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MICROCOPY RESOLUTION TEST CHART
NATIONAL BUREAU OF STANDARDS -1963
PAN
N
CORNV-2329
MASTER
CURRENT STATUS OF IRRADIATION TESTING OF THORIUM FUELS AT OAK RIDGE NATIONAL
LABORATORY*
CHOTI [POSS
A. R. Olsen and J. H. Coobs
Metals and Ceramics Division
J. W. Ullmann
Chemical Technclogy Division
Oak Ridge National Laboratory
Oak Ridge, Tennessee
00. Vi
AUG 10 1966
2.00; MN_.50
CONF660524-10
INTRODUCTION
Knowledge of the irradiation behavior of fuel materials is one of the
basic requirements in the design of reactors for economical power production.
Evaluation studies?, 2 of existing designs utilizing thorium indicate that
only the molten salt converter and gas-cooled reactors currently are
competitive with reactors fueled with natural or low enrichment uranium
fuels. Other studies3,4 indicate that thorium-base fuels using by-product
plutonium from light water power reactors for fissile enrichment, in a
heavy water moderated reactor may offer an economic means of establishing
:
the thorium-233 uranium fuel cycle. The basic material characteristics5
and the irradiation behavior of the thorium alloys anà compounds are
strongly dependent on the processing and fabrication techniques used.
1
These characteristics must be defined to provide the designers with the
2
.
*Research sponsored by the U.S. Atomic Energy Commission under
contract with the Union Carbide Corporation.

RELEASED FOR ANNCUNCEMENT
IN NUCLEAR 5 10:03 ASS FACTS
*
I la
maximum latitude in developing a specific economical design for any of
L'it
the reactor concepts.
--*
. Fr
.
.
.-*
*
The ORNL sol-gel process offers a unique means of providing pure or
*
.z.
*
.
s.
5
mixed compounds of thorium, uranium, and plutonium as oxides and carbides. 5
.
-
.
These fuels have been combined with fabrication procedures such as vibratory
compaction for metal-clad bulk oxide fuels or coating procedures utilizing
pyrolytic carbon as a fission-product retention material with and without
subsequent incorporation into fueled graphite loadings.
LEGAL NOTICE
This report was prepared as an account of Government sponsored work. Nallber the Unito
Slatos, aor the Como ission, nor say person acting on behalf of the Commission:
A. Makes any warranty or representation, expressed or implied, with respect to the accu-
racy, completeness, or usefulness of the information contained in this report, ur that the we
of any information, apparatus, method, or process disclosed in this report may not infringe
privately owned rights; or
B. Assumes any labuiues with resrect to the use of, or for damages resulting from the
use of any information, apparatus, method, or process disclosed in this report.
As used in the above, "person acting on behali of the Commission" Includes any sm-
ployoo or contractor of the Commission, or omployee of such contractor, to tão extent that
such employeo or contractor of the Commission, or employee of such contractor prepares,
disseminates, or provides acceso to, any information pursuant to die cmployment or contract
with the Commission, or bio employment with such contractor.
:
.
.
:
.
.
..
?
.
&
.
Ei
9
.
y
Thus, the full stange of processing and fabrication techniques are incorporated into a
OINI - AEC-OF!!Cini
*
material for irradiation evaluation. The irradi.ation behavior of solid thorium-
base fuels has been under investigation at ORNL for several years. Most of the
-
-
-
data have been reported as it developed. -
Inis paper will only summarize our
findings and establish the current status of the programs. The irradiation testing
can be subdivided into the two basic fuel types.
Metal-Clad Pulk Oxides
A series of noninstrumented fuel rods have been irradiated in the process
water of the Materials Test Teactor (MTR), Engineering Test Reactor (ETR), and
the Chalk River National Research Experiment Reactor (NRX) during the past six
years in the metal-clad bulk oxide program. Two more sophisticated test groups,
ANA
one with instrumented capsules and one loop bundle, were irradiated in the Oak
Ridge Research Reactor (ORR). The programa has been concentrated on the sol.-gel-
,
derived fuels with low fissile content (approximately 5% UO in Tho,) fabricated
into fuel rods by vibratory compaction.
