3.08

Advanced Concepts in TRISO Fuel

K. Minato and T. Ogawa
Japan Atomic Energy Agency, Tokai-mura, Ibaraki, Japan

ß 2012 Elsevier Ltd. All rights reserved.

3.08.1

Introduction

216

3.08.2
3.08.2.1
3.08.2.2
3.08.2.3
3.08.2.4
3.08.2.4.1
3.08.2.4.2
3.08.2.4.3
3.08.2.4.4
3.08.2.4.5
3.08.3
3.08.3.1
3.08.3.2
3.08.3.2.1
3.08.3.2.2
3.08.4

3.08.4.1
3.08.4.2
3.08.4.3
3.08.4.3.1
3.08.4.3.2
3.08.5
3.08.5.1
3.08.5.2
3.08.6
References

ZrC-Coated Particle Fuel
Designs of ZrC-Coated Particle Fuel
Fabrication of ZrC-Coated Particle Fuel
Characterization Techniques for ZrC-Coated Particle Fuel
Performance of ZrC-Coated Particle Fuel
Irradiation performance
Resistance to chemical attack by fission products
High-temperature stability
Retention of fission products
Behavior under oxidizing conditions
ZrC-Containing TRISO-Coated Particle Fuel
Designs of ZrC-Containing TRISO-Coated Particle Fuel
Performance of ZrC-Containing TRISO-Coated Particle Fuel
Irradiation performance
Retention of fission products
SiC-Containing TRISO-Coated Particle Fuel
Designs of SiC-Containing TRISO-Coated Particle Fuel
Fabrication of SiC-Containing TRISO-Coated Particle Fuel
Performance of SiC-Containing TRISO-Coated Particle Fuel

Irradiation performance
Behavior under simulated conditions
TiN-Coated Particle Fuel
Designs of TiN-Coated Particle Fuel
Fabrication of TiN-Coated Particle Fuel
Outlook

216
216
217
219
220
220
221
222
225
227
227
227
229
229
230
231
231
232
233
233
233
234
234

234
235
235

Abbreviations
DB-MHR
DBa(ZrC)
DCe(ZrC)
DCs(SiC)
DCs(ZrC)
DRu(ZrC)
DSr(ZrC)
GA
GAC

Deep-burn modular helium reactor
Diffusion coefficient for Ba in the ZrC
coating layer
Diffusion coefficient for Ce in the ZrC
coating layer
Diffusion coefficient for Cs in the SiC
coating layer
Diffusion coefficient for Cs in the ZrC
coating layer
Diffusion coefficient for Ru in the ZrC
coating layer
Diffusion coefficient for Sr in the ZrC
coating layer
General Atomics
General Atomic Company


GFR
HFIR
HTGR
ICP-AES
IPyC
JAEA
JAERI
JMTR
JRR-2
LANL
LASL
LMFBR
MTS
OPyC

Gas fast reactor
High Flux Isotope Reactor
High-temperature gas-cooled reactor
Inductively coupled plasma-atomic
emission spectrometry
Inner dense PyC
Japan Atomic Energy Agency
Japan Atomic Energy Research
Institute
Japan Materials Testing Reactor
Japan Research Reactor-2
Los Alamos National Laboratory
Los Alamos Scientific Laboratory
Liquid metal fast breeder reactor

Methyltrichlorosilane
Outer dense PyC

215


216

ORR
PyC
R/B
VHTR

Advanced Concepts in TRISO Fuel

Oak Ridge Research Reactor
Pyrolytic carbon
Release-to-birth ratio
Very-high-temperature reactor

3.08.1 Introduction
The TRISO-coated fuel particle consists of a microspherical fuel kernel and coating layers of porous
pyrolytic carbon (PyC), inner dense PyC (IPyC),
silicon carbide (SiC), and outer dense PyC (OPyC).
The chemical form of the fuel kernel can be oxide,
carbide, or a mixture of the two. The function of
these coating layers is to retain fission products
within the particle. The porous PyC coating layer,
called the buffer layer, attenuates fission recoils and
provides void volume for gaseous fission products

