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TLR-RES/DE/REB-2024-14: Examining Graphite Degradation in Molten Salt Environments: a Chemical, Physical and Material Analysis

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Technical Letter Report TLR-RES/DE/REB-2024-14 Examining Graphite Degradation in Molten Salt Environments: A Chemical, Physical, and Material Analysis Date:

August 2024 Prepared in response to Task 4c in User Need Request NRR-2022-009, by:

Veerappan Prithivirajan Idaho National Laboratory NRC Project Manager:

Joseph Bass Reactor Engineer Reactor Engineering Branch Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

DISCLAIMER This report was prepared as an account of work sponsored by an agency of the U.S. Government.

Neither the U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product, or process disclosed in this publication, or represents that its use by such third party complies with applicable law.

This report does not contain or imply legally binding requirements. Nor does this report establish or modify any regulatory guidance or positions of the U.S. Nuclear Regulatory Commission and is not binding on the Commission.

INL/RPT-23-75892 Examining Graphite Degradation in Molten Salt Environments: A Chemical, Physical, and Material Analysis August 2024 Veerappan Prithivirajan Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840

DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed 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. References herein to any speci"c commercial product, process, or service by trade name, trade mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not nec-essarily state or re"ect those of the U.S. Government or any agency thereof.

INL/RPT-23-75892 Examining Graphite Degradation in Molten Salt Environments: A Chemical, Physical, and Material Analysis Veerappan Prithivirajan Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 August 2024 Idaho National Laboratory Computational Mechanics and Materials Department Idaho Falls, Idaho 83415 Prepared for the U.S. Department of Energy Of"ce of Nuclear Energy Under DOE Idaho Operations Of"ce Contract DE-AC07-05ID14517

EXECUTIVE

SUMMARY

Molten-salt
reactors
(MSRs)
are
Generation
IV
nuclear
reactors
that
use
liquid
salt
as
a
coolant
and/or

fuel.
In
several
MSR
designs,
graphite
serves
as
a
moderator
and/or
re"ector.
However,
due
to
limited
exper-imental
data
and
operational
experience,
our
understanding
of
graphite
behavior
in
molten
salt
environments

remains
incomplete.

This
report
aims
to
identify
the
degradation
mechanisms
of
nuclear
graphite
in
MSRs,
detail
the
mecha-nisms
of
each
factor,
and
provide
an
initial
assessment
of
their
impact
on
the
structural
integrity
of
graphite

components.
This
assessment
is
based
on
an
extensive
literature
review
and
insights
from
subject
matter

experts.
Furthermore,
given
the
limited
data,
a
modeling
strategy
using
existing
Grizzly
software
is
pro-posed
for
a
more
thorough
analysis
where
appropriate.
Additionally,
it
presents
mitigation
strategies
where

applicable.

The
report
covers
physical
degradation
mechanisms
such
as
in"ltration,
erosion,
and
abrasion,
as
well

as
chemical
degradation
mechanisms
including
"uorination,
intercalation,
corrosion,
and
oxidation.

Molten
salt
can
infiltrate
the
porous
structure
of
graphite,
leading
to
several
detrimental
effects.
Entrap-ment
of
fissile
products
within
the
graphite
pores
can
cause
radiation
damage
and
DPVME
pose
challenges
in

the
handling
and
disposal
of
contaminated
components.
The
differential
thermal
expansion
between
the

in"ltrated
salt
and
graphite,
along
with
internal
stress
from
pressurized
molten
salt
and
volumetric
heating,

can
compromise
the
structural
integrity
of
graphite.
To
mitigate
these
eects,
employing
ultra-"ne
graphite

grades
and
applying
sealants
and
coatings
are
eective
strategies.
A
computational
model
based
on
coupled

solid
mechanics
and
heat
transfer
phenomena
could
be
used
to
predict
the
internal
stresses
using
Grizzly

software.

In
pebble-bed
MSRs,
graphite
fuel
pebbles
can
cause
abrasion
against
reactor
components
due
to
friction

and
wear.
The
severity
of
wear
is
in"uenced
by
various
factors
such
as
temperature,
environment,
and
the

presence
of
lubricants.
Tribological
studies
reveal
that
higher
temperatures
and
molten
salt
environments,

such
as
FLiBe,
signi"cantly
r educe
wear
r ates
c ompared
t o
d ry
c onditions.
A dditionally,
t he
chemical

composition
of
the
salt
can
further
optimize
graphites
tribological
performance.
Long-term
wear
eects

can
be
modeled
by
incorporating
surface
defects
into
the
geometry
and
predict
stresses
under
thermal
and

radiation
eects
using
Grizzly
software.

Chemical
degradation
of
graphite
in
a
molten
salt
environment
can
occur
through
"uorination
and

intercalation.
Fluorination
can
occur
via
replacement
of
hydrogen
or
oxygen
atoms,
or
at
the
active
sites,

but
does
not
cause
structural
degradation.
Intercalation,
on
the
other
hand,
can
lead
to
exfoliation,
where

layers
of
graphite
separate
and
peel
away,
damaging
the
graphite.
Protective
coatings
can
enhance
graphites

resistance
to
intercalation.

Graphite
generally
exhibits
good
chemical
stability
in
molten
salt
environments,
though
it
can
corrode

under
speci"c
conditions,
particularly
in
the
presence
of
impurities
or
o xidants.
Studies
have
shown
that

protective
coatings,
such
as
plasma-sprayed
partially
stabilized
zirconia
(PSZ),
can
eectively
prevent
such

degradation.
Corrosion
behavior
varies
signi"cantly
with
dierent
graphite
grades
and
coating
applications,

underscoring
the
need
for
detailed
studies
on
uncoated
and
coated
graphite
to
understand
and
mitigate

corrosion
mechanisms
in
MSRs.
Research
indicates
that
the
presence
of
oxidants
and
impurities
can

accelerate
graphite
degradation
in
molten
salts,
making
it
essential
to
explore
acceptable
impurity
limits.

Oxidation
is
another
critical
degradation
mechanism,
leading
to
weight
loss
and
structural
damage
due

to
the
formation
of
CO
and
CO2
from
the
reaction
of
carbon
atoms
with
oxygen.
This
process
creates
new

porosity
and
compromises
graphites
integrity.
While
extensive
research
on
graphite
oxidation
has
been

conducted
for
gas-cooled
reactors,
studies
speci"c
to
MSRs
are
l imited.
Findings
from
the
coal
industry i

suggest
that
molten
alkali
metal
salts
can
accelerate
graphite
oxidation,
a
hypothesis
worth
exploring

for
fluoride
salts
in
MSRs.
Understanding
oxidation
behavior
in
MSRs
is
vital
for
developing

protective
measures.