Table 1 is a listing of the irradiation tests. While most of the origina).
irradiation tests utilized stainless steel cladding, the more recent rods have
been fabricated using Zircaloy-2 cladding materials and the tests have been
designed to evaluate the effects of semiremote fabrication and different sol-
OIHI - AIC - OISICIAL
gel calcination atmospheres. Pure thoria without any
gel calcination atmospheres. Pure thoria without any initial fissile additive
MEDVOMIYI
-
-
-
.-
-
IV121510 - DIY - IMO
?VIDiiso-23V-INIO
Table 1
SUMMARY OF THORIUM FUEL CYCLE PROGRAM IRRADIATION OF POWDER-PACKED RODS *
Y-65933
Designation
No, of
Rods
Type of Oxide
Density
(% theor.)
Linear
Heat Rating
(w/cm)
Peak Burnup
!Mwd/tonne metal)
Objective
wa A o to
MTR-1
7 Arc-fused 86 to 87 390 15,000 to 100,000 Provide base-line data to use in comparing
Sol-Gel E
sol-gel and arc-fused oxide
MTR-II *** 2 Sol-Gel S 88 to 89 600
100,000 Obtain higher heat rating by increasing
enrichment
MTR-111 *** 6 Soi-Gel 35 86 to 820
100,000
Compare oxide calcining atmospheres and
higher heat ratings obtained by increasing
diameter
ETRA
4 Sol-Gel 35 86 to 89
22,000
Same as for MTR-III
NRX-
8 Soi-Gel A & B 86 to 87 160
16,000 Provide base-line data
NRX-11
Sol-Gel C 83 to 86 . 210
5,000
Study effect of increased length
Arc-fused
NRX-111
6 Sol-Gel S 88 to 89
270
23,000 Study effect of increased length
NRX-VII
Sol-Gel .74 to 76** 260
22,000
Study ThOn-PuO2 oxide and lower packed
ThO2-puog
density
ORR Loop 3 Sol-Gel 26
: 2,100
Study in pressurized wuter at 260°C and 1750 psi
ORR Poolside 2 Sol-Gel D 85**
5,000 Measure effec.ive thermal conductivity using a
central thermocouple in Na-k at 315 psi, 540
and 705°C
6 BNL-Sol-Gel 90
30,000 to 100,000 Study effects of remote fabrication and oxide
recalcining
ETR-III
7 Soi-Gel Thoa , 88
770
10,000 to 70,000 Study Thon blanket material with gradually
increasing heat rating and provide high Pa
low-fission-product material for chemical
processing.
*All rods were clad with type 304 SS except for Groups ETR II, III, and ORR Loop, which were clad with Zircaloy-2.
**Tamp packed, all others vibratorily compacted.
odgoriche mit der Basingen
ORNL - AEC - OFFICIAL
500
340
ETRHI
030
is being irradiated in the most recent tests to investigate the effects of
increasing power level.
The initial objectives on the irradiation program were to compare the
characteristics of the chemically produced sol-gel (Thu), fuel with those of
arc-fused material and to compare the performance of vibratorily compacted rods
with others containing pressed and sintered pellets. The results of these early
comparison tests have been favorable. In Table 2, for example, we have listed
a group of tests all of which were conducted at, moderate linear heat ratings
between 300 and 460 w/cm (10 to 14 kw/ft), which is similar to current water reactor
elements using vo, as a fuel.
These tests were taken to high burnups where as
much as 70% of the fissions occurred in the bred SDU. Even with these burnups,
in excess of 100,000 Mwa/Tonne of heavy metal, the fission-gas release rates were
generally less than 20%. Estimates of the swelling of the pellet fuels based
upon the fact that there was no change in diameter or length of the self-supporting
claddings at the high burnup levels indicate the change in volume was less than
0.5%
per 102° fissions/cm). The postirradiation examination of vibratorily
compacted and pellet fuels showed essentially no difference in the macroscopic
appearance after extended irradiations. By comparing Fig. 1 with Fig. 2, where
typical cross sections of a pellet rod and a sol-gel vibratorily compacted powder
rod are shown, it can be seen that without the titles there is no easy way of
.
A
ES
F
F
...
.
.
