NAT_MSRmaterials.txt

NEW DEVELOPMENTS IN MATERIALS

 

FOR MOLTEN-SALT REACTORS

H. E. McCOY, R. L. BEATTY, W. H. COOK, R. E. GEHLBACH
C. R. KENNEDY, J. W. KOGER, A. P. LITMAN, C. E. SESSIONS

and J. R. WEIR  Metals and Cervamics Division,
Oak Ridge National Laboratory

Received August 4, 1969
Revised October 10, 1969

 

Opevating experience with the Molten-Salt Re-
actor Experiment (MSRE) has demonstrated the
excellent compatibility of the gvaphite-Hastelloy-

N-fluoride salt system at 650°C. Sewveral im-
provements in wmaterials are needed for a
molten-salt breeder reactor with a basic plant life
of 30 years; specifically: Hastelloy-N with im-
proved vesistance to embrittlement by thermal
neutvons; graphite with better dimensional stq-
bility in a fast neutron flux; gvaphite that is sealed
to obtain a surface permeability of <10~ cm?/sec;
and a secondary coolant that is inexpensive and
has a melting point of ~400°C. A brief description
is given of the materials work in progress to
satisfy each of these requivements.

 

INTRODUCTION

Our present concept of a molten-salt breeder
reactor' utilizes graphite as moderator and re-
flector, Hastelloy-N for the containment vessel
and other metallic parts of the system, and a
liquid fluoride salt containing LiF, BeF,, UF,, and
ThF, as the fertile-fissile medium. The fertile-
fissile salt will leave the reactor vessel at a
temperature of ~700°C and energy will be trans-
ferred to a coolant salt which in turn is used to
produce supercritical steam.

Experience with the Molten-Salt Reactor Ex-
periment (MSRE) has demonstrated the basic
compatibility of the graphite-Hastelloy-N2 -fluo-
ride salt (LiF-BeF,-ZrF,-UF,) system at 650°C.
However, a breeder reactor will impose more
stringent material requirements; namely: the de-
sign life of the basic plant of a breeder is 30
years at a maximum operating temperature of

156 NUCLEAR APPLICATIONS & TECHNOLOGY

Oak Ridge, Tennessee 37830

KEYWORDS: radiation effects,

molten-salt reactors, breeder

reactors, power reactors,
reactor core, fused salts,
coolants, chromium alloys,
molybdenum alloys, nickel al-
loys, corrosion, brittleness,
thermal neutrons, sodium fluo-
rides, sodium borides, mixing,
graphite, fast neutrons, sta-
bility, expansion, porosity,
Hastelloy, embrittlement, mix-
tures, surfaces

700°C; the power density will be higher in a
breeder and will require the core graphite to
sustain higher damaging neutron flux and fluence;
and neutron economy is of utmost importance in
the breeder and the retention of fission products
(particularly **°Xe) by the core graphite must be
minimized. Each of these factors requires a
specific improvement in the behavior of materials.

Experience has shown that the mechanical
properties of Hastelloy-N deteriorate as a result
of thermal-neutron exposure and a method must
be found of improving the mechanical properties
of this material to ensure the desired 30-year
plant life.

Similarly, graphite is damaged by irradiation.
Although the core graphite can be replaced, the
allowable fast neutron fluence for the graphite has
an important influence on the economics of
molten-salt breeder reactors. Thus, a program
has been undertaken to learn more about irradi-
ation damage in graphite and to develop graphites
with improved resistance to damage.

A big factor in neutron economy is reducing the
quantity of '*Xe that resides in the core. This
gas can be removed by continuously sparging the
system with helium bubbles, but the transfer by
this method probably will not be rapid enough to
prevent excessive quantities of '*Xe from being
absorbed by the graphite. This can be prevented
by reducing the surface diffusivity to <102 cm?/
sec, and we feel that this is best accomplished by
carbon impregnation by internal decomposition of
a hydrocarbon.

 

“Hastelloy-N is the trade name of UCC for a nickel-
base alloy containing 169 Mo, "% Cr, 5% Fe, 0.05% C.
This alloy was originally developed at ORNL specifi-
cally for use in molten-salt systems, It has been ap-
proved by the ASME for use in pressure vessels under
code cases 1315 and 1345.

VOL. 8 FEBRUARY 1970
A new secondary coolant is also needed that
will allow us greater latitude in operating temper-
ature. Sodium fluoroborate has reasonable phys-
ical properties for this application, and the
compatibility of Hastelloy-N with this salt is being
evaluated.

Our work in each of these areas will be de-
scribed in some detail.

EXPERIENCE WITH THE MSRE

Other papers in this series have elaborated on
the information gained from the MSRE regarding
operating experience, physics, chemistry, and
fission-product behavior. Additionally, valuable
information has been gained about the materials
involved.?™*

There are surveillance facilities exposed to the
salt in the core of the reactor and outside the
reactor vessel, where the environment is nitrogen
plus ~2% O,. Hastelloy-N tensile rods and sam-
ples of the grade CGB graphite® used in the core
of the MSRE are exposed in the core facility.
The components are assembled so that portions
can be removed in a hot cell, new samples added,
and the assembly returned to the reactor. Sam-
ples were removed after 1100, 4400, and 9000 h of
full-power (~8 MW) operation at 650°C. As shown
in Fig. 1, the physical condition of the graphite
and metal samples was excellent; identification
numbers and machining marks were clearly visi-
ble. The peak fast fluence received by the graph-
ite has been 4.8 x 10*° n/cm® (>50 keV) and the
dimensional changes are <0.1%. Pieces of graph-
ite from the MSRE have been sectioned and most
of the fission products were found to be located on
the surface and within 10 mils below the surface.
However, a few of the fission products have
gaseous precursors and penetrated the graphite to
greater depths. The microstructure of the Has-
telloy-N near the surface was modified to a depth
of ~1 mil, but a similar modification was found in
samples exposed to static nonfissioning salt for an
equivalent time. The near-surface modification
has not been positively identified, but its presence
is likely of no consequence. The very small
changes in the amounts of chromium and iron in
the fuel salt also indicate very low corrosion
rates and support our metallographic observa-
tions.