and carbon monoxide in the cases of oxide and oxycarbide fuels. The IPyC coating layer acts as a containment to gases during irradiation and protects
the fuel kernel from the reaction with the coating
gases during the SiC coating process. The SiC coating layer provides mechanical strength for the particle and acts as a barrier to the diffusion of metallic
fission products, which diffuse easily through the
IPyC layer. The OPyC coating layer protects the
SiC coating layer mechanically.
The recent interest in the coated particle fuel
concept includes its application outside the past
experience of the high-temperature gas-cooled reactor (HTGR)1:
1. Very-high-temperature reactor (VHTR) with a
gas outlet temperature of 1273 K for supplying
both the electricity and the process heat for hydrogen production, as proposed in the Generation-IV
International Forum.2
2. Actinide burning in deep-burn modular helium
reactor (DB-MHR) with fuel kernels consisting of
high concentrations of transuranium elements.3,4
3. Advanced gas fast reactor (GFR) with nitride fuel,
which aims at a more improved performance compared with the conventional liquid metal fast
breeder reactor (LMFBR) and/or the efficient
actinide burning.1,5
Although SiC has excellent properties, it gradually
loses its mechanical integrity at very high temperatures, especially >1973 K.6–8 The annealing temperature during fuel element fabrication is limited to
2073 K for 1 h. Higher temperatures should result

in a porous structure due to the b-SiC to a-SiC
transformation and the thermal dissociation of SiC,9
which leads to an extensive release of fission products
from the TRISO-coated fuel particles. The fuel temperatures were limited to well below 1973 K during
the design-basis accidents in HTGR designs.10–12
Chemical interaction of the SiC coating layer with

fission products is one of the possible performance
limitations of the TRISO-coated fuel particles. The
fuel performance of TRISO is described in Chapter
3.07, TRISO-Coated Particle Fuel Performance.
The fission product of palladium is known to react
with the SiC layer. Corrosion of the SiC layer could
lead to fracture of the coating layers or provide a
localized fast diffusion path, which degrades the
fission-product retention capability within the particle. Since the fission yield for palladium from 239Pu is
about tenfold that from 235U,13 a careful particle fuel
design should be made in the actinide burning.
The PyC layer develops gas permeability with
increasing fast neutron doses. Intactness of the IPyC
layer is crucial in keeping the integrity of the SiC
layer in the oxide fuel. When the IPyC layer fails, or
develops gas permeability, the SiC layer will react
with CO gas to form volatile SiO.14 The PyC layer
will also develop anisotropy above 2173 K, which is
deleterious to its irradiation behavior.
To improve the high-temperature performance of
the TRISO-coated fuel particles, a new material other
than SiC is needed. Zirconium carbide is a candidate,
and ZrC-coated particles, where the SiC layer was
replaced by a ZrC layer, have been tested. New configurations of the coating layers, with a layer containing SiC or ZrC added to the TRISO coating, have
been proposed and tested to improve the chemical
stability of the TRISO-coated particles. For the application to the fast reactor fuel, TiN coating layers have
been proposed and tested instead of PyC layers.
The following sections summarize the designs
and the research and development of the advanced
concepts in TRISO fuel: (1) ZrC-coated particle

fuel, (2) ZrC-containing TRISO-coated particle fuel,
(3) SiC-containing TRISO-coated particle fuel, and
(4) TiN-coated particle fuel.

3.08.2 ZrC-Coated Particle Fuel
3.08.2.1

Designs of ZrC-Coated Particle Fuel

Zirconium carbide (ZrC) is known as a refractory
and chemically stable compound, which melts eutectically with carbon at 3123 K. The properties of


Advanced Concepts in TRISO Fuel

ZrC are summarized in Chapter 2.13, Properties
and Characteristics of ZrC. To improve the hightemperature stability, the resistance to chemical
attack by fission products, and the retention of
fission products, the ZrC coating layer is a candidate that can replace the SiC coating layer of
the TRISO-coated fuel particle; the resulting particle is termed a ZrC-TRISO-coated fuel particle.
The apparent drawback of the ZrC-TRISO coating
may be that ZrC does not withstand the oxidation in
such an accident as a massive air-ingress accident
though it is highly hypothetical in the modern
HTGR designs.
Historically, several coating designs have been
tested in the United States15,16: (1) ZrC-TRISOcoated particles, (2) ZrC-TRISO type coated
particles without OPyC layer, (3) ZrC-coated particles with ZrC-doped OPyC layer, and (4) ZrCcoated particles with graded C–ZrC layer(s). In
the graded C–ZrC layer, the compositions were
changed gradually from the pure PyC through