Y-65935
85kr
x 1020
Release
(%)
2.4
292
7.2
707
267
Tables
COMPARATIVE OPERATING CONDITIONS FOR VARIOUS ThO2-UO, FUELS
Irradiation Time Maximum Time Averaged
Fuel
(Reactor Fu!! Burnup
Peak Linear
kdo
10
Power Days) (fissions/cm) Heat Rating
(w/cm) (w/cm)
Arc-fused ThO 2-4.5% UO, vicº 110
2.5
33.8
Vic
8.6
• Vic
'14.4
Sol-gel ThO2-4.5% O2 Vic
2.9
37.8
Vic
8.2
32.8
ViC.
16.5
34:6
Pressed and sintered
26.4
44.8
ThO 2-4.5% UO2
Pellet
21,1
48.2
Pellet 497
8.1
31.5
Pellet
406
11.0
42.6
6.4
375
0.5
13.2
17.0
Pellet
905
....
420
22.8
.
660
461
.
270
"Vibratorily compacted to a density 85 to 87% of theoretical.
Pellets pressed from coprecipitated powders and sintered to 93% of theoretical density.
Gas samples diluted with air in sampling and partially lost.
ORNI - AEC - OFFICIAL
-
-
..
..
-
-
.
..
..
ORNOtsexc :?2915851 .


ORNL-DWG 65-2476
0.97401
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OOD
A. METALLOGRAPHY SECTION
B. BURNUP SECTION
C. LONGITUDINAL SECTION
D. NEUTRON ACTIVATION SPECIMEN
Pellet Rod No. 712 Macroscopic Views of Selected Section Locations, as Cut.
S
2
.
ORNL-DWG 65-2480

2012
2210
--28.705 cm -
NO. END
18A
1.91cm
2.54 cm
2.54 cm
/ 4.27 cm
,1.27 cm
4.128 cm
* niny
1.2217
Rain
min
A
A. METALLOGRAPHY SECTION
B. BURNUP SECTION
C. LONGITUDINAL SECTION
D. SPARE LONGITUDINAL SECTION
E. NEUTRON ACTIVATION SECTION
MTR Group I -Capsule Z-8-Macroscopic Views of Selected Section Locations, as Cut.
videos1:10; ;;,
! SISTEM
separating the initial starting fuel form.
Both rods show typical thermally
fragmented fue...
The current irradiation program is investigating the effects oť process
parameters such as semiremote fabrication or the sol-gel calcining atmosphere,
and at the same time attempting to establish the maximum fuel performance
characteristics. In one group of tests in this series, we have succeeded in
developing central voids but with no evidence of melting. Figure 3 shows typical
macroscopic sections similar to those of the lower heat rated rods but showing
the central voids. Figure 4 is a compilation of radial micrographs from three
of the highly rated rods.
The rods opereted at peak heat retings in excess of
1000 w/cm, while the time-averaged heat ratings are those shown under each
micrograph. In these highly rated rods less than 30% of the fission gas was
released from fuels fired in nitrogen or in the siandard reducing atmosphere
of argon-4% hydrogen, while approximately 40% was released from the fuels calcined
in air.
The burnup level on these rods: was 22,000 Mwa/Tonne Th+U. There was
no evidence of swelling as determined by post-irradiation measurements of the
free-standing cladding.
The apparent effect of the firing atmosphere on increased
fission-gas release and microstructural changes will not be resolved until other
tests using identical fuels but at lower linear heat ratings are examined. These
tests are currently under irrandiation and are scheduled to go to approximately
,
.
.
.
-
.
Y
NKARA
E
.
ORNL-DWG 65-2470
V
- 30.505 cm ---
NO. END
2.54 cm
2.54 cm
2.54 cm
2.54 cmT
2.54 cm
2.54 cm 17.91cm
uz

1
A. METALLOGRAPHY SECTION
B. BURNUP SECTION
C. LONGITUDINAL SECTION
D. SPARE LONGITUDINAL SECTION
ETR Group 1 - Capsule 8 - Macroscopic Views of
Selected Section Locations, as Cut,
حی
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همه دسته سه دهه مه لهم، مممم ۱۰۰۰ وعده ماه ۰۰۰ سه ماهه 4
n?!! .
همه و همه... همه مه....امحه مه مه مهم
معه.عمه


-
e
.
100,000 Mwa/Tonne of heavy metal burnup.
