ORNL-TM-2511.txt

 

h - —— — — — — = — —— — _ —

CENTRAL R
ESEARCH LI
DOCUMENT COLLECTIBC?P? i

OAK RIDGE NATIONAL LABORATORY
operated by
UNION CARBIDE CORPORATION m
NUCLEAR DIVISION
for the
U.S. ATOMIC ENERGY COMMISSION

[ g

   

)

-

| 445k 05

 

1] MATERIALS FOR MOLTEN-SALT REACTORS

| .

| H. E. McCoy W. H. Cook R. E. Gehlbach
J. R. Weir, Jr. C. R. Kennedy C. E. Sessions
R. L. Beatty A. P. Litman J. W. Koger

 

|

o

[s

| NOTICE This document contains information of a preliminary nature
and was preparec primarily for internal use at the Oak Ridge Mational
Laboratory. It is subject to revision or correction and therefore does
not represent a final report.

 
1

 

 

 

— LEGAL NOTICE

 

This repart wos prepared os an account of Government sponsared work. Neither the United States,

nor the Commission, nor any persen octing on behalf of the Commission:

A. Mokes ony waorranty or representation, expressed or implied, with respect to the occurocy,
completeness, or usefulness of the informotion contained in this report, or that the use of
any information, apparatus, method, or process disclosed in this report moy nel infringe
privately owned rights; or

B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of
any information, opparatus, method, or process disclosed in this report.

As used in the above, ‘person acting on behalf of the Commission" includes any employee or

contractor of the Commission, or employee of such contracter, to the extent that such employee

or contractor of the Commission, or emplovee of such contractor prepares, disseminates, or
pravides access fo, any infermation pursuant ta his employment or contract with the Commission,
or his employment with such contractor.

 

 

 
ORNL-TM-2511

Contract No. W-7405-eng-26

METALS AND CERAMICS DIVISION

MATERTALS FOR MOLTEN-SALT REACTORS

H. E. McCoy W. H. Cook R. E. Gehlbach
J. R. Weir, Jr. C. R. Kennedy C. E. Sessions
R. L. Beatty A. P. Litman J. W. Koger

Paper submitted to the Journal of Nuclear Applications
MAY 1969

OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee
operated by
UNION CARBIDE CORPORATION
for the
- U.S. ATOMIC ENERGY COMMISSION ' - -

| LOCKHEED MARTIN ENERGY RESEARCH LIBRARIES
t

| 3 445k 0513k7kL 9

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 
iii

CONTENTS

Abstract . . . . . . . . .. 00 .

Introduction .

Experience With The MSRE

Development of A Modified Hastelloy N With Improved Resistance

to Irradiation Damage . . . . . . . .
Irradiation Damage in Graphite . . . . .
SUMMATY v v ¢ & o o o o o o o o o o o o o o o « o« o o«

References . .

14
35
36
MATERIALS FOR MOLTEN-SALT REACTCRS

H. .E. McCoy W. H. Cook R. E. Gehlbach

J. R. Weir, Jr. C. R. Kennedy C. E. Sessions

R. L. Beatty A. P. Litman J. W. Koger
ABSTRACT

Operating experience with the Molten-Salt Reactor
Experiment (MSRE) has demonstrated the excellent compat-
ibility of the graphite-Hastelloy N-fluoride salt system
at 650°C. Several improvements in materials are needed
for a molten-salt breeder reactor with a basic plant life
of 30 years; specifically, (1) Hastelloy N with improved
resistance to embrittlement by thermal neutrons,

(2) graphite with better dimensional stability in a fast
neutron flux, (3) graphite that is sealed to obtain a
surface permeability of less than 10”8 cm?/sec, and (4) a
secondary coolant that is inexpensive and has a melting
point of about 400°C. A brief description is given of
the materials work in progress to satisfy each of these
requirements. Our studies presently indicate that
significant improvements can likely be accomplished in
each area.

 

INTRODUCTION

Our presenf concept of & molten-salt breeder reactor (described
in detail by Bettis et 5;.) utilizes graphite as moderator and reflector,
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 about 700°C and energy will be trans-
ferred to a coolant salt which in turn is used to produce supercritical
steam.

Our experience with the Molten-Salt Reactor Experiment (MSRE) has

demonstrated the basic compatibility of the graphite-Hastelloy N-fluoride
salt (LiF-BeF,-ZrF,-UF,) system at 650°C. However, a breeder reactor
will impose more stringent material requirements; namely, (1) the design
life of the basic plant of a breeder is 30 years at a maximum operating
temperature of 700°C, (2) the power density will be higher in a breeder
and will require the core graphite to sustain higher damaging neutron
flux and fluence, and (3) neutron economy is of utmost importance in

the breeder and the retention of fission products {particularly *°Xe)
by the core graphite must be minimized. ZEach of these factors requires
a specific improvement in the behavior of materials.

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

Similarly, we have found that graphite 1s damaged by irradiation.
Although we can replace the core graphite, we find that the allowable
fast neutron fluence for the graphite has an important infliuence on the
economics of our reactor. Thus, we have undertaken a program to learn
more about irradiation damage in graphite and to develop graphites with
improved resistance to damage.

A big factor in neutron economy is reducing the quantity of 135y%¢
that resides in the core. We will remove this gas 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
135Ye from being absorbed by the graphite. This can be prevented by
reducing the surface diffusivity to < 1078 cm®/sec, and we feel that
this is best accomplished by carbon impregnation by internal decomposi-

tion of a hydrocarbon.

 
We are also searching for a new secondary coolant that will allow
us greater latitude in operating temperature. Sodium fluoroborate has
reasonable physical properties for this application, and we'ére
evaluating the compatibility of Hastelloy N with this salt.

We shall describe our work in each of these areas in some detail.

EXPERIENCE WITH THE MSRE

 

Other papers in this series have elaborated on the information
ga;ned from the MSRE regarding operating experience, physics, chemistry,
and fission-product behavior. Additionally, valuable information has
been gained about the materials involved.(l_B)

We haye'a surveillance facility exposed to the salt in the core
of the reactor and one outside the reactér vessel, where the environment
is nitrogen plus about 2% 0,. Hastelloy N tensile rods and samples 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 sémples added, and returned to the reactor.
We have removed samples after llOO,I44OO, and 9100 hr 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 cledrly visible. The peak fast fluence
received by the graphite has been 4.8 x 1029 neutrons/cm® (> 50 keV)
and the dimensional changes are less than 0.1%. Pieces of graphite

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.

