3.07

TRISO-Coated Particle Fuel Performance

D. A. Petti, P. A. Demkowicz, and J. T. Maki
Idaho National Laboratory, Idaho Falls, ID, USA

R. R. Hobbins
RRH Consulting, Wilson, WY, USA

Published by Elsevier Ltd.

3.07.1

Introduction

153

3.07.2
3.07.2.1
3.07.2.1.1
3.07.2.1.2
3.07.2.1.3
3.07.2.1.4
3.07.2.1.5
3.07.2.1.6
3.07.2.1.7
3.07.2.1.8
3.07.2.1.9
3.07.2.1.10
3.07.2.2

3.07.2.2.1
3.07.2.2.2
3.07.2.2.3
3.07.2.2.4
3.07.2.2.5
3.07.2.2.6
3.07.2.2.7
3.07.2.2.8
3.07.2.2.9
3.07.2.3
3.07.2.3.1
3.07.2.3.2
3.07.2.3.3
3.07.2.3.4
3.07.2.3.5
3.07.2.3.6
3.07.2.3.7
3.07.2.3.8
3.07.2.3.9
3.07.2.3.10
3.07.2.3.11
3.07.2.3.12
3.07.2.3.13
3.07.2.4
3.07.2.5
3.07.2.6
3.07.2.7
3.07.2.7.1
3.07.2.7.2
3.07.2.7.3

3.07.2.7.4

Irradiation Performance
Overview of Irradiation Facilities and Testing
BR-2
IVV-2M
HFR Petten
HFIR
ATR
SAFARI
TRISO-coated particle fuel irradiation testing
Thermal and physics analysis considerations
Gas control system considerations
FPMS considerations
German Experience
R2-K12 and R2-K13
BR2-P25
HFR-P4
SL-P1
HFR-K3
FRJ2-K13
FRJ2-K15
FRJ2-P27
HFR-K6 and HFR-K5
US Experience
F-30
HRB-4 and HRB-5
HRB-6
OF-2
HRB-14

HRB-15B
R2-K13
HRB-15A
HRB-16
HRB-21
NPR-1 and NPR-2
NPR-1A
AGR-1
European Experience
Chinese Experience
Japanese Experience
Irradiation Performance Summary
Heavy metal contamination
In-service failures
Failure mechanisms
Acceleration effects

154
154
154
154
155
155
155
155
156
156
158
159
160

160
161
162
162
163
163
164
164
165
165
166
167
169
170
171
173
174
174
175
176
177
178
178
180
181
184
185
185
186
186

187
151


152

TRISO-Coated Particle Fuel Performance

3.07.3
3.07.3.1
3.07.3.1.1
3.07.3.1.2
3.07.3.1.3
3.07.3.1.4
3.07.3.2
3.07.3.3
3.07.3.3.1
3.07.3.3.2
3.07.3.3.3
3.07.3.3.4
3.07.3.4
3.07.3.4.1
3.07.3.4.2
3.07.3.5
3.07.4
References

Safety Testing
Facility Overview
Ku¨FA at ITU

INL’s FACS
ORNL’s Core Conduction Cooldown Test Facility
KORA
German Experience
European Experience
AVR 73/21
AVR 74/18
HFR K6/3
HFR K6/2
US Experience and Future Plans
Past experience
Future plans
Japanese Experience
Conclusions

Abbreviations
AGR
ATR
AVR
BAF
BISO
BOL
BR-2
CCCTF
CVD
DOE
EFPD
EOL
FACS
FIMA

FPMS
FRJ
GETR
HEU
HFEF
HFIR
HFR
HRB
HTGRs
HTR-10
HTTR
IFEL
IMGA

Advanced Gas Reactor
Advanced Test Reactor
Arbeitsgemeinschaft
Versuchsreaktor
Bacon anisotropy factor
Bi-structural isotropic
Beginning of life
Belgian Reactor 2
Core Conduction Cooldown Test
Facility
Chemical vapor deposition
Department of Energy
Effective full-power day
End of life
Fuel accident condition simulator
Fissions per initial metal atom

Fission product monitoring system
Research Reactor Juelich
General Electric Test Reactor
Highly enriched uranium
Hot Fuel Examination Facility
High-Flux Isotope Reactor
High-Flux Reactor
HFIR Removable Beryllium
High-temperature gas-cooled
reactors
High Temperature Reactor 10
High-temperature test reactor
Irradiated fuel examination laboratory
Irradiated microsphere gamma
analyzer

INET

189
189
189
190
192
193
193
199
199
199
200
200

202
202
205
206
209
212

Institute of Nuclear and New Energy
Technology
INL
Idaho National Laboratory
IPyC
Inner pyrolytic carbon
ITU
Institute for Transuranium
Elements
JMTR
Japan Material Test Reactor
KuFA
Cold finger apparatus (in German)
LEU
Low-enriched uranium
LHTGR
Large High Temperature Gas
Reactor
LTI
Low temperature isotropic
MOL
Middle of life
NE-MHTGR Commercial version of NP-MHTGR

NGNP
Next Generation Nuclear Plant
NP-MHTGR New Production Modular
High-temperature Gas-Cooled
Reactor
ORNL
Oak Ridge National Laboratory
ORR
Oak Ridge Research Reactor
PIE
Postirradiation examination
R&D
Research and development
R/B
Release to birth ratio
SiC
Silicon carbide
TRIGA
Training research and isotope
production, General Atomics
TRISO
Tristructural isotropic
UCO
Uranium oxycarbide
Uranium dioxide
UO2
VHTR
Very-high-temperature reactors
VXF
Vertical experimental facility

WAR
Weak acid resin


TRISO-Coated Particle Fuel Performance

3.07.1 Introduction
For all high temperature gas reactors (HTGRs), tristructural isotropic (TRISO)-coated particle fuel
forms the heart of the concept. Such fuels have
been studied extensively over the past four decades
around the world, for example, in countries including
the United Kingdom, Germany, Japan, United States,
Russia, China, and more recently, South Africa. In
early gas-cooled reactors, the coated particle fuel
form consisted of layers of carbon surrounding the
fissile kernels. Highly enriched uranium (HEU) and
thorium carbides and oxides were used as fissile and
fertile kernels. Ultimately, the carbon layer coating
system (termed BISO for bistructural isotropic) was
abandoned because it did not sufficiently retain fission
products, leading to the development of the current
three-layer coating system (termed TRISO for tristructural isotropic). In TRISO-coated fuel, a layer of
silicon carbide (SiC) is sandwiched between pyrolytic
carbon layers. This three-layer system is used to both
provide thermomechanical strength to the fuel and
contain fission products. In addition, for operational
and economic reasons, the fuel kernel of choice today
is low-enriched uranium (LEU) uranium dioxide
(UO2) for the pebble bed design and uranium oxycarbide (UCO) for the prismatic design.
In both pebble bed and prismatic gas reactors,

the fuel consists of billions of multilayered TRISOcoated particles ($750–830 mm in diameter) distributed
within fuel elements in the form of circular cylinders
(12.5 mm in diameter and 50 mm long) called ‘compacts’
or spheres called ‘pebbles’ (6 cm in diameter). The
active fuel kernel is surrounded by a layer of porous
carbon, termed ‘the buffer’; a layer of dense carbon,
termed ‘the inner pyrolytic carbon layer’; a layer of
SiC; and another dense carbon layer, termed ‘the outer
pyrolytic carbon layer.’ These collectively provide for
accommodation and containment of fission products
generated during operation. The buffer layer is designed
to accommodate fission recoils, volumetric swelling of
the kernel, and fission gas released under normal operation. The inner pyrolytic carbon layer protects the kernel from reactive chlorine compounds produced during
SiC deposition in the chemical vapor deposition (CVD)
coater. The SiC layer provides structural strength to the
particle. The outer pyrolytic carbon layer protects the
particles during formation of the fuel element. Under
normal operation, radiation damage causes shrinkage of
the pyrolytic carbon layers, which induces compressive
stresses in the SiC layer to counteract tensile stresses
associated with fission gas release. All three layers of

153

the TRISO coating system exhibit low permeability.
These fuel constituents are extremely stable and are
designed not to fail under normal operation or anticipated accident conditions, thereby providing effective
barriers to the release of fission products. Figure 1 is a
montage of TRISO fuel used in both prismatic and
pebble bed high-temperature gas reactors.

Rigorous control is applied at every step of the
fabrication process to produce high-quality, very lowdefect fuel. Defect levels are typically on the order
of one defect per 100 000 particles. Specifications
are placed on the diameters, thicknesses, and densities
of the kernel and layers; the sphericity of the particle;
the stoichiometry of the kernel; the isotropy of the
carbon; and the acceptable defect levels for each
layer. Statistical sampling techniques are used to demonstrate compliance with the specifications usually at
the 95% confidence level. For example, fuel production for German reactors in the 1980s yielded only
approximately 100 defects in 3.3 million particles produced. This remains the standard for gas-cooledreactor fuel production today.1,2
Irradiation performance of high-quality, lowdefect coated particle fuels has been excellent. In
Section 3.07.2, a detailed review of the state of the
art in irradiation testing, capabilities of existing fission reactors worldwide to irradiate TRISO fuel, and
the irradiation behavior of modern TRISO-coated
particle fuel around the world will be discussed.
Testing of German fuel under simulated accident
conditions in the 1980s has demonstrated excellent
performance. Section 3.07.3 describes the accident
behavior of TRISO-coated particle fuel largely on the
basis of the German database and the plans to perform
similar testing for the current generation of TRISOcoated fuels. Additional limited testing of TRISOcoated particle fuel performed under air and water
ingress events and under reactivity pulses has been
reported elsewhere3 and will not be repeated here.
The outstanding irradiation and accident simulation testing results obtained by German researchers
form the basis for fuel performance specifications
used in gas-cooled-reactor designs today. Specifications for in-service failure rates under irradiation and
accident conditions are very stringent, typically on
the order of 10À4 and 5 Â 10À4, respectively.
Significant research and development (R&D)
related to TRISO-coated fuels is underway worldwide

as part of the activities of the Generation IV International Forum on Very-High-Temperature Reactors
(VHTRs). The focus is largely on extending the capabilities of the TRISO-coated fuel system for higher


154

TRISO-Coated Particle Fuel Performance

Pyrolytic carbon
Silicon carbide
Uranium dioxide or oxycarbide kernel

Prismatic

Pebble

Particles

Matrix

Compacts

Fuel element

TRISO-coated fuel particles (left) are formed into fuel compacts
(center) and inserted into graphite fuel elements (right) for the
prismatic reactor

Kernel
Buffer layer


5 mm graphite layer
Coated particles imbedded
in graphite matrix

Inner PyC-layer
Fuel-free shell
SiC-layer
Fueled zone
Outer PyC-layer

Fuel sphere
Dia 60 mm

Half
section

TRISO-coated fuel particles are formed
into fuel spheres for pebble bed reactor
Figure 1 TRISO-coated particle fuel and compacts and fuel spheres used in high temperature gas reactors.

burnups (10–20%) and higher operating temperatures
(1250  C) to improve the attractiveness of hightemperature gas-cooled reactors as a heat source for
large industrial complexes where gas outlet temperatures of the reactor would approach 950  C.4 Of greatest concern is the influence of higher fuel temperatures
and burnups on fission product interactions with the
SiC layer leading to degradation of the fuel and the
release of fission products. Activities are also underway
around the world to examine modern recycling techniques for this fuel and to understand the ability of gas
reactors to burn minor actinides.5,6


3.07.2.1.1 BR-2

The Belgian Reactor 2 (BR-2) reactor is a materials test
reactor in Mol, Belgium7 that produces very fast
(3.5 Â 1014 neutrons cmÀ2 sÀ1 [E > 1 MeV]) and thermal neutron fluxes (1012 neutrons cmÀ2 sÀ1). The facilities have irradiation test rigs ($15 mm ID and 400 mm
long) that can be used to irradiate coated-particle gas
reactor fuel forms. They have adequate flux, fluence,
and temperature characterization for the capsule,
and have the infrastructure needed for capsule disassembly and postirradiation examination (PIE). The
capsule size precludes irradiation of pebbles; however, it could handle approximately six to eight fuel
compacts.