the ZrC with excess carbon and into the pure ZrC.
The graded layer was applied to either the inside
or the outside surface of the ZrC layer. Propylene
was used to produce the pure PyC and to provide
the carbon for the graded portion of the codeposited
carbon and ZrC.
ZrC-coated fuel particles are being developed in
Japan17 since the early 1970s. The ZrC-coated fuel
particles at the early stage of the development were
characterized by a thick ZrC layer with a composition
of C/Zr > 1.0 and by the absence of the OPyC layer.
A ZrC layer of this kind was called ‘zirconium-carballoy,’ meaning ZrC–C alloy. Later, it was found that
the retention of metal fission products, especially
90
Sr, by the zirconium-carballoy was rather poor,18
presumably owing to a short circuit through the free
carbon phase. It was also felt from the irradiation
experiences that the presence of the OPyC layer
was essential for the mechanical integrity of the
coated fuel particle. The emphasis was, therefore,
placed on the development of ZrC-TRISO-coated
particles with the stoichiometric composition of
C/Zr ¼ 1.0.
Although most of the reported work on the use of
ZrC in coated fuel particles has been directed toward
the development of a replacement for the SiC barrier
layer, ZrC was also tested as a fission product and
oxygen getter,19,20 in which ZrC was deposited over
the fuel kernel or dispersed throughout the buffer
layer. These types of the coated particle fuels are

described in Section 3.08.3.

217

3.08.2.2 Fabrication of ZrC-Coated
Particle Fuel
The coating layers of ZrC and ZrC–C were produced
by chemical vapor deposition, in which the pyrolytic
reaction of zirconium halide with hydrocarbon in the
presence of hydrogen was used in principle. Mainly,
two different processes have been developed in supplying zirconium halide to the coater: (1) using ZrCl4
powder and (2) using in situ generation of zirconium
halide vapor.
The chemical vapor deposition of ZrC has been
studied using a gas mixture of CH4, H2, ZrCl4, and Ar
at Los Alamos Scientific Laboratory (LASL; now Los
Alamos National Laboratory, LANL).16,21–23 A key
development in the ZrC coating project proved to be
the ZrCl4 powder feeder for metering ZrCl4 into the
coater. ZrCl4 is a solid at room temperature and
sublimes at 625 K. In this process, hygroscopic
ZrCl4 powder was supplied from the powder feeder,
whose rate was controlled by the auger speed and
metered by the output of the load cell on which the
powder feeder was hung. The powder was swept by
Ar to the coater base where it was mixed with the
other coating gases supplied from a gas manifold. The
ZrCl4 powder in the gas stream was vaporized in the
coater base before entering the coating chamber.22
Figure 1 shows an apparatus for ZrC deposition by

the process using ZrCl4.22
The effects of varying CH4 and H2 concentrations
and particle bed area on the coating rate, the appearance, and the composition of the ZrC were studied
using the ZrCl4 powder feeder.22 Increases in CH4
Coater

Induction
heater

ZrCI4 powder
Auger
Mixing
chamber

Drive
motor

Gas
manifold

Argon
Figure 1 Experimental apparatus for ZrC deposition by
the process using ZrCl4. Reproduced from Wagner, P.;
Wahman, L. A.; White, R. W.; et al. J. Nucl. Mater. 1976, 62,
221–228.