ORNL - ALC - OFFICIAL
Ini
A
Since only recently has a sol-gel process for Puo, been developed, our
irradiation experience with the so-called mixed progeny fuels is limited. Three
rods containing sol-gel-derived Tho, mixed, before calcining with precipitated
2
5.2% Pub, were tamp-packed to approximate-y 75% of theoretical density and exposed at
a linear heat rating of 200 to 245 w/cm for a peak burnup of 29,000 Mwa/Tonne (Th+Pu).
The fission-gas release rates were less than 5% and the microstructures as shown in
Fig. 5 were similar to those of Tho -5% vo, exposed under the same conditions.
The center temperatures estimated by using a thermal conductivity of 0.017 w/cm,
which was the effective thermal conductivity derived in the instrimente: capsule
experiments, were approximately 1500°c in rod p-4 and 1300°c in rod P-6. There is
some incipient sintering and grain growth in the fuel at the higher heat rating
while with rod P-6, except for accentuated distributed porosity in the center, the
fuel is essentially unchanged from the preirradiation appearance.
Although the data on Tho, -base fuels are still somewhat limited, a
preliminary comparison with Uo, fuels is in order.
The data presented in Table 3
compare powder-compacted vo, and Thoq-uo, fuels using the fkde values for specific
structural changes most commonly used in comparing bulk oxide fuels.
The Uo, work
451
ORNIO AEC - OFFICIAL
was done as part of the Maritime Reactor Program and was reported previously.10 The
INTR-I data were included to provide a value for the Skde for void formation in the
ES5
*
-
2
11a
--
-
-
-
-
!
A
:
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.
3
-
-
)
...
-
-
-.
*
11131330 - )3 V - INTO
Table w
COMPARISON OF UO2 AND ThOg-UO, VIBRATORILY COMPACTED FUEL RODS
Inside
Surface
Temperature
of Cladding
Burnup
Mwd/tonne
Heavy
Element
Fuel
Material
- 85kr
Fuel
Density
Experiment
Linear
Heat
Rating
(w/cm)
pog reg
Js kdo J's kdo
(w/cm)_(w/cm) (w/cm)
Release
(%TD)
(°C)
(w/cm)
(%) _
VO2
434
12.4
ORR Loop
7NI
701
7P1
87.1
87.1
86.9
4,780
5,140
6.290
32.5
34.8
361
465
20.7
21.1
20.0
47.0
72.0
383
566
41.2
347
8N1
801
8P1
85.6
85.8
85.5
4,810
5,040
6,880
351
404
423
578
30.9
31.3
40.8
22.8
21.2
23.9
17.5
16.1
20.3
77.0
25.0
25.0
387
ThO2-6% UO2
85.2
2.3
ORR Loop
LIA
LIB
LIC
84.1
1,600
2,100
1,730
368
381
499
410
30.3
39.7
32,6
No void
No void
No void
36.6
84.1
17.7
20.7
248
3.9
ETR-1
18
ThO2-5% UO2
211
870
63,0
88.1
89.4
85.1
20,400
20,000
22,000
865
209
218
66.4
61.3
35.8
24.6
28.2
34.7
19.8
25.3
38.0
28.0
21.0
914
quo, fuel was arc-fused crushed and vibratorily compacted; oxygen-to-uranium ratio 2.002:2.003; enriched in 235U 5 to 6%.
ThO2-UO2 fuel was sol gel material crushed and vibratorily compacted; U enriched in 2300 to 93%.
pºis center to surface of fuel; m
is void to surface of fuel; reg is the limit of columnar graingrowth to surface of
JS
fuel; and
9 is the limit of discernible equiaxed grain growth to surface of fuel.
.
.
:
.
thoria-base fuels even though the surface temperature of the cladding was 150°C
.
-
>
-
-
"
ORNLOAstfi 9021KOSL
lower, the loop experiments were run in the same facility. There are uncertainties
..-
.
--
-
- -
in the peak-to-average ratios of the heat flux in these experiments. Nevertheless,
-
-
-
-
.
a comparison of the heat ratings to produce similar microstructural changes and
.
.
-
-
- -
fission gas release rates in the vibratorily compacted Thoria base fuels and the
-
.
..
-
urania fuels indicates the thoria fuels can accomodate approximately 40% higher
heat ratings.
