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Effects of fast-neutron irradiation on pyrolytic silicon carbide
JOURNAL
OF NUCLEAR
33
MATERIALS
17-22.
(1969)
EFFECTS OF FAST-NEUTRON
0 NORTH-HOLLAND
IRRADIATION
PUBLISHINQ
ON PYROLYTIC
CO., AMSTERDAM
SILICON CARBIDE
R. J. PRICE Gulf General
Atomic
Incorporated,
John
Jay
San Diego, Received
Samples
of pyrolytic
were irradiated at
decreased from
between
showed
460
an isotropic
with
increasing
and
parameter
scopic accuracy. X-ray
changes
dimensional Irradiation
caused
line broadening
generation
of internal
change in Young’s
were
changes that
“C.
qui Btait attribuable internes.
to the macroexperimental increase
was attributable
Proben wurden
de carbure
4,2x
1021 nvt
comprises
(E>0,18
entre
presentaient
une
460
MeV) et
aus
avec la temperature
de 0,35%
8. 460 “C a 0,05%
parametres
1.
Introduction
“C.
Les
d’irradiation
qui
und Temperaturen
zwischen zeigten
Btaient
Bgales
die mit steigender
x 1021 nvt,
B-Sic
0,18 MeV)
460 und 1040 “C unter-
eine isotrope
lineare
Aus-
Bestrahlungstemperatur
der
Messgenauigkeit
stimmen
Dimensionsanderungen
iiberein.
lung
Verbreiterung
trat
eine
leichte
zurtickgefiihrt
aux
wird.
die Pinde-
mit den makroskopischen Durch
reflexe ein, die auf die Entstehung
croissante,
a 1040 “C. Les variations
reticulaires
polykristallinem
(2,0-4,2
sucht. Die Proben
Innerhalb
Bchantillons lineaire
pyrolytischem
rungen der Gitterparameter
temperatures
isotrope
significative
de rupture.
von 0,35 y0 bei 460 “C auf 0,05 O/obei 1040 “C abfiel.
,!l pyrolytique
a des doses de 2 a 8. des
1040
dilatation
diminuait, des
de silicium
ou du module
mit Neutronen
dehnung, ont 6th irradies
de deformations
pas de variation
in
to the
modulus or the modulus of rupture.
Des echantillons
d’Young
erreurs un leger
des raies des rayons X
a la creation
avait
aux
produisait
de l’elargissement
11 n’y
du module
strains. There was no significant
et polycristallin
Science,
macroscopique,
L’irradiation
accroissement
equal
slight
p&s.
The
within a
de dimension
that
1040 “C. The
and Applied
1969
d’experience
temperature
at
for Pure
USA
variations
MeV)
expansion
irradiation
0.35 y0 at 460 “C to 0.05%
lattice
1040
linear
Laboratory
27 February
/l-silicon carbide
to 2.0 to 4.2 x 1021nvt (E>O.lS
temperatures
samples
polycrystalline
Hopkins
California 92112,
BestrahRontgen-
innerer Spannungen
Bezeichnende
Anderungen
Elastizitiits-
oder Bruchmoduls
traten
pyrolytic
silicon
are
carbide
die der
nicht
des
auf.
therefore
of
considerable interest. Previous work has dealt with the effects of low neutron exposures (less than 102i nvt) on the dimensions and lattice parameters of single crystals of oc-silicon carbide s-5). The effects of intermediate neutron exposures (up to 1.8 x 1021 nvt) at 250, 475, and 700 “C on the dimensions, lattice parameters, thermal diffusivity, and Young’s modulus of low-density self-bonded polycrystalline /?-silicon carbide have also been reported 6). In the present work, pyrolytic, polycrystalline /l-silicon carbide was irradiated to 2.0 to 4.2x 1021 nvt (E>0.18 MeV) at temperatures between 460 and 1040 “C. The
Pyrolytic silicon carbide is used as a coating for the particles of uranium or thorium oxide or carbide that form the fuel for high-temperature gas-cooled nuclear reactors 1). The properties of silicon carbide that make it suitable for use as a fuel-coating material are its hightemperature strength, low neutron absorption cross-section, and low permeability to fission products. Pyrolytic silicon carbide has also been considered as a coating to protect against the accidental oxidation of moderator graphite in high-temperature nuclear reactors 2). The effects of fast-neutron irradiation at high temperatures on the structure and physical properties of 17
18
R.
changes
in dimensions,
density,
meter, X-ray line-broadening, and modulus structure
of rupture
and physical
lattice
J.
para-
Young’s modulus,
were measured. properties
The
of the un-
irradiated material are reported elsewhere 7).