3.07.2 Irradiation Performance
3.07.2.1 Overview of Irradiation Facilities
and Testing
This section provides a brief overview of irradiation
facilities that are available today to perform TRISOcoated particle irradiations.

3.07.2.1.2 IVV-2M

The IVV-2M is a 15-MW water-cooled reactor that
has been used in Russia for a variety of coatedparticle testing.8 Four different test rigs have been
used to test specimens ranging from particles, to
compacts, to spheres. The coated particle ampoule


TRISO-Coated Particle Fuel Performance

is a noninstrumented rig that can hold 10–13 graphite
disks (15 mm in diameter and 2 mm thick), each of

which can hold 50 particles. The rig can also hold
coated particles in axial holes, 1.2 mm in diameter,
and a uniform volume of coated particles, 12–18 mm
in diameter and 20–255 mm high, in a graphite
matrix. Another rig, termed a ‘CP hole,’ is 27 mm in
diameter and that can handle six to eight capsules.
A third rig, identified as ASU-8, is a 60-mm hole that
can handle three compacts. The largest channel available is Vostok, which is 120 mm in diameter and
contains four cells. All of these rigs can irradiate
fuel at representative temperatures, burnups, and
fluences for HTGRs. There is a large degree of
flexibility in the testing options at IVV-2M. Their
rigs can handle particles, compacts, and spheres.
3.07.2.1.3 HFR Petten

The High Flux Reactor (HFR) in Petten, Netherlands,
is a multipurpose research reactor with many irradiation locations for materials testing.9 The HFR has
two different types of irradiation rigs/locations in the
facility: one that can accommodate compacts and
another that can accommodate spheres. Rigs for
spheres are multicell capsules, 63–72 mm in diameter
that can handle 4–5 spheres in up to 4 separate cells.
For compacts rigs/locations are $32 mm in diameter
and 600 mm in useful length. They can handle three
or four parallel channels of compacts. For the threechannel configuration, approximately 30 compacts
could be irradiated in the rig. There is a large axial
flux gradient across the useable length (40% spread
maximum to minimum) that must be considered in
the design of any experiment.
3.07.2.1.4 HFIR


The High Flux Isotope Reactor (HFIR) at Oak Ridge
National Laboratory (ORNL) is a light-water-cooled,
beryllium-reflected reactor that produces high neutron fluxes for materials testing and isotope production.10 Two specific materials irradiation facilities
locations are available for gas reactor fuel testing:
(a) the large RB positions (eight total) that are
46 mm in diameter and 500 mm long, and can accommodate capsules holding up to 24 compacts (three in
each graphite body, eight bodies axially) in a single
swept cell; and (b) the small vertical experimental
facility (VXF) positions (16 total) that are 40 mm in
diameter and 500 mm long, and can accommodate
capsules holding up to 16 compacts (eight in each
graphite body, two bodies axially) in a single swept
cell. Capsules can be irradiated in the lower flux small

155

VXF positions and then moved to the higher flux
removable beryllium positions. Neither of these positions can accommodate pebbles. A third facility, the
large VXF positions (six total), are farther out in
the reflector (and therefore have lower fluxes), but
are 72 mm in diameter and also 500 mm long. As with
the HFR, there is a large axial flux gradient that must
be considered in the design of any experiment in any
of these facilities.
3.07.2.1.5 ATR

The Advanced Test Reactor (ATR) at Idaho National
Laboratory (INL) is a light-water-cooled, berylliumreflected reactor fuel in a four-leaf clover configuration to produce high neutron fluxes for materials
testing and isotope production.11 The clover leaf

configuration results in nine very high flux positions,
termed ‘flux traps.’ In addition, numerous other holes
of varying size are available for testing. Several positions can be used to irradiate coated-particle fuel.
The 89-mm-diameter medium I position (16 total)
and the 100–125-mm-diameter flux traps can accommodate pebbles. Specifically, the use of a medium
I position early in the irradiation, required because
of the enrichment of the fuel, followed by transfer of
the test train to the northeast flux trap can provide
irradiation conditions representative of a pebble bed
reactor. Approximately 10–12 pebbles in five or six
individually swept cells can be envisioned in the test
train. The large B positions in ATR (four total) are
38 mm in diameter and 760 mm in length. They can
accommodate six individually swept cells, with two
graphite bodies per cell, containing up to three 2-in.
long compacts per body. Thus, 36 full-size US compacts can be irradiated in this location. Of special
note, here is the very flat burnup and fluence profile
available axially in the ATR over the 760 mm length.
This allows for nearly identical irradiation of large
quantities of fuel.
3.07.2.1.6 SAFARI

The SAFARI Reactor in Pelindaba, Republic of
South Africa, is an isotope production and research
reactor.12 The core lattice is an 8 Â 9 array, consisting
of 28 fuel assemblies, 6 control rods, and a number
of aluminum and beryllium reflector assemblies.
The reactor is cooled and moderated by light water
and operates at a maximum power level of 20 MW.
In-core irradiation positions include six high-flux

isotope production positions: two hydraulic, two
pneumatic, and two fast transfer systems that are
accessible during operation. Several other irradiation


156

TRISO-Coated Particle Fuel Performance

positions can also be accessed when the reactor
is shut down. A large poolside facility allows for a
variety of radiation applications. An intermediate
storage pool and a transfer canal allow for easy and
safe transport of activated materials to a hot cell.
3.07.2.1.7 TRISO-coated particle fuel
irradiation testing

The historical experience in irradiation testing of
coated particle fuels suggests that multicell capsules
wherein fuel can be tested in separate compartments
under different temperature, burnup, and fluence conditions allow for tremendous flexibility and can actually
save time and money in an overall fuel qualification
program. Although there are differences in details of
the test trains used in each of the reactors, they share a
number of important similarities in the state of the art
with irradiation testing of this fuel form. In this section,
these important similarities are presented to highlight
the technical considerations in executing this type of
testing.
Because of the differences in neutron flux spectrum between a gas reactor and a light-water materials test reactor, simultaneous matching of both the

rate of burnup and the rate of accumulation of fast
neutron fluence is difficult to achieve. In addition, the
traditional 3-year fuel cycle of high-temperature gas
reactors makes real-time irradiation testing both timeconsuming and an expensive part of an overall fuel
development effort. To overcome these shortcomings,
irradiations in material test reactors have historically
been accelerated relative to those in the actual reactor.
Usually, the time acceleration is focused on achieving
the required burnup in a shorter time than would be
accomplished in the actual reactor, with the value of
the fast fluence left as a secondary variable that must
fall between a minimum and maximum value.
The level of acceleration can also impact the
potential for fuel failure during irradiation. The
level of acceleration at a given test reactor power,
coupled with fuel loading in the experiment, results
in a power density for the fuel specimen in the
experiment. The power density peaks at the beginning of the irradiation when the fissile content is
highest and decreases as the fissile material is burned
out of the fuel. As the level of acceleration increases,
the temperatures in the fuel kernels increase above
that in the fuel matrix because of the thermal resistances associated with the coatings of the particle,13
and the potential for high temperature, thermally
driven failure mechanisms to play a deleterious role
in fuel performance becomes more important.

As discussed in Section 3.07.2.7, the irradiation
performance database suggests that modest levels of
acceleration (1.5–3Â) appear to be acceptable without jeopardizing fuel performance in the irradiation,
and should be a baseline requirement for future gas

reactor irradiations. This acceleration level can be
translated into a maximum power per fuel body or
power per particle that can be used by experimenters
in the design of the irradiation capsule.
Given the limitations of materials test reactors
around the world, the TRISO-coated particle irradiation database contains results from tests conducted
under a range of accelerations. Successful German
TRISO-coated particle fuel irradiations in the
European HFR-Petten reactor were conducted using
an acceleration of less than a factor of three. By comparison, other German irradiations in the Forschungzentrum Reaktor Juelich (FRJ) reactor at Ju¨lich had
a neutron spectrum that was too thermalized. This
resulted in the fuel receiving too little fast fluence
to be prototypic of a high-temperature gas reactor.
Similarly, historic US irradiations in ORNL’s HFIR
reactor had too high a thermal flux resulting in significant burnup acceleration of the irradiation. On the basis
of these considerations, the large B positions (38 mm
diameter) in the ATR (see Figure 2) were chosen for
the US Department of Energy’s (DOE) Advanced Gas
Reactor (AGR) Program fuel irradiations because the
rate of fuel burnup and fast neutron fluence accumulation in these positions provide an acceleration factor
of less than three times that expected in the hightemperature gas reactor.
3.07.2.1.8 Thermal and physics analysis
considerations

Given the complexity of the capsules currently being
designed, the extensive review by safety authorities of
the thermo-mechanical stresses, and the importance of
each capsule in terms of irradiation data for fuel qualification, three-dimensional physics and thermal analyses are essential in irradiation capsule design. These
analyses are critical to ensure that the fuel reaches the
intended burnup, fluence, and temperature conditions.

To achieve high burnups with these fuels requires
detailed physics calculations to determine the time to
reach full burnup. Given the concerns about severely
accelerated irradiations, it is not uncommon for such
irradiations to take approximately 2 years to reach full
burnup in LEU TRISO-coated particles. In addition,
because thermocouples should not be attached directly
to the fuel, thermal analysis is used to calculate the fuel
temperature during the irradiation.


