Reactors," Brookhaven National Laboratory, WASH-1097, June 1969, Section 5.3.

[4] E. S. Bettis et al., "The Aircraft Reactor Experiment-Design and Construction," Nucl. Sci. Eng 2, 804 (1957).

[5] P. N. Haubenreich and J. R. Engel, "Experience with the Molten-Salt Reactor Experiment," Nucl. Appl. Technol. 8, 118 (1970).

[6] J. Serp et al., "The Molten Salt Reactor (MSR) in Generation IV: Overview and Perspectives," Prog. Nucl. Energy 77, 308 (2014).