218

Advanced Concepts in TRISO Fuel


Scrubber

Quartz

Carbon
wool
Graphite

Induction
coil

Alumina
Zr Br4

Quartz

Zr

Furnace

Ar

Ar
Br2 + Ar

CH4 + H2
Ar

and H2 concentrations were found to be effective in

increasing the linear coating rate of ZrC. Increases in
the ratio of CH4 to ZrCl4 in the coating gas resulted
in a decreased metallic appearance of the coating and
an increase in the C/Zr in the deposit. Increases in
H2 inhibited these effects.
The ZrC coating layers were prepared using a gas
mixture of C3H6, H2, ZrCl4, and Ar with the same
coater and ZrCl4 powder feeder as described earlier.23 In general, ZrC coating layers made with
CH4 and C3H6 were similar and were affected similarly by variations in the hydrocarbon and hydrogen
concentrations. The coating layers produced using
C3H6 were more sensitive to changes in hydrogen
concentration than those produced using CH4.
The coating processes based on the in situ generation of zirconium halide vapor are developed at Japan
Atomic Energy Research Institute (JAERI; now Japan
Atomic Energy Agency, JAEA) to avoid the handling
of highly hygroscopic halide powder.17 Several processes were studied: the chloride process,24 the methylene dichloride process,25 the iodide process,26,27
and the bromide process.28–31 Among these processes,
the bromide process proved to be the most convenient
and reliable; in this process, ZrBr4 is produced by
reacting bromine with zirconium sponge inside the
coater. ZrBr4 is preferred to ZrCl4 since the reaction
of excess chlorine with hydrogen is a potential explosion hazard.
Figure 2 shows an apparatus for ZrC deposition
by the bromide process.32 In this process, bromine,
which is liquid at room temperature, was carried by
argon onto the heated zirconium sponge beneath the
spouting nozzle and reacted to generate ZrBr4 vapor,
which was mixed with the other coating gases of
CH4 and H2 before entering the chamber. Propylene
could be used instead of methane. The flow rate of

ZrBr4 was controlled successfully by controlling the
flow rate of Ar passing through liquid bromine at
273 K and maintaining the temperature of zirconium
sponge at 873 K.
Along with the deposition experiments of ZrC by
the bromide process, thermochemical analyses were
performed to find the optimum deposition condition.31 The multiphase equilibrium in the system was
analyzed based on the minimization of the total free
energy of the system. The analyses predicted that the
ZrC monophase region exists in a wide composition
range of the feed gas mixture. It was concluded from
the analyses that the deposition rate could be controlled through the methane flow rate, and the composition of the deposit through the ZrBr4 flow rate in

Figure 2 Experimental apparatus for ZrC deposition by
the bromide process. Reproduced from Ogawa, T.; Ikawa, K.
Deposition of LTI pyrolytic carbon by a nozzle without water
cooling, JAERI-M 9568; Japan Atomic Energy Research
Institute, 1981.

the presence of excess hydrogen. The experimental
results agreed with the predicted results. By adjusting
the deposition condition, stoichiometric ZrC layers
were obtained with the bromide process.
Recently, a new ZrC coater was installed and
ZrC coating experiments were carried out at JAEA.33
The ZrC coater was designed with the maximum batch
size of 0.2 kg, which is about 10 times larger than the
previous one. The ZrC coater mainly consists of the
gas supply equipment, the coater, and the off-gas combustion equipment. The coater is composed of the
lower and the upper heaters with in-line configuration.



Advanced Concepts in TRISO Fuel

The lower one is for the reaction of bromine with
zirconium sponge and the upper one, for the chemical
vapor deposition of ZrC at 1873 K in the maximum.
The off-gas treatment equipment removes soot, hydrogen bromide, and residual hydrogen.
3.08.2.3 Characterization Techniques for
ZrC-Coated Particle Fuel
It is important to characterize the key basic properties of the coating layers that are critical to the fuel
performance. Although most of the characterization
techniques used for the ordinary TRISO-coated particle fuels can be applied to the ZrC-coated particle
fuels, some techniques have to be developed primarily for the ZrC-coated particle fuels.
In the case of the ordinary TRISO-coated fuel
particles, the PyC layers are burnt off to recover the
SiC fragments for characterization, such as density,
composition, and strength measurements. However,
it is almost impossible to separate the ZrC from the
PyC layers by the same method since ZrC, in contrast
with SiC, does not form a protective oxide layer,
resulting in oxidation of ZrC to ZrO2 when exposed
to air at high temperatures. To begin with, a method
of obtaining fragments of the ZrC coating layer from
the ZrC-coated particles is needed.
The plasma oxidation method was developed to
obtain the ZrC fragments from the coating layers containing the PyC.34 The difference in the oxidation rates
between PyC and ZrC is very large. In this method, the
samples were set in the plasma oxidation apparatus,
where low-pressure oxygen was ionized by high frequency induction coupling. Plasma reaction was monitored by a color analyzer and an optical power meter.