 

*Trade name of Union Carbide Corporation for the needle-coke graphite
used in the MSRE.
W R-42962

et

  

 

Fig. 1. Graphite and Hastelloy N Surveillance Assembly Removed from
the Core of the MSRE After 72,400 Mvhr of Operation. Exposed to flowing
salt for 15,300 hr at 650°C.

 
However, a few of the fission products have gaseous precursors and
penetrated the graphite to greater depths. The microstructure of the
Hastelloy N near the surface was modified to a depth of about 0.001 in.,
but we found a similar modification.in samples exposed to static non-
fissioning salt for an equivalent time. We have not positively
identified the near-surface modification, but find its presenée 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 observations.

Hastelloy N samples were removed from the surveillance facility
outside the reactor vessel after 4400 and 9100 hr of full-power operation.
This enviromment is oxidizing, and we have found that an oxide film
about 0.002 in. thick was formed on the surface after the longer exposure.
There was no evidence of nitriding, and the mechanical properties of
these samples were not affected adversely by the presence of the thin
oxide film.

Thus, our experience with the MSRE has proven in service the excel-
lent compatibility of the Hastelloy N-graphite-fluoride salt system.

DEVELOPMENT OF A MODIFIED HASTELLOY N WITH IMPROVED RESISTANCE
TO TIRRADIATION DAMAGE

 

 

Since the MSRE was constructed, we have found that Hastelloy N, as
well as most otfier iron- and nickel-base alloys, is subject to a type of
high-temperature irradiation damage that reduces the creep-rupture
strength and the fracture strain.(4_9) 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
ORNL-DWG 68-4200

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

70
N HEAT NOS.
N o -5065
\ A-5067
' o - 5085
60 . N
' "~ PRETEST ANNEAL-IHR AT 1177°C
\\ SOLID POINTS-IRRADIATED <150°C
50 N OPEN POINTS-IRRADIATED 500—
. 650° G
* AN
o N§f . \<TYP UNIRRADIATED
o N
O \
4 °
7 N N
8 o Te | \\
E qe \\o o |d N
wn 30 o S -
N
o o \#\ \
20 4By |
10
0
N I 10 100 1,000 10,000

RUPTURE TIME, HR

Fig. 2. Creep-Rupture Properties of Hastelloy N at 650°C After Irradiation to a
Thermal Fluence of About 5 X 10°% neutrons/cm?.
ORNL-DWG 68-4199R2

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

273
i
14 )
/
HEAT NOS. ° [
3 0 - 5063 1
A - 5067 ¥
o -5085 ; / ¢
12 /
PRETEST ANNEAL-IHR AT1177°C . / ‘
SOLID POINTS-IRRADIATED < 150°C /
OPEN POINTS-IRRADIATED 500-650°C / o
1 ; 4
’V
/
10 ;
g/ /
&/ /
I L
9
| /
y
» / 1/
-8
/
: 17
o
-
= , /
&7 / /
P / /
o o
=
L /
6
/Y
A LA
/]
5 7
/L
4 , /]
4 ,, /
'\ // / ~—TENSILE TESTS
3\ / /
Q\ A / //
|
¢ \\l o / / /]
2 ’
.\:o N )j //
N B
\. o EN / //
b\\ | p
I g T -
Al f{"""‘%- cr/
) I
001 o I 1 10 100 1000

MINIMUM CREEP RATE, %/HR

Fig. 3. Fracture Strain of Hastelloy N at 65g°C After Irradiation
to a Thermal Fluence of About 5 X 10°C neutrons/em®.
and converge for stresses below 20,000 psi. The property change that
concerns us most in reactor design and operation is the reduction in
fracture strain. The postirradiation fracture strain is shown in Fig. 3
as & function of strain rate. TIn tensile tests the strain rate 1is a
controlled parameter, and for the creep tests we have used the minimum
creep rate. The data are characterized by a curve with a minimum at a
strain rate of about 0.1%/hr, with rapidly increasing fracture strain
as the strain rate is increased, and slowly increasing fracture strain
as the strain rate is decreased. Thus, under normal operating conditions
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 condi-
tions that would impose higher stresses or require that the material
absorb thermally induced strains could cause failure of the material.
Therefore, it is desirable that we have a material with improved proper-
ties in the irradiated condition and have embarked upon a program with
this as our goal.*

The changes in high-temperature properties of iron- and nickel-

base alloys during irradiation in thermal reactors have been shown

 

*¥Although the fast neutron flux will be quite high in the core of
our proposed 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 less than
1 X 102! neutrons/cm?, and we do not feel that fast-neutron displace-
ment damage is of concern. Experiments will be run to confirm this
point.
rather conclusively 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,a)”Li transmutation.(lo_l2) The mechanical
properties are only affected under test conditions that produce inter-
granular 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 irradiated material fractures and
the unirradiated material continues to deform. Thus, the main influence
of irradiation is to enhance intergranular fracture.

A logical cure for 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 concentrations obtainable by
commercial melting practice are in the range of 1 to 5 ppm. This
approach seems even more hopeless when we look carefully at the low
helium levels that have caused the creep-rupture properties to deterio-

. rate in Hastelloy N. For example, we found in some in-reactor tube
burst tests at 760°C 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.(g) 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 tem-.

perature is high enough for grain boundary deformation to occur
10

(above about half the absolute melting temperature for meny 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 devas-
tating effects. We reasoned that the addition of an element that formed
stable borides would fesult in the boron being concentrated in discrete
precipitates rather than being distributed uniformly along the grain
boundaries. The transmuted helium would likely remain associated with
the precipitate and be less detrimental. Additionally, certain precipi-
tate morphologies and allbying elements are beneficial in improving the
resistance of alloys to intergranular fracture.

Following this reasoning we have made small additions of Ti, HT,
and Zr to Hastelloy N and have found the postirradiation properties to

be improved markedly.(13s14)

We have chosen the titanium-modified alloy
for development as & structural material for a molten-salt breeder
experiment (MSBE). A further modification made in the composition was
reducing the molybdenum content from 16 to 12%. This change was prompted
by the observation that the additional molybdenum was used in forming'
large carbide particles that made it difficult to control the grain size.