The
analysis
of
post-irradiated
graphite
from
the
MSRE
experiment
demonstrated
exceptional
chemical

compatibility
with
molten
fluoride
salt,
suggesting
that
the
extent
of
chemical
attack
on
graphite

largely
depends
on
the
salts
infiltration
capability.
Therefore,
the
use
of
ultra-fine
grade
graphite

could
help
mitigate
chemical
degradation
effects.
Existing
oxidation
modeling
capabilities
in
Grizzly,

which
model
graphite-air
interactions
using
reaction-diffusion
equations,
cPVME
CF
BEBQUFE
UP
TJNVMBUF
UIF

DIFNJDBM
EFHSBEBUJPO
FGGFDUT
PG
HSBQIJUF
JO
NPMUFO
TBMU
FOWJSPONFOUT

ii

ACKNOWLEDGMENTS The
author
thanks
the
following
persons
for
their
active
participation
in
discussions,
thoughtful
questions,

and
valuable
feedbackall
of
which
significantly
contributed
to
this
report:
U.S.
Nuclear
Regulatory

Commission
(NRC)
staff,
Drs.
Joseph
Bass,
Raj
Iyengar,
as
well
as
Matthew
Gordon,
and
Alexander

Chereskin,
along
with
Idaho
National
Laboratory
(INL)
colleagues
Drs.
William
Windes
and
Benjamin

Spencer.

This
manuscript
was
authored
by
Battelle
Energy
Alliance,
LLC
under
contract
no.
DE-AC07-05ID14517

with
DOE.
The
U.S.
Government
retains
a
nonexclusive,
paid-up,
irrevocable,
worldwide
license
to
publish

or
reproduce
the
published
form
of
this
manuscript,
or
allow
others
to
do
so,
for
U.S.
Government
purposes.

iii

CONTENTS 1

3 3

3 10 13 14 15 16 17 18 20 iv

1 INTRODUCTION Molten-salt reactors (MSRs) are a type of Generation IV nuclear reactor that utilizes liquid salt as a coolant. The liquid salt is typically composed of "uoride or chloride salts, and the nuclear fuel can be either dissolved in the salt or used in the form of solid particles such as pebbles. Table presents some of the MSR systems currently being developed, highlighting the type of salt used, the moderator material utilized, and other relevant information.

MSRs oer several advantages over light-water reactors (LWRs). For instance, they can operate at higher temperatures, approximately 700°C, resulting in improved electricity generation eciency and greater process heat opportunities. Moreover, MSRs operate at near-atmospheric pressures compared to LWRs. This eliminates the need for large, expensive reactor pressure vessels and secondary containment structures, and could minimize the risk of major coolant leaks during accidents, thus improving overall reactor safety.

Liquid-fuel-based MSRs do not require fuel assembly production reducing the need for fuel fabrication.

Another advantage of MSRs is their ability to refuel online, meaning fuel can be added to the reactor while it is still running at full power. LWRs, on the other hand, require shutdowns to move or introduce new fuel.

Online refueling capabilities improve the capacity factor and economics of MSRs.

Table 1: Selected MSR systems under development along with fuel and moderator information No.

Company Country Type Fuel Salt Moderator Potential application(s)

Ref.

1 Thorcon USA Liquid reactor Uranium NaF-BeF2-UF4 Graphite Energy

[,

]

2 Flibe Energy USA LFTR Uranium LiF-BeF2-UF4 Graphite Energy, shipping,nuclear medicine

[ ]

3 Terrestrial Energy Canada IMSR Uranium Molten Fluoride +

Uranium Graphite Energy, ammonia, green Hydrogen

[ ]

4 Kairos Power USA KP-FHR TRISO fuel pebbles FLiBe Graphite (ET-10)

Energy

[ ]

5 Terra Power USA Fast reactor Uranium NaCl-MgCl2 Energy

[,

]

6 Moltex Canada,UK SSR-U Uranium/Sodium Fluoride eutectic salts Graphite (Not in contact with the fuel)

Energy

[,

]

7 Moltex Canada,UK SSR-W Higher actinides from conventional oxide fuel Chloride fuel salt

[ ]

8 Copenhagen atomics Denmark SMR Thorium F7LiThPu Unpressurized heavy water Energy

[

]

9 Exodys Energy (Elysium)

USA MCSFR DU, LEU, SNF, RGPu, WGPu, Th, Unat Chloride fuel salt Energy

[

]

10 Abilene Christian University (ACU)

USA MSR Uranium LiF-BeF2-UF4 Graphite (ultra"ne grade)

Research reactor

[

]

The ratio of the net electricity generated (over the time considered) to the energy that could have been generated at continuous full-power operation during the same period.

RGPu (Reactor-Grade Plutonium), WGPu (Weapons-Grade Plutonium), and Unat (Natural Uranium).

1

MSRs oer many other advantages in addition to the ones mentioned above; however, they also carry certain limitations. One of their most notable drawbacks is the potential for radioactive "ssion products to deposit on components throughout the primary coolant system, potentially resulting in issues such as material degradation stemming from direct contact with the molten fuel salt. In MSRs that use lithium-based salts, tritium could be produced, which is both radioactive and mobile. Furthermore, remote maintenance becomes necessary due to components that are contaminated with highly radioactive "ssion products in the reactor, leading to higher operating costs.

This report speci"cally focuses on the degradation aspects of nuclear graphite material in a molten salt environment. Nuclear graphite is commonly used as a moderator and a re"ector in MSR systems. In this report, we present details on various potential degradation mechanismsinformation gathered from the literature and via consultations with subject matter experts. Figure concisely summarizes the potential degradation mechanisms, incorporating details from reference [ ] along with additional information. There are two main categories of degradation mechanisms: physical and chemical. In"ltration and abrasion/erosion are categorized as physical degradation mechanisms. The chemical category includes processes such as "uorination, intercalation, corrosion, and oxidation. A detailed review of each of the above mechanisms is provided in later sections of this report. Furthermore, the existing modeling approaches for a few critical graphite degradation mechanisms are brie"y described, and potential modeling approaches are proposed where none exists.

  

 

















 












 



































 


 






  






















 








 




 








 



 



 






 











 


 




 

 







Figure 1: Summary of potential degradation mechanisms for graphite in a molten salt environment.