TRISO-Coated Particle Fuel Performance

157

North

ON-8

ON-9 ON-10 ON-11 ON-12

ON-3

ON-4

ON-5

ON-6

ON-1


Fuel
elements

ON-7
ON-2

I-19

I-20

I-1

I-2

H positions

I-3

I-4

I-18
I-17

Small B
position

I-5
I-6


I-16
I-15

I-7
I-8

I-14

I-13

I-12

I-11

I-10

OS-6

OS-1

I-9

East large B
position location
for AGR-1

In-pile
tube
OS-2


OS-3

OS-4

OS-5

OS-7

OS-8

OS-9

OS-10 OS-11 OS-12

Control drum

OS-13 OS-14 OS-15 OS-16 OS-17

I positions
OS-18 OS-19 OS-20 OS-21 OS-22

Figure 2 Schematic of ATR showing fuel and select irradiation positions.

Examples of a test train for fuel compacts used in
INL’s ATR and the pebbles used in HFR-Petten are
shown in Figures 3 and 4 respectively.
These irradiation capsules have extensive instrumentation to measure temperature, burnup, and fast
fluence at multiple locations in the test train. Traditional commercial thermocouples have been used
extensively in past irradiations, but thermocouples
can suffer from drift and/or de-calibration in the

reactor. Redundancy in thermocouple measurements
is another consideration in light of the low reliability
of thermocouples at high temperatures and long
times in neutron fields typical of TRISO-coated particle fuel irradiations. Melt wires are inexpensive and
have been used as a backup to thermocouples where
space was available in the capsule. However, melt
wires only indicate that a certain peak temperature
has been reached, and not the time of that peak.

Direct temperature measurements of the coated
particles are problematic because direct metal contact (e.g., thermocouple wires or sheaths) with the
fuel element is not recommended as the metals can
attack the TRISO fuel coatings. Thus, temperatures
must be calculated on the basis of thermocouples
located elsewhere in the capsule. Thermocouples
are generally located as close as possible to the
fuel body to minimize the uncertainties on the
calculated fuel temperatures related to irradiationinduced dimensional change and thermal conductivity changes of the materials in the capsule.
Encapsulating the fuel element in a graphite sleeve
or cup and inserting thermocouples into the graphite
has been used successfully in many designs. The
high conductivity of graphite minimizes the effect
of irradiation-induced dimensional changes on the
calculated fuel temperature.


158

TRISO-Coated Particle Fuel Performance


Gas line

Fuel stack

Thermocouple

SST holder

Thermocouple
Purge gas pipe
Radiation shields

Hafnium shield

Graphite cup
Test fuel element

Capsule
spacer nub

Figure 3 Schematic of capsule used in US INL AGR
program.

Historically, metal sleeves have not been allowed
to touch fuel elements because of past experiences
in which SiC was attacked by transition metals (Fe,
Cr, and Ni). Although quantitative data on transport
rates of such metals into the fuel element and corrosion rates of the SiC are unknown, 2 or 3 mm thickness of graphite between the fuel element and the
metallic components (e.g., graphite sleeve) has been
found to be effective in minimizing the potential for

interaction.
These irradiation experiments typically include
both thermal and fast fluence wires. A number of
different fluence wires have been used successfully
to measure thermal and fast neutron fluences in coated
particle fuel irradiations. The specific type of wire to
be used will depend on the measurement need (fast
or thermal), the temperature it will experience during the irradiation, and compatibility with the material of encapsulation. Quartz encapsulation is not
recommended for high-temperature, high-fluence
applications. Neutronically, transparent refractories
(e.g., vanadium) are a good alternative material of
encapsulation. Inert gas filling of the flux wire

Figure 4 Schematic of pebble irradiation experiment
used by the Germans.

encapsulation is recommended to ensure no oxygen
interaction with the flux wire. Although fission
chambers and self-powered neutron detectors have
been used extensively in other reactor irradiations,
they may not be practical in the space-constrained
capsules expected for TRISO-coated particle fuel
qualification tests.
3.07.2.1.9 Gas control system considerations

Automated gas control systems – designed to change
the gas mixture in the experiment to compensate for
the reduction in fission heat and changes in thermal
conductivity with burnup – minimize human operator error and have proven to be a reliable method of
thermal control during these long fuel irradiations.

The temperature of each experiment capsule is controlled by varying the mixture of two gases with
differing thermal conductivities in a small insulating
gas jacket between the specimens and the experiment
containment. A mixture of helium and argon has been
used in the past and provides a wide temperature
control band for the experiments. Unfortunately,


TRISO-Coated Particle Fuel Performance

argon cannot be used in fuel experiments where
online fission product monitoring is used because
the activated argon will reduce detectability of the
system. Therefore, helium and neon are used instead.
Computer-controlled mass flow controllers are typically used to automatically blend the gases (on the
basis of feedback from the thermocouples) to control
temperature. The gas blending approach allows for a
very broad range of control. Automatic gas verification (e.g., by a thermal conductivity analyzer) has
been implemented in some experiments to prevent
the inadvertent connection of a wrong gas bottle. Gas
purity is important and an impurity cleanup system
should be implemented during each irradiation. Flow
rates and gas tubing should be sized to minimize
transit times between the mass-flow controllers and
the experiment, as well as between the experiment
and the fission product monitors.
3.07.2.1.10 FPMS considerations

In addition to thermal control, sweep gas is used to
transport any fission gases released from the fuel to a

fission product monitoring system (FPMS). A number of techniques have been used historically to
quantify the release of fission gases from the fuel in
these irradiation capsules. Techniques include gross
gamma monitoring, online gamma spectroscopy, and
offline gamma spectroscopy of grab samples. Online
gross gamma monitoring of the effluent gas in the

experiment using ion chambers and sodium iodide
detectors is an excellent means to capture any
dynamic failures of the coated particles associated
with the instantaneous release upon failure. Grab
samples can provide excellent noble gas isotopic
information. The temporal resolution and the number
of isotopes that can be measured depend on the frequency of the grab samples and the delay time
between acquisition of the grab sample and offline
analysis. Weekly grab samples are typical in most
irradiations, although daily or even hourly samples
are possible if failure has occurred, assuming operation and associated analysis costs are not too high.
Typical isotopes to be measured include 85mKr, 87Kr,
88
Kr, 131mXe, 133Xe, and 135Xe. Measurement of verylong-lived isotopes (e.g., 85Kr) would be useful in
elucidating fission product release mechanisms from
the kernel, but would also require waiting for the
decay of the shorter lived isotopes in the sample.
Online gamma spectroscopy, although the most
expensive in terms of hardware costs, can provide
the most detailed real-time information with detailed
isotopic spectrums as often as necessary subject to
data storage limitations of the system. An example of
the system used for the US AGR program is shown in

Figure 5. With such systems, transit times from the
experiment to the detector should be minimized to
allow measurement of short- and medium-lived isotopes, but must remain long enough to allow decay of

Temperature control
gas mixing system

Vessel wall

He

Filter

6
5
Silver
zeolite

4 Capsules
in-core
3
2
1

Fission product
monitoring system
Grab sample

Figure 5 Integrated fission product monitoring system used in US AGR program irradiations.


H and V exhaust

Ne
Particulate
filters

159


160

TRISO-Coated Particle Fuel Performance

any short-lived isotopes associated with the sweep
gases ($2–3 min). With this delay time, 89Kr, 90Kr,
135m
Xe, 137Xe, 138Xe, and 139Xe should also be capable of being measured online. Measurements of
xenon gas-release during reactor outages are recommended to provide information on iodine release
behavior from the decay of xenon precursors. Multiple
options for fission gas-release measurements should be
considered for long irradiations where reliability of the
overall fission gas measurement system can be a concern. Redundancy is also recommended for online
systems so that failure of a spectrometer does not
jeopardize the entire experiment.
On the basis of the online concentration data, a
release-to-birth ratio (R/B), a key parameter used in
reactor fuel behavior studies,14 can be calculated and
provide some insight into the nature of any particle
failures. Because these instruments are online during
the entire irradiation, a complete time history of gas

release is available. Gas release early in the irradiation (i.e., from the start of the irradiation) is indicative
of initially failed particles or contamination outside
of the SiC layer. Release later during the irradiation is
indicative of in situ particle failure. The timing of the
failure data can then be correlated to temperature,
burnup, and/or fluence that can be used when coupled with PIE to determine the mechanisms responsible for the fuel failure.
3.07.2.2

German Experience

Previously, particle fuel development was conducted
by German researchers in support of various HTGR
designs that employed a pebble bed core. These
reactors were intended to produce process heat or
electricity. The sequence of fuel development used
by German researchers followed improvement in
particle quality and performance and was largely
independent of developments in reactor technology.
German fuel development can be categorized
according to the sequence of fuels tested as provided
in Table 1.
German irradiation test conditions generally
covered projected fuel operating conditions, where
fuel was to reach full burnup at fast fluences
of 2.4 Â 1025 n mÀ2 and operate at temperatures
up to 1095  C for process-heat applications and
up to 830  C for electrical production applications.
With the exception of irradiation duration, the various experiments performed bounded expected nominal conditions or were purposely varied to meet
other test objectives. In order to obtain results in a


Table 1

German particle fuel development sequence

Date of design
consideration

Fuel form

1972
1977

BISO coated (Th, U)O2
Improved BISO coated (Th, U)O2
TRISO-coated UCO fissile particles with
ThO2 fertile particles
TRISO-coated (Th, U)O2
LEU TRISO-coated UO2

1981

timely manner, tests conducted by German researchers were generally accelerated by factors of 2–3.
The following sections present irradiation experiment summaries for fuels of ‘modern’ German design.1
For these experiments, this definition extends to
high-enriched (Th, U)O2 TRISO-coated particles
fabricated since 1977, and low-enriched UO2 TRISOcoated particles fabricated since 1981. Table 2 provides the physical attributes of the fuel used in these
tests. Mixed oxide fuel test summaries are presented
first, followed by the UO2 tests.
3.07.2.2.1 R2-K12 and R2-K13


The R2-K12 and R2-K13 cells were irradiated in the
R2 reactor at Studsvik, Sweden. The main objective
of the R2-K12 experiment was to test mixed oxide
(Th, U)O2 and fissile UC2/fertile ThO2 fuel elements, whereas for R2-K13, the main objective was
to test mixed oxide (Th, U)O2 fuel elements and
supply fuel for subsequent safety tests.
In R2-K12, four full-size spherical fuel elements
were irradiated in four independently gas-swept cells.
Two cells contained mixed oxide fuel spheres, while
the other two contained fissile/fertile fuel spheres. As
the German researchers did not develop the twoparticle fissile/fertile system further, only the mixed
oxide results were reported. R2-K13 was a combined
experiment with the United States. Four independently gas-swept cells were positioned vertically on
top of one another. The top and bottom cells each
contained a full-size German fuel sphere. The middle
two cells contained US fuel and will be discussed in
Section 3.07.2.3. Configuration and irradiation data
from both experiments are given in Tables 3 and 4.
Cold gas tests on each fuel sphere during PIE
indicated that all the particles had remained intact
in both R2-K12 and R2-K13. These tests are conducted after the fuel has been stored (for $14 days) at
room temperature and a quasi-steady-state release of
fission gas has been reached. The fuel is then swept
with a carrier gas that is monitored for various fission