The color changed from the pale violet of pure oxygen
to pale blue during vigorous oxidation of free carbon,
and again to pale violet when the PyC was completely
removed and a very thin oxide scale was formed on the
ZrC. The brightness also changed dramatically during
the reaction. The obtained ZrC fragments were examined by Raman spectroscopy and X-ray diffraction.
It was confirmed that the bulk of the ZrC remained
unaffected by the plasma oxidation.35
The physical grinding technique was also developed to obtain the ZrC fragments.33 In this technique,
quartz powder having the Mohs hardness of 7 was
used since the Mohs hardness for ZrC is about 8–9
and that for graphite is 6. The fragments of the
combined layers of ZrC and PyC were ground with
quartz powder. After grinding, the fragments of the
ZrC layers without PyC were separated from the rest

219

in liquid tetrabromoethane (C2H2Br4) by the density
difference. The specific gravity of tetrabromoethane
is 2.965 Mg mÀ3.
The density of the SiC layers is measured by
the sink-float technique or a liquid gradient column,
in which a liquid having the same density as the
sample is needed. The density of the SiC layers
is around 3.21 Mg mÀ3, and a liquid mixture of
methylene iodide (CH2I2) having a density of
3.325 Mg mÀ3 and benzene (C6H6) having a density
of 0.8785 Mg mÀ3, for example, is used for the measurement of density. In the case of the ZrC density
measurement, no suitable liquid is present since the

density of the ZrC layers is around 6.6 Mg mÀ3, and
so, other techniques are needed. Gas pycnometry,
which required at least 100 mg of the samples,33 was
developed for the ZrC density measurement. The
stoichiometry of the ZrC layer affects the thermal
conductivity,36 fission product retention,18,37 etc. The
properties of ZrC are summarized in Chapter 2.13,
Properties and Characteristics of ZrC. Analysis of
the free carbon is important for controlling the quality
of the ZrC coating. Plasma oxidation with emission
monitoring was also applied to the quantitative analysis
of the free carbon in ZrC powder.38 The emission was
monitored with an optical color analyzer and was calibrated with standard samples of ZrO2 + C mixtures.
The oxidation rates of the free and the combined
carbons are so different that it is possible to estimate
the amount of the former from the emission. With
powdered ZrC of about 10 mg, free carbon of <1 wt%
could be easily determined. Without this method, the
composition of the ZrC was estimated by burning the
ZrC and PyC together, weighing ZrO2 and CO2, and
then subtracting the contribution of PyC from the total
amount of CO2.34
A new method to measure ZrC stoichiometry was
developed using the infrared light absorption during
combustion in oxygen and the inductively coupled
plasma-atomic emission spectrometry (ICP-AES).
With this method, accuracy of the C/Zr atom ratio
was on the order of 0.01.33
The defective SiC layer fraction is a very important indicator to show the quality of the ordinary
TRISO-coated particle fuels. This is measured by

the burn-leach method, where the fuel compact
or the coated fuel particles are heated at 1173 K in
air to oxidize the graphite matrix of the compact and
the OPyC layers, followed by the acid leaching
of the exposed uranium. The defective SiC coating
layer exposes uranium during burning. However, this
method cannot be applied to the ZrC-coated particle


220

Advanced Concepts in TRISO Fuel

fuels since the intact ZrC layer is oxidized and exposes
the actinide oxides to be leached by nitric acid solution.
A new method of measuring the defective ZrC layer
fraction is needed.
3.08.2.4 Performance of ZrC-Coated
Particle Fuel
3.08.2.4.1 Irradiation performance

Although systematic irradiation experiments on the
ZrC-coated fuel particles have not been completed,
some promising data have been obtained; the ZrC
layer is less susceptible to chemical attack by fission
products and fuel kernels, and the ZrC-coated fuel
particles perform better than the ordinary TRISOcoated particles at high temperatures, especially above
1873 K. Some early irradiation tests showed poor performance of the ZrC-coated particles, which may be
attributed to the fact that the fabrication conditions
had not been optimized.

Irradiation tests on several coating designs of the
ZrC-coated particles were carried out in the High Flux
Isotope Reactor (HFIR) and the Oak Ridge Research
Reactor (ORR).39–45 The irradiated ZrC-coated fuel
particles, which were made at LASL, were (1) ZrCTRISO-coated particles, (2) ZrC-TRISO type coated
particles without OPyC layer, (3) ZrC-coated particles
with ZrC-doped OPyC layer, and (4) ZrC-coated
particles with graded C–ZrC layer(s). Prior to the
tests on the fuel-kerneled coated particles, carbonkerneled inert coated particles were tested to determine the stability of the ZrC at temperatures of 1173
and 1473 K and fast neutron fluencies of 3.5 Â 1025 to
10.7 Â 1025 mÀ2.39,40 The irradiation results of the
inert coated particles were encouraging. In the test of
the fuel-kerneled particles, it was found that the graded
coatings were cracked and it was postulated that the
cracking was associated with the low PyC deposition
rate and was not related to the ZrC.41 The ceramographic examination showed that the performance of
the ZrC-coated particles in the HRB-12 capsule
Table 1