2

2 DEGRADATION MECHANISMS This section explores the mechanisms that can lead to the degradation of nuclear graphite in molten salt environments. These mechanisms are categorized as either physical or chemical in nature.

2.1 Physical Factors Within the realm of physical degradation, two signi"cant mechanisms are observed: in"ltration and erosion/abrasion.

This subsection delves into the speci"cs of each mechanism, outlining their nature, potential impact on graphite, and modeling methodologies.

2.1.1 In"ltration Molten salt directly contacts the moderator material in most MSR designs. Nuclear graphite, which is commonly used as a moderator, has a porous microstructure due to its manufacturing process. Graphite pores that are interconnected in a network that opens to the surface are known as open porosity, while others that are not connected to the surface are referred to as closed porosity. Some common nuclear graphite grades, along with the related microstructure and porosity information, are provided in Table Table 2: Dierent graphite grades, along with microstructure and porosity information [

]

No.

Grade Type Grain size Bulk density Open pores volume Porosity Pore dia (m)

(g.cm3)

(cm3.g1)

(%)

(m) 1 POCO ZXF-5Q micro"ne 1

1.8 0.083 20 0.5 2

POCO AXF-5Q ultra"ne 5

1.73 0.102 23 0.9 3

POCO TM super"ne 10 1.73 0.102 23 2

4 IG-110 super"ne 10 1.76 0.079 21 3.9 5

2114 super"ne 13 1.81 0.071 19 3.5 6

ETU-10 super"ne 15 1.74 0.098 22 3.6 7

NBG-25 "ne 60 1.81 0.068 19 5.1 8

CGB medium-"ne 1.86 0.003 17 0.2 9

PGX medium-"ne 460 1.76 0.055 22 5.6,30 10 NBG-17 medium-"ne 800 1.85 0.042 17 3,12,51 11 PCEA medium-"ne 800 1.77 0.065 21 64 12 NBG-18 medium-coarse 1600 1.86 0.044 17 12 As per ASTM D8075-16; No information was found on CGB; ORNL data from physical measurements according to ASTM C559-163; Calculated as per the volume-mass-density relation: 0 = (

1

1

); Calculated as per the equation 100%; Calculated using mercury porosimetry.

Utilizing data compiled from various sources (as presented in reference [

]), it is inferred that most molten salts display a non-wetting behavior when in contact with graphite, characterized by a contact angle that exceeds 90°. Consequently, such non-wetting liquids do not spontaneously in"ltrate the graphite pores but require a minimum pressure dierential () for in"ltration, as described by the Young-Laplace equation:

= 4cos

(1)

It should be noted that the characteristic pore diameters of NBG-17 exhibit a tri-modal distribution.

3

 


   












 



 
 

Figure 2: Schematic of microstructural features aecting the "ow of molten salt.

It is essential to note that the magnitude of depends on the surface tension (), contact angle (), and diameter of the pore (). Equation predicts whether a given salt will in"ltrate the open pores in graphite.

However, once the salt in"ltrates the pores, its "ow characteristics (e.g., velocity and stoppage) also depend on the features of the graphites internal microstructure. Figure provides a 2D schematic of important features of the graphites internal microstructure that can in"uence the "ow characteristics. It should be noted that these are simpli"ed representations of the graphite microstructures and only intended to aid in this discussion of salt "ow mechanisms. In Figure

, the gray colored region represents the solid skeleton of the graphite material, and the blue colored region represents a porous channel within the graphite material that is completely "lled with the molten salt. Four dierent features are identi"ed that can in"uence the "ow characteristics. They are as follows:

• Connectivity: In a microstructure, a porous channel may not extend entirely through the material, potentially terminating within the microstructure itself. In this scenario, "ow stops once an open channel ends inside the microstructure.

• Tortuosity: Due to gravity and curvature eects, tortuosity aects the "ow.

• Surface roughness: Due to friction, surface roughness reduces the "ow velocity.

§Surface tension is highly dependent on the composition and purity of the material [

]. Therefore, chemistry control is a crucial factor to consider in conjunction with salt in"ltration.

¶Interested readers can refer to reference [

] for an understanding of the actual 3D graphite microstructures obtained through X-ray computed tomography.

4

• Morphology: Changes in pore morphology aect the "ow. For example, as shown in Figure (d), if the pore constricts, the "ow may decrease in velocity or otherwise stop altogether, as higher pressures are required to drive the "ow, as per Equation.

In"ltration curves can be either obtained via mercury porosimetry or through direct salt intrusion studies

[

]. Figure gives a plot of mercury intrusion in various graphite grades, displaying the cumulative intrusion volume of mercury as a function of mercury pressure. Notably, higher pressures are required to in"ltrate the smaller pores, as implied by Equation

, which indicates an inverse relationship between the pressure dierential and the pore diameter. Moreover, this relationship is evident in Figure

. For example, when comparing the plots of PCEA and ZXF-5Q, it becomes apparent that PCEA, with a median pore diameter of 64 m, initiates in"ltration at approximately 1 psia, whereas ZXF-5Q, having a pore diameter of 0.5 m, begins in"ltration at around 300 psia. Also, note that at very high pressures, the internal microstructure can be damaged as indicated inside the red box as depcited in Figure. However, such high pressures (greater than 10000 psia) are unlikely in the normal operation of MSRs.

Figure 3: Mercury intrusion studies and the average pore diameters of various graphite grades [

].

Based on Equation, we can predict molten salt in"ltration curves based on the results from mercury intrusion experiments, with the following equation [

]:

=

cos cos (2) where,, and cos represent the pressure dierential, surface tension, and wetting angle of mercury on graphite, respectively. Similarly,,, and cos depict the same quantities for salt with salt in"ltration.

Mercury intrusion porosimetry is supported by international standards and can be rapidly performed in specialized laboratories. Thus, this technique is highly convenient for those engaged in developing, testing, and manufacturing new graphite grades for MSRs, especially during the selection and production phases.

However, note that the reliability of the correlation between the mercury intrusion data and the molten salt in"ltration behavior depends on the accuracy of the molten salt data with regard to surface tension and contact angle [

]. Caution should be exercised in interpreting these correlations, as they have not yet been fully validated.

5

Multiple research groups have conducted direct salt in"ltration studies, employing various graphite grades and quantifying the weight change and volume of in"ltrated salt [

].

