1. Alternative Containment Materials for Molten
Salt Reactors
May 7, 2016
Kegan Mckinnon

2. Introduction
In the pursuit of safer nuclear energy technology, generation IV reactors have been designed
with efficiency and safety as the primary strategic goals. The attraction of potentially reusing spent
nuclear fuel as well as utilizing thorium as a fuel source has not gone unnoticed. The United States,
Russia, and China have all made forays of some kind into molten salt nuclear reactors [1]. The research
relating to this paper began with the Molten Salt Reactor Experiment (MSRE), and the subsequent Oak
Ridge National Lab (ORNL) reports. It was then decided that while Hastelloy N is a good material to start
with, but modern materials science has developed many materials since the dawn of the Cold War. A
two pronged attack was decided upon, consisting of further development along the lineage of nickel
based superalloys, and exploration into lower cost materials that may have been overlooked, specifically
316 Stainless Steel.
Background
The motivation for this study of materials derives from the molten salt propensity to corrode
piping at elevated operating temperatures. While this was discussed in detail in a previous paper, it is
still worth noting the corrosion mechanisms and issues [2]. Molten salts can be very corrosive if
contained in an oxidizing environment, thus reactors are designed to be reducing to limit corrosion.
ORNL discovered early on that 316 Stainless Steel (SS) is susceptible to high amounts of mass transfer
due to chromium dissolution from the piping. According to researchers at the University of Wisconsin,
“There are three driving forces for corrosion in molten fluorides: impurities, temperature gradients, and
activity gradients” [3]. A comparison of activity levels for different fluoride compositions can be seen in
Figure 1, with Mo, Ni, and Fe having the highest free energy of formation. Chromium dissolution due to
all three of the aforementioned driving forces would cause Stainless Steels to fail on a timescale of tens
of hours, two orders of magnitude less than nickel based alloys.

3. Figure 1 Gibbs free energy of formation for various fluoride compositions. Taken from [3].
ORNL’s response to the shortcomings of 316 SS was the design of Hastelloy N, a nickel
molybdenum system optimized for creep and corrosion resistance. Further review of Hastelloy N
samples indicated tellurium embrittlement, tellurium entering the system as a decay byproduct.
Tellurium was controlled by adjusting the ratio of UF3 to UF4 at roughly 2% UF3 [2]. These
embrittlement issues are important to keep in the back of one’s mind as SS materials are explored in
order to design mechanisms to defeat embrittlement on the front end of the alloy design process. The
end state is to be able to operate a reactor utilizing thermal spectrum neutrons in the temperature
range from 500-800 C. An additional benefit to using a Stainless Steel is that ASME Section III regulation
is already in place for use in high temperature applications, possibly allowing for earlier deployment to
test other designs like heat exchangers or pumps [4].
Much Ado About Salt Composition
Through the course of this research many different salt compositions were encountered, and it
would prove folly to attempt to list each individual composition. The main salts used were FLiBe and
FLiNaK salts. There are positive and negative attributes to any salt composition chosen, and each
responds differently to containment materials and temperatures. In a broader sense there are three
variables to consider when choosing salts in addition to considering their compatibility with container
materials, moderation capabilities, and ease of processing. Those variables are beryllium toxicity, lithium
availability, and the inherent hygroscopicity of salts.
Beryllium in FLiBe salts is toxic to humans with the potential for causing lung cancer and Chronic
Beryllium Disease. Thus the use of a FLiBe salt comes with the additional cost of yearly checkups for
workers in contact with beryllium as the effects are nearly instantaneous in some individuals, while in

4. others it may take years for effects to be seen [4], [5], and [6]. With alternative battery technology,
worldwide lithium demand for lithium batteries will increase greatly. Devices such as cell phones,
laptops, and any item coming from electric car manufacturer Tesla’s Gigafactory will use lithium, which
increases demand for the element. Lithium used in FLiBe must be separated from natural lithium in
order to obtain pure Li-7, a process that yields Li-6, which can be used in the production of tritium,
causing proliferation issues [5]. Lastly molten salts are hygroscopic in nature. Adequate precautions
must be taken when processing salts and operating the reactor to ensure minimal moisture content in
the salt, as this creates hydroxides
and H+ ions as well as hydrofluoric acid gas that increases corrosion
[1].
The Case for Stainless Steel
As noted previously, modified versions of nickel molybdenum alloys are being studied for MSR
applications, however 316 SS should be considered for a few reasons. 316 SS was known to plug
convection loops in tens of hours, and this plugging occurred in a salt containing roughly 4 wt% uranium.
The uranium present in the salt can be attributed to some portion of the mass transfer [7]. MIT found
that in the absence of a neutron field and in a clean salt, 316 SS is compatible with FLiBe salts [8]. In MSR
applications nickel alloys experience embrittlement in neutron fluence above 1020
n/cm2
, while steels
were found to experience less embrittlement under similar conditions [9]. This may be due to smaller
cross sections in the alloying contents, or a difference in the free space in between grain boundaries for
steels versus nickel molybdenum systems. According to MIT’s analysis of FHRs (Fluoride High
Temperature Reactors), a 316 SS vessel protected by graphite reflectors and shields should allow for
operation at high temperature (700C) with a low dose rate. This is extrapolated due to the fact that
these temperatures are not seen in a LWR (Light Water Reactor), which is the source of a large amount
of 316 SS data [8]. A collection of defect data across a broad temperature range is indicated in Figure .
With increasing temperature, the metal appears to reduce the number of irradiation induced vacancies
and voids in the lattice by two orders of magnitude. This may be due to phonons making both vacancies
and voids thermally unstable, allowing for a process similar to annealing to occur. Dislocations created
by cold working are annealed over a wider temperature range than vacancies and voids, but are
annealed at a higher temperature regime than vacancies and voids, suggesting a possible self-healing
ability to some degree with irradiated 316 SS at elevated temperatures.
316 Stainless Steel also experiences swelling and creep when irradiated. It swells due to helium
bubble formation combined with voids formed from atomic displacement. Using the EBR-II design for
neutron spectrum measurements, it was estimated that there is marginal damage to the lattice below
15 dpa. Between 15 to 40 dpa the lattice begins to increase in size, swelling to around 0.5%. The swelling
continues until it achieves a near linear rate around 60 dpa, rapidly increasing in deformation. This
swelling phenomenon is illustrated in Figure 3.