Future tests are planned to investigate the sol-gel-derived (Th-Puno, fuels
with emphasis on low-energy vibrational compaction of microspheres.
In addition,
tests utilizing thin-walled claddings which are not free-standing are planned
to investigate the swelling characteristics of thoria-base fuels.
Coated particles
The irradiation program on coated particle · fuels is of more recent origin
than the program on bulk oxides.
The fuel has all, been in the form of small
particles (150-350 H diam) of (Th,U) C, and (Th, u)o, coated with pyrolytic
-45
carbon for fission product retention. Both loose beds of coated particles and fueled-
graphite elements containing coated particles have been irradiated.
Much of the
5.-
irradiation testing has been designed to evaluate the effects of changing coating
ORHI - AEC - OFFICIAL
parameters on the particle behavior as discussed by Prados et al. in another paper
wa
.
M
t
at this symposium.dIn order to obtain accelerated burnup, high concentrations
...;"
...hr
.
21
.
3.
V
IN
of uranium have been used for most of the test with this fuel form.
The dicarbide
particles have generally been obtained from the carbon bed melting process, while
the oxides are from conventional sintering processes or the sol-gel microsphere
process.
Table 4 is a partial listing of the tests utilizing fuel kernels containing
thorium dícarbide. Some properties of these coated particles are given in Table 5.
There were no observed coating failures in any of these tests. As can be seen
from Table 4, the irradiation test temperature has varied from approximately
1000°c up to a maximum of 1500°C and the heavy metal fissioning has been carried
to as high as 27 at. % for the low thorium-to-uranium ratio particles. Testing
of the higher thorium to uranium ratios involves longer periods of exposure and
only recently have coated fuel particles with the higher thorium-to-uranium ratios
of 2.2 and 4.9 been tested at significant levels of fissioning. Preirradiation
and postirradiation micrographs of batch GA-314, which has a thorium-to-uranium
:
ratio of 2.2:1, are shown in Fig. 6. After 8 at.% heavy metal fissioning the
two outer layers of the triplex coating appear unchanged, indicating that the
porous inner coating has absorbed all direct fission recoil damage and also
accommodated any swelling of the fuel particle. Such good behavior would probably
be predicted by the mathematical model discussed by Prados -- if detailed properties
of the individual layers of the coating were well documented.
SA
.
.
Y-65934
ORNL -MEC - OFFICIAL

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MLADIATION ASI CONDENONS ANO FISSION OASIAN
DAIA FOI QAND IT, UXPunais
:.
:
Shalom
:.
Y-71303
UNCL
LS 13989
1.3.104
C
18
1173
1125
130
1800
- 1000
ind
NCC-214
MOC
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NCC-100
MCC 30
MOC-210
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1400
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ORNL - AEC - OFFICIAL
.
.
1.2
TATTOOS
Table 5. Continued.
Particle Batch Member
NCC-222
HB-23
OR-182
OR-205
OR-206
Core material
102
Ta02
(ThỄU)O2
3.1
17.8
(Th,U) C2
5.55
25.4
93
(Th, UFO3
3.3
39.8
33.8
36.3
93
97
93
407
500
406
148
243
206
129
128
Uranium content, wt%
Thorium content, wt %
Enrichment, at. %
Particle dimensions, u
Total particle, av
Core particle, av
Coating thickness, av
Coating structure
Pour density, g/cm3
Particle weight, av (8 x 10-4)
Coated-particle density," g/cm3
Crushing load, av (g)
Surface contamination, % U x 10-3
Fuel removal by leaching
% U X 10-3
Duplex
3
437
214
112
Duplex
1.66
1.15
2.67
1780
8.7
1.7
Triplex
1.54
0.82
2.43
1210
96
Duplex
1.7
1.1
a
1270
20.2
a
100
Duplex
1.8
1.3
2.94
2730
6.7
2.87
3210
-
0.16
3.1
2.3
0.01
275
0.3
2.4
enot measured
'Measured with helium pycnometer
**
.
_.":11:2
.:'
=
4
ry_
Table 5. General Information on Pyrolytic-Carbon-Coated Particles Used in Irradiation Tests
11736-121ECB
(U, Th) C2
19.9
12.2.