appeared to be poor in comparison with the ordinary
TRISO-coated particles. However, there was no evidence of palladium attack on any of the ZrC layers.42
The ZrC-TRISO-coated UC2, UO2, and UCxOy
particles made at General Atomic Company (GAC;
now General Atomics, GA) were irradiated in
HFIR.43–45 The irradiation conditions of the ZrCTRISO-coated fuel particles in the United States are
listed in Table 1.42–45 The performance of the ZrCTRISO-coated particles was very favorable. But there
was evidence that the ZrC-TRISO-coated UC2 particles in the HRB-7 and HRB-8 capsules might not be as
effective in retaining fission products as the ordinary
TRISO-coated particles; the electron probe microanalysis showed the rare-earth fission products on the
outside of the ZrC coating, whereas all cesium was

retained within the coating.43 The ZrC coating layers
in the ZrC-TRISO-coated UO2 and UC2 particles in
the HRB-15A capsule suffered no fission product
attack, while the ordinary TRISO-coated fuel particles
showed some degree of SiC-fission product interaction. The ZrC-TRISO-coated UC2 particles, however,
showed poor retention of silver and europium, with
great variety from particle to particle in this respect.44
The irradiation tests on the ZrC-coated UO2 particles characterized by the zirconium-carballoy layer
without the OPyC made at JAERI were carried out in
the Japan Materials Testing Reactor (JMTR).17 The
particles in the 73F-12A and 73F-13A capsules experienced very high temperatures exceeding 1873 K,
where the failure fractions of the ZrC-coated particles were rather high though they were on par with
those of the ordinary TRISO-coated particles or
below. The ceramographic examination revealed no
interaction between the UO2 kernel and the zirconium-carballoy layer when they came into contact
with each other due to the kernel migration, showing
better performance of the zirconium-carballoy layer
than the SiC layer against chemical attack.
The ZrC-TRISO-coated UO2 particles made at
JAERI were irradiated in JMTR and the Japan

Irradiation tests of ZrC-TRISO-coated fuel particles in the United States

Capsule

Fuel kernel

Temperature (K)

Burnup (% FIMA)


Fast neutron fluence (mÀ2, E > 29 fJ)

Reference

HRB-7
HRB-8
HRB-12
HRB-15A
HRB-15A
HRB-16
HRB-16

UC2
UC2
UC4.6O1.1
UC2
UO2
UC2
UCxOy

1498
1498
1523
1328–1403
1348–1398
1353
1433

84.4

84.4
84–86
24.9–28.3
27.2–28.8
20.9
20.3

4.50 Â 1025
5.92 Â 1025
(4.4–6.9) Â 1025
(4.9–6.3) Â 1025
(6.0–6.2) Â 1025
4.1 Â 1025
3.8 Â 1025

[43]
[43]
[42]
[44]
[44]
[45]
[45]


Advanced Concepts in TRISO Fuel

Table 2

Irradiation tests of ZrC-TRISO-coated fuel particles in Japan


Capsule

Fuel kernel

78F-4A
80F-4A
ICF-26H
VOF-8H
VOF-14H
88F-3A

UO2
UO2
UO2
UO2
UO2
UO2

Temperature (K)
1373
1173
1673–1773
1643 (15 K mmÀ1)
1873 (15 K mmÀ1)
1673–1923

221

Burnup (% FIMA)


Fast neutron fluence (mÀ2, E > 29 fJ)

Reference

4.0
1.5
1.8
1.6
1.6
4.5

2.2 Â 10
1.2 Â 1025
1.0 Â 1025
<1.0 Â 1024
<1.0 Â 1024
–

[46]
[46]
[46]
[46]
[46]
[47]

25

through-coating failure fractions revealed better irradiation performance of the ZrC-TRISO-coated fuel
particles. Figure 3 shows a typical optical micrograph
of the polished cross-section of the ZrC-TRISOcoated fuel particle after irradiation.47 Optical microscopy and electron probe microanalysis on the polished

cross-section of the ZrC-TRISO-coated fuel particles
revealed no interaction of palladium with the ZrC
coating layer or accumulation of palladium at the
inner surface of the ZrC coating layer, whereas severe
corrosion of the SiC coating layer was observed in the
ordinary TRISO-coated fuel particles.
100 µm
Figure 3 A ceramograph of the ZrC-TRISO-coated UO2
particle after irradiation at 1673–1923 K to 4.5% FIMA.
Reproduced from Minato, K.; Ogawa, T.; Sawa, K.; et al.
Nucl. Technol. 2000, 130, 272–281.