Figure shows a compilation of plots of cumulative salt-in"ltrated volume against the absolute salt pressure for select graphite grades. Both FLiNaK and FLiBe salts were used to generate the data in the "gure. In the Molten-Salt Reactor Experiment (MSRE), 0.5 vol.%

salt in"ltration into graphite was the design goal, and 4 vol.% and above was deemed unacceptable for reactor control [

]. If we compare the 4 vol.% criteria to experimental in"ltration results shown in Figure

, it is evident that of the select grades only ZXF-5Q, UGG-2, NG-CT-50, and G2 would have experienced acceptable levels of in"ltration [

].

Figure 4: In"ltrated salt volume as a function of absolute salt pressure for select graphite grades, replotted from the "gure provided in reference [

].

Molten salt in"ltration into graphite pores can result in the following degradation eects:

(i) Entrapment of "ssion products Molten salt can carry "ssionable materials and "ssion products into the graphite pores. Fission and radioactive decay of those materials can in"ict radiation damage on the graphite. Additionally, these processes can act as local heat sources and alter the heat transfer and mechanical deformation of graphite. Furthermore, the entrapped "ssion products can pose signi"cant challenges in the handling and disposal of contaminated components.

(ii) Internal stress Internal stresses in MSR graphite components can stem from several factors, including the coecient of thermal expansion (CTE) mismatch between salt and graphite, pressurized molten salt in the pores, volumetric heating (see point [i]), and crystalline pressure. H. Zhoutang et al. conducted experiments Salt intrusion studies are conducted using ASTM std. 8091-16.

Volume percent is with respect to the total volume of the graphite specimen [

].

6

to measure the bulk thermal expansion of both pristine graphite and graphite in"ltrated with salt across a range of dierent temperatures [

]. The resulting thermal expansion behavior is illustrated in Figure

. Similar studies were also reported in references [

]. Close examination of Figure reveals that the thermal expansion of pristine graphite follows a linear trend. However, in the case of graphite in"ltrated with salt, a distinct bilinear behavior is observed, as characterized by a noticeable kink that occurs precisely at the melting point of the FLiNaK salt (i.e., 450°C). This behavior arises from the signi"cant dierence in CTE between solidi"ed salt (3.8e-5 K1) and graphite (4.5e-6 K1),

with salt expansion being an order of magnitude greater below its melting point [

]. Consequently, below this temperature, the thermal expansion curve for in"ltrated graphite is steeper than for pristine graphite. Above the melting point, the curves slope aligns with that of pristine graphite as a result of additional salt melting and leaking from the pores, bringing the CTE of in"ltrated graphite closer to that of pristine graphite [

].

Thus, before reaching the melting point, the dierence in CTE between graphite and molten salt can generate thermal stresses in the graphite material [

]. Furthermore, W. Qi et al. discuss the crystallization pressure, with crystallization of FLiNaK in the graphite pore creating pressure against the pore wall as a result of thermal dynamics [

]. This pressure is corroborated by a decrease in the d(002) spacing of the graphite, as measured using x-ray diraction [

].

During a freeze-thaw cycle that could occur in the circumstance that the reactor core is drained, cooled, and then reheated, the thermal stresses and crystalline pressure can potentially induce internal stresses.

In contrast to the above discussion, which suggests that thermal cycling can in"ict damage on salt-in"ltrated graphite, studies conducted during the MSRE era indicate that thermal cycles do not impact the integrity of in"ltrated graphite. According to reference [

], samples of CGB graphite, a "ne-grade graphite with sub-micron pore size, were in"ltrated with MSRE fuel salt at a gauge pressure of 1 kPa, and subjected to 100 freeze-thaw cycles ranging from 200°C to 700°C. No damage was indicated when analyzing the quantity and size of graphite cracks following the cycling process [

]. This disparity is likely attributable to the microstructure of the CGB graphite, which resulted in minimal salt in"ltration under those speci"c conditions.

Figure 5: Comparison of the thermal expansion of graphite before and after molten salt in"ltration [

]

(Reproduced with permission from Elsevier).

Crystallization pressure refers to the pressure exerted on the walls of a porous material due to the crystallization of an overcooled melt (such as salt) within its pores.

7

(iii) Decrease of compressive strength Figure 6: In"uence of salt in"ltration on the compressive strength of (a) IG-110 and (b) NG-CT-10 graphite grades, and tensile strength of (c) IG-110 and (d) NG-CT-10 graphite grades (Replotted from reference [

]).

C. Zhang et al. analyzed the eects of molten salt in"ltration on tensile and compressive strength, and investigated the mechanism of failure under compressive loads [

]. Their study, which utilized FLiNaK as the salt, examined two graphite grades: IG-110 and NG-CT-10. The graphite samples were degassed before being immersed in the molten salt. The in"ltration pressure, controlled by the pressure of the argon blanket gas, varied from 450 kPa to 1000 kPa. Once the salt was heated to 700°C and melted, the graphite samples were immersed and subjected to dierent pressures for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> to allow for in"ltration. Before the molten salt solidi"ed, the graphite samples were removed to separate them from the salt. The pressure vessel was then cooled to room temperature while maintaining the gas overpressure. To minimize graphite oxidation during the experiment, high-purity argon (99.999%)

was used.The weight change after in"ltration was determined by weighing the graphite in a dry (H2O

< 1 ppm), low oxygen (O2 < 1 ppm) argon atmosphere glove box.

Mechanical testing was conducted using a High-Temperature Universal Testing Machine (HTUTM).

The samples, which had been in"ltrated with molten salt, were removed from the glove box and transferred to the HTUTM in a hermetically sealed bag.

To minimize graphite oxidation, high-purity argon (99.999%) was used throughout the experiment.Before testing, the high-temperature chamber was evacuated and then "lled with argon. The samples were subsequently heated at a rate of approximately 10°C/min to 700°C and maintained at this temperature for about 60 minutes. Following this, the samples were subjected to uniaxial tension and compression. The results for compressive strength and tensile strength, as a function of in"ltration pressure, are presented in Figures (a) and (b), respectively.

From Figure

, the following observations can be made: (a) the compressive strength of IG-110 decreased by approximately 20% and the compressive strength of NG-CT-10 decreased by approx-imately 10% in comparison to the non-in"ltrated samples, and (b) the tensile strength was reduced by approximately 8% for both grades in comparison to the non-in"ltrated sample. Furthermore, the 8

Figure 7: Images of reassembled major fragments and schematic representations of failure mechanisms observed during uniaxial compression testing: (a) virgin IG-110, (b) virgin NG-CT-10, (c) schematic of shear fracture, (d) molten in"ltrated IG-110, (e) molten in"ltrated NG-CT-10, and (f) schematic of longitudinal splitting fracture [

] (Reproduced with permission from Elsevier).

authors analyzed the fracture types of the above graphite grades during compression failure. Figure shows images of reassembled major fragments and provides schematic representations of the failure mechanisms observed during uniaxial compression testing. It also illustrates the transition of the fracture mechanism from shear fracture in non-in"ltrated graphite samples to longitudinal splitting fracture in in"ltrated samples.