The observed low corrosion rate of Hastelloy-N
in the MSRE does not come as a surprise, since
thousands of hours of corrosion tests preceded

 

brrade name of Union Carbide Corporation for the
needle-coke graphite used in the MSRE.

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

McCoy et al. MATERIALS

 

Fig. 1.

Graphite and Hastelloy-N surveillance assembly
removed from the core of the MSRE after

72 400 MWh of operation.
salt for 15 300 h at 650°C.

Exposed to flowing

construction of the reactor. Hastelloy-N is nickel
based and contains about 16% Mo, 7% Cr, and 5%
Fe. Under normal operating conditions, the fuel
salt cannot oxidize (form fluorides) any of these
elements except Cr. Since Cr is present in very
small concentrations in the alloy, the corrosion is
limited by the diffusion of Cr to the metal surface.
Corrosion can be reduced even further by con-
trolling the oxidation state of the salt, thus
reducing the rate of the corrosion reaction at the
metal-salt interface. The oxidation state of the
salt in the MSRE is controlled by the addition of
beryllium metal.

Hastelloy-N samples were removed from the
surveillance facility outside the reactor vessel
after 4400 and 9000 h of full-power operation.
This environment is oxidizing, and an oxide film
~92 mils thick was formed on the surface after the
longer exposure. There was no evidence of ni-
triding, and the mechanical propertiies of these
samples were not affected adversely by the pres-
ence of the thin oxide film.

Thus, experience with the MSRE has proved in
service the excellent compatibility of the Hastel-
loy-N-graphite-fluoride salt system.

DEVELOPMENT OF A MODIFIED HASTELLOY-N WITH
IMPROVED RESISTANCE TO IRRADIATION DAMAGE

Since the MSRE was constructed, Hastelloy-N,
as well as most other iron- and nickel-base
alloys, was found to be subject to a type of high-
temperature irradiation damage that reduces the

FEBRUARY 1970 157
McCoy et al. MATERIALS
stress rupture life and the fraction strain.’-'°
This effect is characterized in Figs. 2 and 3 for a
test temperature of 650°c. The rupture lives for
irradiated and unirradiated materials differ most
at high stress levels and are about the same for
below 20 000 psi. The property change of most
concern in reactor design and operation is the
reduction in fracture strain. The postirradiation
fracture strain is shown in Fig. 3 as a function of
strain rate; the scatter band is based upon test
results for three different heats of metal., The
plot includes results from both tensile and creep
tests. In tensile tests the strain rate is a
controlled parameter and the test results are
plotted directly. In creep tests the stress is
controlled and the strain measured as a function
of time. The minimum strain rate was used in
constructing Fig. 3. The data are characterized
by a curve with a minumum at a strain rate of
~0.1%/h, with rapidly increasing fracture strain
as the strain rate is increased, and slowly in-
creasing fracture strain as the strain rate is
decreased. Thus, under normal operating con-
ditions for a reactor where the stress levels (and
the strain rates) are low, the rupture life will not
be affected significantly (Fig. 2), but the fracture
strain will be only 2 to 4% (Fig. 3). However,
transient conditions that would impose higher
stresses or Trequire that the material absorb
thermally induced strains could cause failure of
the material, Therefore, a material is desired
‘that has improved properties in the irradiated
condition and a program with this as its goal has
been embarked upon.©

The changes in high-temperature properties of
iron- and nickel-base alloys during irradiation in
thermal reactors have been shown rather con-
clusively to be related to the thermal fluence and
more specifically related to the quantity of helium
produced in the metal from the thermal '°B(n,0)"Li
transmutation.”™** The mechanical properties
are only affected under test conditions that pro-
duce intergranular fracturing of the material.
Under these conditions both creep and tensile
curves for irradiated and unirradiated materials
are identical up to some strain where the irradi-
ated material fractures and the unirradiated ma-
terial continues to deform. Thus, the main in-
fluence of irradiation is to enhance intergranular
fracture.

 

“Although the fast-neutron flux will be quite high in the
core of our propased breeder, the neutrons reaching
the Hastelloy-N vessel will be reduced in energy by the
graphite present. Thus, the fast fluence seen by the
vessel during 30 years will be <1 x 10%2'n/cm?, and we
do not feel that fast-neutron displacement damage is of
concern. Experiments will be run to confirm this point.