Research Reactor-2 ( JRR-2). The irradiation conditions are summarized in Table 2.46,47 The releaseto-birth ratio (R/B) of 88Kr was measured during
irradiation in the 80F-4A capsule, which showed no
through-coating failure of the ZrC-TRISO-coated
UO2 particles. In the postirradiation examination,
no coating failure was detected by stereomicroscopy,
X-ray microradiography, and acid leaching. The
ZrC-TRISO-coated UO2 particles in the VOF-14H
capsule were irradiated at 1873 K in a steep temperature gradient of 15 K mmÀ1. The ceramographs of the
particles from the VOF-14H showed carbon deposits
at the colder end of the fuel kernel accompanied
by kernel migration up the temperature gradient. In
the ordinary TRISO-coated fuel particles, palladium
attack of the SiC layer was occasionally found at the
colder side of the particle.48 However, there was no
indication of coating deterioration in the ZrC layer.
In the 88F-3A capsule, the ZrC-TRISO-coated
UO2 particles and ordinary TRISO-coated UO2
particles were irradiated under identical conditions.47 The postirradiation measurement of the


3.08.2.4.2 Resistance to chemical attack by
fission products

The better performance of the ZrC coating layer
than the SiC coating layer against chemical attack
by fission product palladium has been demonstrated
in out-of-reactor experiments and irradiation tests.
The out-of-reactor experiments of the chemical
reactions of ZrC and SiC with palladium were
performed, where the ZrC-TRISO and ordinary
TRISO-coated particles were heated in either palladium powder or vapor.49 The coating layers would be
attacked by the fission product palladium from inside
in irradiation, while palladium was supplied from
outside the particles in the out-of-reactor experiments to simulate the situation. The experiments on
the reactions in the mixture of SiC, ZrC, Pd, and C,
and the reaction of ZrC with Ag–Pd alloy were also
studied. Reaction morphology was observed by ceramography and the reaction products were identified
by X-ray diffraction and electron probe microanalysis. When the ZrC-TRISO-coated particles were
heated in the palladium powder, ZrPd3 and C were
formed. However, no reaction was found on the ZrCTRISO-coated particles heated in the palladium
vapor at 1830–2150 K, whereas the SiC layers were
attacked severely. It was revealed that ZrC did react
with palladium at a sufficiently high palladium activity, but the reaction could not occur at a low palladium activity, such as in the fuel particles.49


222

Advanced Concepts in TRISO Fuel


As described briefly in Section 3.08.2.4.1, the
comparatively better performance of the ZrC coating
layer than the SiC coating layer against chemical
attack by fission product palladium was confirmed
in the irradiation tests. For example, the ZrCTRISO-coated UC4.6O1.1 particles irradiated at
1523 K to 86% FIMA in the HRB-12 capsule had
no evidence of palladium attack on the ZrC layers,
and the ZrC-TRISO-coated UO2 and UC2 particles
irradiated at 1328–1403 K to 24.9–28.8% FIMA in
the HRB-15A capsule suffered no fission product
attack, while the ordinary TRISO-coated particles
showed effects of the SiC-fission product interaction.
The irradiation tests described earlier that were
carried out in the United States were characterized
by low irradiation temperatures and high burnup
compared with those in Japan. The irradiation test
of the ZrC-TRISO-coated UO2 particles in the 88F3A capsule in Japan is a good example, where the
irradiation temperature was 1673–1923 K and the
burnup was 4.5% FIMA.47
Optical microscopy and electron probe microanalysis on the polished cross-section of the ZrCTRISO-coated particles irradiated in the 88F-3A
capsule revealed no interaction of palladium with
the ZrC coating layer or accumulation of palladium
at the inner surface of the ZrC coating layer, as shown
in Figure 4(a).47 Optical microscopy on the polished
cross-section of the ordinary TRISO-coated fuel particles irradiated under identical conditions, on the
other hand, showed severe corrosion of the SiC coating layer. Figure 4(b) shows an example,47 which is a
typical feature of the corrosion of the SiC coating
layer by the fission product palladium.48,50
The fission product behavior inside the IPyC
coating layer should be the same regardless of the