We also adopted the vacuum-melting practice to reduce the concentrations

of other residual elements thought to be deleterious.

 
11

The stress-rupture properties at 650°C of several heats of the
titanium-modified alloy (0.5% Ti) are summarized in Fig. 4. The proper-
ties 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 about 10% by irradiation at 650°C to a thermal fluence of
5 x 10%° neutrons/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 & very well-defined ductility
minimum as a function of strain rate, but the minimum strain is about
3% compared with O.?% for the standard alloy.

Electron microscopy has shown that the titanium-modified alloy
forms very fine-grain boundary precipitates when annealed at 650°C.
These precipitates are only a few tenths of a micron in size and are
spaced along the grain boundaries at about 2-u intervals. They have a
face-centered cubic crystal structure with a lattice parameter of‘about
4.24 R and are likely complexes involving Mo, Cr, Ti, C, N, and B. This
microstructure should lead to trapping of the helium as we 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 MopC carbides and
thatlthe postirradiation properties of material irradiated at 760°C are
very poor. Since we planned to operate the MSBR at about 700°C, this
difference in precipitate morphology and subsgquent deterioration of
properties was of utmost concern to us. Further studies have shown that
the desirable structu;e can be stabilized by the addition of Nb, Si, or
Hf, and we are confident of having properties at least as good as those

shown in Figs. 4 and 5.
 

70

60

O
O

STRESS, l?oo PS|
O

W
QO

20

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

ORNL-DWG 68-4202
STD. UNIRRADIATED |
\4 HEAT NOS.
Y o -66-548 (0.5% Ti)
SN
4 //V o -21545 (0.5% Ti)
NG
“C/ PRETEST ANNEAL-I HR AT 1177°C
?%3 IRRADIATED AT 500-650°C
A//E%’?\
- I \4\4 /‘2 ]
i ~. % \<<
\‘ ~
\‘L\L.\ ] }
e adil
~ N
~I Qé//
I~ N
I §
~ L N
1~ \
STD. IRRADIATED] i N\
N
| | 10 100 1000 10,000

RUPTURE LIFE, HR
Fig. 4. Creep-Rupture Properties of Several Heats of Modified Hastelloy N at 650°C. Samples

irradiated to a thermal fluence of about 5 x 102° neutrons /em?.

¢t
13

ORNL-DWG 68-4203

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

181t A16.45 D19.8
14 b4 {RRAD, TEMP., | TEST TEMP !
SYMBOL | HEAT e o
o 6252 €50 650
A ® le252 760 760
13 o Bal 650 650
& [sa 760 760
a losas 500-650 650
A 21545 500-650 760 !
12 D |e6-548| 500-6%0 650
m  [|66-548| 500-630 760
A ANNEAL | HR AT 1177°C BEFORE TESTING o
i
A
10 ! \ 4
at \ ] ;
9
= | .
3 ' i
21545
+
& A \ /
- \ O /
5
3 N ? /
7 66-548
-
\( )
N
. A \ 1!
o\ \ 117
\l \ 650°C / o
5 / /
0
\\ \\ 760°C /
"N \p / /
4 m N
/
¢ \ // /
& \—’/ d’/‘ /
3 A Tt /{'650"(:
K\\ F STANDARD //
2 /)
. \\ o o /
o N o | ¢ .
~L < ‘
W 760°¢ | ||| >
| ;\\ a8 &
< < \\ ® ,/ __.-—'//
{’ T"h- ::_,4—
0

 

 

 

 

 

 

 

 

 

 

 

.00 0t A . | 10 100
MINIMUM CREEP RATE, %/HR
Fig. 5. Variation of Fracture Strain with Strain Rate for Several
Hastelloy N Type Alloys. Samples irradiated to a fluence of about
5 x 10?° neutrons/cm?® prior to testing.
14

TRRADIATION DAMAGE IN GRAPHITE

 

Neutron irradiation alters the physical properties of graphite,
but our major concerns arise because of the dimensional changes that

15,16)

occur.( These dimensional changes are illustrated in Fig. € where

the data of Henson et 2&,(17)

are presented for an 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(lg) and we have shown that 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 decreasé
in density with potential crack and void formation. At some fluence
this will cause the mechanical properties to deteriorate and the perme-
ability to salt and fission products to increase. We feel that the
properties will be acceptable, at least until the material returns to
its original volume, and have defined this fluence as the lifetime. A
fourth problem is that the dimensional changes are dependent on temper-
ature 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

 
ORNL-DWG 67 —10504A

 

10

 

o

 

550~-600° C

 

/

VOLUMETRIC CHANGE (%)
O

.

v

e

 

l
O,

 

 

 

 

 

\\ _/_// 400-440°C ——

 

 

 

-10
0 10 20 30 40

50 (x402)

NEUTRON FLUENCE (neutrons/cm?) (£ >50 keV)

Fig. 6. Volume Change in Isotropic Graphite Dounreay Fast Reactor Irradiations.

GT
16

creep in graphite geometries of interest to us. Therefore, we consider
the onset of rapid growth to be the primary problem and the initial
dimensional changes of secondary importance.

We anticipate graphite temperatures between 550 and 750°C and
would like to operate with a fast flux (> 50 keV) of about
1 X 10'° neutrons cm™? sec™!. Data in Fig. 6 show that this flux will
cause this particular graphite to expand rapidly after a fluence of
approximately 3 X 10°° neutrons/cm? is reached (about 1 year of operation).
We can reduce the flux by decreasing the power density, but this is done
only with increases in the fuel inventory and doubling time. Hence, it
is quite desirable that we use graphites with better resistance to irra-
diation damage than the graphite shown in Fig. 6. We examined the data
available on current reactor graphites irradiated to high fluences and
found that the results described a fairly consistent picture. The
fluences required for graphite to reach its minimum volume were strongly
temperature dependent (decreased with increasing temperature), but were
not appreciably different for any of the graphites studied to date.
Although this observation is discouraging, we have pressed the problem
further and have found that better graphites already exist and that
others can probably be developed with only small changes in present mate-
rials and processing. ILet us look briefly at a simple description of
the origin of the dimensional changes and then return to our specific
observations.

Graphite, after being well graphitized at temperatures 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

 
17

within the basal planes (a-direction) and very weak van der Waals' forces
between atoms in adjacent basal planes (c-direction). This anisotropy

in atomic density and bond strength is reflected by very anisotropic
properties.