All the above-mentioned degradation aspects relate to the graphites structural integrity, which is poten-tially compromised by pressurized molten salt in the pores of the graphite, volumetric heating of the molten salt (if fuel is also dissolved) in those pores, the dierence in the coecient of thermal expansion (CTE) between the salt and graphite, and the reduced strength of the graphite. A computational model based on coupled solid mechanics and heat transfer phenomena could be used to predict the internal stresses due to in"ltration eects using Grizzly software [

]. Furthermore, the eects of irradiation could also be added to assess the graphite components performance in an integrated manner using Grizzly software.

Based on the preceding discussion, it becomes evident that utilization of ultra"ne graphite grades with micron-sized pores is a potential strategy for minimizing salt in"ltration. Another avenue explored in the literature involves the application of sealings and coatings. This includes pyrolytic carbon (PyC), glassy carbon (GC), and silicon carbide (SiC) coatings, as well as composite approaches combining multiple barrier types. However, delamination of coating is a major concern due to the potential mismatch in thermally-or irradiation-induced dimensional changes between the coating and graphite. As an alternative to coatings, researchers are exploring pore-"lling techniques (e.g., "xed-bed deposition and resin impregnation) to block open pores in graphite materials [

].

9

Multiple studies have shown the eectiveness of coatings and pore-"lling techniques in preventing in-

"ltration. In investigating the use of isotropic PyC as a barrier coating for IG-110, He et al. observed a signi"cant reduction (23-fold) in FLiNaK in"ltration as compared to uncoated graphite [

]. Similar "ndings were documented by Song et al. in their research on rough laminar PyC [

]. Bernadet et al. concluded that employing a dual coating layer comprised of a PyC interlayer and a GC top layer resulted in minimal in"ltration [

]. Additionally, He et al. reported that graphite coated with SiC exhibited less than 1.1%

weight gain in FLiNaK after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exposure at 650°C [

]. Lastly, He et al. showed that phenolic resin impregnation completely prevented in"ltration into IG-110 graphite [

].

2.1.2 Erosion and Abrasion In pebble-bed molten salt reactors, resinated graphite fuel pebbles can rub against each other and the systems structural components due to coolant circulation and cycling of pebbles through the reactor. This occurs within both the molten salt environment and in areas where the pebbles lack salt coverage. The literature oers two comprehensive studies on the tribology of graphite-graphite and graphite-stainless steel (SS), as covered in the subsections below.

(i) Tribology of Graphite-Graphite In their study, Vergari et al. conducted tribological experiments to investigate the friction and wear characteristics of dierent types of graphite, both in a dry argon environment and a wet FLiBe (molten salt) environment [

].

In the "rst set of studies, experiments were conducted in an argon environment at two dierent tem-peratures: room temperature (RT) and 600°C (HT). This study focuses on ET-10 nuclear graphite.

Tribological measurements were carried out using a pin-on-disk tribometer, and various characteriza-tion techniques were employed to analyze wear spots. Before conducting the tests, the spheres (located at the tip of the pins) and the discs were degassed at 600°C to eliminate any adsorbed oxygen and moisture.

The results reveal a signi"cant contrast between the RT and HT tests. The RT tests showed higher average coecients of friction (COFs)at 0.55whereas the HT conditions resulted in a 1.5-fold reduction in COF, for an average of 0.33. Speci"c wear rates exhibited a similar trend, with RT wear rates being 0.4 ug/Nm, signi"cantly higher than the 0.06 ug/Nm observed in the HT tests, marking a 6.7-fold reduction. Pro"lometry analysis identi"ed the presence of a tribo-"lm (self-lubricating "lm) whose thickness was 10 microns for RT and 50 microns for HT wear spots, respectively.

The authors provide the following points to explain the discrepancy between RT and HT: At HT, the results indicate a more stable and mechanically resilient tribo-"lm, characterized by increased thickness and continuity. This enhanced "lm at HT provides improved lubrication, reducing the COF and wear in contrast to RT. The existence of temperature hysteresis further supports the argument, showing that the "lm maintains its eectiveness at intermediate temperatures before "nally degrading at RT.

In the second set of experiments, ET-10 and IG-11 graphite grades were tested in a FLiBe environment.

These experiments employed a ball-on-three-plates con"guration within an argon glovebox, with a rotating shaft coming into contact with the three graphite plates. Prior to testing, all samples underwent a degassing process for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 600°C to remove absorbed moisture and oxygen.

Tests were performed at four dierent temperatures, ranging from 500°C to 600°C. The COFs and normalized wear rates (NWRs) were determined. Benchmark tests were also conducted in argon (dry sliding).

10

Results at 600°C indicated that the average COF values were approximately 1.3-1.55 times lower in a FLiBe environment than in an argon one, and the NWR was about 1.5-2 times lower in FLiBe than in argon. The temperature-dependent tests suggested the COF to be independent of temperature, whereas the NWR increased with temperature. The impact of salt chemistry was assessed by adding 1 wt%

Cr2, 1 wt% AgF, and 1 wt% Be. The results showed signi"cant changes in both NWR and COF.

The test involving Cr2 resulted in dry sliding due to the lack of salt on the contact points, and this test is not discussed in this report. However, a lower COF and NWR were observed for the other two compositions, with Be having the lowest values. In comparison to FLiBe, the addition of Be resulted in average COFs and NWRs that were 4 and 2 times lower, respectively.

By applying the Hamrock and Dowson equation [

], the authors calculated the thickness of the FLiBe "lm at the contact interface and compared it to the composite roughness. This analysis revealed that sliding in FLiBe occurs in a boundary lubrication regime. Boundary lubrication is known to reduce adhesion forces between surfaces but is typically too thin to prevent mechanical abrasion caused by surface asperities. This suggests that FLiBe is particularly eective at reducing friction and wear, potentially more so than the formation of a graphite "lm.