PRICE
diffraction Norelco
measurements diffractometer
Experimental
was then
computed
for systematic
The material was prepared by the decomposition
of methyltrichlorosilane-hydrogen
mix-
made
with
copper
a
Ka
radiation, and traces were obtained for peaks with indices up to (333). The lattice parameter from
using an extrapolation 2.
were using
the
the peak positions
procedure
that corrects
errors in sample positioning
di~ractometer
geometry 9). X-ray
in
peak
widths were measured and corrected for doublet
tures in a fluidized bed of particles at 1400 “C. The total gas flow rate was 10 000 ml/min with
and
instrumental
broadening
using
Jones’
carbide was out into strips ~neasuring about 0.6 cm x 0.1 cm x 0.01 am. The lengths of the strips were measured with
carbide strip samples were then estimated by calculating their difference in temperature from the fuel samples. The mean sample temperatures are considered accurate to within i 50 “C and in addition the temperatures were subject to fluctuations of & 50 “C during the course of the irradiation. The neutron fluxes were calculated from the analysis of nickel and iron dosimeter wires.
method 10). A potassium chloride standard was used. The correction curves appropriate for a 5 to 10% of the hydrogen carrier gas passing line profile intermediate in shape between a througl~ a n~ethyltrichlorosilane bubbler at Gaussian and a Lorenzian and for a diffractoroom temperature. The fluidized bed was meter with a line source were used ii). contained in a 3.8 cm diameter graphite tube The modulus of rupture and Young’s modulus and the initial bed surface areas were between were measured in 3-point bending at room 500 and 2000 cma. The reactant input flux was between 2 and 10 x IO-6 moles of silane~rni~~~crnz temperature, using techniques described previously 12). of bed surface. Other details of the deposition About ten strip samples were included in two technique are given elsewhere 8). or three cells in each of four irradiation The material consisted of /I-silicon carbide capsules. The neutron exposures and mean and had a density of 3.17 to 3.20 g/cm3 operating temperatures of each group of samples (theoretical density: 3.21 g/cma). The grain size are shown in table 2. The irradiations were was less than 1 pms). X-ray diffraction patterns carried out in the Engineering Test Reactor showed no evidence for excess silicon or carbon. (ETR) at Idaho Falls, Idaho, in capsules whose A typical impurity analysis is shown in table 1. primary purpose was to test coated-particle Other structural properties are reported elsefuel, and whose design has been described 13). where 7). The operating temperatures of the coatedSamples for irradiation were obtained by particle fuel samples were monitored with placing graphite discs 0.7 cm diameter x0.1 cm thermocouples and regulated by control of the thick in the fluidized bed. After coating, the gas composition in heat-transfer gaps around discs were cut with a diamond wheel and the the crucibles. The temperatures of the silicon graphite removed by grinding. The silicon
a machinist’s microscope to an accuracy of 0.02%. Densities were measured by suspension in methylene iodide-benzene mixtures. X-ray
TABLE
Spectrographic
Mn 4 *
Measured
1
impurity analysis of pyrolytic carbide (ppm) Mg 8
Fe 20
Ni (10
Al
Cu
Cl*
60
20
87
by neutron activation
silicon
analysis.
3.
Results and discussion
3.1.
CHARGES
IN
DIiVfEBSIONS
AND
LATTICE
PARAMETER
The mean linear dimensional
expansions
and
EFFECTS
OF
FAST-NEUTRON
19
IRRADIATION
TABLE 2 Change in linear dimensions, lattice parameter and X-ray Irradiation
P-13-F
I
1
P-13-H P-13-J P-13-K
y0
S.D.)
(f
Increase in RMS internal
Mean 1 temperature
Cell no.
of silicon carbide during irradiation
Mean expansion,
conditions
Veutron exposure Capsule no.
line-broadening
Linear
Lattice
strain
dimensions
parameter
(x 104)
!
(“Cl
(
630
0.24 f
0.02
0.20 f
0.02
5.0
2.8 x 1021
1020
0.08 i
0.04
0.03 * 0.02
4.0
2.7 x 1021
1010
0.06 & 0.04
0.05 f
0.02
3.0
3.8 x 1021
700
0.30 + 0.02
0.33 f
0.02
3.5
4.2 x 1021
980
0.07 * 0.05
0.07 f
0.01
3.5
3.8 x 1021
1040
0.06 f
0.03
0.10 + 0.04
7.0
2.7 x 1021
460
0.36 f
0.03
0.34 & 0.04
3.5
2.7 x 1021
620
0.23 f
0.03
0.19 f
0.03
5.0
2.0 x 1021
1010
0.03 f
0.04
0.05 f
0.02
3.0
2.8x
1021
1
i
changes in lattice parameter are shown in table 2. The standard deviations of the mean
Lattice Parameter
Source
0
macroscopic expansions were obtained from the experimental spread among about ten similar strips irradiated under each set of conditions. Lattice-parameter measurements were made after lumping the similar strips together, and the standard deviation of the estimate was obtained from the curve-fitting statistics of the extrapolation procedure. The lattice parameter expansions were the same as the macroscopic
Pyrolytic
1
0.