92.9
93
93
93
93
432
426
200
Particle Batch Number NCC-208 NCC-210 NCC-214 GA-310 GA-314
Core material
(U,Th)C2 (U, Th) C2 (U,Th) C2 (U, Th) C2 (U, Th) C2
Uranium content, wt%
19.6 - 23.4 23.0
14.4 18.9
Thorium content, wt%
13.4 13.6 13.4 9.0 38.7
Enrichment, at. %
93
Particle dimensions, u
Total particle, av
445
424 470 550
Core particle, av
213
216
218 175
349
coating thickness, av
117
108
103
148
101
Coating structure
Duplex Duplex Duplex Triplex Duplex
Pour density, g/cm3
1.6
1.8
1.7
1.5
2.2
Particle weight, av (8 x 10-4) 1.2 1.2
1.2 1.4
3.5
Coated-particle density, g/cm3 2.79 2.95 2.80 2.37 3.58
Crushing load, av (8)
1473
1320 1456
1598
1531
Surface contamination,
% U x 10-3
1.8 3.9
2.0 1 2.4 7.9
Fuel removal by leaching,
% U X 10-3
<0.5 0.5 17.0 0.2 0.?
113
Triplex
a
a
2.70
1290
:
-
AR
Perhaps the most significant development in the area of coated fuels during
the past year has been the excellent performance of pyrolytic-carbon coated sol-gel-
derived oxide microspheres. The depos:ition conditions used for coating sol-gel
oxide microspheres are given in Table 6, while properties of the coated particles
were presented in Table 5.
Table 7 is a listing of the irradiation tests on these
fuels which have been completed to date.
The thorium-to-uranium ratio in the
mixed oxide particles has been 12 to 1 and, consequently, the fuel burnup has
been restricted. However, the test of fully enriched vo, sol-gel microspheres
coated with a thick three-layer coating, which was exposed at 1600°C to a burnup
of 25 at. % heavy metal fissioned, clearly shows the excellent performance of
the oxide particles. Again there were no observed failures in any of these tests.
of greater importance than the lack of any failed particles is the fact that
the fission-gas release rates, as defined by the R/B ratios, are as low or slightly
lower than those from the coated carbide particles shown in Table 4. We attribute
the lower release rates to the reduced coating contamination normally observed with
carbon-coated oxide particles. Contamination of the coatings by fuel usually
occurs during deposition of the high-density outer coating at high temperatures
(31800°c).
The fission-gas release rates from many previous experiments have
_
indicated a correlation with coating contamination. In a recent series of sweep-
capsule irradiations using both coated to, and (Th U)c, particles we were able to

RNL - AEC - OFFICIAL

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Y-71302
Y-71
HADIATION TEST CONDITIONS AND PASSION GAS RUAS DATA
FOR UPISUPPORTED COATED SOL-GEL OXIOC MICROSIERS
om
aple Type of fuel Casting
-
Designation
Temperature Hoery Motel
la
Particle Structure
Fliend
01-27 . (Th,U10, Duplex
0.4
OR-706 (th, U10, Duplex 120g
2.7
OR-205 Thon Dupler
OH-IR (Th,Uso, Duplex 170 0.23 .
HB-23 NO₂ Potetor 1600 2 5
1.0.10..
120
2.3x 10+
2.5x 1005
LS 13988
1000
-
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Thill rotto in all mined fusto = 12:1.
"Not insoared - static experiment. .
-
TAPE 7.
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Tablo 6. Deposition Conditions of Coatings on Sol-Gel Oxide Microspheres
Sample I
OR-20%
pation
OR-206
OR-182
HB-23
(Th, Uloz.
Thoz
(Th, uloga
voz
217
Type of fuel particle
Average fuel particle
diameter, u
First Layer
Coating temperature,°c
243
206
148
1400
1400
1400
980-1050
CA4 flow rate, cm°min-1
0.83
0.83
2.3
0.15
54
50
1800
1800
1800
1800
1800
1800
2320
1320
Coating thickness, h
Second Layer
Coating temperature, °C
CH_flow rate, cm3 min-
cm-2
Coating thickness,
Third Layer
Coating temperature, °C
CH4 flow rate, cm*min-1
0.33
0.17
0.17
1.20
20
1900
cm-2
0.05
Coating thickness,
Particles contain 8 wt% , highly enriched
"Acetylene was used for inner coating on HB-23
4
:
-
verify that the observed release is a direct function of the amount of fissiona'yle
material in the coatings, as shown in Fig. 7. Such a correlation applics only if
no particles are broken.