158 NUCLEAR APPLICATIONS & TECHNOLOGY

70

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

\ TTTI
HEAT NO.
60 NG a 5067
N o 5085
] il
N
\\

>0 'KTYPICAL UNIRRADIATED

—~ N

a \\\~ { \\

S 40 T N\

S S~ 3

* ° Iy ™ Th

2 50 P~ al N

g) ?“‘u& \o\

F i\ M
20 Ehas
PRETEST ANNEAL-1h AT 1177°C
10 SOLID POINTS-IRRADIATED < 150°C
OP‘EN POllNT‘S—IITRADIATEDl 500-6|50°(':
100 10t 102 103 104
RUPTURE TIME (h)

Fig. 2. Creep-rupture properties of Hastelloy-N at
650°C after irradiation to a thermal fluence of
~ 5% 10%°n/cm?,

©27.3

022.2

a22.2

14 021.2
13 HEAT NO.
© 5065
12 s 5067

0 5085

PRETEST ANNEAL-1 h AT 4177°C
10 - SOLID POINTS-IRRADIATED < 150°C
OPEN POINTS-IRRADIATED 500-
650°C

FRACTURE STRAIN (%)

     

10~ 100 To} 102 103
STRAIN RATE (% h)

Fig. 3. Fracture strain of Hastelloy-N at 650°C after
irradiation to a thermal fluence of ~5 x 10%°

n/cm?,

VOL. 8 FEBRUARY 1970
A logical solution to this problem would be to
remove the boron from the alloy. However, boron
is present as an impurity in most refractories
used for melting, and the lowest boron concen-
trations obtainable by commercial melting prac-
tice are in the range of 1 to 5 ppm. The low
‘helium levels that have caused the creep-rupture
properties to deteriorate in Hastelloy-N make this
approach very unattractive. For example, in-
reactor tube burst tests at 760°C showed that the
rupture life was reduced by an order of magnitude
and that the fracture strain was only a few tenths
of a per cent when the computed helium levels
were in the parts-per-billion range.'’ Thus, we
have concluded that the properties of Hastelloy-N
cannot be improved solely by reducing the boron
level.

Another very important observation has been
that the properties are altered by irradiation at
elevated temperatures only when the test temper -
ature is high enough for grain boundary deforma-
tion to occur (above about half the absolute
melting temperature for many materials). Thus,
the role of helium must be to alter the properties
of the grain boundaries so that they fracture more
easily.

The size of boron lies intermediate between the
sizes of small atoms such as carbon that occupy
interstitial lattice positions and the larger metal
atoms, such as nickel and iron, that occupy the
normal lattice positions. For this reason boron
concentrates in the grain boundary regions where
the atomic disorder provides holes large enough
to accommodate the boron atoms. Thus, the
transmuted helium will be generated near the
grain boundaries where it will have its most
devastating effects. We reasoned that the addition
of an element that formed stable borides would
result in the boron being concentrated in discrete
precipitates rather than being distributed uni-
formly along the grain boundaries. The trans-
muted helium would likely remain associated with
the precipitate and be less detrimental. Addition-
ally, certain precipitate morphologies and alloying
elements are beneficial in improving the resis-
tance of alloys to intergranular fracture.

Following this reasoning, small additions of Ti,
Hf, and Zr have been made to Hastelloy-N and the
postirradiation properties were improved mark-
edly.'*’*® The titanium-modified alloy was chosen
for development as a structural material for a
molten-salt breeder experiment (MSBE). A fur-
ther modification made in the composition was
reducing the molybdenum content from 16 to 12%.
This change was prompted by the observation that

McCoy et al. MATERIALS
tice was adopted to reduce the concentrations of
other residual elements thought to be deleterious.

The stress-rupture properties at 650°C of
several heats of the titanium-modified alloy (0.5%
Ti) are summarized in Fig. 4. The properties in
the absence of irradiation are improved over
those of standard Hastelloy-N and the rupture life
of the modified alloy is not reduced more than
~10% by irradiation at 650°C to a thermal fluence
of 5% 10%° n/cm® The postirradiation fracture
strains of the titanium-modified alloy are also
improved over those of the standard alloy (Fig. 5).
The modified alloy has a very well-defined duc-
tility minimum as a function of strain rate, but the
minimum strain is ~3% compared with 0.5% for
the standard alloy.

Electron microscopy has shown that the
titanium-modified alloy forms very fine-grain
boundary and matrix precipitates of the MC type
when annealed at 650°C. These precipitates are
only a few tenths of a micron in size and those in
the grain boundaries are spaced at ~2-u intervals.
They have a face-centered cubic crystal structure
with a lattice parameter of ~4,24 A and are likely
complexes involving Mo, Cr, Ti, C, N, and B.
These complex compounds also form precipitates
in the matrix. This microstructure should lead to
trapping of some of the helium as proposed
earlier and should also inhibit fracture along the
grain boundaries. However, further studies have
shown that the precipitates which form during long
exposure at 760°C are relatively coarse MoxC
carbides and that the postirradiation properties of

70

 

 

 

 

 

 

 

 

 

 

 

STD. UNIRRADIATED
\< HEAT NOS.
/7 . o -66-548 (0.5% Ti)
r
60 /\ /;:7\ o -21545 (0.5% Ti)
N PRETEST ANNEAL-I h AT 1177°C
??\? IRRADIATED AT 500-650°C
_ LR
w ~ ‘\ 0
-t
o ~
O N
Q40 \\\ N
w ™1 ¥ é 1 N
[2] ~. q
w g N
3 ™~ N
| g . N
n 30 - .
~ \

STD. |RRADlATED] TN .

20

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

o L 1

1 10 100 1000
RUPTURE LIFE, h

10,000

the additional molybdenum was used in forming  Fig, 4. Creep-rupture properties of several heats of
large carbide particles that made it difficult to modified Hastelloy-N at 650°C. Samples irra-
control the grain size. The vacuum-melting prac- diated to a thermal fluence of ~5x 10%° n/cm?
NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8 FEBRUARY 1970 159
McCoy et al. MATERIALS
HE AT IRRADIATED TEST
TEMPERATURE °C TEMPERATURE °C

© 6252 650 650

® 6252 760 760

© 5911 650 650

® 5911 760 760

a 24545 500-650 650

A 21545 500-650 760

0 66-548 500-650 650

® 66-548 500-650 760
ANNEAL 41 h AT 41477°C BEFORE TESTING

18.11a 516.45 19.8

6 MODIFIED
(05% Ti)

FRACTURE STRAIN (%)

TENSILE

2 STANDARD TESTS

 

10~ 100 10! 102
STRAIN RATE (%/h)

0
10-3 10~2

Variation of fracture strain with rate for sev-
eral Hastelloy-N type alloys. Samples irra-
diated to a fluence of ~5 x10?® n/cm? prior to
testing.