3:17
p=
- 3.20
[present work]
a Single
cl
[Primak
Crystals
er~_I.~)]
A
a Single
Crystals
0
a Single
Crystals
[Ealarin4)]
[Pravdyuka_al.‘)] Self-bonded
V
6.
p-
2.1
-
2.2
[Thorne&A_.6)]
expansions within the accuracy of the measurements. Measurements of the change in density showed that the changes were isotropic. Previous work 5, a) has shown that radiation-induced
expansion
the
of silicon carbide
saturates and becomes independent of exposure after a low neutron exposure (1 to 3 x lo20 nvt). The expansion level decreases with increasing irradiation temperature. The expansions obtained in the present work are plotted as a function of irradiation temperature in fig. 1, together with other values reported in the literature for exposures high enough for saturation to have occurred. The linear expansion decreases from about 1% for irradiations near room temperature to about 0.05% at 1000 “C. There does not appear to be any systematic dependence on the gram size or porosity of the material, and 01 (hexagonal) silicon carbide behaves similarly to fi (cubic) silicon carbide.
”
400
200
Irradiation
Fig.
1.
Saturation
600
800
temperature
(‘C)
radiation-induced
IO00
expansion
of
silicon carbide as a function of irradiation temperature (neutron
exposures > 1020 nvt).
There is no measurable difference between the lattice expansion and the macroscopic expansion. The saturation of the expansion as the neutron exposure increases, and the monotonic decrease in expansion with increasing irradiation temperatures are typical of the behavior of isotropic or near-isotropic ceramic materials 14). Irradiation with fast neutrons creates equal
20
R.
numbers of interstitials
and vacancies,
in various
stages
occurs
Saturation
defect production retention
I
400
311 220
331
222
I
422
420
333
changes
rate equals the
rate. The amount of damage
decreases with increasing
temperature
200 111
property
of the property
when the annihilation
Reflection
The
of aggregation
and are responsible for the observable changes.
PRICE
most of
which are removed by mutual annihilation. rest remain
J.
because
irradiation
the annihilation
rate is
thermally activated, while the defect production rate is fixed by the neutron flux. Because the dilatation associated with an interstitial atom
7
is greater than the relaxation around a vacancy, an overall expansion results. Single point defects
c
and small point-defect clusters change both the lattice parameter and the bulk dimensions whereas large clusters of defects equally, condense to form dislocation loops that cause a larger fractional change in the bulk dimensions than in the lattice parameter. In the present work the experimental results do not show a significant difference between the lattice para-
0
b-
"0
(b)
Irradiated at 1040°c
to 3.8
X
IO
21
"Yt
0 /
4
x 0
3
B 92
I 0 6
meter expansion and the macroscopic linear expansion (table 2). The lack of a difference indicates that most of the point defects are present as single defects or small clusters.
(c)
Irradiated
to 2.7
X
IO
However, those data do not exclude the possibility that dislocation loops also are formed. Both interstitial and vacancy loops could be present without changing the equality between the
bulk
provided
and
lattice
dimensional
changes,
that the total loop area is equal for
each type.
If, as seems more probable,
inter-
stitial loops predominate, up to 10e3 of the atoms could be located in such loops before the resulting linear expansion would exceed the lattice parameter change by more than 0.03%, which is the aocuracy of the present measurements. The possible existence of dislocation loops will be referred to again in the following section. 3.2.
X-RAY
LINE BROADENING
If the corrected width p of an X-ray diffraction peak is measured as a function of the Bragg angle 0, a plot of B cos 8 against sin 8 may be used to estimate the mean internal
0
1.0
0.5 Sin
0
Fig. 2. Plots of j9 co9 0 versus sin 0 for pyrolytic silicon carbide (a) unirradiated, (b) irradiated to 3.8x 1021 nvt (E>0.18 MeV) at 1040 “C, and (c) irradiated to 2.7~ 1021 nvt (E>O.18 MeV) at 460 “C. /I = corrected peak width, 0 = Bragg angle.
strain 15). For material where particle-size broadening is negligible, such a plot gives a straight line through the origin whose slope is equal to 2127~ times the root-mean-square (RMS) strain. Line broadening in pyrolytic silicon carbide is all attributable to internal strain 7). When the corrected peak widths from the silicon carbide irradiated in the present experi-
EFFECTS
OF FAST-NEUTRON
21
IRRADIATION
TABLE3 Mechanical property
Cell no.
1 2 and 3
changes in irradiated pyrolytic
Mean irradiated property
Mean Neutron exposure 1 temperature , (n/cm21 j (E>0118 MeV) 1 i0c, 2.8 x loal 2.8 x 1021
Xean unirradiated property >
-j&D.
I-___ __.~.