The changes that take place in a single crystal of graphite during
irradiation are shown schematically in Fig. 7. A neutron having an
energy above about 0.18 eV can displace a carbon atom from a close-packed
basal plane with a reasonable probability 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 planes (a-
direction) — see Fig. 7(b). This new configuration leads to a slight
increase in volume [Fig. 7(e)].

Polycrystalline graphites are not initially of theoretical density.
The voids present in the material are a result of shrinkage of the binder
during grephitization and from separation or fracture of layer planes
during cooling from the graphitization temperature. Initially,vduring
irradiation the porosity within the material tends to be filled as the
crystallites expand in the c-direction. This produces a densification
illustrated by the bottom curve in Fig. V(d). 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 the c¢-~direction. Thus, a minimum volume is obtained as shown in
18

ORNL —DWG 68— 42011A

PERFECT SINGLE CRYSTAL

 

 

 

 

 

 

 

 

DEFECT - *
PLANES g —
AL av
Lo ® N,
BASAL .
PLANES
(@) (b) (¢)
POLYCRYS TAL
DEFECT PLANES
+ + OBSERVED
AV | AV
Vo T, 2
PREDICTED
- DENSIFICATION -
(d) (e)

 

Fig. 7. Graphite Dimensional Changes Due to Irradiation.

 
19

Fig. 7(e). During the subsequent expansion, the material either remains
internally contiguous, in which case the volume change rate of the poly-
crystalline 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 crystallites of differing orienta-
tion (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 expec£ed if the material remained internally contiguous.
One further consideration helps ué to understand why the unpredicted
rapid growth takes place. A schematic representation of several coke
particles and binder after graphitizatién is shéwn in Fig. 8 (ref. 19).
Each coke particle consists of several crystals with a very high degree
of preferred orientation. Although we make a large piece of graphite
in which the coke particles are arranged randomly, there are still inter-
faces 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 graph-
ites undergo large dimensional changes at high fluences indicates that
these interfaces or boundaries are fracturing. As indicated by the
sketch in Fig. &, these boundaries are made up largely of the graphitized
binder materials. Thus, the properties of these boundaries are influenced
largely by the nature of the binder material and its interaction with
the coke particles.
We have set about making graphites with known filler and binder
materials, but our work in this area has not progressed very far. We

are also studying the properties of several commercial graphites that

we feel may be potentially useful for MSBR applications and others that
GRAPHITE

Petroleum coke

Pitch coke crystallites crystallites

 

 

 

 

 

Petroleum /

coke particle

Fig. 8. Proposed Arrangement of Crystallites in Graphitized Stock. (Reprinted

from High-Temperature Materials and Technology, ed. by I. E. Campbell and E. M. Sherwood,,

Wiley, New York, 1967, p. 195).

Qc
21

.should give us some basic information about irradiation damage in graph-
ite. Our graphite irradiations have been done at 705 + 10°C in the
High Flux Isotope Reactor (HFIR) Whefe the peak flux (> 50 keV) is

1 X 101° neutrons cm™? see™?. Thus, samples can be irradiated to
fluences of 1 X 1072 neutrons/cm2 in about 4 months.

A summary of our results to date is shown in Fig. 9. Several mate-
rials show a significant deviation from the "typical" behavior illustrated
in Fig. 6. The POCO* graphites show excellent resistance to irradiation
with very small positive dimensional changes out to fluences of
2.5 X 10%? neutrons/cm®. Thus, these data give us confidence that the
"typical" behavior of graphite can be improved markedly, but our results

have not been extended to fluences high enough to determine the exact

magnitude of this improvement.

SEALING GRAPHITE

 

Entry of fuel éalt into graphite can be prevented by keeping the
entrance_diameter of the accessible porosity smaller than 1 p. Although
this does require some extra care during processing, it can be accom-
plished routinely on large shapes. In fact, the grade CGB graphite
obtained 5 years ago for the MSRE satisfies this requirement.(2°) How-
ever, the graphite structure must be much more restrictive to prevent
gaseous fission products, particularly '3°Xe, from diffusing into the
graphite. We presently plan 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

 

*POCC Graphite, Inc., Garland, Texas.
4OOIn(eV +1)
0

Fig. 9.

22

ORNL-DWG68-10063R3

 

 

J
)

 

Y-12-3

 

BY-12

/

/]
/

—

 

POCO GRADES

T

/
o

 

N

 

\_/
N

 

N\

UK-ISO\\

NAT.
ATJ-SG oA H-315A I

7

%7‘"

 

 

/
//
7

(BNWL)

 

AGOT

/

 

i425>\

ATJ—S%
/<CSF(BNWL)
u

A

 

 

 

 

 

 

 

 

 

10

5

20

FLUENCE (neutrons/cm2) (£>50kev)

25

30
(X102

Volume Changes of Graphite Irradiated at 705°C.

 
23

heavily on the size of bubbles that can be injected and cifculated 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 !?°Xe will result if this gas
can enter the graphite surface at a high rate. Our present calculétions
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
resistance to gas flow only when its surface diffusivity to 135%e is

less than 1078 cmz/sec._

The best grades of commercial graphites presently available have
bulk diffusivities in the range of 10"1 o 1074 cm2/sec, and we feel
that it is unreasonable 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 impregnations
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 impreg-
nants to diffuse out of the material and the times required to reach
the graphitizing temperature become excessive. Thus, we presently feel
that it is more reasonable to reduce the surface diffusivity by &a post-
fabfication surface-sealing process involving gaseous impregnation.
Since the pyrolytic carbon that would be deposited and the graphite sub-
strate will change dimensions differently under irradiation, it is
imperative that the pyrolytic carbon be linked with the substrate struec-

ture and not deposited as a surface layer that can be sheared easily.
24

The task that we have in sealing the graphite is illustrated by
the photomicrograph in Fig. 10. We must adjust our processing parameters
so that the volds 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 processing times. We have used a different method to
accomplish penetration which involves pulsing the sample environment
between a rich hydrocarbon environment and vacuum. The vacuum cycle
removes the reaction products (primarily hydrogen) and allows more hydro-
carbon gas to enter the void. Specifically, we have used 1,3 butadiene
at 20 psig, deposition temperatures of 800 to 1000°C, and cycle times
of about 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 to 1000°C because higher temperatures result
in a surface coating not penetrating the pore structure and lower tem-
peratures 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 required for the process will be very important
in determining the cost.