Fig. 5.

material irradiated at 760°C are very poor. Since
some designs require the vessel of the MSBR to
operate at ~700°C, this difference in precipitate
morphology and subsequent deterioration of prop-
erties was of concern.

Further work has shown that the desired MC -
type carbide can be stabilized at higher service
temperatures. This carbide is favored by in-
creasing amounts of Ti, Hf, and Nb and decreasing
concentrations of Si. Several alloys have been

160 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

prepared that retain good postirradiation prop-
erties after irradiation at 760°C. Several of the
alloy additions that appear satisfactory are 1.2%
Ti, 1% Hf, 1% Hf plus 1% Ti, 0.5% Ti plus 2% Nb,
and 2% Hf. The results of tests on these alloys
give encouragement that a commercial alloy can
be developed that has properties at least as good
as those shown in Figs. 4 and 5.

IRRADIATION DAMAGE IN GRAPHITE

Neutron irradiation alters the physical prop-
erties of graphite, but the dimensional changes
that occur'®'” are of major concern. These
dimensional changes are illustrated in Fig. 6
where the data of Henson et al.’ are presented for
isotropic graphite. With increasing fluence the
graphite first contracts and then begins to expand
at a very high rate. Several potential problems
arise as a result of these dimensional changes.
First, the initial contraction will change the
volume occupied by fuel salt and change the
nuclear characteristics of the reactor. These
dimensional changes seem small enough for most
isotropic graphites that the nuclear effects may be
accommodated by design. A second problem is
stress generation due to flux gradients across a
piece of graphite. Graphite creeps under irra-
diation™ and this creep is large enough to reduce
the stress intensities to quite acceptable values.
The third and most serious problem is that the
rapid growth rate represents a rapid decrease in
density with potential crack and void formation.
At some fluence this will cause the mechanical
properties to deteriorate and the permeability to
salt and fission products to increase. We feel that
properties will be acceptable, at least until the
material returns to its original volume, and have

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

defined this fluence as the lifetime. A fourth
10
;\f 5
W 550 - 600° C
2
I
o O
o /
@
L_J N . /
= \ /
3-5 B — 400-440° C-
</
-10 | 5
o 10 20 30 40 50  (x10°")
NEUTRON FLUENCE (n/cm?) (£ >50 kev)
Fig. 6. Volume change in isotropic graphite Dounreay

fast reactor irradiations.

FEBRUARY 1970
problem is that the dimensional changes are
dependent on temperature and the curve in Fig. 6
is shifted up and to the left for increasing
temperature. Thus, stresses develop in a part
having a temperature gradient since segments of
the part are seeking different dimensions. Again,
this stress is relieved by the irradiation-induced
creep in graphite geometries of interest. There-
fore, we consider the onset of rapid growth to be
the primary problem and the initial dimensional
changes of secondary importance.

Graphite temperatures between 550 and 750°C
are anticipated in MSBR’s, and operation with a
fast flux (>50 keV) as high as 1 x 10" n/(cm? sec)
is desired. Data in Fig. 6 indicate that this flux
will cause this particular graphite to expand
rapidly after a fluence of ~3x 10% n/cm® is
reached (~1 year of operation). The flux can be
reduced by decreasing the power density, but this
usually increases the fuel inventory and doubling
time. Hence, it is quite desirable that graphite be
used with better resistance to irradiation damage
than the graphite shown in Fig. 6. The data
available on current reactor graphites irradiated
to high fluences were examined and the results
described a fairly consistent picture. The flu-
ences required for graphite to reach its minimum
volume were strongly temperature dependent (de-
creased with increasing temperature), but were
not appreciably different for any of the graphites
studied to date. Although this observation is

McCoy et al. MATERIALS
discouraging, current experiments show that bet-
ter graphites already exist and that others can
probably be developed with only small changes in
present materials and processing. Let us look
briefly at a simple description of the origin of the
dimensional changes and then return to our spe-
cific observations.

Graphite, after being well graphitized at tem-
peratures above 2000°C, has a hexagonal close-
packed crystal structure consisting of close-packed
layers, (basal planes) of carbon atoms with very
strong covalent bonds within the basal planes (a
direction) and very weak Van der Waals’ forces
between atoms in adjacent basal planes (¢ direc-
tion). This anistropy in atomic density and bond
strength is reflected by very anisotropic prop-
erties.

The changes that take place in a single crystal
of graphite during irradiation are shown schemat-
ically in Fig. 7. A neutron having an energy above
~0.18 eV can displace a carbon atom from a
close-packed basal plane with a reasonable prob-
ability of creating a vacancy in a basal plane and
an interstitial carbon atom between the basal
planes. Repetition of this process and diffusion at
elevated temperatures can result in the formation
of defect clusters, specifically partial planes of
atoms between the basal planes and vacancy
clusters within the original planes. This leads to
an expansion perpendicular to the basal planes (c
direction) and a contraction within the layer

 

 

 

 

 

 

PERFECT SINGLE CRYSTAL
i C
DEFECT —_— * *
PLANES <o
| —-q AL o 4V @
— o — Lo Vo
BASAL a
PLANES -
(a) (b) (c)
POLYCRYSTAL
DEFECT PLANES |
+ + OBSERVED
AV _4av
Vo Vo ,o—

- DENSIFICATION

(d)

 

Fig. 7.