The thermal stability of pyrolytic-carbon coated oxides is remarkably good
at temperatures much higher than the irradiation tests listed in Table 7.
An
extensive series of heat treatments on coated Tho, and vo, particles at 1900°c
and 2000°c demonstrated that high-density outer coatings successfully contain
CO formed by reaction of the oxides with carbon.
Since the reaction is thus
inhibited, little or no carbide is formed and migration of fuel into the coating
is virtually undetectable.
fuel migration
The striking difference in/behavior of vo, and UC,
with similar coatings is illustrated in Fig. 8 (x-71686). Both types of particles
have porous inner coatings (density < 1.0 glcm) deposited from acetylene.
The obvious process advantage of being able to handle the uncoated particles
without the use of an inert atmosphere, when combined with the lower fuel swelling
of the oxide, indicates a potential advantage for these fuels. The basis for
predicting reduced swelling as developed from the examination of the bulk oxide
fuel has been discussed elsewhere.
This same discussion indicates an advantage
for thorium-containing fuels over those containing only fissile uranium.
These
fuel characteristics together with the variety of coating parameters have been
.
i
.
included in the model discusseä in the paper by Prados et al, which describes
7
...
Fig. A Relationship Between Fission-Gas Release and Coating-
Surface Contamination.
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ORNL-OWG 65-12343
STTT
CCA SULE C1-15, BATCH GA-314,
IRRADIATED AT 1400°C —

_
CAPSULE 89-26, BATCH OR-354,
IRRADIATED AT 1350°C -
ď
=
R/B FOR 88kr, RATIO OF RELEASE RATE TO BIRTH RATE
-CAPCULE CI-16, BATCH OR - 343,
IRRADIATED AT 1400°C
-CAPSULE 89-27, BATCH OR-348,
IRRADIATED AT 1500°C
fiume.
1000
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1005
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CONTAMINATION Ig of uranium in coating) BASED ON ALPHA COUNT
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Pyrolytic-Carbon Coated UO2,
Heat Treated 500 hr at 2000°C
Pyrolytic-Carbon Coated UCZ,
As-Coated at 2000°C (8 hor)
Comparison of Euel Migration at 2000 °C. in
Payotiy tie Carbon Cartet vos ne uca Parte NL - AEC - OFFICIAL

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the prediction of coated particle performance. A series of noninstrumented capsule
2000 °C. in
NL - AEC - OFFICIAL
irradiations was recently initiated to test the model. The results of postirradiation
examination of the first capsule in the series have been included in that paper.
A later experiment in the series is specifically designed to investigate the
effects of the thorium-to-uranium ratio on fuel particle behavior.
Migration at
Euel
woty tell Castor Catete vais ad uca Parte
Companion of
To date, none of our irradiation tests have shown any significant difference
-
in the performance of dicarbide or oxide fuel particles regardless of thorum content
1
As we have seen in Fig. 6 and as listed in Table 4, there are (Th, U)c, particles
with both two- and three- layer coatings that have shown good performance to high
burnup levels. A direct comparison of the performance of carbide and oxide fuel
particles with and without thorium as part of the fuel is shown in Fig. 9.
In
this comparison we have chosen the only available experiment for the thorium-
uranium dicarbides which had a similar burnup to the thorium-uranium oxide fuels
and also a similar inner coating layer.
The greater damage to the coated (Th,U),
particles must mean that the laminar inner coating on these particles, while
similar in appearance and structure, has a lower density.
In the case of the
W4i
mixed-oxide particles, the irradiation experiment in which the coated (Th,U)o,
-hot
s
-
particles were tested also contained coated thoria particles. At the burnup level
ORNL - AEC - OFFICIAL
attained, we were unable to distinguish between the uranium-bearing and the pure
.
thoria particles.
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Conclusions
In summary, then, the irradiation tests on thorium fuels at Oak Ridge
...