TRISO-Coated Particle Fuel Performance

Table 2


161

Characteristics of modern German TRISO fuel particles

Particle batch

EUO 2308

EUO 2309

HT 354–383

EO 1607

EO 1674

Experiments irradiated in:

FRJ2-K13
FRJ2-P27
HFR-P4
HFR-K3
SL-P1
UO2
9.82
497 Æ 3%
10.81
94
41
36

40
895
1.00
[1.9]
3.20
1.88
1.053
1.019

FRJ2-P27
HFR-P4

FRJ2-K15

R2-K12
BR2-P25

R2-K13

UO2
9.82
497 Æ 3%
10.81
93
37
51
38
922
1.00
[1.9]

3.20
1.87

UO2
16.76
501 Æ 10.8%
10.85
92 Æ 14.3
38 Æ 3.4
33 Æ 1.9
41 Æ 3.8
906 Æ 28.8
1.013
[1.9]
3.20
1.88
1.029
1.020

(Th, U)O2
89.57
494 Æ 3%
10.12
85
39
37
39
888
1.09
1.93

3.20
1.93

(Th, U)O2
89.01
496 Æ 3%
10.10
89
37
33
39
890
1.06
1.90
3.19
1.90

Kernel form
U enrichment (%)
Kernel diameter (mm)
Kernel density (g cmÀ3)
Buffer thickness (mm)
IPyC thickness (mm)
SiC thickness (mm)
OPyC thickness (mm)
Particle diameter (mm)
Buffer density (g cmÀ3)
IPyC density (g cmÀ3)
SiC density (g cmÀ3)
OPyC density (g cmÀ3)

IPyC BAF
OPyC BAF
235

Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum
Ju¨lich GmbH: Ju¨lich, Germany, 1990.
Notes: The Æ entries are one standard deviation. Entries in square brackets, [ ], are estimated values.
BAF is the Bacon anisotropy factor for the layer, where values closer to one are more isotropic.

Table 3

R2-K12 and R2-K13 configuration

Number of cells
Number of fuel spheres
Spherical fuel element
diameter
Fuel zone diameter
Fuel type
Particle batch
235
U enrichment
235
U per fuel element
232
Th per fuel element
Heavy metal per fuel
element
Number of particles per
spherical fuel element

Defective SiC layersa
(U/U-total)

R2-K12

R2-K13

2
2
59.9 mm

2
2
59.77 mm

47 mm
HEU (Th, U)O2
LTI – TRISO
EO 1607
89.57%
1.002 g
4.961 g
6.076 g

47 mm
HEU (Th, U)O2
LTI – TRISO
EO 1674
89.01%
1.02 g

10.125 g
11.27 g

10 960

19 780

<1 Â 10À5

<5 Â 10À6

a
Defective SiC layer fractions reported for German fuel are per
pebble with the exception of loose particle experiments that are
per particle batch.
Source: Gontard, R.; Nabielek, H. Performance Evaluation of
Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90;
Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.

gases (usually 85mKr) and heated to $60  C. Sudden
increases in the amount of fission gas detected indicate failed particles. The amount of increase is

proportional to the gas source, and in a calibrated
system, indicates the number of failed particles.
The fuel sphere from R2-K12 Cell 1 was partially
deconsolidated and visual inspection revealed two
kernels ‘without coating.’ Segments from each of the
two fuel spheres were also metallographically examined; those examinations revealed a reaction zone on
the inner side of the buffer layer, as well as tangential
cracks between the buffer and the inner pyrocarbon

layer. Only one particle exhibited a radial crack in the
buffer layer beyond the reaction zone. All of the SiC
and PyC layers examined remained intact.
3.07.2.2.2 BR2-P25

The BR2-P25 capsule was irradiated in the BR2 reactor at Mol, Belgium. The primary objective of this
experiment was to test (Th, U)O2 mixed oxide fuel.
One independently gas-swept cell contained 12 compacts. Each compact was cylindrical in shape and
contained a small fuel sphere. Configuration and irradiation data are given in Tables 5 and 6, respectively.
During PIE, Compacts 3 and 7 were electrolytically
deconsolidated with no particle failures being evident.
Ceramographic examination of cross-sections from
Compacts 4 and 8 revealed some radial cracks in
the buffer layers; however, no defective particles
were found.


162
Table 4

TRISO-Coated Particle Fuel Performance

R2-K12 irradiation data

Start date
End date
Duration (full power days)
Cell
Burnup (% FIMA)
Fast fluence (1025 n mÀ2, E > 0.10 MeV)

Center temperature ( C)
Surface temperature ( C)
BOL 85mKr R/B
EOL (report date) 85mKr R/B

R2-K12

R2-K13

28 November 1978
12 February 1980
308
1
11.1
5.6
1100
950
3.9 Â 10À9
3.0 Â 10À7

22 April 1980
19 September 1982
517
1
10.2
8.5
1170
960
2.2 Â 10À9
7.0 Â 10À8


2
12.4
6.9
1280
1120
4.6 Â 10À9
2.0 Â 10À7

4
9.8
6.8
980
750
1.5 Â 10À9
5.0 Â 10À8

Source: Gontard, R.; Nabielek, H. Performance Evaluation of Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90; Forschungszentrum
Ju¨lich GmbH: Ju¨lich, Germany, 1990.

Table 5

BR2-P25 configuration

Number of cells
Number of compacts
Cylindrical compact diameter
Cylindrical compact height
Diameter of spherical fuel zone
Fuel type

Particle batch
U enrichment
235
U per fuel compact
232
Th per fuel compact
Heavy metal per fuel compact
Number of particles per compact
Number of particles per cell
Defective SiC layers (U/U-total)
235

Table 7
1
12
26.58–27.74 mm
29.87–30.03 mm
20 mm
HEU (Th, U)O2
LTI – TRISO
EO 1607
89.57%
0.136 g
0.6744 g
0.8264 g
1490
17 880
<1 Â 10À5

Source: Gontard, R.; Nabielek, H. Performance Evaluation of

Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90;
Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.

Table 6

BR2-P25 irradiation data

Start date
End date
Duration (full power days)
Burnup (% FIMA)
Fast fluence (1025 n mÀ2, E > 0.10 MeV)
Maximum temperature ( C)
Minimum temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

30 October 1978
19 December 1981
350
13.9–15.6
6.2–8.1
1070
1010
2 Â 10À7
1 Â 10À6

Source: Gontard, R.; Nabielek, H. Performance Evaluation of
Modern HTR TRISO Fuels; Tech. Rep. HTA-IB-05/90;
Forschungszentrum Ju¨lich GmbH: Ju¨lich, Germany, 1990.


HFR-P4 configuration

Number of cells
Number of compacts per cell
Cylindrical compact diameter
Cylindrical compact height
Diameter of spherical fuel zone
Fuel type
Particle batch – cells 1 and 3
Particle batch – cell 2
235
U enrichment
Number of particles per compact
Number of particles per capsule
Defective SiC layers (U/U-total)

3
12
23–29 mm
32 mm
20 mm
LEU UO2 LTI – TRISO
EUO 2308
EUO 2309
9.82%
1630
19 600
<1 Â 10À6


with 36- and 51-mm-thick SiC layers irradiated at
1000  C, beyond burnups of 12% fissions per initial
metal atom (FIMA), and beyond fast fluences of
6 Â 1025 n mÀ2 (E > 0.10 MeV). The performance of
the 36 mm SiC layer fuel was also to be evaluated at an
irradiation temperature of 1200  C. Three independently gas-swept cells each contained 12 compacts
that were cylindrical in shape and contained a small
fuel sphere in each. Configuration and irradiation
data are given in Tables 7 and 8, respectively. Note
that the burnup and fast fluence goals were met, while
the irradiation temperature goals were not. PIE
revealed that the test articles remained intact. However, some failures caused by the thermocouples and
gas inlet tubes were found on the upper compacts.
3.07.2.2.4 SL-P1

3.07.2.2.3 HFR-P4

The HFR-P4 capsule was irradiated at the HFR
in Petten. The main objective of this experiment
was to compare the fuel performance of particles

The SL-P1 experiment was irradiated at the Siloe¨
Reactor in Grenoble, France. The objective of the
experiment was to test reference LEU fuel up to the
potential limits for burnup and fast fluence at 800  C.


TRISO-Coated Particle Fuel Performance

Table 8


HFR-P4 irradiation data

Start date
End date
Duration (full
power days)
Capsule
SiC layer thickness
(mm)
Maximum
temperature ( C)
Minimum
temperature ( C)
Maximum burnup
(% FIMA)
Peak fast fluence
(1025 n mÀ2,
E > 0.10 MeV)
BOL 85mKr R/B
EOL 85mKr R/B

Table 9

Table 10

10 June 1982
28 November 1983
351
1

36

2
51

3
36

940

945

1075

915

920

1050

14.7

14.9

14.0

8.0

8.0


8.0

3.5 Â 10À9
8 Â 10À8

–
8 Â 10À8

3.6 Â 10À9
8 Â 10À9

SL-P1 configuration

Number of cells
Number of compacts
Cylindrical compact diameter
Cylindrical compact height
Diameter of spherical fuel zone
Fuel type
Particle batch
235
U enrichment
Number of particles per
compact
Number of particles per cell
Defective SiC layers (U/U-total)

1
12
30.1 mm

30.8 mm
20 mm
LEU UO2 LTI – TRISO
EUO 2308
9.82%
1634
19 600
<1 Â 10À6

One gas-swept cell contained 12 compacts. Each
cylindrical compact contained one small fuel sphere.
Configuration and irradiation data are provided in
Tables 9 and 10, respectively. The operational objectives for this experiment were met. PIE revealed that
none of the compacts showed mechanical failure.
3.07.2.2.5 HFR-K3

The HFR-K3 capsule was irradiated at the HFR in
Petten. The primary objective of this experiment was
to determine the performance of reference LEU fuel
from an accelerated test. Four full-size spherical fuel
elements were irradiated in three independently gasswept cells. The cells were vertically positioned on
top of one another, with the middle cell containing
two fuel spheres. To minimize flux gradient effects
on the test fuel, the entire test rig was rotated 90
several times during the irradiation. Configuration

SL-P1 irradiation data

Start date
End date

Duration (full power days)
Burnup (% FIMA)
Fast fluence (1025 n mÀ2, E > 0.10 MeV)
Compact mean temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

Table 11

24 June 1982
23 December 1983
330
8.6–11.3
5.0–6.8
743–794
5.8 Â 10À7
1.2 Â 10À6

HFR-K3 configuration

Number of cells
Number of fuel spheres
Spherical fuel element diameter
Fuel zone diameter
Fuel type
Particle batch
235
U enrichment
Number of particles per spherical
fuel element

Defective SiC layers (U/U-total)

Table 12

163

3
4
59.98 mm
47 mm
LEU UO2 LTI – TRISO
EUO 2308
9.82%
16 400
4 Â 10À5

HFR-K3 irradiation data

Start date
End date
Duration (full
power days)
Cell/sphere
Burnup (% FIMA)
Fast fluence
(1025 n mÀ2,
E > 0.10 MeV)
Center
temperature ( C)
Surface

temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

15 April 1982
5 September 1983
359
A/1
7.5
4.0

B/2
10.0
5.8

B/3
10.6
5.9

C/4
9.0
4.9

1200

920

920

1220


1020

700

700

1020

1 Â 10À9
2 Â 10À7

9 Â 10À10
1 Â 10À7

9 Â 10À10
1 Â 10À7

2 Â 10À9
3 Â 10À7

and irradiation data are given in Tables 11 and 12,
respectively. Subsequent PIE reported no failures.
3.07.2.2.6 FRJ2-K13

FRJ2-K13 cells were irradiated at the DIDO reactor in
Ju¨lich, Germany. The main objective of this test was to
supply irradiated reference fuel for subsequent safety
tests. Fuel performance was also to be examined under
the controlled irradiation conditions of significant

burnup with negligible fast neutron fluence. Four fullsize spherical fuel elements were irradiated in two


164

TRISO-Coated Particle Fuel Performance

independently gas-swept cells. The cells were vertically
positioned on top of each other, with the fuel spheres
similarly positioned within the cells. Configuration and
irradiation data are given in Tables 13 and 14, respectively. Subsequent PIE reported no failures.