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

PREDICTED

 

@)

Graphite dimensional changes due to irradiation.

FEBRUARY 1970 161
McCoy et al. MATERIALS

planes (a direction) as shown in Fig. 7b. This
new configuration leads to a slight increase in
volume (Fig. Tc).

Polycrystalline graphites are not initially of
theoretical density. The voids present in the
material are a result of shrinkage of the binder
during graphitization and from separation or frac-
ture of layer planes during cooling from the
graphitization temperature. Initially, during irra-
diation the porosity within the material tends to be
filled as the crystallites expand in the ¢ direction.
This produces a densification illustrated by the
bottom curve in Fig. 7d. As the porosity fills, the
shrinkage saturates and the dimensional behavior
begins to be dominated by the volume expansion
due to the growth of the crystallites in thel ¢
direction. Thus, a minimum volume is obtained
as shown in Fig. T7e. During the subsequent
expansion, the material either remains internally
contiguous, in which case the volume change rate
of the polycrystalline material should be similar
to the small rate of expansion exhibited by the
crystallites themselves, or fractures internally
due to the stresses generated between the crys-
tallites of differing orientation (causing a higher
rate of growth to occur). Observations to date
indicate that most graphites increase in volume at
a faster rate at high fluences than expected if the
material remained internally contiguous.

One further consideration helps to explain why
the unpredicted rapid growth takes place. A

schematic representation of several coke par-
ticles and binder after graphitization is shown in
Fig. 8.° Each coke particle consists of several
crystals with a very high degree of preferred
orientation. Although the coke particles are ar-
ranged randomly in a large piece of graphite,
there are still interfaces between particles of
widely different orientations. As each particle
changes dimensions, these interfaces must be
strong and able to shear large amounts without
fracturing. The observation that graphites under -
go large dimensional changes at high fluences
indicates that these interfaces or boundaries are
fracturing. As indicated by the sketch in Fig. 8,
these boundaries are made up largely of the
graphitized binder materials. Thus, the prop-
erties of these boundaries are influenced largely
by the nature of the binder material and its inter-
action with the coke particles.

We are making graphites with known filler and
binder materials, but our work in this area has
not progressed very far. This work also includes
a study of the properties of several commercial
graphites that may be potentially useful for MSBR
applications and others that should give some
basic information about irradiation damage in
graphite. Graphite irradiations have been done at
705 + 10°C in the High Flux Isotope Reactor
(HFIR) where the peak flux (>50 keV) is 1 X
10" n/(cm® sec). Thus, samples can be irradiated
to fluences of 1 x 10** n/em? in ~4 months.

GRAPHITE

Pitch coke crystallites

 

Fig. 8.

162

 

NUCLEAR APPLICATIONS & TECHNOLOGY

Petroleum coke
crystallites

 

 

 

Petroleum /
coke particle

Proposed arrangement of crystallites in graphitized stock.

VOL. 8 FEBRUARY 1970
A summary of the results obtained to date is
shown in Fig. 9. Several materials show a
significant deviation from the ‘‘typical’’ behavior
illustrated in Fig. 6. The POCOd graphites show
excellent resistance to irradiation with very small
dimensional changes out to fluences of 2.5 X
10%2 n/em®. Thus, these data support the suppo-
sition that the typical behavior of graphite can
be improved markedly, but test results have not
been extended to fluences high enough to deter-
mine the exact magnitude of this improvement.

SEALING GRAPHITE

Entry of fuel salt into graphite can be pre-
vented by keeping the entrance diameter of the
accessible porosity smaller than 1 p. Although
this does require some extra care during pro-
cessing it can be accomplished routinely on large
shapes. In fact, the grade CGB graphite obtained
5 years ago for the MSRE satisfies this require-
ment.? However, the graphite structure must be
much more restrictive to prevent gaseous fission
products, particularly '*Xe, from diffusing into
the graphite. We presently propose to strip '*Xe
from the fuel salt by purging with helium. Helium
bubbles will be injected and later removed in a
gas-liquid separator. The efficiency of this
purging depends very heavily on the size of
bubbles that can be injected and circulated and the
mass transfer of ®*Xe from the salt to the helium
bubbles. Both of these factors are presently
uncertain, and we must anticipate that large
quantities of '*Xe will be available to the graphite
surfaces and that excessive (>0.5%) retention of
135x%e will result if this gas can enter the graphite
surface at a high rate. Present calculations show
that the accessibility of '*Xe to the graphite
surfaces will be impeded by a laminar salt film
and that the graphite offers an additional resis-
tance to gas flow only when its diffusivity to '®Xe
is <~10-% cm?/sec.

The best grades of commercial graphites pres-
ently available have bulk diffusivities in the
range of 107! to 10~* cm?/sec, and it is unreason-
able to expect that techniques can be developed for
making massive shapes with such a restrictive
structure. The techniques used for reducing the
porosity of graphite involve multiple impreg-
nations of the material with liquid hydrocarbons
and then firing to graphitize this material. As the
bulk diffusivity decreases, it becomes progres-
sively more difficult for the gases released by the
decomposing impregnants to diffuse out of the
material and the times required to reach the

 

dpOCO Graphite, Inc., Garland, Texas.