Natiomal Laboratory closely coupled with process variables and fuel characteri-
.
zations, indicate that to date:
i
The metal-clad bulk oxide fuel rods using vibratorily compacted
sol-gel derived thoria-base fuels with less than 5% fissile additions will
operate at moderate heat ratings to burnups in excess of 100,000 Mwa/Tonne
..
of Th+U with no evidence of breakaway swelling or sudden increases in fission-
-
•
gas release, and that linear heat ratings in excess of 2000 w/cm can be accom-
modated without central melting or high fission-gas-release rates.
Secondly, the coated-particle irradiations show no significant difference
in the performance of fuels with and without thorium. Pyrocarbon coated oxide
fuels perform as well as, and possibly better, than coated dicarbide fuels at
burn up levels as high as 25 atom percent fissioning.
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11
References
1. Rosenthal, M. W., Adams, R. E., Bennett, L. L., Carter, W. L., Douglas, D. A. Jr.,
Hoskins, R. E., Lawson, C. G., Lotts, A. L., Olson, R. C., Perry, A. M. ,
Roberts, J. T., Salmon, R., and Vondy, D. R., A Comparative Evaluation of
Advanced Converters, USAEC Report ORNL-3686, Oak Ridge National Laboratory,
January, 1965.
2. Rosenthal, M. W., Bauman, H. F., Bennett, L. L., Carlsmith, R. S., and
Vondy, D. R., "The Technical and Economic Characteristics of Thorium Reactors,"
paper presented at the IAEA Panel on "Utilization of Thorium in Power Reactors,"
Vienna, Austria, June 14-18, 1965.
3. Duret, M. F., and Halsall, M. J., "A Preliminary Assessment of Thorium
as a Fuel for Thermal Reactors," paper presented at the IAEA Panel on
"Utilization of Thorium in Power Reactors," Vienna, Auatria, June 14-18,
1965.
4. Carlsmith, R. S., Private Communication February, 1966.
5. Peterson, S., Adams, R. E., and Douglas, D. A. Jr., "Properties of Thorium,
Its Alloys, and Its Compounds," paper presented at the IAEA Panel on
"Utilization of Thorium in Power Reactors," Vienna, Austria, June 14-18,
1965.
6. R. G. Wymer and J. H. Cocbs, "Preparation, Coating, Evaluation, and
Irradiation Testing of Sol-Gel Oxide Microspheres", to be published in
J. British Ceramics Society.
INFO
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.
T
P
7. Kabin, S. A., Osborne, M. F., Clinton, S. D., and U].lmann, J. W.,
.
"Thorium Fuel Cycle Irradiation Program at the Oak Ridge National
Laboratory," Proceedings of the Thorium Fuel Cycle Symposium,
Gatlinburg, Tennessee, December 5-7, 1962, USAEC Report TID-7650,
Bk II, P. 643, AEC Technical Information Division.
8. Olsen, A. R., Trauger, D. B., Harms, W. O., Adams, R. E., and
Douglas, D. A., "Irradiation Behavior of Thorium-Uranium Alloys
and Compounds," paper presented at the IAEA Panel on "Utilization
of Thorium in Power Reactors," Vienna, Austria, June 14-18, 1965.
9. Rabin, S. A., Ullmann, J. W., Long, E. L., Jr., Osborne, M. F.,
and Goldman, A. E., Irra diation Behavior of High-Bumup Tho, 4.5%
VO,_Fuel Rods, USAEC Report ORNL-3837, Oak Ridge National Laboratory
October, 1965.
10. Osborne, M. F., and Long, E. L., Jr., Post Irradiation Examination
of Maritime-ORR Loop Experiments 5, 7, and 8, USAEC Report ORNL-TM-921,
Oak Ridge National Laboratory, October, 1964.
11. Prados, J. W., Beatty, R. L., Beutler, H., Coobs, J. H., Olsen, A. R.,
and Scott, J. L., "Development of Coated Particle Fuels for Advanced
Gas Cooled Reactors, Thorium Symposium, May, 1966.
12. Hamner, R. L., Beatty, R. L., and Beutler, H., "Effects of Heat
Treatment on Pyrolytic-Carbon Coated Oxide Particles," pp. 43-50 in
GCRP Semiềnn. Progr. Rept., Sept. 30, 1965, USAEC Report, ORNL-3885.
.
EE
T.
-.
13. ORNL-TM-1297.
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END
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9 / 15 / 66







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