The observed behavior in tests at dierent temperatures supports the hypothesis that FLiBe physically interposes itself between mating surfaces, thereby reducing adhesion. Additionally, there is evidence of chemical interactions between FLiBe and graphite, resulting in the formation of covalent C-F bonds at crystallite edges. This chemical reaction hypothesis is supported by the COF behavior during consecutive runs at dierent temperatures, showing hysteresis and irreversible changes in COF values.

Furthermore, the research suggests that salt chemical reactivity can be in"uenced by the addition of metal "uorides, which impacts the salts redox potential and subsequently its lubricating properties.

The study also identi"es a dierent lubrication mechanismpotentially related to salt wetting or the precipitation of Be metalthat may be at play when Be is added to FLiBe. Further investigations, including contact angle measurements, energy dispersive spectroscopy (EDS), and x-ray photoelectron spectroscopy (XPS) analysis, are needed to explore these mechanisms in greater detail.

(ii) Tribology of Graphite-Stainless Steel In reference [

], lab-scale experiments were performed to understand the tribocorrosion behavior of graphite sliding against FLiNaK-lubricated 316H SS in an argon environment. The tribological tests were performed using a pin-on-disk con"gurationwith a graphite pin and a 316H SS diskon a high-temperature tribometer equipped with an environmental control chamber. Three factors temperature, sliding velocity, and salt volume (i.e., amount of lubrication)were of primary interest in regard to impacting tribological behavior. The below paragraphs detail the role of each parameter.

(i) Role of temperature Studies were performed at two dierent temperatures (i.e., 550°C and 650°C), with the other two parameters (i.e., sliding velocity and salt volume) being held constant. Figure displays the obtained measurements of the COF and the wear of the graphite pin and SS "at. The outcomes indicate that the levels of friction and wear are lower at 550°C. The COF ranges between 0.1 and 0.15 at 550°C, and varies from 0.1 to 0.25 at 650°C. At 550°C, the pin and "at wear volumes were 0.012 and 0.009 mm3, respectively. At 650°C, the wear volume doubled and the "at volume seemingly tripled. The authors hypothesize that three factors are behind the additional friction and wear at 650°C:

• Insucient lubrication at 650°C, attributed to a 40% decrease in viscosity.

11

Figure 8: Role of temperature on the (a) COF, (b) wear of graphite pins, and (c) wear of SS "ats [

]

(Reproduced with permission from Elsevier).

• Enhanced corrosion of SS at 650°C in comparison to 550°C. In contrast, graphite is relatively chemically inert.

• Softening of SS at higher temperatures (ii) Role of sliding velocity Material testing was performed at three dierent sliding velocities (1, 10, and 100 mm/s), with the other two parameters being held constant. It is important to note that these tests were conducted over a total duration of 10000 seconds, resulting in varying sliding distances for each test case.

The results for the COF and the wear of the graphite pin and SS "at are shown in Figure

. COFs of 0.15, 0.15, and 0.1 were obtained for sliding velocities of 1, 10, and 100 mm/s, respectively.

This is potentially due to the thicker lubrication at higher sliding velocities, per the Hamrock and Dowson formula [

]. The wear volume for both the graphite pin and SS "at is higher at 100 mm/s due to there being a larger sliding distance; however, the wear per unit distance is smaller.

The graphite wear is higher than that of the SS "at wear at higher velocities, due to graphites brittle behavior leading to more micro-fracture wear caused by vibration-induced impact.

(iii) Role of salt volume To study the eect of salt volume on lubrication performance, three tests were carried out: (a) salt-less sliding (no salt, dry condition), (b) "ooded lubrication (18 g of salt, fully submerged contact interface), and (c) starved lubrication (~6 g salt, partially submerged contact interface).

The COF and wear characteristics from this study are given in Figure

. The steady-state COF for all the salt quantities are within a range of 0.1-0.25. The dry condition led to material transfer from the graphite to the SS "ats, whereas both lubricated cases experienced wear in the graphite pin as well as the SS "at. The starved lubrication resulted in more wear than the other cases, 12

Figure 9: Role of sliding velocity on the (a) COF, (b) wear of graphite pins, and (c) wear of SS "ats [

]

(Reproduced with permission from Elsevier).

hypothetically due to the molten salt failing to provide eective lubrication and thus interfering with graphite deposition. As a result, a stable boundary lubrication or transfer deposit "lm could not be established, leading to poor wear performance.

For MSRs in which the nuclear fuel is dissolved in the salt, there is a possibility of erosion due to rolling contact wear that could occur as a result of precipitated particles carried by the molten salt from other areas of the reactor. However, it is important to note that no evidence of abrasion or erosion was found in the post-irradiation examination of graphite from the MSRE [

]. Therefore, more experimental evidence is needed to conclusively determine whether erosion/abrasion can be a limiting factor in MSRs containing fuel dissolved in the salt.

Phenomena such as abrasion and erosion cause wear to the graphite, leading to material loss and the formation of surface defects over the components lifetime. These surface defects can be modeled using a diuse interface approach and coupled with mechanical, thermal, and irradiation eects to understand the stress concentrations caused by wear. Grizzly software can be used for this integrated analysis, providing insights into the impact of these defects on the components performance.

2.2 Chemical factors In the domain of chemical degradation aecting nuclear graphite in molten salt environments, four major mechanisms play a key role: "uorination, intercalation, corrosion, and oxidation. This section explains each mechanism with their potential impact on the graphite material.

13

Figure 10: Role of salt volume on the (a) COF, (b) wear of graphite pins, and (c) wear of SS "ats [

]

(Reproduced with permission from Elsevier).

2.2.1 Fluorination Fluorination of graphite in molten salt environments is a possibility. A few possible mechanisms for this, as provided in the literature, are as follows:

(i) Fluorination through hydrocarbon and oxygen functional structure In nuclear graphite, C-H bonds are partly residual following graphitization of the raw materials. In a study by Yang et al. [

] on IG-110 grade nuclear graphite, CK-edge x-ray absorption near-edge structure revealed that the C-H bonds were uniformly distributed throughout the sample, and indicated that immersing the graphite in molten "uoride salts at 500°C for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> resulted in the partial replacement of existing C-H bonds with C-F bonds. The chemical reactions of "uoride salts with graphite are generally thermodynamically unfavorable, so graphite is expected to be inert. To address this discrepancy, the authors of [

] hypothesized that it is uncertain whether degassing graphite at temperatures exceeding 800°C in order to remove chemisorbed hydrogen would result in "uorination.