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

McCoy et al.  MATERIALS

RY—12-297//
6
/ /\RY-12-31
a /

BY-12 ‘/

/ /
/

—_

o

Q // /
NAT. FLAKE
ATJ-SG A H-315A / / AGOT (BNWL)
- 7/ /
_6 CSF (BNWL)

4425—>\

 

 

 

 

-
N

 

+1)

POCO GRADES

N

/

AV
Vo

 

 

{00In(
S

 

 

 

 

>
=
o

ow

AGOT

I

 

\

 

 

 

 

 

 

 

 

 

0 5 10 15 20 25 30
21
FLUENCE (n/cm2) (£>50keV) (X10°7)

Fig. 9. Volume changes in graphite irradiated at 705°C.

graphitizing temperature become excessive.
Thus, it is more reasonable to reduce the surface
diffusivity by a postfabrication surface-sealing
process involving gaseous impregnation. Since
the pyrolytic carbon, that would be deposited, and
the graphite substrate will change dimensions
differently under irradiation, it is imperative that
the pyrolytic carbon be linked with the substrate
structure and not deposited as a surface layer that
can be sheared easily.

The task of sealing the graphite is illustrated
by the photomicrograph in Fig. 10. The pro-
cessing parameters must be adjusted so that the
voids are filled internally in preference to closing
the voids near the surface and forming a coating.
This can be accomplished by using a flowing
stream of hydrocarbon at low partial pressure and
temperature appropriate to maintain very low
deposition kinetics, but this requires long pro-
cessing times. We have used a different method

FEBRUARY 1970 163
McCoy et al.

MATERIALS

 

—

i~

 

0.005 INCHES
750X

fo

1

 

 

Fig. 10. Photomicrograph of the edge of graphite showing a pore that has been partially coated and then sealed over

with pyrocarbon,

to accomplish penetration which involves pulsing
the sample environment between a rich hydro-
carbon environment and vacuum. The vacuum
cycle removes the reaction products (primarily
hydrogen) and allows more hydrocarbon gas to
enter the void. Specifically, we have used 1, 3
butadiene at 20 psig, deposition temperatures of
800 and 1000°C, and cycle times of ~1 min for the
vacuum and a fraction of a second for the hydro-
carbon. Butadiene was chosen because it is a gas
at room temperature and because of its high
carbon yield per molecule. The temperature
range is restricted to 800 and 1000°C because
higher temperatures result in a surface coating
not penetrating the pore structure and lower
temperatures yield intolerably low deposition
rates. The lengths of the vacuum and pressure
periods are very important because they not only
influence the processing rate, but also the depth
of penetration of the impregnant. The time re-
quired for the process will be important in
determining the cost.

Two commercial graphites have been used in
this work—-viz., AXF made by POCO and ATJ-SG

164

NUCLEAR APPLICATIONS & TECHNOLOGY

made by UCC. These materials had widely dif-
ferent accessible pore spectra; nearly all the
pores in the AXF material were <0.8 p in
diameter while the ATJ-SG grade had pores in all
size ranges up to 17 p. Thus, the sealing charac-
teristics of the two materials were widely dif-
ferent. The results of some parameter studies
are shown in Fig. 11 where the vacuum-hydro-
carbon cycle times were varied at 850°C. The
initial slopes are proportional to the surface area
being coated and the slope is much steeper for the
AXF graphite than for the ATJ-SG material. The
sharp break in the curves for the AXF graphite
indicates that the pores have been filled or closed
off and that the surface area being coated is
reduced. The sharpness of this break attests to
the uniform pore size of the AXF graphite. The
horizontal portions of the curves represent, essen-
tially, surface coating, and the data suggest that
some finite amount of surface coating is neces-
sary to attain the MSBR permeability specifi-
cation of <10-° cm%sec for '*Xe at 700°C. (A
¥°Xe permeability at 700°C of 4 x 10~ cm?/sec is
approximately equal to a helium permeability at

VOL. 8 FEBRUARY 1970
0.10

0.08

0.06

004

MATERIAL: POCO AXF GRAPHITE

0.02

010

0.08

TOTAL WEIGHT GAINED (q)

006

004

 

McCoy et al. MATERIALS

VACUUM (sec) HYDROCARBON (sec)
a 60 Yo
s 60
o 30
e 30
o 15
v 10x10°8cm?2/sec (HELIUM)
v <10°9%m2/sec (HELIUM)

  

5x mm Hg (AIR 2.5x mm

2.5x1072 .

Hg (Al Hg (AIR

75x1072
Hg (AIR)

MATERIAL: ATJ-SG GRAPHITE

0.02

0 4 8 12 16 20

 

24 28 32 36 40 46

TOTAL PROCESSING TIME (h)

Fig. 11, Impregnation rate of graphite using 1, 3 butadiene at 850°C.

25°C of 1 x 1078 cm?/sec.) Helium permeability
measurements were made and the goal was a
helium permeability of <1 x 10~% cm?/sec (denoted
on these curves by a ‘v’ mark). The ATJ-SG
graphite was not sealed to the desired level under
the conditions shown in Fig. 11 and the slope
changes very gradually due to the wide variation
in the pore sizes. |

The data indicate that the processing time
could be reduced by shortening the length of the
vacuum cycle. Another interesting feature of the
process for the AXF graphite was that the final
total weight of carbon deposited was increased by
shortening the vacuum pulse. This indicates that
the depth of penetration of carbon into the ma-
terial was increased. Thus, shortening the vac-
uum pulse accelerated the process and improved
the product, both very desirable characteristics.

These studies are not yet extensive enough to
optimize the deposition conditions but are suffi-
cient to make us optimistic about being able to
reduce the diffusivity of graphite to the desired
level. The remaining question of prime impor-

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

tance is the integrity of the seal after exposure to
high neutron fluences.