We have worked with two commercial graphites — viz, AXF made by
POCO and ATJ-SG made by UCC. These materials had widely different
accessible pore spectra; nearly all the pores in the AXF material were

less than 0.8 p in diameter while the ATJ-SG grade had appreciable pores
25

 

Fig. 10. Photomicrograph of the Edge of Graphite Showing a Pore
that has been Partially Coated and then Sealed over with Pyrocarbon.

 

 
26

in all size ranges up to 17 p. Thus, the sealing characteristics of

the two materials were widely different. The results of some of our
parameter studies are shown in Fig. 11 where we have varied the vacuum-
hydrocarbon cycle times 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 essentially
surface coating, and the data suggest that some finite amount of surface
coating is necessary to attain the MSBR permeability specificdtion of
less than 1078 cm®/sec for 135%e at 700°C. This is approximately equiva-
lent to a helium permeability of 108 cmz/sec at ambient temperature.

We have used the latter criterion in our studies (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 material was
increased. Thus, shortening the vacuum pulse accelerated the process

and improved the product, both very desirable characteristics.

 
27

0.10

0.08

0.06

0.04
MATERIAL: POCO AXF GRAPHITE

o
O
N

ORNL-DWG 68-12030

VACUUM (sec) HYDROCARBON (sec)

60
60
30
30
15

10x10°8cm?2 /sec (HELIUM)

<1079 m2/sec (HELIUM)

 

 

 

 

 

TOTAL WEIGHT GAINED (g)
o

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

0
O T _4| T T T T T T
M5X1O mm Hg(AIR) 2.5%x102mm
| L oHg(AIR)
o —425x1072mm = 48x1072 mm
—*" |Hg (AIR) Hg (AIR)
008 A .,/ " '—-_J-_-—-T
( -t o] /‘//x -2
A o At 1 75x10" “mm
> o L= e Hg (AIR)
/o/ / /1,__4/
006 -.,“ 1/ /“/ V
p 7 > L~ o
./.,j/ /‘/‘/
'/ ,/
A |
0.04 ol yed
;‘ ‘/n /
. //
_ Ve MATERIAL: ATJ-SG GRAPHITE
002 AN
/o
Y/
/7
0
0 4 8 12 16 20 24 28 32 36 40 46

TOTAL PROCESSING TIME (hr)

Fig. 11. Impregnation Rate of Graphite Using 1,3 Butadiene at

850° C.
28

Our studies are not yet extensive enough to optimize the deposition
conditions but are sufficient to make us optimistic about being able to
reduce the surface diffusivity of graphite to the desired level. The
remaining question of prime importance is the integrity of the seal

after exposure to high neutron fluences.

CORROSION IN FLUORIDE SALT SYSTEMS

 

(21-29) ang our experience with the

Two decades of corrosion testing
MSRE(2’3) have demonstrated the excellent compatibility of Hastelloy N
and graphite with fluoride salts containing LiF, BelF,, ThF,, and UF,.

Our fertile~fissile salt will contain these éame fluorides, so only
proof-testing will be required for the primary reactor circuit. How-
ever, we desire a lower melting coolant salt in the secondary coolant
circuit than the LiF-BeF, salt presently used in the MSRE and have
chosen a sodium fluoroborate salt (NaBF,—8 mole % NaF) for further study.
This salt is inexpensive (< $0.50/1b) and has a low melting point of
380°C. A significant characteristic of this salt is that it has an
appreciable equilibrium overpressure of BF; gas (e.g., 180 mm at 600°C).

Much of our present corrosion work is concerned with the compati-
bility of Hastelloy N with sodium fluoroborate. Some earlier thermal
convection loop studies involving a relatively impure salt of composition
NaBF;~4 mole % NaF—6 mole % KBF, showed that a Croloy 9M loop plugged
after 1440 hr at a maximum temperature of 607°C and a tempgrature
difference of 145°C, and that a Hastelloy N loop was partially plugged
after 8765 hr of operation under the same temperature conditions.(3°)

The plug in the Croloy loop was comprised primarily of pure iron
29

crystals and the partial plug in the Hasteiloy N loop was made up of a
compact mass of green single crystals of Na;CrF,. The salt charge from
the Hastelloy N loop contained large amounts of Cr, Fe, Ni, and Mo, all
major alloying elements in Hastelloy N. -

In our more recent tests we have used a fluoroborate salt of compo-
sition NaBF,—8 mole % NaF of a higher purity than the first salt that we
worked with.(31’32) We are using a modified thermal convection loop
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 (NCIL-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 hr of opera-
tion and a large variation after 4200 hr 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 chromium concentrations have continued to
increase at a rate pfoportional to (time)%, indicating that the process

is controlled by diffusion in the metal. The nickel and molybdenum con-

centrations in the salt have remained very low except for times when
30

ORNL -DWG68- 12034

 

 

 

 

 

6
4 COLDEST SPECIMENS ——
A 460°C
/ | o
A 0/
A\ e
2 A,A——-'A"‘ S —
W/ o—T
e

 

 

 

 

 

 

WEIGHT CHANGE (mg/cm?)
t
N
‘>/ 7

 

  
 

 

 

-6 HOTTEST SPECIMENS -
607°C

 

 

 

 

 

 

 

 

 

 

 

 

-8
O—TITANIUM MODIFIED HASTELLOY N- NCL-14
-{0 —A—STANDARD HASTELLOY N- NCL-{3 \o
-2 ,
0o { 2 3 4 5

2
(x103)
TIME ( hr)

Fig. 12. Comparison of the Weight Changes of Hastelloy N Specimens
Inserted in NaBF,-NaF {92—8 mole %) Thermal Convection Loops.