/Modulus of rupture/ Young’s
630 1020
ments were analyzed in this way it was found that the RMS strain increases slightly during irradiation. Examples of plots of p cos 0 against sin 0 before and after irradiation are shown in fig. 2. The increase in internal strain during irradiation was obtained from the irradiated peak widths by using the peak widths of unirradiated material in place of potassium chloride as the standard. The irradiation-induced line-broadening was assumed to be caused solely by internal strain, i.e. the lines in plots 2(b) and 2(c) were drawn through the origin. However, the scatter in the data is such that some particle-size broadening could have been present; this would correspond to the lines making a positive intercept on the ordinate axis. The increase in RMS internal strain for various irradiation conditions is shown in table 2. The increase in strain was very small in all cases, and was less than the strain in the as-deposited material (10-s). Since the crystal structure of ,?Gsiliconcarbide is cubic, there is no possibility of intergranular stresses arising from anisotropic dimensional changes of the grains. Single point defects or small defect clusters only increase the X-ray background intensity without causing line broadening. Therefore it is probable that the Iine broadening in the present experiments arises from the presence of dislocation loops ra)_ As stated in the previous section, the experimental uncertainty in the measurement of the lattice and bulk expansions allows for the possibility th&up to about 10-3 of the lattice atoms may be present in interstitial loops.
silicon carbide
1.18 & 0.19
1
1.04 & 0.25
3.3.
/
modulus
0.98 & 0.06 1.03 -& 0.05
MECHANICAL PROPERTIZS
Young’s modulus and the modulus of rupture of groups of samples irradiated in capsule P-13-F to 2.8 x 1021 nvt (_E>O.lS MeV) at 630 and 1020 “C were measured. The ratios of the irradiated values to the unirradiated values are shown in table 3. There was no significant change in mechanical properties after irradiation. The lack of any strength reduction is not surprising in view of the cubic structure and the absence of a mechanism for the generation of large intergranular strains. Also the amount of helium produced by (n, a) reactions in silicon carbide with the low impurity content of the present material would be very small.
4.
conclusions
I. Pyrolytic ~-silicon carbide irradiated to 2.0 to 4.2x 1021 nvt (E>0.18 MeV) at temperatures between 460 and 1040 “C undergoes an isotropic expansion. The linear expansion decreases from 0.35% at 460 “C! to 0.05~* at 1040 “C. 2. The increase in lattice parameter agrees with the increase in macroscopic dimensions within the accuracy of the measurements. 3. Irradiation causes a slight increase in X-ray peak width. The change is attributable to an increase in RMS internal strain of about 5 x 104. 4.
Irradiation causes no significant change in the elastic modulus or modulus of rupture of the samples.
R.
22
J.
PRICE
Acknowledgements The
author
is grateful
to
W.
H.
Ellis,
B)
17. J. Gagnon, and J. M. Dixon for experimental help, R. (3;. Mills for supervising the preparation of the material, Irradiation
and members
Group who constructed and operated
the irradiation capsules. The work was supported by the U.S. Atomic Energy Commission under Contract No. AT(04-3)-167, Project Agreement No.
7)
of the Capsule 1; 10)
17.
References 1) G. Hauser, Energ. Nucl. (Paris) 7 (1965) 100 2) J. L. Jackson, USAEC Report No. BNWL-
242 (Battelle-Northwest Laboratories, Richland, Washington 1966) 3) W. Primak, L. H. Fuchs and P. P. Day, Phys. Rev. 103 (1956) 1184 4) M. Bela&, Phys. Status Solidi, 11 (1965) K 67 Short Notes -7 N. F. Pravdyuk, V. A. Nikolaenko, V. I. Kapuchin and V. N. Kusnetsov, Properties of Reactor Materials and the Effects of Radiation Damage
12) 13)
14) 15) 16)
Proceedings, (ed. D. J. Littler; Butterworths, London 1962) p. 57 R. P. Thorne, V. C. Howard and B. Hope, Proc. Brit. Ceram. Sot. 7 (1967) 449 R. J. Price, USAEC Report No. GA-8883 (Gulf General Atomic Incorporated, San Diego, Calif., 1968) T. D. Gulden, J. Am. Ceram. Sot. 51 (1968) 424 R. E. Vogel, and C. P. Kempter, Acta Crystallogr. 14 (1961) 1130 F. W. Jones, Proc. Roy. Sot. (London) Ser. Al66 (1938) 16 H. P. Klug and L. E. Alexander, X-ray Diffraction Procedures for Polycrystalline and Amorphous Materials (John Wiley & Sons, New York, 1954) 494 J. C. Bokros, and R. J. Price, Carbon 3 (1966) 503 F. D. Carpenter, J. 0. Barner, B. B. Spillane and W. P. Wallace, International Symposium on Developments in Irradiation Capsule Technology, p. 4.2.1 (ed. D. R. Hoffman) (USAEC Report CONF-660511, Washington, D.C.) (1966) R. S. Wilks, J. Nucl. Mat. 26 (1968) 137 G. K. Williamson and W. H. Hall, Acta Met. 1 (1953) 22 M. A. Krivoglaz and K. P. Ryaboshka, Phys. Metals Metallogr. 16 no. 5 (1963) 1
OF NUCLEAR
33
MATERIALS
17-22.