CORROSION IN FLUORIDE SALT SYSTEMS

Two decades of corrosion testing”?™®° and ex-
perience with the MSRE®* have demonstrated the
excellent compatibility of Hastelloy-N and graphite
with fluoride salts containing LiF, BeF,, ThF,,
and UF,. The fertile-fissile salt for an MSBR will
contain these same fluorides, so only proof-
testing will be required for the primary reactor
circuit. However, a coolant salt is needed with a
lower melting temperature than the LiF-BeF; salt
presently used in the MSRE; a sodium fluoro-
borate salt (NaBF,-8 mole% NaF) has been chosen
as a potential coolant salt for further study. This
salt is inexpensive (<$0.50/1b) and has a low
melting point of 385°C. A significant characteris-
tic of this salt is that is has an appreciable
equilibrium overpressure of BF; gas (e.g., 180
mm at 600°C).

FEBRUARY 1970 165
McCoy et al. MATERIALS

Much of our present corrosion work is con-
cerned with the compatibility of Hastelloy-N with
sodium fluoroborate. Some earlier thermal con-
vection loop studies involving a relatively impure
salt of composition NaBF,-4 mole% NaF-6 mole%
KBF4 showed that a Croloy-9M loop plugged after
1440 h at a maximum temperature of 607°C and a
temperature difference of 145°C, and that a Has-
telloy-N loop was partially plugged after 8765 h of
operation under the same temperature condi-
tions.® The plug in the Croloy was comprised
primarily of pure iron crystals and the partial
plug in the Hastelloy-N loop was made up of a
compact mass of green single crystals of
NazCrFe. The salt charge from the Hastelloy-N
contained large amounts of Cr, Fe, Ni, and Mo, all
major alloying elements in Hastelloy-N.

In more recent tests a purer fluoroborate salt
of composition NaBF,-8 mole% NaF has been
used.””** A thermal convection loop is being used
from which we can remove salt samples for
chemical analysis and metal samples for weighing
without interrupting operation of the loop. The
weight changes for the hottest and coldest samples
are shown in Fig. 12 for the two loops presently in
operation. The loops are constructed of identical
materials, but the removable samples in one loop
(NCL-13) are standard Hastelloy-N and those in
the other loop (NCL-14) are a modified Hastelloy-
N containing 0.5% Ti and only 0.1% Fe (standard
Hastelloy-N contains 4% Fe). As shown in Fig. 12,
the weight changes of the modified Hastelloy-N
are smaller, and this is later shown to be due
primarily to the lower iron content of the modified
material. The rate of weight change was steady
except for a small perturbation after 1500 h of
operation and a large variation after 4200 h of
operation. These times corresponded to times
when moist air inadvertently came in contact with
the salt. The changes in chemistry shown in Fig.
13 also reflect the admission of air at these times
since the oxygen and water levels in the salt
increased. The iron and chronium concentrations
have continued to increase at a rate proportional
to (time)7?, indicating that the process is con-
trolled by diffusion in the metal. The nickel and
molybdenum concentrations in the salt have re-
mained very low except for times when moist air
was inadvertently contacted with the salt. These
results show qualitatively that the corrosion rates
increased when the oxygen and water levels in-
creased. Capsule tests in which sodium fluoro-
borate containing 1400 ppm O, and 400 ppm H,O
was contacted with Hastelloy-N for 6800 h at 607°C
exhibited very low corrosion rates (<0.1 mil/year).
Future work will be directed toward defining the
oxygen and water levels that result in acceptable
corrosion rates.

166 NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

6 ]
: COLDEST SPECIMENS
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N-NCL-14 °
A STANDARD HASTELLOY \
-0 —  N-NCL-13
| T
2 J
0 { 2 3 4 5 6 (x103)

TIME (h)

Fig. 12, Comparison of the weight changes of Hastelloy-
N specimens inserted in NaBF4NaF (92-8
mole%) thermal convection loops. (Assuming
uniform corrosion, a weight change of 20 mg/
cm? in 8750 h is equivalent to a corrosion rate
of ~1 mil/year.)

The information obtained on the changes in salt
composition and the weight change of samples
located at various points (and temperatures)
around the loop is sufficient to attempt a mass
balance for the system. The weight of metal lost
must equal the weight of metal deposited plus the
weight of metal in the salt-i.e.,

AI/Vloss-—'AI/Vdeposited-*-AI/Vsa.lt ° (1)
A weight change vs temperature profile is con-
structed based on the removable samples and the
assumption is made that each segment of the loop
wall follows this same curve. This procedure
results in mass balances, Eq. (1), that close
within 10%.

Diffusion theory can be used for further anal-
ysis. As mentioned earlier, chemical analyses
(Fig. 13) indicate that the iron and chromium
concentrations in the salt are increasing, so it is
assumed that the salt selectively removes these
elements from the alloy. The modified Hastelloy-
N tested here is relatively free of iron, so the

FEBRUARY 1970
 

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250

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200

 

 

 

 

 

 

CONCENTRATION (ppm)

 

 

 

 

   

 

 

 

 

 

 

 

 

100 ——— Mo - Ni
50 |
o\rw‘r:r O O O [ O [ o
0 1000 2000 3000 4000 5000 6000 7000 8000 9000

TIME (h)

Fig. 13, Variation of impurities with time in NaBFs-
NaF (92-8 mole%), thermal convection loop
NCL-14,

weight loss of this sample should be due primarily
to the removal of chromium. The quantity of
material removed by diffusion under conditions
where the surface concentration of the diffusing
element is zero is given by:

AM =Co VDt (2)

where
A M = material removed, g/cm’
Co = bulk concentration of diffusing species,
g/cm’
D = diffusivity, cm®/sec

t = time, sec.