 
31

ORNL-DWG 68—-11768R

 

3000 ' ‘

 

2000 >0

 

 

 

CONCENTRATION {(ppm)
:‘:.
0
»

 

 

 

1000 F———j NI T ~
— . . - .
® ®
0
400 O

 

350 A- "

 

 

 

 

A/
300 e b
A
*
E . //////
S 250
=
9 H
g 200
o
- [ ]
=
L
g 150 *
8 s

 

 

 

 

100 7' //wc \QT,M

N

50 HY C//
)——0——0 O -

0
0 1000 2000 3000 4000 5000 6000 7000 8000 9000
TIME (hr)

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Fig. 13. Variation of Impurities with Time in NaBF,-NaF (928 mole %)
Thermal Convection Loop NCL-14.
32

moist air was inadvertently contacted with the salt. These results
show qualitatively that the corrosion rates increased when the oxygen
and water levels increased. Capsule tests in which sodium fluoroborate
containing 1400 ppm O, and 400 ppm H,O was contacted with Hastelloy N
for 6800 hr at 607°C exhibited very low corrosion rates. Future work
will be directed toward defining the oxygen and water levels that result
in acceptable corrosion rates.

The information that we obtained on the changes in salt composition
and the weight ;hanges of ten samples located at various points (and tem-
peratures) around thé loop enabled us 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.,

Awloss - Awdeposited * AwSalt ’ (1)

We construct a weight change versus temperature profile based on the-
removable samples and assume 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 analysis. 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 is relatively free of iron, so the weight loss of this sample
should be due primarily to the removal of chromium. The quantity of mate-
rial removed by diffusion under conditions where the surface concentration

of the diffusing element is zero is given by:

 
where

: 2
material removed, g/cm 5

&

Cp = bulk concentration of diffusing species, g/cm3,

o
I

diffusivity, cm®/sec

ct
Il

time, sec.

Using the data of Grimes et a_l.(33) for diffusion of chromium in
Hastelloy N, we found by this analysis that the quantity of chromium
removed by diffusion 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 temper-
atures or by impurities (likely HF formed by water ingestion) that lead
to some general attack of the metal that is nof diffusion controlled.
Two observations argue against the latter possibility. First, we have
not been able to see by microprobe analysis any transfer of nickel or
molybdenum to the colder surfaces of the loop except during a brief
period after 4200 hr 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 the standard and modified Hastelloy N during
the first 4000 hr of operation. Rewriting of Eq. (2) in terms of a

reaction rate constant, K, instead of the diffusion coefficient yields
M=C, JKt . (3)
We have already indicated that this equation predicts the material trans-

ported by diffusion to be too low for the modified alloy when K equals D,

but let K take on a value so that the predicted and observed weight losses
34

for a given time of operation agree with Cg equal to 7% Cr. Now consider
the standard alloy in which Cy corresponds to 7% Cr + 4% Fe. We find that
the same K chosen for the modified alloy predicts the observed weight
change for the standard alloy. Thus, the difference in the weight losses
between the modified and sténdard alloy is due principally to the iron
content. Had much general corrosion occurred, this adjustment'in Co
should not have worked. In fact, we found this analytical procedure to
be entirely unsatisfactory for the short time period after 4200 hr 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 con-
centration of iron and chromium at zero. Thus, even though the process
remains diffusion controlled, the rate can be increased by impurities.
Although some of the curves in Figs. 12 and 13 have quite large
slopes, the corrosion rates are not very high. Using the rather inac-
curate method of converting the weight losses to corrosion depths indi-
cates that the average rate during 8000 hr of operation has been
0.7 mil/yr. The rate has decreased to about 0.3 mil/yr for long time
periods in which operation was not disturbed. In scaled-up MSBR systems
we probably will use a cold trapping technique to remove some of the
corrosion products so that their solubilities are not exceeded. We
presently feel that the sodium fluoroborate salt will provide a satis-

factory and economical secondary coolant for molten salt reactors.

 
35

SUMMARY

OQur experience with the MSRE has proven the basic compatibility of
the graphite-Hastelloy N-fluoride salt system at elevated temperatures.
However, a molten-salt breeder reactor will impose more stringent
operating conditions, and we need to make some improvements in the
graphite and Hastelloy N for this system. The mechanical properties of
Hastelloy N deteriorate under thermal neutron irradiation, and we have
found that the addition of titanium in combination with strong carbide
formers such as niobium 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 presently limits the lifetime
of the core graphite. Although we can replace the core graphite as often
as neceséary, these replacements influence the economics of our reactor,
and we have embarked on a program to find a better graphite. Our studies
to date indicate that graphites can be developed that have better
resistance to irradiation damage than conventional nuclear graphites.

We plan to seal the graphite used in the core 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. Our corrosion studies
are currently concentrated on evaluating the compatibility of Hastelloy N
with a potential coolant salt, sodium fluoroborate. Cur studies indi-
cate that the corrosion rate is acceptable as long as the éalt does not

contain large amounts of impurities, such as HF and H;O.
10.

11.

36

REFERENCES

W. H. Cook, Molten-Salt Reactor Program Semiann. Progr. Rept.

 

August 31, 1965, ORNL-3872, Oak Ridge National Laboratory,

 

pp. 87-92.

H. E. McCoy, An Evaluation of the Molten-Salt Reactor Experiment

 

 

Hastelloy N Surveillance Specimens — First Group, ORNL-TM-1997,
Oak Ridge National Laboratory (November 1967).

H. E. McCoy, An Evaluation of the Molten Salt Reactor Experiment

 

Hastelloy N Surveillance Specimens — Second Group, ORNL-TM-2359,

 

Oak Ridge National Laboratory, in press.

 

D. R. Harries, J. Brit. Nucl. Energy Soc. 5, 74 (1966).

W. R. Martin and J. R. Weir, pp. 251-267 in Flow and Fracture of

 

Metals and Alloys in Nuclear Environments Spec. Tech. Publ. 380,
American Society for Testing and Materials, Philadelphia, 1965.

J. T. Venard and J. R. Weir, p. 269 in Flow and Fracture of Metals

 

and Alloys in Nuclear Enviromments Spec. Tech. Publ. 380, American
Society for Testing and Materials, Philadelphia, 1965.

W. R. Martin and J. R. Weir, Nucl. Appl. 1(2), 160-167 (1965).

W. R.- Martin and J. R. Weir, Nucl. Appl. 3, 167 (1967).

H. E. McCoy and J. R. Weir, Nucl. Appl. 4, 96 (1968).

P.C.L. Pfeil and D. R. Harries, p. 202 in Flow and Fracture of Metals

 

and Alloys in Nuclear Environments Spec. Tech. Publ. 380, American

 

Society for Testing and Materials, Philadelphia, 1965.

P.C.L. Pfeil, P. J. Barton, and D. R. Arkell, Trans. Am. Nucl. Soc.