11 h. The 85mKr R/B ratio from each capsule increased
to a maximum of $10À8 at the start of the transient and
then dropped back to the pretransient levels after the
temperature was returned to the nominal test condition.
3.07.2.2.8 FRJ2-P27

FRJ2-K15 cells were irradiated at the DIDO reactor
in Ju¨lich, Germany. The main objectives of this test
were to demonstrate the high burnup potential of
reference fuel used in AVR reload 21-1 and to perform in-core temperature transient tests. Fuel performance was also to be examined under the controlled
irradiation conditions of significant burnup with negligible fast neutron fluence. Three full-size spherical
fuel elements were irradiated in three independently
gas-swept cells. Configuration and irradiation data
are given in Tables 15 and 16, respectively.
Capsules 2 and 3 underwent a temperature transient
test at a burnup of $10% FIMA. The temperature of
the sphere surfaces was raised to 1100  C and held for


FRJ2-P27 cells were irradiated at the DIDO reactor
in Ju¨lich, Germany. The main objectives of this test
were to investigate fission product release at various
cyclic temperatures and to determine the effectiveness of thicker SiC layers on the retention of 110mAg.
Each of the three independently gas-swept cells
contained three compacts and two coupons (trays).
The compacts were cylindrical in shape and contained
an (unspecified) outer fuel-free zone. The coupons
were graphite disks with holes, annularly spaced, for
the insertion of 34 particles. Of the two coupons that
contained the thicker SiC particles (51 mm vs. 36 mm),
one was placed in Cell 1, and the other in Cell 3.
Configuration and irradiation data are provided in
Tables 17 and 18, respectively.

Table 13

Table 15

3.07.2.2.7 FRJ2-K15

FRJ2-K13 configuration

Number of cells
Number of fuel spheres
Spherical fuel element diameter
Fuel zone diameter
Fuel type
Particle batch
235

U enrichment
Number of particles per spherical
fuel element
Defective SiC layers (U/U-total)

Table 14

2
4
59.98 mm
47 mm
LEU UO2 LTI – TRISO
EUO 2308
9.82%
16 400
4 Â 10À5

FRJ2-K13 irradiation data

Start date
End date
Duration (full
power days)
Cell/sphere
Burnup (% FIMA)
Fast fluence
(1025 n mÀ2,
E > 0.10 MeV)
Center
temperature ( C)

Surface
temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

Number of cells
Number of fuel spheres
Spherical fuel element diameter
Fuel zone diameter
Fuel type
Particle batch
235
U enrichment
Number of particles per spherical
fuel element
Defective SiC layers (U/U-total)

Table 16

24 June 1982
12 February 1984
396
A/1
7.5
0.2

A/2
8.0
0.2


B/3
7.9
0.2

B/4
7.6
0.2

1125

1150

1150

1120

985

990

990

980

2 Â 10À9
2 Â 10À8

2 Â 10À9
2 Â 10À8


8 Â 10À10
7 Â 10À9

8 Â 10À10
7 Â 10À9

FRJ2-K15 configuration
3
3
60.04 mm
47 mm
LEU UO2 LTI – TRISO
HT 354–383
16.76%
9500
<5 Â 10À5

FRJ2-K15 irradiation data

Start date
End date
Duration (full
power days)
Cell
Burnup (% FIMA)
Fast fluence
(1025 n mÀ2,
E > 0.10 MeV)
Center
temperature ( C)

Surface
temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

4 September 1986
20 May 1990
590
1
14.1
0.2

2
15.3
0.2

3
14.7
0.1

970

1150

990

800

980


800

2.0 Â 10À10
1.0 Â 10À8

2.47 Â 10À10
5.0 Â 10À9

2.0 Â 10À10
3.0 Â 10À9


TRISO-Coated Particle Fuel Performance

Table 17

FRJ2-P27 configuration

Number of cells
Number of compacts per cell
Number of coupons per cell
Cylindrical compact diameter
Cylindrical compact height
Coupon diameter
Coupon height
Diameter of coupon fuel annulus
Fuel type
Particle batch for compacts and four
coupons
Particle batch for two coupons (thick SiC)

235
U enrichment
Number of particles per compact
Number of particles per coupon
Number of particles per cell
Defective SiC layers (U/U-total)

Table 18

Table 19
3
3
2
27.9–28.03 mm
29 mm
27 mm
2.2 mm
23 mm
LEU UO2
LTI – TRISO
EUO 2308
EUO 2309
9.82%
2424
34
7340
<3 Â 10À6

17 February 1984
10 February 1985

232
1
7.6
1.4

2
8.0
1.7

3
7.6
1.3

1080

1320

1130

880

1220

1080

À6

1.0 Â 10
1.6 Â 10À6


À7

8.6 Â 10
1.0 Â 10À5

U enrichment
Number of particles per
spherical fuel element

HFR-K6

HFR-K5

4
4
60 mm

4
4
60 mm

LEU UO2 –
TRISO
10.6%
14 600

LEU UO2 –
TRISO
10.6%
14 600


were irradiated in four independently gas-swept
cells. A typical reactor temperature history was
simulated in the test with 17 temperature cycles
(corresponding to 17 passes through the core). For
one-third of a cycle, the fuel sphere center temperature was held at 800  C; for the other two-thirds
of the cycle, the center temperature was 1000  C.
In addition, three temperature transients (sphere center temperature held at 1200  C for 5 h) were performed at beginning of life (BOL), middle of life
(MOL), and end of life (EOL). Limited configuration
and irradiation data are given in Tables 19 and 20,
respectively. There were no particle failures reported
as a result of the irradiations.

FRJ2-P27 irradiation data

Start date
End date
Duration (full
power days)
Cell
Burnup (% FIMA)
Fast fluence
(1025 n mÀ2,
E > 0.10 MeV)
Center
temperature ( C)
Surface
temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B


HFR-K6 and HFR-K5 configurations

Number of cells
Number of fuel spheres
Spherical fuel element
diameter
Fuel type
235

165

3.07.2.3
À8

2.0 Â 10
1.2 Â 10À7

PIE revealed that all specimens and components
were in excellent condition. Cold gas tests of all
compacts and coupons determined that there was
only one defective/failed particle present. This particle was from a Capsule 2 coupon (with nominal SiC
thickness). Ceramographic examination revealed that
the particle was inserted in the coupon ‘without
coating’ and that kernel interactions led to a compression of the inner side of the buffer to a thickness
of $10 mm.
3.07.2.2.9 HFR-K6 and HFR-K5

The HFR-K6 and HFR-K5 capsules were irradiated
at the HFR in Petten.1,9 These experiments were a

proof test for HTR MODUL reference fuel. In each
experiment, four full-size spherical fuel elements

US Experience

Historical US particle fuel development effort
(through the mid 1990s), which included design and
testing, coincided with the development of various
HTGRs. This sequence of development is listed
in Table 21, and identifies the main fuel forms
under consideration at that time. US gas reactors
were designed to use prismatic graphite blocks
containing fuel compacts, and were primarily
intended to produce electricity with the exception
of the New Production Modular High-temperature
Gas-Cooled Reactor, which was designed to produce
tritium. Over the years, the design has also supported
steam cycle, direct cycle, process heat, and weapons
material disposition applications. More recently, DOE
established the AGR Fuel Development and Qualification Program to provide a baseline fuel qualification
data set at a peak fuel centerline temperature of
1250  C15,16 in support of the licensing and operation
of the Next Generation Nuclear Plant (NGNP).
Irradiation test conditions employed by the United
States generally covered projected fuel operating conditions. US fuel was to operate at temperatures as


166

TRISO-Coated Particle Fuel Performance


Table 20

HFR-K6 and HFR-K5 irradiation data

Start date
End date
Duration (full
power days)
Cell
Burnup (% FIMA)
Fast fluence
(1025 mÀ2,
E > 0.10 MeV)
Temperature
EOL 85mKr R/B

Table 21

HFR-K6

HFR-K5

1990
4 May 1993
634

1991
16 May 1994
564


1
7.2
3.2

2
9.3
<4.8

3
9.7
4.8

4
9.2
<4.8

1
6.7
2.9

2
8.8
<4.3

3
9.1
4.3

4

8.7
<4.3

Cycled
3 Â 10À7

Cycled
3 Â 10À7

Cycled
3 Â 10À7

Cycled
3 Â 10À7

Cycled
3 Â 10À7

Cycled
3 Â 10À7

Cycled
3 Â 10À7

Cycled
3 Â 10À7

Historical US particle fuel development and testing sequence

Date of design

conception

Reactor/status

Major fuel form tested

1960
1964

Peach Bottom built
Fort St. Vrain built

1967

LHTGR design only

1984

NE-MHTGR commercial design only

1989
1995

NP-MHTGR government design only
GT-MHR commercial design only

2005

NGNP design


BISO-coated (Th, U)C2
TRISO-coated (Th, U)C2 fissile
TRISO-coated ThC2 fertile
TRISO-coated UC2 fissile
BISO and TRISO-coated ThO2 fertile
TRISO-P coated UCO fissile
TRISO-P coated ThO2 fertile
TRISO-P coated UCO
TRISO-coated UCO fissile
TRISO-coated UCO and/or UO2 fertile fuel not yet tested
TRISO-coated UCO fissile

high as 1400  C and reach full burnup (commensurate
with 235U enrichment and kernel composition) at fast
fluences of 4 Â 1025 n mÀ2. With the exception of irradiation duration, the various experiments performed
either bounded expected nominal conditions or were
purposely varied to meet other test objectives. In order
to obtain results in a timely manner, US tests were
accelerated by factors of 3–10.
The particle fuel irradiation experiments and PIE
results summarized below consider only selected
tests of key US fuel types. These fuel types include
TRISO fissile/BISO fertile particles, weak acid resin
(WAR) TRISO fissile/BISO fertile particles, TRISO
fissile/TRISO fertile particles, and TRISO-P fissile
particles (conventional TRISO-coated particles with
an additional ‘protective’ pyrolytic carbon layer above
the outer pyrolytic carbon layer) as well as TRISO
fissile particles. General Atomics and Babcock &
Wilcox manufactured the majority of the kernel and

coating batches. However, some of the batches were
manufactured by ORNL. The following US

experiment summaries are listed in chronological
order and are not grouped by fuel type. Listed configuration and irradiation data are actual values, not
specification values or ranges. Interpretations of PIE
results are from the original sources and no overt
attempt has been made to reinterpret the results.
3.07.2.3.1 F-30