Using the data of Grimes et al.®* for diffusion of
chromium in Hastelloy-N, this analysis showed
that the quantity of chromium removed by dif-
fusion cannot account for the total weight lost.
This discrepancy can be accounted for by
short-circuit diffusion mechanisms enhancing the
rate of chromium removal at these low tempera-
tures or by impurities (likely HF formed by water
ingestion) that lead to some general attack of the

NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

McCoy et al. MATERIALS
metal that is not diffusion controlled. Two obser-
vations argue against the latter possibility. First,
microprobe analyses have not shown any transfer
of nickel or molybdenum to the colder surfaces of
the loop except during a brief period after 4200 h
of operation in which we knew that large amounts
of impurities were present. A second and more
convincing argument is based on the relative
behavior of standard and modified Hastelloy-N
during the first 4000 h of operation. Rewriting of
Eq. (2) in terms of a reaction rate constant, K,
instead of the diffusion coefficient yields

AM = CoVKI . (3)

This equation predicts the material transported by
diffusion to be too low for the modified alloy when
K equal D, but let K take on a value so that the
predicted and observed weight losses for a given
time of operation agree with Co equal to ™% Cr.
Now consider the standard alloy in which Co
corresponds to 7% Cr + 4% Fe. The same K
chosen for the modified alloy predicts the ob-
served weight change for the standard alloy.
Thus, the difference in the weight losses between
the modified and standard alloy appears due
principally to the iron content. Had much general
corrosion occurred, this adjustment in Co should
not have worked. In fact, this analytical proce-
dure was entirely unsatisfactory for the short
time period after 4200 h when the water and oxygen
levels were high and nickel and molybdenum were
being removed (Figs.. 12 and 13). A further
possible role of impurities is to provide the
oxidizing potential necessary to keep the surface
concentration of iron and chromium at zero.
Thus, even though the process remains diffusion
controlled, the rate can be increased by im-
purities.

Although some of the curves in Figs. 12 and 13
have quite large slopes, the corrosion rates are
not very high. Using the rather rough method of
converting the weight losses to corrosion depths
indicates that the average rate during 8000 h of
operation has been 0.7 mil/year. The rate has
decreased to ~0.3 mil/year for long time periods
in which operation was not disturbed. In scaled-
up MSBR systems, a cold trapping technique will
likely be used to remove some of the corrosion
products so that their solubilities are not ex-
ceeded. We presently feel that the sodium fluoro-
porate salt will provide a satisfactory and
economical secondary coolant for molten-salt re-
actors.

SUMMARY

Experience with the MSRE has proven the basic
compatibility of the graphite—Hastelloy-N-fluoride

FEBRUARY 1970 167
McCoy et al. MATERIALS
salt system at elevated temperatures. However,
-a molten-salt breeder reactor will impose more
stringent operating conditions, and some improve-
ments in the graphite and Hastelloy-N for this
system are needed. The mechanical properties of
Hastelloy-N deteriorate under thermal-neutron
irradiation, but the addition of titanium in combi-
nation with strong carbide formers such as nio-
bium and hafnium makes the alloy more resistant
to this type of irradiation damage. Graphite
undergoes dimensional changes due to exposure to
fast neutrons, and the possible loss of structural
integrity due to these dimensional changes pres-
ently limits the lifetime of the core graphite.
Although the core graphite can be replaced as
often as necessary, these replacements influence
the economics of the reactor, and a program to
find a better graphite has been initiated. Studies
to date indicate that graphites can be developed
that have better resistance to irradiation damage
than conventional nuclear graphites. The graphite
used in the core will be sealed with pyrocarbon to
reduce the amount of '*Xe that is absorbed.
Techniques have been developed for this sealing,
and studies are in progress to determine whether
the low permeability is retained after irradiation.
Corrosion studies indicate that the corrosion rate
of Hastelloy-N in sodium fluoroborate is accep-
table as long as the salt does not contain large
amounts of impurities, such as HF and H,O.

ACKNOWLEDGMENTS

This research was sponsored by the U.S. Atomic
Energy Commission under contract with the Union
Carbide Corporation.

REFERENCES

1. E. S. BETTIS and R. C. ROBERTSON, “The Design
and Performance Features of a Single-Fluid Molten-Salt
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2. W. H. COOK, ‘‘Molten-Salt Reactor Program Semi-
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3. H. E. McCOY, ‘““An Evaluation of the Molten-Salt
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4. H. E. McCOY, ‘““An Evaluation of the Molten Salt
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NUCLEAR APPLICATIONS & TECHNOLOGY

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10. H. E. McCOY and J. R. WEIR, ‘“‘Stress-Rupture
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NUCLEAR APPLICATIONS & TECHNOLOGY VOL. 8

McCoy et al. MATERIALS

TABOADA, and G. M. TOLSON, Fluid Fuel Reaclovs,
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31. J. W. KOGER and A. P. LITMAN, ‘‘Compatibility
of Hastelloy N and Croloy 9M with NaBFs-NaF-KBF4
(90-4-6 mole%) Fluoroborate Salt,”” ORNL-TM-2490,
Oak Ridge National Laboratory, in preparation,

32. J. W. KOGER and A. P. LITMAN, ‘‘Molten-Salt
Reactor Program Semiannual Progress Report, Febru-
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National Laboratory,

33. ‘‘Molten-Salt Reactor Program Semiannual Pro-
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National Laboratory, in press.

34, W. R. GRIMES, G. M. WATSON, J. H. DeVAN, and
R. B. EVANS, in Conference on the Use of Radioiso-
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FEBRUARY 1970 169