 

8, 120 (1965).
12.

13.

14,

15.

16.

17.

18.

19.

20.

21.

22.

37

P.R.B. Higgins and A. C. Roberts, Nature 206, 1249 (1965).

H. E. McCoy Jr., and J. R. Weir, Jr., Materials Development for

 

Molten-Salt Breeder Reactors, ORNL-TM-1854, Oak Ridge National

 

Laboratory (June 1967).

H. E. McCoy, Jr., and J. R. Weir, Jr., "Development of a Titanium-

Modified Hastelloy with Improved Resistance to Radiation Damage,"

Proceedings of a Symposium on the Effects of Radiation on Structural

 

Metals, San Francisco, Calif., June 23-28, 1968, to be published.

 

R. E. Nightingale, Nuclear Graphite, Academic Press, New York, 1962.

 

J.H.W. Simmons, Radiation Damage in Graphite, Pergamon Press, New

York, 1965.

R. W. Henson, A. J. Perks, and J.H.W. Simmons, Lattice Parameter and

 

Dimensional Changes in Graphite Irradiated Between 300 and 1350°C,

 

AFRE-R 5489, Atomic Energy Research Establishment, p. 33 (June 1967).

C. R. Kennedy, Gas Cooled Reactor Program Semiann. Progr. Rept.

 

Mar. 31, 1964, ORNL-3619, Oak Ridge National Laboratory, pp. 151—-154.

2 J

W. C. Riley, High-Temperature Materials and Technology, ed. by

 

I. E. Campbell and E. M. Sherwood, Wiley, New York, 1967, p. 188.

w. H. Cbok, Molten-Salt Reactor Program Semiann. Progr. Rept.

 

July 31, 1964, ORNL-3708, Oak Ridge National Laboratory, p. 377.

—)

L. S. Richardson, D. C. Vreeland, and W. D. Manly, Corrosion by

Molten Fluorides, ORNL-1491, Oak Ridge National Laboratory

 

(March 17, 1953).

 

G. M. Adamson, R. S. Crouse, and W. D. Manly, Interim Report on

Corrosion by Alkali-Metal Fluorides:

Oak Ridge National lLaboratory.

Work to May 1, 1953, ORNL-2337,
38

23. G. M. Adamson, R. S. Crouse, and W. D. Manly, Interim Report on

 

~Corrosion by Zirconium-Base Fluorides, ORNL-2338, Oak Ridge

 

National Laboratory (Jan. 3, 1961).
24. W. B. Cottrell, T. E. Crabtree, A. L. Davis, and W. G. Piper, Dis-

assembly and Postoperative Examination of the Aircraft Reactor

 

Experiment, ORNL-1868, Oak Ridge National Laboratory (April 2, 1958).
Z25. W. D. Manly, G. M. Adamson, Jr., J. H. Coobs, J. H. DeVan, D. A. Douglas,

E. E. Hoffman, and P. Patriarca, Aircraft Reactor Experiment —

 

Metallurgical Aspects, ORNL-2349, Oak Ridge National Laboratory

 

(Dec. 20, 1957), pp. 2-24.
26. W. D. Manly, J. H. Coobs, J. H. DeVan, D. A. Douglas, H. Inouye,

P. Patriarca, T. K. Roche, and J. L. Scott, Progr. Nucl. Energy

 

Ser. IV 2, 164-179 (1960).
27. W. D. Manly, J. W. Allen, W. H. Cook, J. H. DeVan, D. A. Douglas,
H. Inouye, D. H. Jansen, P. Patriarca, T. K. Roche, G. M. Slaughter,

A. Taboada, and G. M. Tolson, Fluid Fuel Reactors, ed. by

 

James A. Lane, H. G. MacPherson, and Frank Maslan, Addison-Wesley,
Reading, Pa., 1958, pp. 595-604.

28. Molten-Salt Reactor Program Status Report, ORNL-CF-58-5-3, Oak

 

Ridge National Laboratory (May 1, 1958), pp. 112-113.
29. J. H. DeVan and R. B. Evans III, pp. 557-579 in Conference on

Corrosion of Reactor Materials, June 48, 1962, Proceedings Vol. II,

 

International Atomic Energy Agency, Vienna, 1962.

30. J. W. Koger and A. P. Litman, Compatibility of Hastelloy N and

 

Croloy 9M with NaBF;-NaF-KBF, (90—4—6 mole %) Fluorcborate Salt,

 

ORNL-TM-2490, Oak Ridge National Laboratory, in preparation.

 
39

31. J. W. Koger and A. P. Litman, Molten-Salt Reactor Program Semiann.

 

Progr. Rept. Feb. 29, 1968, ORNL-4254, Oak Ridge National Laboratory,

 

pp. 221-225. .

32. Molten=-Salt Reactor Program Semiann. Progr. Rept. Aug. 31, 1968,

 

ORNL-4344, Oak Ridge National Laboratory, in press.
33. W. R. Grimes, G. M. Watson, J. H. DeVan, and R. B. Evans, p. 559 in

Conference on the Use of Radioisotopes in the Physical Sciences

 

and Industry, Sept. 6—17, 1960, Proceedings Vol. III, International

 

Atomic Energy Agency, Vienna, 1962.

 
»—

1-3.
=5,

6—15.
16.
17.
18.
19.
20.
21.
22.
23.
2.
25.
26.
27.
28.
29.

30-34.
35.
36.
37.
38.
39.
40.
41.
42.
43,
et
45,
46.
47,
48.
49,
50.
51.
52.
53.
5.
55.
56.
57.
58.
59,
60.
61.

HEHGORANDOENERUQONPEHGAPNOHNOREENEO0NGOR S D aw

41

INTERNAL DISTRIBUTION

Central Research Library
ORNL — Y-12 Technical Library
Document Reference Section
Laboratory Records Department

Laboratory Records, ORNL RC
ORNL Patent Office

Adams

Adamson, Jr.