(1969)
EFFECTS OF FAST-NEUTRON
0 NORTH-HOLLAND
IRRADIATION
PUBLISHINQ
ON PYROLYTIC
CO., AMSTERDAM
SILICON CARBIDE
R. J. PRICE Gulf General
Atomic
Incorporated,
John
Jay
San Diego, Received
Samples
of pyrolytic
were irradiated at
decreased from
between
showed
460
an isotropic
with
increasing
and
parameter
scopic accuracy. X-ray
changes
dimensional Irradiation
caused
line broadening
generation
of internal
change in Young’s
were
changes that
“C.
qui Btait attribuable internes.
to the macroexperimental increase
was attributable
Proben wurden
de carbure
4,2x
1021 nvt
comprises
(E>0,18
entre
presentaient
une
460
MeV) et
aus
avec la temperature
de 0,35%
8. 460 “C a 0,05%
parametres
1.
Introduction
“C.
Les
d’irradiation
qui
und Temperaturen
zwischen zeigten
Btaient
Bgales
die mit steigender
x 1021 nvt,
B-Sic
0,18 MeV)
460 und 1040 “C unter-
eine isotrope
lineare
Aus-
Bestrahlungstemperatur
der
Messgenauigkeit
stimmen
Dimensionsanderungen
iiberein.
lung
Verbreiterung
trat
eine
leichte
zurtickgefiihrt
aux
wird.
die Pinde-
mit den makroskopischen Durch
reflexe ein, die auf die Entstehung
croissante,
a 1040 “C. Les variations
reticulaires
polykristallinem
(2,0-4,2
sucht. Die Proben
Innerhalb
Bchantillons lineaire
pyrolytischem
rungen der Gitterparameter
temperatures
isotrope
significative
de rupture.
von 0,35 y0 bei 460 “C auf 0,05 O/obei 1040 “C abfiel.
,!l pyrolytique
a des doses de 2 a 8. des
1040
dilatation
diminuait, des
de silicium
ou du module
mit Neutronen
dehnung, ont 6th irradies
de deformations
pas de variation
in
to the
modulus or the modulus of rupture.
Des echantillons
d’Young
erreurs un leger
des raies des rayons X
a la creation
avait
aux
produisait
de l’elargissement
11 n’y
du module
strains. There was no significant
et polycristallin
Science,
macroscopique,
L’irradiation
accroissement
equal
slight
p&s.
The
within a
de dimension
that
1040 “C. The
and Applied
1969
d’experience
temperature
at
for Pure
USA
variations
MeV)
expansion
irradiation
0.35 y0 at 460 “C to 0.05%
lattice
1040
linear
Laboratory
27 February
/l-silicon carbide
to 2.0 to 4.2 x 1021nvt (E>O.lS
temperatures
samples
polycrystalline
Hopkins
California 92112,
BestrahRontgen-
innerer Spannungen
Bezeichnende
Anderungen
Elastizitiits-
oder Bruchmoduls
traten
pyrolytic
silicon
are
carbide
die der
nicht
des
auf.
therefore
of
considerable interest. Previous work has dealt with the effects of low neutron exposures (less than 102i nvt) on the dimensions and lattice parameters of single crystals of oc-silicon carbide s-5). The effects of intermediate neutron exposures (up to 1.8 x 1021 nvt) at 250, 475, and 700 “C on the dimensions, lattice parameters, thermal diffusivity, and Young’s modulus of low-density self-bonded polycrystalline /?-silicon carbide have also been reported 6). In the present work, pyrolytic, polycrystalline /l-silicon carbide was irradiated to 2.0 to 4.2x 1021 nvt (E>0.18 MeV) at temperatures between 460 and 1040 “C. The
Pyrolytic silicon carbide is used as a coating for the particles of uranium or thorium oxide or carbide that form the fuel for high-temperature gas-cooled nuclear reactors 1). The properties of silicon carbide that make it suitable for use as a fuel-coating material are its hightemperature strength, low neutron absorption cross-section, and low permeability to fission products. Pyrolytic silicon carbide has also been considered as a coating to protect against the accidental oxidation of moderator graphite in high-temperature nuclear reactors 2). The effects of fast-neutron irradiation at high temperatures on the structure and physical properties of 17
18
R.
changes
in dimensions,
density,
meter, X-ray line-broadening, and modulus structure
of rupture
and physical
lattice
J.
para-
Young’s modulus,
were measured. properties
The
of the un-
irradiated material are reported elsewhere 7).