The F-30 experiment was irradiated in the General Electric Test Reactor (GETR) at Pleasanton,
California.17 The primary objective of this experiment was to demonstrate the irradiation performance
of Fort St. Vrain production fuel. Five independently
gas-swept cells contained the fuel. Cells 1, 3, and 4
contained only fuel compacts, Cell 2 contained only
loose particles, and Cell 5 contained both fuel compacts and loose particles. Configuration and irradiation data are given in Tables 22 and 23, respectively.
Postirradiation metallographic examination of
seven fuel compacts containing fissile and fertile
particles was performed. In addition, five sets of


TRISO-Coated Particle Fuel Performance

167

loose fissile particles and five sets of loose fertile
particles were examined. Fissile particle failure,
defined as a crack completely through the SiC layer,
ranged between 0% and 6.1%, while fertile particle
failure ranged between 0% and 15.1%. A typical

photomicrograph of SiC failure in an F-30 fissile
particle is presented in Figure 6. Metallography
revealed that inner pyrolytic carbon layers had
remained bonded to the SiC layer throughout irradiation. Figure 7 displays a typical photomicrograph of
a fissile particle with an IPyC layer crack and a
densified buffer.
50 mm

3.07.2.3.2 HRB-4 and HRB-5

The HRB-4 and HRB-5 capsules were irradiated in
HFIR at ORNL.18 The main objective of these
experiments was to test candidate fuel materials
and manufacturing processes for the proposed large
HTGR. Each test involved a single gas-swept cell
containing six fuel compacts vertically positioned.

Table 22

F30 configuration

Number of cells
Total number of fuel compacts
Cylindrical fuel compact diameter
Cylindrical fuel compact lengths
Fissile fuel type
Nominal Th/U ratio
235
U enrichment
Fissile particle diameter

Fertile fuel type
Fertile particle diameter
Number of fissile particle batches
Number of fertile particle batches
Defective SiC layer
fraction – fissile particles
Defective SiC layer
fraction – fertile particles

Table 23

Figure 6 A typical SiC layer crack in an F-30 fissile fuel
particle. Reproduced from Scott, C .B.; Harmon, D. P.
Post Irradiation Examination of Capsule F-30; GA-A13208,
UC-77; General Atomics Report, 1975.

5
13
12.45 mm
18.54 and 49.28 mm
HEU (Th, U)C2 TRISO
4.25
93%
429–560 mm
ThC2 TRISO
648–771 mm
7
9
<5 Â 10À4–1.5 Â 10À3
3 Â 10À4–1.0 Â 10À3


100 mm
Figure 7 A typical IPyC layer crack in a fissile F-30 fuel
particle. Reproduced from Scott, C. B.; Harmon, D. P.
Post Irradiation Examination of Capsule F-30; GA-A13208,
UC-77; General Atomics Report, 1975.

F30 irradiation data

Start date
End date
Duration (full power days)
Cell
Fissile burnup (% FIMA)
Fertile burnup (% FIMA)
Fast fluence (1025 n mÀ2, E > 0.18 MeV)
Time average peak temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

15 May 1972
5 April 1973
269
1
15
3
8
1100
2 Â 10À6
8 Â 10À6


2
19
4.5
10.5
1100
7 Â 10À7
1 Â 10À4

3
20
5
11.5
1120
8 Â 10À7
1 Â 10À5

4
18
4
9.5
1100
7 Â 10À7
2 Â 10À5

5
12
1.5
12
1200

2 Â 10À6
2 Â 10À5


168

TRISO-Coated Particle Fuel Performance

Table 24

HRB-4 and HRB-5 configurations

Number of cells
Number of fuel compacts
Cylindrical fuel compact
diameter
Cylindrical fuel compact
lengths
Fissile fuel type
Fertile fuel type
U enrichment
Fissile particle diameter
Fertile particle diameter
Fissile particle batch
Fertile particle batch
Total number of fissile
particles
Total number of fertile
particles


235

Table 25

HRB-4

HRB-5

1
6
12.4 mm

1
6
12.4 mm

25.4 mm

25.4 mm

WAR UC2
TRISO
ThO2 BISO
5.99%
639 mm
805 mm
OR52A
T01424BIL
17 780


WAR UC2
TRISO
ThO2 BISO
5.99%
639 mm
805 mm
OR52A
T01424BIL
17 780

4180

4180

Figure 8 Typical HRB-4 fissile particle irradiated to 27.7%
FIMA and 10.5 Â 1025 n mÀ2 fast fluence. Reproduced
from Scott, C. B.; Harmon, D. P. Post Irradiation
Examination of Capsules HRB-4, HRB-5, and HRB-6;
GA-A13267, UC-77; General Atomics Report, 1975.

HRB-4 and HRB-5 irradiation data

Start date
End date
Duration (full power
days)
Peak fissile burnup
(% FIMA)
Peak fertile burnup
(% FIMA)

Peak fast fluence
(1025 n mÀ2,
E > 0.18 MeV)
Peak temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

HRB-4

HRB-5

8 October 1972
26 June 1973
244

8 October 1972
3 February 1973
107

27.7

15.7

13.4

4.3

10.5

4.7


1250
1.4 Â 10À5
3.2 Â 10À4

1250
3 Â 10À6
1 Â 10À4

Configuration and irradiation data are given in
Tables 24 and 25.
Metallographic examinations were performed on
each fuel compact. A typical photomicrograph of
an irradiated HRB-4 fissile particle is presented in
Figure 8, which shows the formation of gas bubbles
in the kernel and the densification of the buffer. IPyC
layers of the examined fissile particles had remained
bonded to the SiC. The examination indicated that the
fissile particles had failed between 0% and 6% of the
SiC layers. These failures consisted primarily of radial
cracks through the SiC layer. Between 4% and 73% of

Figure 9 Photomicrographs of typical fission product
attack in irradiated HRB-4 fissile particles. Reproduced
from Scott, C. B.; Harmon, D. P. Post Irradiation
Examination of Capsules HRB-4, HRB-5, and HRB-6;
GA-A13267, UC-77; General Atomics Report, 1975.

the OPyC layers failed during irradiation. There were
no tabulations of IPyC layer failures reported.

Several of the fissile particles examined displayed
evidence of fission product attack. This attack mostly
occurred in large concentrations at the IPyC–SiC
interface and where fission products in smaller concentrations had diffused up to 25 mm into the SiC.
Figure 9 presents typical photomicrographs of fission product attack in HRB-4 fissile particles.
In HRB-5, IPyC layers of the examined fissile
particles had remained bonded to the SiC. There


TRISO-Coated Particle Fuel Performance

Table 26

HRB-6 configuration

Number of cells
Number of fuel compacts
Cylindrical fuel compact diameter
Cylindrical fuel compact length
Fissile fuel type
Nominal Th/U ratio
235
U enrichment
Fissile particle diameter
Fertile fuel type
Fertile particle diameter
Fissile particle batch
Fertile particle batch
Defective SiC layer
fraction – fissile particles


Figure 10 Typical HRB-4 fissile particle irradiated to
27.7% FIMA and 10.5 Â 1025 n mÀ2 fast fluence.
Reproduced from Scott, C. B.; Harmon, D. P. Post
Irradiation Examination of Capsules HRB-4, HRB-5,
and HRB-6; GA-A13267, UC-77; General Atomics Report,
1975.

were no tabulations of IPyC layer failures reported.
There was no evidence of fission product attack as
seen in the HRB-4 fissile particles. However, the
examination indicated that between 0.4% and
17% of the SiC layers in fissile particles had failed.
These failures consisted primarily of radial cracks
through the SiC layer. A typical photomicrograph
of irradiated HRB-5 fissile particles with cracked
SiC layers is presented in Figure 10. This photomicrograph is also representative of HRB-4 fissile particles with cracked SiC layers. It was reported that a
large fraction of these cracked SiC layers were due to
metallographic preparation and not a result of fast
neutron exposure or fuel burnup effects.
3.07.2.3.3 HRB-6

The HRB-6 capsule was irradiated in HFIR at
ORNL.18 Fissile fuel particles used in HRB-6 came
from the same production batch as used in the first core
of Fort St. Vrain and were one of the batches previously
irradiated in the F-30 experiment. This test involved a
single gas-swept cell containing six fuel compacts vertically positioned. During operation, the sweep gas flow
rate was reduced because of high activity in the sweep
lines. Because of this gas flow reduction, in-pile fission

gas-release data were not obtained. The irradiation of
HRB-6 in HFIR coincided with part of the HRB-4
irradiation. Configuration and irradiation data are
given in Tables 26 and 27.

Table 27

169

1
6
12.4 mm
25.4 mm
HEU (Th, U)C2 TRISO
4.25
93.15%
556 mm
ThO2 BISO
888 mm
CU6B-2427
T01451BIL-W
<5 Â 10À4

HRB-6 irradiation data

Start date
End date
Duration (full power days)
Peak fissile burnup (% FIMA)
Peak fertile burnup (% FIMA)

Peak fast fluence (1025 n mÀ2,
E > 0.18 MeV)
Peak temperature ( C)
Minimum TRIGA BOL 85mKr R/B
Maximum TRIGA EOL 85mKr R/B

27 February 1973
8 September 1973
183
26.6
9.3
7.9
1100
5.0 Â 10À7
2.7 Â 10À4

PIE included gas-release measurements of each
fuel compact performed in the Training Research
and Isotope Production, General Atomics (GA)
(TRIGA) reactor. However, during the unloading of
the HRB-6 capsule, fuel compacts 2A, 2B, and 2C
were damaged and as many as 30 broken fuel particles were observed. Therefore, the TRIGA gasrelease measurements at EOL for these compacts
would be higher than in-pile sweep line measurements had they been performed.
A typical photomicrograph of an irradiated
HRB-6 fissile particle is presented in Figure 11,
which shows the formation of gas bubbles in the
kernel and densification of the buffer. The photomicrograph also shows an incipient crack in the IPyC
layer. No tabulations of IPyC layer failures were
reported. IPyC layers of the examined fissile particles
had remained bonded to the SiC, and there was no

evidence of fission product attack. However, the
examination indicated that the fissile particles had
failed between 0% and 2% of the SiC layers. These
failures do not include the fissile particles broken
during capsule unloading. It was reported that a


170

TRISO-Coated Particle Fuel Performance

Table 28

OF-2 configuration

Number of cells
Total number of fuel compacts
Cell 1 cylindrical fuel compact
dimensions (16 compacts)
Cell 2 cylindrical fuel compact
dimensions (48 compacts)
Cell 2 cylindrical fuel compact
dimensions (24 compacts)
Fissile fuel type

235

Figure 11 Typical HRB-6 fissile particle irradiated to
26.5% FIMA and 7.9 Â 1025 n mÀ2 fast fluence. Reproduced
from Scott, C. B.; Harmon, D. P. Post Irradiation

Examination of Capsules HRB-4, HRB-5, and HRB-6;
GA-A13267, UC-77; General Atomics Report, 1975.