As gas composition measurements were not taken, it cannot be ruled out that the "uorination reported by Yang et al. [

] was caused by H2O being present in their vacuum oven during the experiment. In the presence of LiF(g) and NaF(g) vapors, H2O could have produced hydrogen "uoride (HF), which may have been responsible for the observed "uorination of the graphite. Furthermore, XPS results led the authors [

] to deduce that "uorination occurred as a result of replacing oxygen atoms in their study with graphite in FLiBe. The graphite sample initially had 11.1% relative peak area as oxygenated groups, which decreased to 0-7.9% after exposure to molten salt.

(ii) Fluorination at active sites (edge sites) 14

Figure 11: Possible reactive carbon sites leading to C-F bond formation upon FLiBe exposure [

]

(Reproduced with permission from Elsevier).

Using XPS, Wu et al. observed "uorination of IG-110 graphite when immersed in molten FLiBe at 700°C for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> [

]. They argued that certain carbon sites are more reactive than others, and that the "uorination may have occurred on those reactive sites (Figure

). Analysis of the XRD and Raman data also revealed changes in the graphites structure. These "ndings led them to propose that "uorination could introduce additional defects and active sites, potentially enhancing the chemisorp-tion of tritium. This mechanism, however, is not known to cause any structural degradation of graphite.

(iii) Intercalation Fluorination can also happen through intercalation, which is a mechanism discussed in detail in the following subsection.

2.2.2 Intercalation Graphite intercalation compounds (GICs) can form when atoms or molecules (intercalants) are inserted between the layers of graphite, potentially leading to structural damage through exfoliation. Wu et al. sug-gested the possibility of "uorine GICs, based on their XPS analysis of molten FLiBe immersed in IG-110 graphite at 700°C for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> [

]. Another study, this time led by Bernardet et al., involved experiments on nuclear graphite in powdered and disc forms, both in their raw state and with various coatings. These dierent graphite samples were impregnated with a Li-, Na-, and Zr-based "uoride molten salt at 500°C for 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> [

]. The results showed that, post-treatment, the c lattice parameter had increased from 6.758 to 6.771 Å for both the powdered and raw disc graphite, with a much smaller increase in the coated disc samples. However, in all cases, the increase in the lattice parameter was deemed too insigni"cant to con"rm the formation of a "uorine-intercalated compound. Previous studies reported that the interlayer spacing for stage-1 "uorine-intercalated compounds typically falls within 9.38-9.44 Å; for stage-2, it is approximately 12.7 Å [

]. In their study, Sure et al. observed GIC formation when conducting experiments on uncoated These are the same sites that are also susceptible to oxidation [

]. Potentially, "uorination and oxidation might be competing reactions. However, more studies need to be performed to fully understand this aspect.

15

high-density (HD) graphite in a molten LiCl-KCl eutectic salt at 600°C for up to 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. In contrast, when the graphite was coated with plasma-sprayed partially stabilized zirconia (PSZ), no intercalation com-pounds were observed [

]. Based on the available evidence, it is reasonable to conclude that coated graphite exhibits greater resistance to GIC formation than does uncoated graphite. The variation between the results found in references [

] and [

] may be attributable to the dierent graphite grades and exposure times.

Furthermore, Takashima and Watanabe reported that, under speci"c conditions, graphite could react with a mixture of "uorine and hydrogen "uoride gases to form GICs [

].

2.2.3 Corrosion The corrosion behavior of materials in the presence of molten salts at high temperatures has been extensively studied. Multiple research studies have revealed graphite to be chemically stable in molten salt environments, whereas a few other studies have shown that graphite can exhibit corrosion under certain conditions.

Multiple research articles have reported on the corrosion behavior of high-temperature structural alloys in the presence of molten "uoride salts at high temperatures. Some of these experiments were conducted using graphite crucibles; others employed crucibles made of materials such as nickel. Corrosion was noted to occur predominantly via the de-alloying of chromium (Cr) [

]. Weight loss due to corrosion was generally correlated with the initial Cr content of the alloys, and was consistent with the Cr content measured in the salts following the corrosion tests. Furthermore, the use of graphite crucibles accelerated the de-alloying of Cr and its deposition onto graphite, whereas this phenomenon was not very pronounced when other crucibles (e.g., Nickel) were used [

]. It must be highlighted that, while galvanic corrosion can aect the structural integrity of alloys near graphite, it does not signi"cantly impact the structural integrity of the graphite material itself.

However, corrosion was observed in graphite, as reported in references [

]. Sure et al. conducted a detailed study on the corrosion behavior of HD graphite and plasma-sprayed PSZ-coated HD graphite in a molten LiCl-KCl eutectic salt environment at 600°C over varying time periods [

]. Uncoated HD graphite exhibited weight loss due to corrosion. The molten salt corrosion mechanism in graphite involves three key processes: the formation of intercalation compounds, the adherence and diusion of the salt into the graphite, and the "lling of surface porosity with molten salt. This process begins when the molten salt penetrates through the pores of the graphite, interacts with the carbon atoms on the surface, and dislodges the atoms from its lattice. In contrast, PSZ-coated graphite showed weight gain due to salt deposits on the surface, but demonstrated excellent corrosion resistance, with no observable degradation on the coatings surface. This indicates the eectiveness of PSZ coatings in protecting graphite against a chemical attack by molten salt environments. Sure et al. pointed out that several oxidants and impurities in the molten salts also accelerate the degradation of HD graphite in LiCl-KCl salt by dissolution and leaching of the elements.

This makes investigating the acceptable limits of impurities in graphite and molten salt a promising avenue for future research.

Additionally, Kamali et al. investigated microstructural changes in graphite by heating a mixture of synthetic polycrystalline graphite and lithium chloride to 1250°C [

]. Dierent forms of corrosion attack on the graphite occurred, leading to the formation of dierent microstructures comprised of exfoliated carbon sheets and nanosheets, pitted particles, and carbon nanorods. It is important to note that 1250°C is beyond the operating temperature range of MSRs. In another study, Vacik et al. investigated the interaction of molten FLiNaK with GC, pyrographite, and reactor-grade graphite at 540°C for up to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Researchers found that reactor-grade graphite completely dissolved in FLiNaK after 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, GC underwent a 0.5% weight 16

loss, and pyrographite remained completely stable [

]. However, the speci"c grade of reactor graphite used in the study is not speci"ed, and no other research group has reported such an observation, leaving the exact cause of its complete dissolution unclear. Hence, further experimental research on molten "uoride salt corrosion of graphite is essential to elucidate the chemical stability of graphite against "uoride salts at various temperatures.