PRICE
diffraction Norelco
measurements diffractometer
Experimental
was then
computed
for systematic
The material was prepared by the decomposition
of methyltrichlorosilane-hydrogen
mix-
made
with
copper
a
Ka
radiation, and traces were obtained for peaks with indices up to (333). The lattice parameter from
using an extrapolation 2.
were using
the
the peak positions
procedure
that corrects
errors in sample positioning
di~ractometer
geometry 9). X-ray
in
peak
widths were measured and corrected for doublet
tures in a fluidized bed of particles at 1400 “C. The total gas flow rate was 10 000 ml/min with
and
instrumental
broadening
using
Jones’
carbide was out into strips ~neasuring about 0.6 cm x 0.1 cm x 0.01 am. The lengths of the strips were measured with
carbide strip samples were then estimated by calculating their difference in temperature from the fuel samples. The mean sample temperatures are considered accurate to within i 50 “C and in addition the temperatures were subject to fluctuations of & 50 “C during the course of the irradiation. The neutron fluxes were calculated from the analysis of nickel and iron dosimeter wires.
method 10). A potassium chloride standard was used. The correction curves appropriate for a 5 to 10% of the hydrogen carrier gas passing line profile intermediate in shape between a througl~ a n~ethyltrichlorosilane bubbler at Gaussian and a Lorenzian and for a diffractoroom temperature. The fluidized bed was meter with a line source were used ii). contained in a 3.8 cm diameter graphite tube The modulus of rupture and Young’s modulus and the initial bed surface areas were between were measured in 3-point bending at room 500 and 2000 cma. The reactant input flux was between 2 and 10 x IO-6 moles of silane~rni~~~crnz temperature, using techniques described previously 12). of bed surface. Other details of the deposition About ten strip samples were included in two technique are given elsewhere 8). or three cells in each of four irradiation The material consisted of /I-silicon carbide capsules. The neutron exposures and mean and had a density of 3.17 to 3.20 g/cm3 operating temperatures of each group of samples (theoretical density: 3.21 g/cma). The grain size are shown in table 2. The irradiations were was less than 1 pms). X-ray diffraction patterns carried out in the Engineering Test Reactor showed no evidence for excess silicon or carbon. (ETR) at Idaho Falls, Idaho, in capsules whose A typical impurity analysis is shown in table 1. primary purpose was to test coated-particle Other structural properties are reported elsefuel, and whose design has been described 13). where 7). The operating temperatures of the coatedSamples for irradiation were obtained by particle fuel samples were monitored with placing graphite discs 0.7 cm diameter x0.1 cm thermocouples and regulated by control of the thick in the fluidized bed. After coating, the gas composition in heat-transfer gaps around discs were cut with a diamond wheel and the the crucibles. The temperatures of the silicon graphite removed by grinding. The silicon
a machinist’s microscope to an accuracy of 0.02%. Densities were measured by suspension in methylene iodide-benzene mixtures. X-ray
TABLE
Spectrographic
Mn 4 *
Measured
1
impurity analysis of pyrolytic carbide (ppm) Mg 8
Fe 20
Ni (10
Al
Cu
Cl*
60
20
87
by neutron activation
silicon
analysis.
3.
Results and discussion
3.1.
CHARGES
IN
DIiVfEBSIONS
AND
LATTICE
PARAMETER
The mean linear dimensional
expansions
and
EFFECTS
OF
FAST-NEUTRON
19
IRRADIATION
TABLE 2 Change in linear dimensions, lattice parameter and X-ray Irradiation
P-13-F
I
1
P-13-H P-13-J P-13-K
y0
S.D.)
(f
Increase in RMS internal
Mean 1 temperature
Cell no.
of silicon carbide during irradiation
Mean expansion,
conditions
Veutron exposure Capsule no.
line-broadening
Linear
Lattice
strain
dimensions
parameter
(x 104)
!
(“Cl
(
630
0.24 f
0.02
0.20 f
0.02
5.0
2.8 x 1021
1020
0.08 i
0.04
0.03 * 0.02
4.0
2.7 x 1021
1010
0.06 & 0.04
0.05 f
0.02
3.0
3.8 x 1021
700
0.30 + 0.02
0.33 f
0.02
3.5
4.2 x 1021
980
0.07 * 0.05
0.07 f
0.01
3.5
3.8 x 1021
1040
0.06 f
0.03
0.10 + 0.04
7.0
2.7 x 1021
460
0.36 f
0.03
0.34 & 0.04
3.5
2.7 x 1021
620
0.23 f
0.03
0.19 f
0.03
5.0
2.0 x 1021
1010
0.03 f
0.04
0.05 f
0.02
3.0
2.8x
1021
1
i
changes in lattice parameter are shown in table 2. The standard deviations of the mean
Lattice Parameter
Source
0
macroscopic expansions were obtained from the experimental spread among about ten similar strips irradiated under each set of conditions. Lattice-parameter measurements were made after lumping the similar strips together, and the standard deviation of the estimate was obtained from the curve-fitting statistics of the extrapolation procedure. The lattice parameter expansions were the same as the macroscopic
Pyrolytic
1
0.