5. Figure 2 Experimental saturation density of irradiation induced microstructural defects vs temperature. The term ‘black spots’
indicates vacancies or interstitials while ‘cavities’ indicates bubbles or voids formed. Image taken from [8].
Figure 3 Swelling vs. dpa for 20% cold worked 316 SS irradiated in EBR-II and HFIR. Taken from [8] .
Creep experienced due to irradiation of the steel can occur before the component begins to
show signs of swelling. Creep can also be initiated by swelling due to the expansion of a restrained
component. This expansion can cause stress to build up and initiate creep while at operating
temperatures. A comparison of creep at different strain rates with and without radiation is shown in
Figure . Irradiation hardening below 0.5Tm decreases the stress range at which dislocation glide

6. Figure 4 Creep maps for 316 SS with strain rate appear in the upper right corner of each plot. The left plot is un-irradiated and
the right plot is irradiated. Circled is the projected range of operation for the FHR. Taken from [8].
occurs in the irradiated sample, while irradiation softening increases the range at temperatures greater
than 0.5Tm [8]. The difference in the size of Coble and Nabarro-Herring creep between the two samples
is due to differences in strain rate as these creep mechanisms are unaffected by irradiation damage [8].
It is estimated that at temperatures of approximately 650 C a container built of 316 SS will experience
significantly greater thermally activated creep than irradiation creep, therefore thermally activated
creep should take top design priority in regards to the two creep mechanisms. Lastly, the creep to
rupture time was taken into consideration for a 316 SS vessel designed to operate for 40 years. As seen
in Figure , after 100,000 hours (11.4 years) the rupture strength of the material has reduced to about 90
MPa, with an estimated 40-year strength of 20 MPa [8]. This rupture strength should be sufficient as the
estimated TMSR pressure calculated for MSU was 6 MPa.
Figure 5 Creep rupture stress for 316 SS containment vessel at 10,000 and 100,000 hours. Taken from [8].

7. Tellurium production as a fission byproduct can embrittle steels at grain boundaries, increasing
intergranular attack and potentially leading to a point of failure in the containment system. When the
Tellurium concentration reaches a certain solubility limit, NiyTex or CrxTey compounds form. These form
at grain boundaries and induce intergranular fracture [4], [10]. As exposure temperature increases,
room temperature ductility and tensile strength decrease [10]. This likely relates to the diffusion
distance of the tellurium atoms into the grain boundaries. Potentially other fission byproducts will form
compounds with the metal, weakening it further.
Coatings and Additives
With the propensity to cause high mass transfer with chromium, a well-known issue with 316 SS,
many different techniques have been implemented in order to reduce attack. Coatings and changes to
the salt mixture have been attempted in order to increase the life of the containment vessel. Graphite
as a protective shielding is one such approach. Graphite in a FLiNaK salt is very resistant to corrosion,
however it greatly increases the corrosion of the 316 SS that it was trying to shield by as much as two
orders of magnitude [3], [4]. Galvanic corrosion is the reason this occurs, as graphite acts as a cathode
and the container material the anode, increasing the driving force for mass transfer. Graphite can act as
a chromium sink, collecting those atoms that have diffused from the metal. It also acts as an alloying
element, becoming coated in carbides such as Cr7C3 up to 10 μm thick [4].
Additional coating attempts include nickel electroplating, molybdenum thermal spray, a
diamond like coating, and a SiC ceramic. The molybdenum and diamond coatings had spalling issues,
preventing them from successfully protecting the underlying vessel material. The nickel electroplating
proved most promising as it greatly reduced the chromium corrosion caused by mass transfer. The “SiC
and pyrolytic carbon coating showed virtually no signs of corrosion”, however in other composite
samples, regions of pure Si were attacked [4].
One element added to the salt mixture is beryllium. Beryllium helps to bring the salt to a
reducing condition, lessening its attack on 316 SS. This is because Be forms a stable fluoride, BeF2. It was
reported that the addition of Be metal to the salt has the potential to reduce corrosion of 316 SS to less
than 2 μm per year. The Be is selectively oxidized before the elements in the stainless steel are [8] [11].
Beryllium rods can also be used to control the ratio of UF4
to UF3
[4]. In order to take advantage of
beryllium’s passivating effect it would need to be added continuously, or incrementally to maintain the
salt in a reducing state [11].
Alternative Materials
Alternative materials to stainless steel include a carbon fiber reinforced composite (CFRC) and
variations of Hastelloy. The CFRC composites are being considered for applications in gas-cooled
reactors, but face issues with anisotropy. Under irradiation dimensional instability arises as one
direction may densify while another direction swells. This instability can even lead to the CFRC pulling
itself apart. The Carbon Fiber Reinforced Composite material exhibits continuous strengthening over
increasing dpa measurements up to 32 dpa. Unfortunately, this strengthening is accompanied by mass
loss, increased volume, and a reduction in the elastic modulus [4]. If neutron doses are kept below 10