U enrichment
Fissile particle diameter
Fertile fuel type
Fertile particle diameter
Number of fissile particle batches
Number of fertile particle batches

Table 29

large fraction of these failures were due to metallographic preparation.
3.07.2.3.4 OF-2

The OF-2 capsule was irradiated in the Oak Ridge
Research Reactor (ORR).19 The main objectives of
the test were to investigate the irradiation performance of various particle fuel forms (mostly WAR
UCO with different stoichiometries) and to compare
the performance of fuel particles fabricated from
different coaters. OF-2 consisted of 88 fuel compacts
(and several sets of loose inert particles) contained in
a single capsule that was divided into two independently gas-swept cells. Various combinations from 15
fissile batches, 16 fertile batches, and 4 compact
matrix compositions comprised the fuel compacts
(each compact contained fuel from only one fissile
batch and one fertile batch). Configuration and irradiation data are given in Tables 28 and 29.
Postirradiation metallography was performed
on three fuel compacts from Cell 1 and on 27 fuel
compacts from Cell 2. A significant level of OPyC

layer failures was observed in the fissile TRISOcoated particles from Cell 1. However, there were
no observed SiC layer failures or any layer failures
in the BISO-coated fertile and inert particles in these
compacts. Examination of 11 fuel compacts from
Cell 2, containing the same three fissile particle
batches as in Cell 1, also indicated significant levels
of OPyC layer failures. The fissile particle batch
with the highest OPyC anisotropy (optical Bacon

2
88
15.75 mm diameter,
25.4 mm long
15.75 mm OD,
3.30 mm ID,
12.70 mm long
15.75 mm diameter,
50.8 mm long
WAR UCxOy TRISO
(Th, U)O2 TRISO
UC2 TRISO
Not reported
600–753 mm
ThO2 BISO
806–889 mm
15
16

OF-2 irradiation data


Start date
End date
Duration (full power days)
Cell
Burnup (% FIMA)
Fast fluence (1025 n mÀ2,
E > 0.18 MeV)
Maximum temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

21 June 1975
1 August 1976
352
1
2
75.9–79.6
50.0–79.5
5.86–8.91
1.94–8.36
1350
2 Â 10À5
1 Â 10À4

1350
1 Â 10À4
5 Â 10À6

anisotropy factor (BAF) ¼ 1.069) had 100% OPyC
layer failure, while the other two batches with lower

anisotropy (optical BAF of 1.035 and 1.030) had
0–33% OPyC layer failures.
Of the 30 fuel compacts metallographically examined, only one compact (that contained WAR UCO
fissile particles) displayed cracked SiC layers. Among
the 27 fissile particles observed in this compact,
16 displayed cracked SiC layers. These cracks were
identified as artifacts of polishing. However, no
photomicrographs of these cracks were presented to
support this conclusion. The metallographic examinations also revealed typical WAR UCO behavior of
kernel and buffer densification. This densification
was also accompanied by varying degrees of kernel
migration.
Figure 12 presents a typical WAR UCO photomicrograph that displays kernel and buffer densification, and OPyC layer failure. Examination of OF-2
particles also indicated several incidences of fission


TRISO-Coated Particle Fuel Performance

product accumulation at the IPyC and SiC interface.
A typical photomicrograph of fission product accumulation is presented in Figure 13.

Figure 12 Photomicrograph of irradiated OF-2 fissile
WAR UCO particle. Reproduced from Tiegs, T. N.;
Thoms, K. R. Operation and Post Irradiation Examination
of ORR Capsule OF-2: Accelerated Testing of HTGR
Fuel; ORNL-5428; 1979. Courtesy of Oak Ridge National
Laboratory, U.S. Department of Energy.

3.07.2.3.5 HRB-14


The HRB-14 capsule was irradiated in HFIR at
ORNL.20 The main objectives of this experiment
were to test LEU particles and to demonstrate
reduced matrix–OPyC layer interactions by using
cure-in-place fuel compacts. This test involved a
single gas-swept cell equally divided among 20 fuel
compacts vertically positioned and molded planchets
(wafers) containing BISO-coated ThO2 fertile particles. Online fission gas-release measurements were
not reported. Also, irradiation results from the BISOcoated fertile particles were reported separately and
are not included in this summary. Configuration and
irradiation data are given in Tables 30 and 31.
Disassembly of the HRB-14 capsule after irradiation produced five fuel compacts with no remaining
structure; in essence, there were five collections of
loose particles, four compacts that were partially
intact, nine compacts that were intact but displayed
significant amounts of debonding, and only two compacts in relatively good shape.

Table 30

Lower half of HRB-14 configuration

Number of cells
Total number of fuel compacts
Cylindrical fuel compact diameter
Cylindrical fuel compact length
Fissile fuel type

235

U enrichment

Fissile particle diameter
Fertile fuel type
Fertile particle diameter
Number of fissile particle batches
Number of fertile particle batches
Defective SiC layer
fraction – fissile particles

Table 31

Figure 13 Photomicrograph of irradiated OF-2 fissile
fuel particles displaying fission product accumulation
at IPyC–SiC interface. Reproduced from Tiegs, T. N.;
Thoms, K. R. Operation and Post Irradiation Examination of
ORR Capsule OF-2: Accelerated Testing of HTGR Fuel;
ORNL-5428; 1979. Courtesy of Oak Ridge National
Laboratory, U.S. Department of Energy.

171

1
20
12.50 mm
9.52 mm
UCxOy TRISO
(Th, U)O2 TRISO
UO2 TRISO
19.18–19.66%
760–813 mm
ThO2 TRISO

786–882 mm
5
8
7.0 Â 10À7–
1.3 Â 10À4

Lower half of HRB-14 irradiation data

Start date
End date
Duration (full power days)
Peak fissile burnup (% FIMA)
Peak fertile burnup (% FIMA)
Peak fast fluence (1025 n mÀ2,
E > 0.18 MeV)
Maximum temperature ( C)
Minimum temperature ( C)
Minimum TRIGA BOL 85mKr R/B
Maximum TRIGA EOL 85mKr R/B

20 May 1978
4 January 1979
214
28.6
8.5
8.3
1190
895
3.8 Â 10À7
3.0 Â 10À4



172

TRISO-Coated Particle Fuel Performance

Metallographic examination was performed on 15
fuel compacts, and 8 of them contained fissile particles. A few fissile particles were reported to have SiC
layer cracks but these cracks were attributed to
metallographic preparation. It should be noted that
visual inspection of each compact during capsule
disassembly indicated that between 0% and 9% of
the visible particles (from compact surfaces and loose
particles that had fallen off) had failed SiC layers.
However, this visual inspection did not distinguish
between fissile and fertile particles.
The metallographic examination of fissile particles revealed that between 0% and 3% of the IPyC
layers had failed (cracked) and that the IPyC layers
had debonded from the SiC in 0% to 7.7% of the
particles. Buffer layers did not crack in the UO2 or
(Th, U)O2 fuel but did crack in 10–71% of the UCO
fuel particles. Kernel extrusion was reported only in
UCO fuel. Figure 14 displays typical kernel extrusion, and Figure 15 presents a typical photomicrograph of kernel migration.
In several particles of each fuel form, high concentrations of fission products were observed in
small, localized regions at the SiC–IPyC layer interface. In addition to fission product accumulation,
localized chemical attack was also observed in the
SiC layers of several (Th, U)O2 and UO2 fuel particles.
This localized attack, which had penetrated $2 mm
into the SiC, was attributed to palladium, and was


Figure 14 Photomicrograph of a UCO particle (batch
6157-08-020) from Compact 10 irradiated at 1040  C to
27.8% FIMA and to a fast fluence (E > 0.18 MeV) of
7.1 Â 1025 n mÀ2 displaying kernel extrusion. Reproduced
from Young, C. A. Pre- and Post Irradiation Evaluation of
Fuel Capsule HRB-14; GA-A15969, UC-77; General
Atomics Report, 1980.

observed in 8% of the particles. UCO fuel particles
that did not display localized chemical attack, had
uniform attack along the inner SiC layer (usually on
one side of the particles). This uniform attack was
attributed to rare earth fission products. Figure 16
displays typical uniform fission product attack in a
UCO fuel particle. It should be noted that with
optimized UCO stoichiometry, the kernel retains rare
earth fission products and does not display kernel
migration as found here with non-optimized UCO
kernels containing excess UC2 leading to rare earth
migration.
Metallographic examination of fertile particles
indicated that between 0% and 2.4% of the particles
in each compact had total coating failure, defined as
cracked OPyC and SiC layers. These failures were
attributed to pressure vessel failure. Figure 17 displays a typical failed fertile particle. Separate tallies
of particles where only the SiC layer had failed were
not reported. Other fertile particle observations
include the following:
 1.5–29.1% of the particles had failed OPyC layers
 8–70% of the particles had failed IPyC layers

 11–85% of the particles had IPyC layers debonded
from the SiC
 6–26% of the particles had cracked buffers
 no kernel migration was observed
 a few kernels had extruded into buffer cracks.

Figure 15 Photomicrograph of a UCO particle (batch
6157-08-020) from Compact 10 irradiated at 1040  C to
27.8% FIMA and to a fast fluence (E > 0.18 MeV) of
7.1 Â 1025 n mÀ2. Reproduced from Young, C. A. Pre- and
Post Irradiation Evaluation of Fuel Capsule HRB-14;
GA-A15969, UC-77; General Atomics Report, 1980.