3:17
p=
- 3.20
[present work]
a Single
cl
[Primak
Crystals
er~_I.~)]
A
a Single
Crystals
0
a Single
Crystals
[Ealarin4)]
[Pravdyuka_al.‘)] Self-bonded
V
6.
p-
2.1
-
2.2
[Thorne&A_.6)]
expansions within the accuracy of the measurements. Measurements of the change in density showed that the changes were isotropic. Previous work 5, a) has shown that radiation-induced
expansion
the
of silicon carbide
saturates and becomes independent of exposure after a low neutron exposure (1 to 3 x lo20 nvt). The expansion level decreases with increasing irradiation temperature. The expansions obtained in the present work are plotted as a function of irradiation temperature in fig. 1, together with other values reported in the literature for exposures high enough for saturation to have occurred. The linear expansion decreases from about 1% for irradiations near room temperature to about 0.05% at 1000 “C. There does not appear to be any systematic dependence on the gram size or porosity of the material, and 01 (hexagonal) silicon carbide behaves similarly to fi (cubic) silicon carbide.
”
400
200
Irradiation
Fig.
1.
Saturation
600
800
temperature
(‘C)
radiation-induced
IO00
expansion
of
silicon carbide as a function of irradiation temperature (neutron
exposures > 1020 nvt).
There is no measurable difference between the lattice expansion and the macroscopic expansion. The saturation of the expansion as the neutron exposure increases, and the monotonic decrease in expansion with increasing irradiation temperatures are typical of the behavior of isotropic or near-isotropic ceramic materials 14). Irradiation with fast neutrons creates equal
20
R.
numbers of interstitials
and vacancies,
in various
stages
occurs
Saturation
defect production retention
I
400
311 220
331
222
I
422
420
333
changes
rate equals the
rate. The amount of damage
decreases with increasing
temperature
200 111
property
of the property
when the annihilation
Reflection
The
of aggregation
and are responsible for the observable changes.
PRICE
most of
which are removed by mutual annihilation. rest remain
J.
because
irradiation
the annihilation
rate is
thermally activated, while the defect production rate is fixed by the neutron flux. Because the dilatation associated with an interstitial atom
7
is greater than the relaxation around a vacancy, an overall expansion results. Single point defects
c
and small point-defect clusters change both the lattice parameter and the bulk dimensions whereas large clusters of defects equally, condense to form dislocation loops that cause a larger fractional change in the bulk dimensions than in the lattice parameter. In the present work the experimental results do not show a significant difference between the lattice para-
0
b-
"0
(b)
Irradiated at 1040°c
to 3.8
X
IO
21
"Yt
0 /
4
x 0
3
B 92
I 0 6
meter expansion and the macroscopic linear expansion (table 2). The lack of a difference indicates that most of the point defects are present as single defects or small clusters.
(c)
Irradiated
to 2.7
X
IO
However, those data do not exclude the possibility that dislocation loops also are formed. Both interstitial and vacancy loops could be present without changing the equality between the
bulk
provided
and
lattice
dimensional
changes,
that the total loop area is equal for
each type.
If, as seems more probable,
inter-
stitial loops predominate, up to 10e3 of the atoms could be located in such loops before the resulting linear expansion would exceed the lattice parameter change by more than 0.03%, which is the aocuracy of the present measurements. The possible existence of dislocation loops will be referred to again in the following section. 3.2.
X-RAY
LINE BROADENING
If the corrected width p of an X-ray diffraction peak is measured as a function of the Bragg angle 0, a plot of B cos 8 against sin 8 may be used to estimate the mean internal
0
1.0
0.5 Sin
0
Fig. 2. Plots of j9 co9 0 versus sin 0 for pyrolytic silicon carbide (a) unirradiated, (b) irradiated to 3.8x 1021 nvt (E>0.18 MeV) at 1040 “C, and (c) irradiated to 2.7~ 1021 nvt (E>O.18 MeV) at 460 “C. /I = corrected peak width, 0 = Bragg angle.
strain 15). For material where particle-size broadening is negligible, such a plot gives a straight line through the origin whose slope is equal to 2127~ times the root-mean-square (RMS) strain. Line broadening in pyrolytic silicon carbide is all attributable to internal strain 7). When the corrected peak widths from the silicon carbide irradiated in the present experi-
EFFECTS
OF FAST-NEUTRON
21
IRRADIATION
TABLE3 Mechanical property
Cell no.
1 2 and 3
changes in irradiated pyrolytic
Mean irradiated property
Mean Neutron exposure 1 temperature , (n/cm21 j (E>0118 MeV) 1 i0c, 2.8 x loal 2.8 x 1021
Xean unirradiated property >
-j&D.
I-___ __.~.