8. dpa the CFRCs could prove to be a viable high temperature engineering material for molten salt
reactors.
Another composite material is a SiC/SiC composite, which also requires low irradiation to be
used. SiC/SiC has desirable strength, elastic modulus, and fracture toughness characteristics which are
unaffected or improved with low level neutron radiation. Under irradiation, cubic SiC undergoes an
isotropic dimensional change that occurs near the temperature of use for a high temperature reactor
(800C). Lastly materials properties such as hardness, elastic modulus and material strength only
experience minor changes across the range of irradiation levels, allowing for stable and predictable
behavior [4].
Hastelloy N developed by ORNL was intentionally designed for the MSRE and thus exhibited
superior corrosion resistance compared to any of the steels at the time. This development came with an
increased alloy cost compared to that of the “off the shelf” steels. Work at the Kurchatov Institute in
Russia during recent years has sought to improve the corrosion properties and radiation resistance in
the containment vessel. The issue of helium embrittlement at grain boundaries was solved by alloying
with fine carbide particles, trapping the helium within the grains and preventing migration to the
boundaries. Telluirum embrittlement was next mitigated by maintaining a reducing environment by
controlling the ratio of UF3
to UF4
with the addition of beryllium metal as discussed above.
Approximately 70 different alloy blends of HN80MT were tested with variations of Al, Cu, Mn, Nb, Re, V,
and W. The significant finding was that with a 2.5 wt% aluminum content and a 0.5 wt% reduction in
titanium, a significant improvement in corrosion resistance and mechanical properties was seen. This
limited the intergranular corrosion and chromium mass transfer that the alloy experienced. This alloy
was also tested in FLiBe at a power density of 10 W/cm3
without exhibiting radiation induced corrosion
[2]. A compilation of different alloying compositions is seen in Figure 6 with corresponding corrosion
tests in Table 1.
The most promising alloys were the HN80MT variant in a FLiBe salt, showing less than 3
micrometers of corrosion per year. Conflicting with previously stated research, HN80MT contained no
aluminum, possibly due to the interaction between aluminum and chromium causing an increase in
dissolution at the specified wt%. The Kurchatov Institute identified the need to shift from alloying
modifications of titanium and other rare earth elements to those that are modified with niobium or
aluminum, where further research is being conducted at ORNL and at the Kurchatov Institute
respectively. In addition to wanting additional high-volume testing of theses alloys at longer hours, it is
again mentioned that methods of design and regulation need to be constructed in order to allow
implementation of the alloy, similar to the ASME code for 316 SS.

9. Figure 6 Alloying composition of Nickel based alloys in wt%. Taken from [2].

10. Table 1 Summary of Kurchatov institute corrosion tests for molten fluoride salts. Taken from [2]
Conclusion
If the correct set of variables between salt content, temperature, and application are met, a 316
Stainless Steel containment vessel would suffice for a molten salt reactor. The primary issue of
chromium dissolution would have to be solved by closely monitoring both beryllium concentrations and
UF3
to UF4
ratios to maintain a highly reducing environment. The clean salt would have to be kept free of
moisture, tellurium, graphite and chromium to prevent corrosion. The vessel would need to take into
account the large decrease in yield strength at lifetimes out to 40 years with a minimal corrosion rate of
2 microns per year. A calculation for thermal neutron flux would need to be made in order to ensure
that the vessel does not receive enough radiation to swell the material beyond design specifications. If
stringent design criteria cannot be met, then exploration of alternative alloys such as the HN80 series or
the implementation of composites should be investigated. From the recent studies on the HN80 series,
it is the recommendation of this author to further develop these alloys in lieu of using 316 Stainless
Steel.

11. Bibliography
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