2.2.4 Oxidation At elevated operating temperatures, graphite is susceptible to oxidation from impurities in the coolant or graphite itself. Oxygen has a high anity for unsaturated carbon atoms present at the active edge sites, such as armchair or zig-zag con"gurations. It is at these locations that oxygen chemisorption occurs [

].

The primary consequence of oxidation in regard to graphite is the weight loss that results from the formation of CO and CO2 when carbon atoms react with oxygen. This gasi"cation process leads to the formation of new porosity and networks of pores, making previously sealed porosity accessible. The extent of weight loss, changes in microstructure, and thickness of the region where oxidation occurs depend on graphite grade and temperature [

].

Extensive research has been conducted on graphite oxidation in gas-cooled [

] and high-temperature gas-cooled [

] reactors. Nevertheless, studies on graphite oxidation in regard to MSRs have not gained widespread attention, causing uncertainty over whether molten salt environments would accelerate or decelerate the oxidation process.

The coal industry has, however, studied graphite oxidation in molten sodium carbonates and sulfates.

This body of this research reveals that sodium alkali metal molten salts exhibit a catalytic eect on coal gasi"cation, causing accelerated graphite oxidation when the graphite is immersed in a salt environment [

]. For example, at 900°C, graphite oxidation rates have been shown to surge by approximately 40-fold when the graphite is placed in a lithium carbonate salt environment [

]. While evidence for "uoride-salt-mediated oxidation remains insucient, it represents an avenue worth exploring in future studies.

It is crucial to acknowledge that the analysis of post-irradiated graphite from the MSRE experiment demon-strated exceptional chemical compatibility between the molten "uoride salt and the graphite moderator [

].

The corrosion mechanism, as described by Sure et al. [

], suggests that the extent of chemical attack on graphite by molten salt largely depends on the salts ability to in"ltrate the graphite structure. Given that the MSRE used CGB grade graphite, known for its submicron characteristic pore diameter and consequently low in"ltration rates, the observed compatibility is consistent with expected behavior. Therefore, using ultra-"ne grade graphite could help mitigate chemical degradation eects. The existing oxidation modeling capabilities in Grizzly, which simulate graphite-air interactions using reaction-diusion equations, could be modi"ed to model chemical degradation eects.

17

3

SUMMARY

Degradation of nuclear graphite in molten salt environments is a critical consideration when designing and operating MSRs. This report has examined various degradation mechanisms, including both physical and chemical factors, to better understand the potential challenges associated with maintaining the structural integrity of graphite components in MSRs.

Among the physical factors, salt in"ltration into graphite pores has been identi"ed as one of the potential degradation mechanisms. The in"ltration of molten salt into these open pores can lead to the entrapment of "ssion products, potentially causing radiation damage and raising safety concerns regarding the handling and disposal of contaminated graphite components. Moreover, the pressurization of pores from molten salt in"ltration, the dierential thermal expansion between the salt and graphite, and volumetric heating may induce internal stress, compromising the structural integrity of the graphite. Ideally, a nuclear graphite material that prevents in"ltration would be preferred. Though complete prevention might pose as a challenge under dierent operating conditions, studies indicate that employing ultra-"ne-grade graphite and applying sealants and coatings are eective strategies to minimize salt in"ltration. A computational model capable of predicting salt in"ltration based on graphite microstructure would signi"cantly aid in rapidly assessing a materials susceptibility to in"ltration. Additionally, experiments evaluating the durability of sealants and coatings under various temperatures and radiation levels would be bene"cial. The current body of literature does not provide a de"nitive answer on whether the aforementioned factors will induce internal stress or to what extent. Therefore, a computational model to calculate internal stress based on coupled solid mechanics and heat transfer in Grizzly software is proposed to predict the internal stresses resulting from these factors.

Erosion and abrasion represent additional physical factors that can lead to the structural degradation of graphite components. This degradation often results from the interaction between solid particles, such as dust or fuel pebbles, and the surfaces of graphite components as they circulate within the system.

Tribocorrosion experiments have detailed the impact of these interactions, revealing a complex relationship between material properties, environmental conditions, and lubrication. Notably, tribological studies on graphite have shown that wear and friction are most pronounced at lower temperatures under dry conditions.

Conversely, higher temperature environments with FLiBe, a molten salt, signi"cantly reduce friction and wear rates, highlighting the protective role of FLiBe against the more severe eects observed in dry sliding conditions. Furthermore, the chemistry of the salt was found to aect the tribological performance of graphite, with the addition of metal "uorides resulting in an enhanced performance. This suggests a promising avenue for optimizing the tribological behavior of graphite through careful manipulation of chemical additives in the lubricant. To further advance our understanding and capability in this area, a multi-faceted research approach would be bene"cial. This includes comprehensive tribological testing across a broad spectrum of conditionsencompassing various temperatures, environments, and salt compositions. Through this comprehensive investigation, wear rates could be quanti"ed under dierent conditions.

Additionally, a computational model could be developed by modeling long-term wear eects as surface defects, and Grizzly software could be used to evaluate stresses due to combined thermal and irradiation eects.

Investigations into chemical degradation mechanisms aecting graphite, including "uorination, inter-calation, corrosion, and oxidation, reveal varied impacts. While "uorination does not structurally degrade graphite, intercalation may cause damage through exfoliation. The application of protective coatings has shown promise in enhancing graphites resistance to these processes. Corrosion and oxidation, both exacer-bated by impurities, lead to signi"cant weight loss and structural integrity loss. Studies on graphite oxidation in molten sodium carbonates and sulfates suggest that alkali molten salts accelerate graphite oxidation.

However, there is no available research on "uoride-salt-mediated oxidation and it would be bene"cial to conduct such experiments. Given these insights, it is bene"cial to identify appropriate protective coatings for 18

graphite
to
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19

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[1] D. Wilson, ThorCon: Status 2022.

[2]

[3]

[4]

[5] Graphite material quali"cation for the kairos power "uoride salt-cooled high-temperature reactor.

[6] B. M. Chisholm, Development of TerraPowers Molten Chloride Fast Reactor (MCFR) to en-able low-cost, economy-wide decarbonization.

, 2021.

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[8] I. Scott, Stable salt reactors.

, 2020.

[9] Stable Salt Reactor - Uranium (SSR-U).

, 2022.

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[11] Molten chloride salt fast reactor.

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[13] Abilene christian university molten salt research reactor preliminary safety analysis report.

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