/Modulus of rupture/ Young’s
630 1020
ments were analyzed in this way it was found that the RMS strain increases slightly during irradiation. Examples of plots of p cos 0 against sin 0 before and after irradiation are shown in fig. 2. The increase in internal strain during irradiation was obtained from the irradiated peak widths by using the peak widths of unirradiated material in place of potassium chloride as the standard. The irradiation-induced line-broadening was assumed to be caused solely by internal strain, i.e. the lines in plots 2(b) and 2(c) were drawn through the origin. However, the scatter in the data is such that some particle-size broadening could have been present; this would correspond to the lines making a positive intercept on the ordinate axis. The increase in RMS internal strain for various irradiation conditions is shown in table 2. The increase in strain was very small in all cases, and was less than the strain in the as-deposited material (10-s). Since the crystal structure of ,?Gsiliconcarbide is cubic, there is no possibility of intergranular stresses arising from anisotropic dimensional changes of the grains. Single point defects or small defect clusters only increase the X-ray background intensity without causing line broadening. Therefore it is probable that the Iine broadening in the present experiments arises from the presence of dislocation loops ra)_ As stated in the previous section, the experimental uncertainty in the measurement of the lattice and bulk expansions allows for the possibility th&up to about 10-3 of the lattice atoms may be present in interstitial loops.
silicon carbide
1.18 & 0.19
1
1.04 & 0.25
3.3.
/
modulus
0.98 & 0.06 1.03 -& 0.05
MECHANICAL PROPERTIZS
Young’s modulus and the modulus of rupture of groups of samples irradiated in capsule P-13-F to 2.8 x 1021 nvt (_E>O.lS MeV) at 630 and 1020 “C were measured. The ratios of the irradiated values to the unirradiated values are shown in table 3. There was no significant change in mechanical properties after irradiation. The lack of any strength reduction is not surprising in view of the cubic structure and the absence of a mechanism for the generation of large intergranular strains. Also the amount of helium produced by (n, a) reactions in silicon carbide with the low impurity content of the present material would be very small.
4.
conclusions
I. Pyrolytic ~-silicon carbide irradiated to 2.0 to 4.2x 1021 nvt (E>0.18 MeV) at temperatures between 460 and 1040 “C undergoes an isotropic expansion. The linear expansion decreases from 0.35% at 460 “C! to 0.05~* at 1040 “C. 2. The increase in lattice parameter agrees with the increase in macroscopic dimensions within the accuracy of the measurements. 3. Irradiation causes a slight increase in X-ray peak width. The change is attributable to an increase in RMS internal strain of about 5 x 104. 4.
Irradiation causes no significant change in the elastic modulus or modulus of rupture of the samples.
R.
22
J.
PRICE
Acknowledgements The
author
is grateful
to
W.
H.
Ellis,
B)
17. J. Gagnon, and J. M. Dixon for experimental help, R. (3;. Mills for supervising the preparation of the material, Irradiation
and members
Group who constructed and operated
the irradiation capsules. The work was supported by the U.S. Atomic Energy Commission under Contract No. AT(04-3)-167, Project Agreement No.
7)
of the Capsule 1; 10)
17.
References 1) G. Hauser, Energ. Nucl. (Paris) 7 (1965) 100 2) J. L. Jackson, USAEC Report No. BNWL-
242 (Battelle-Northwest Laboratories, Richland, Washington 1966) 3) W. Primak, L. H. Fuchs and P. P. Day, Phys. Rev. 103 (1956) 1184 4) M. Bela&, Phys. Status Solidi, 11 (1965) K 67 Short Notes -7 N. F. Pravdyuk, V. A. Nikolaenko, V. I. Kapuchin and V. N. Kusnetsov, Properties of Reactor Materials and the Effects of Radiation Damage
12) 13)
14) 15) 16)
Proceedings, (ed. D. J. Littler; Butterworths, London 1962) p. 57 R. P. Thorne, V. C. Howard and B. Hope, Proc. Brit. Ceram. Sot. 7 (1967) 449 R. J. Price, USAEC Report No. GA-8883 (Gulf General Atomic Incorporated, San Diego, Calif., 1968) T. D. Gulden, J. Am. Ceram. Sot. 51 (1968) 424 R. E. Vogel, and C. P. Kempter, Acta Crystallogr. 14 (1961) 1130 F. W. Jones, Proc. Roy. Sot. (London) Ser. Al66 (1938) 16 H. P. Klug and L. E. Alexander, X-ray Diffraction Procedures for Polycrystalline and Amorphous Materials (John Wiley & Sons, New York, 1954) 494 J. C. Bokros, and R. J. Price, Carbon 3 (1966) 503 F. D. Carpenter, J. 0. Barner, B. B. Spillane and W. P. Wallace, International Symposium on Developments in Irradiation Capsule Technology, p. 4.2.1 (ed. D. R. Hoffman) (USAEC Report CONF-660511, Washington, D.C.) (1966) R. S. Wilks, J. Nucl. Mat. 26 (1968) 137 G. K. Williamson and W. H. Hall, Acta Met. 1 (1953) 22 M. A. Krivoglaz and K. P. Ryaboshka, Phys. Metals Metallogr. 16 no. 5 (1963) 1