TRISO-Coated Particle Fuel Performance

Table 32

HRB-15B configuration

Number of cells
Total number of particle trays
Maximum number of loose
particles per tray
Particle tray outer diameter
Particle tray inner diameter
Fissile fuel type

235


U enrichment
Fissile particle diameter
Fertile fuel type
Figure 16 Photomicrograph of a UCO particle (batch
6157-08-020) from Compact 10 irradiated at 1040  C to
27.8% FIMA and to a fast fluence (E > 0.18 MeV) of
7.1 Â 1025 n mÀ2 displaying fission product attack of the
SiC layer. Reproduced from Young, C. A. Pre- and Post
Irradiation Evaluation of Fuel Capsule HRB-14; GA-A15969,
UC-77; General Atomics Report, 1980.

Fertile particle diameter
Number of fissile particle batches
Number of fertile particle batches

The primary objective of the HRB-15B experiment
irradiated in HFIR at ORNL21 was to test a variety
of LEU fissile fuel designs and ThO2 fertile particle
designs. This test involved a single gas-swept cell

22.3–23.6 mm
11.1 mm
UCO TRISO and
silicon-BISO
(Th, U)O2 TRISO and
silicon-BISO
UC2 TRISO and
silicon-BISO
UO2 TRISO and
silicon-BISO

UO2* TRISO and
silicon-BISO
$19.5%
742–951 mm
ThO2 TRISO, BISO
and silicon-BISO
773–836 mm
19
22

HRB-15B irradiation data

Start date
End date
Duration (full power days)
Peak fissile burnup (% FIMA)
Peak fertile burnup (% FIMA)
Peak fast fluence (1025 n mÀ2,
E > 0.18 MeV)
Time average temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

3.07.2.3.6 HRB-15B

1
184
116

Note: Two types of UO2* fuel were tested, one with ZrC dispersed

in the buffer and the other with pure ZrC layer around the kernel.

Table 33

Figure 17 Photomicrograph of a ThO2 fertile particle
(batch 6252-17-010) irradiated at 1130  C to 8.5% FIMA
and to a fast fluence (E > 0.18 MeV) of 8.3 Â 1025 n mÀ2
displaying pressure vessel failure. Reproduced from
Young, C. A. Pre- and Post Irradiation Evaluation of Fuel
Capsule HRB-14; GA-A15969, UC-77; General Atomics
Report, 1980.

173

6 July 1978
4 January 1979
169
26.7
6.0
6.6
815–915
2.9 Â 10À8
5.1 Â 10À6

containing 184 thin graphite trays. Each tray could
accommodate up to a maximum of 116 individual,
unbonded fuel particles. The loose fissile fuel particles included UC2, UCO with four different stoichiometries, (Th, U)O2, UO2, and two types of UO2*
(one type had ZrC dispersed throughout the buffer
layer and the other had a pure ZrC coating around
the kernel). Each fissile fuel type was tested with

both TRISO coating and silicon–BISO coating
which consisted of the kernel surrounded by a buffer
layer, an IPyC layer, and finally a silicon doped
OPyC layer. The loose fertile particles tested
included TRISO-, BISO-, and silicon–BISO-coated
ThO2. Configuration and irradiation data are
provided in Tables 32 and 33.


174

TRISO-Coated Particle Fuel Performance

Postirradiation metallography was performed on
20 different particle types, each consisting of approximately 20 particles. These examinations revealed
considerable gas bubble formation in UC2 and UCO
kernels, and buffer densification in TRISO-coated
particles. Some SiC layer cracking was observed in
each TRISO-coated fuel type, but mostly in the
UCO particles. These cracks were reported to have
occurred during mount preparation because of the
crack orientation and because the visual examination
detected no OPyC cracking. No further tabulation of
layer failures was reported.
3.07.2.3.7 R2-K13

The R2-K13 capsule was irradiated in the R2 reactor
at Studsvik, Sweden.22 The main objective of this
experiment was to test reference UCO fissile particles and ThO2 fertile particles. Four independently
gas-swept cells were positioned vertically on top of

one another. The middle two cells contained US fuel.
The top and bottom cells each contained a full-size
German fuel sphere (discussed in the section on
German irradiation results). Configuration and irradiation data are given in Tables 34 and 35.
Postirradiation metallographic examination was
performed on two fuel compacts. All of the 99
fissile particles examined displayed debonding
between the buffer and IPyC layers. In some
cases, debonding between the buffer, IPyC, and
SiC layers was also observed. Likewise, all of the
68 fertile particles examined displayed debonding
between the buffer, IPyC, and SiC layers. The SiC
layers of all the particles examined were observed
to be intact.
Table 34

R2-K13 US configuration

Number of cells
Total number of fuel compacts
Cylindrical fuel compact diameter
Cylindrical fuel compact length
Total number of piggyback sample
sets
Fissile fuel type
Fertile fuel type
235
U enrichment
Fissile particle diameter
Fertile particle diameter

Fissile particle batches
Fertile particle batches
Defective SiC layer fraction – fissile
particles
Defective SiC layer fraction – fertile
particles

2
12
12.52 mm
25.4 mm
31
LEU UCO TRISO
ThO2 TRISO
19.61%
803 and 824 mm
781–805 mm
2
3
1.9 Â 10À4 and
4.4 Â 10À4
<2 Â 10À6–
1.6 Â 10À5

3.07.2.3.8 HRB-15A

The main objective of the HRB-15A experiment
irradiated in HFIR at ORNL23 was to test several
candidate fuel designs for the proposed Large High
Temperature Gas Reactor (LHTGR). This test

involved a single gas-swept cell containing 20 cylindrical fuel compacts positioned vertically on top of
one another. Interspersed between the fuel compacts
were 17 tray assemblies. Each assembly had a graphite tray holding 54 unbonded particles in separate
holes, and serving as a lid, a graphite wafer containing
54 particles bonded in separate holes with carbonaceous matrix material. Configuration and irradiation
data are given in Tables 36 and 37.
Table 35

R2-K13 US irradiation data

Start date
End date
Duration (full power days)
Cell
Peak fissile burnup (% FIMA)
Peak fertile burnup (% FIMA)
Peak fast fluence (1025 n mÀ2,
E > 0.18 MeV)
Average center temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

Table 36

22 April 1980
19 September 1982
517
2
3
22.5

22.1
4.6
4.5
7.8
7.4
1190
1 Â 10À5
8 Â 10À5

HRB-15A configuration

Number of cells
Total number of fuel compacts
Cylindrical fuel compact diameter
Number of short fuel compacts/length
Number of long fuel compacts/length
Number of bonded wafer/unbonded
tray assemblies
Fissile fuel type

Fertile fuel type
235

985
2 Â 10À7
8 Â 10À6

U enrichment
Fissile particle diameter
Fertile particle diameter

Fissile particle batches
Fertile particle batches
Defective SiC layer fraction – fissile
particles
Defective SiC layer fraction – fertile
particles

1
20
12.54 mm
3/9.53 mm
17/19.05 mm
17
UCO TRISO
UC2 TRISO
UC2 ZrC-TRISO
UO2 TRISO
UO2 ZrC-TRISO
UO2*
ThO2 TRISO
ThO2 silicon-BISO
$19.5%
736–894 mm
713–1014 mm
10
5
1.4 Â 10À5–
7.4 Â 10À2
6.7 Â 10À5–
1.4 Â 10À3


Note: Two types of UO2* fuel were tested, one with ZrC dispersed
in the buffer and the other with pure ZrC layer around the kernel.


TRISO-Coated Particle Fuel Performance

Table 37

HRB-15A irradiation data

Start date
End date
Duration (full power days)
Peak fissile burnup (% FIMA)
Peak fertile burnup (% FIMA)
Peak fast fluence (1025 n mÀ2,
E > 0.18 MeV)
Average center temperature ( C)
BOL 85mKr R/B
EOL 85mKr R/B

Table 38

175

HRB-16 configuration

26 July 1980
29 January 1981

174
29.0
6.4
6.5

Number of cells
Total number of fuel compacts
Cylindrical fuel compact diameter
Cylindrical fuel compact length
Number of loose particle trays
Number of particles per tray

1150
6.96 Â 10À6
3.76 Â 10À4

Fissile fuel type

Fertile fuel type
235

U enrichment
Fissile particle diameter
Fertile particle diameter
Fissile particle batches
Fertile particle batches
Defective SiC layer fraction – fissile
particles
Defective SiC layer fraction – fertile
particles


1
18
12.45 mm
18.70 mm
27
110 (2 particles
per hole)
UCO TRISO
UCO ZrC-TRISO
UC2 TRISO
UC2 ZrC-TRISO
UO2 TRISO
UO2* TRISO
(Th, U)O2 TRISO
ThO2 TRISO
ThC2 BISO
19.20–19.61%
742–884 mm
756 and 786 mm
9
2
4.6 Â 10À7–
4.4 Â 10À4
1.6 Â 10À5 and
5.0 Â 10À4

Note: Two types of UO2* fuel were tested, one with ZrC dispersed
in the buffer and the other with pure ZrC layer around the kernel.


Figure 18 Photomicrograph of a UO2 ZrC–TRISO-coated
particle (batch 6162-00-010) irradiated at 1075  C to
27.2% FIMA and to a fast fluence of 6.0 Â 1025 n mÀ2
(E > 0.18 MeV) displaying ZrC layer cracks. Reproduced
from Ketterer, J.; et al. Capsule HRB-15A Post Irradiation
Examination Report; GA-A16758, UC-77; General Atomics
Report, 1984.

Postirradiation metallographic examination was
performed on five fuel compacts. Between 0% and
5.6% SiC (and OPyC) layer failures were reported
for the UO2 particles but were attributed to sample
preparation. In contrast, the ZrC layer failures
observed in the UO2 ZrC–TRISO-coated particles
were also attributed to sample preparation but were
not tabulated. A photomicrograph of a UO2 ZrC–
TRISO-coated particle displaying a cracked ZrC
layer is presented in Figure 18. No SiC layer failures
were reported for the UCO fuel.
Between 0% and 12.5% of the SiC layers and
between 83% and 92% of the IPyC layers were

reported to have failed in the fertile particles. These
high layer failures for the fertile ThO2 particles were
attributed to the high IPyC BAF values for these particles. The high BAF was a result of intentionally depositing the IPyC layer at low coating rates in an attempt to
produce layers that were impermeable to chlorine
(chlorine trapped in the particle during SiC deposition
may enhance SiC degradation during irradiation).
3.07.2.3.9 HRB-16


The main objective of the HRB-16 experiment conducted in the HFIR at ORNL24 was to test a variety of
LEU fissile particle fuel designs. This test involved a
single gas-swept cell containing 18 fuel compacts
stacked vertically and interspersed with 27 trays of
unbonded particles and several encapsulated fission
product piggyback transport specimens. Configuration
and irradiation data are given in Tables 38 and 39.
Postirradiation metallographic examination was
performed on seven fuel compacts that contained
particles from six different fissile batches and one
fertile batch. For fuel compacts containing multiple
fissile batches, the following visual criteria were used
to identify fuel forms:
 UO2* had the conspicuous, bright ZrC layer next
to the kernel