Affel

Anderson

Apple

Baes

Baker

Ball

Bamberger

Barton

Bauman

Beall

Beatty

. Bell

Bender

E. Bettis

S. Bettis

S. Billington
E. Blanco
F
0
E

SPHEASEGERAIRER

. Blankenship

. Blomeke

. Bloom
Blumberg
G. Bohlmann
J. Borkowski
E. Boyd
Braunstein
Bredig
Briggs
Bronstein
Brunton
. Canonico
Cantor
Carter
Cathers
Cavin
Chandler
Clark
Cobb
Cochran

POPE P

OWEmEWH P

62.
63.
6.
65—69.
70.
71.
72.
73.
T,
75.
76.
77.
78.
79.
80.
g1.
82.
83.
8.
85.
g6.
g87.
g8.
89.
90.
91.
92.
93.
.
95.
96—100,
101.
102.
103.
104.
105.
106.
107.
108.
109.
110.
111.
112.
113,
114,
115.

PP ORI LR RO NP O N PPN NS PN AP RN LD AP SR e O

HENOQTEr NI EONEPE AT P PRI TP NS GE N EOEEE N ROl E SR

ORNL-TM-2511

Collins
Compere
Cook
Cook
Corbin

O
™

Crowley
Culler
Cuneo
Cunningham
Dale
Davis
Davis
DeBakker
DeVan
Ditto
Dudley
Dworkin
Dyslin
Eatherly
Engel
Epler
Ferguson
Ferris
Fraas
Friedman
Frye, Jr.
Furlong
Gabbard
Gallaher
Gehlbach
Gibbons
Gilpatrick
Goodwin
Grimes
Grindell
Gunkel
Guymon
Hammond
Hannaford
Harley
Harman
Harms
Harrill
Haubenreich
Helms
116.
117.
118.
119-121.
122.
123.
124.
125.
126.
127.
128.
129.
130.
131.
132.
133.
134.
135-139.
140.
141.
142.
143,
144,
145-149.
150.
151.
152.
153.
154,
155.
156.
157.
158.
159.
160-164%.
165.
leob.
167.
168.
169.
170.
171.
172.
173.
174,
175.
176.
177.
178.
179.
180.
181-185.

HEEEEEOO I EII R PHQPIRHQEOCAP TN CGEROR R EENERAP I UOEICY T

42

Herndon
Hess
Hightower
Hill
Hoffman
Holmes
Holz
Horton
Houtzeel

L. Hudson

R. Huntley
Inouye
Jordan
Kasten
Kedl
Kelley
Kelly
Kennedy
Kerlin
Kerr
Keyes
Kiplinger
Kirslis
Koger
Korsmeyer
Krakoviak
Kress
Krewson
Lamb

Lane
Larson
Lawrence -
Lin
Lindauer
Litman
Llewellyn
Long, Jr.
Lotts
Lundin
Lyon
Macklin
MacPherson
MacPherson
Mailen
Manning
Martin
Martin
Mateer

. Mauney
McClaln

W. MeClung
E. McCoy

EoREAI=EAQ

LSRR RN R IR EPEEAEDEN SO IED O 0D

186.
187.
188.
189.
190.
191.
192.
193.
194.
195.
196.
197.
198.
199.
200.
201.
202.
203.
204,
205,
206.
207.
208.
209,
210.
211.
212.
213.
214,
215,
216.
217.
218.
- 219.
220224,
225.
226.
227.
228.
229.
230.
231.
232.
233.
234,
235.
236.
237.
238.
239.
240.
241,

CEEPAENOERCEPES TR NERSYR P ES 0O

EROPpREIEIHUOORESA ?‘?*?’F’?

McElroy
McGlothlan
McHargue
McNeese
McWherter
Metz
Meyer
Moore
Moulton
Mueller
Nelms
Nichol
Nichols
Nicholson
Oakes
Patrlarca
Perry
Pickel
Piper
Prince
Ragan
Redford
ichardson
Robbins
Robertson
Robinson
Romberger
Ross
Savage
Schaffer .
. Schilling

CETEPERPOHIEERE

MfiQPFQprPergs

Seagren
Sessions
Shaffer
Sides
Slaughter
Smith
Smith
Smith
Smith
Smith
piewak
Steffy
Stoddart
Stone
Strehlow
sSundberg
Tallackson
Taylor
erry

QP PoEgEERER

S E PP EO0

.
242,
243,
244,
245,
246,
. 247,
24.8.

249,

250.

251.

252.

253.

254—258.

272.
273.
274.
275.
276.
277,
278.
279.
280.
. 281.
282.
283.
284.
285.
286.
287.
288.
289.
290.
291.
292.

293.
29%.
295.
296.
297,
298.
299.
300.
301—315.

—

5

a

GrPQEoOITEDEYE

SREEUSQQUPHEIZQAPIOQDEO

D IOENnE

*

JEIADREOREE R AE

BEEEEOO

C.')

ErQEERSPHED

43

- e P

Thoma, ‘ 259. W. J. Werner
Thomason 260, K. W. West
Toth 261, M. E. Whatley
Trauger 262, J. C. White
Unger 263. R. P. Wichner
Waddell 264, F. W, Wiffen
Watson 265. L. V. Wilson
Watson 266. J. W. Woods
Watts | 267. Gale Young
Weaver 268. H. C. Young
Webster 269. J. P. Young
Weinberg 270. E. L. Youngblood
Weir, Jr. 271. F. C. Zapp

EXTERNAL DISTRIBUTION

Allaria, Atomics International

Asquith, Atomics International

Cope, RDT, SSR, AEC, Oak Ridge National Iaboratory
Cunningham, AEC, Washington

Deering, AEC, OSR, Oak Ridge National Laboratory
Dieckamp, Atomlcs International

1ambusso, AEC, Washington

Haines, AEC, Washington

Johnson, AEC, Washington

Kitterman, AEC, Washington

Larkin, AEC, Oak Ridge Operations

McIntosh, AEC, Washington

Martin, Atomics International

Mason, Atomics International

Matthews, RDT, OSR, AEC, Oak Ridge National lLaboratory
Meyers, Atomics International

Moteff, General Electric, Cincinnati

E.
M.

Reardon, AEC, Canoga Park Area Office
Roth, AEC, Oak Ridge Operations

ohaw, ABEC, Washington

M.

R

Simmons, Division of Reactor Development and Technology,
Washington

Smalley, AEC, Washington

Stamp, AEC, Canoga Park Area Office

Stansbury, the University of Tennessee

Stevens, AEC, Washington

.. Sweek, AEC, Washlngton

Taboada, AEC, Washlngton

F.

Wilson, Atomlcs International

lLaboratory and University Division, AEC, Oak Ridge Operations
Division of Technical Information Extension