ZrC or SiC coating layer as long as the IPyC coating
layer is intact. It is reasonable to assume that the
fission product palladium is released from the kernel
to the ZrC coating layer in a fashion similar to its
release to the SiC coating layer in the ordinary
TRISO-coated fuel particles. According to the outof-reactor experiment, the reaction of ZrC with
palladium will occur when the concentration of palladium is sufficient.49 A probable explanation of the
absence of corrosion on the ZrC coating layer is that
palladium was not stopped by the ZrC coating layer
and never reached a concentration on the surface of
the coating layer to cause the corrosion.47
No data are available on the release behavior of
palladium from the ZrC-TRISO-coated fuel particles, but some data that may be relevant are available

ZrC

IPyC

10 mm

(a)

SiC

IPyC

(b)

20 mm


Figure 4 Polished cross-sections of irradiated particles at
1673–1923 K to 4.5% FIMA; (a) secondary electron image of
the ZrC-TRISO-coated UO2 particle and (b) optical
micrograph of the ordinary TRISO-coated UO2 particle.
Reproduced from Minato, K.; Ogawa, T.; Sawa, K.; et al.
Nucl. Technol. 2000, 130, 272–281.

for ruthenium. It has been reported that ruthenium
was released from the ZrC-TRISO-coated fuel particles during postirradiation heating tests at 1873,
2073, and 2273 K,51,52 while no ruthenium release
was reported under similar conditions from the ordinary TRISO-coated fuel particles.53,54 In addition, in
postirradiation examinations, ruthenium was sometimes found at the inner surface of the SiC coating
layer of the ordinary TRISO-coated fuel particles,
indicating that it does not easily diffuse through the
SiC layer.48
3.08.2.4.3 High-temperature stability

The better mechanical integrity of the ZrC-TRISOcoated fuel particles when compared with the ordinary
TRISO-coated fuel particles at high temperatures has
been revealed in the out-of-reactor heating experiments.


Advanced Concepts in TRISO Fuel

100
Std. design
Calculated
Experimental
Advanced design
Failure fraction (%)


The out-of-reactor high-temperature heating
experiments on unirradiated ZrC-TRISO-coated particles, together with the ordinary TRISO-coated
particles, were carried out in a vertical graphite resistance furnace contained in a stainless steel watercooled jacket.55 The particles were heated at 2073 K
for 1 h to simulate the annealing effect during compact fabrication, and then at the desired temperatures
for 1 h. The ZrC-TRISO-coated UO2 particles withstood the heating at 2723 K for 1 h, though more
than half of the particles failed at 2773 K within 1 h.
The ZrC-TRISO coating layers were expanded
plastically leaving a large gap between the kernel and
the buffer layer by a large internal pressure of CO,
while the SiC coating layers decomposed to lose their
mechanical integrity. The limiting factor of the stability of ZrC-TRISO-coated UO2 particles is not the
chemistry of ZrC but that of the system enclosed
by the ZrC layer. Although the ZrC itself was stable
up to the eutectic point of ZrC–C at about 3123 K,56
the ZrC-TRISO particles could not withstand the
heating above 2773 K. Failure was induced by large
internal pressures.55
The postirradiation heating test of the ZrCTRISO-coated UO2 particles was performed at a rate
of 1 K minÀ1 to the maximum temperature of 2673 K to
clarify the high-temperature stability of the particles.46
The particles were sampled from an irradiated fuel
compact at 1373 K to 4% FIMA after its electrolytic
disintegration. A total of 101 particles were heated
individually, placed in holes of two graphite disks in a
cold walled furnace with a graphite heater. During
heating, the radioactivity in flowing helium gas was
monitored with an ionization chamber. The activity
was due mostly to 85Kr. When a through-coating failure or a pressure-vessel failure occurs, the activity of
fission gas retained within the particle is released.

No failure was detected during the heat-up stage.
An activity burst occurred only after keeping the
particles at 2673 K for about 6000 s (100 min). The
activity burst corresponded to one particle failure
among the 101 particles heated. This interpretation
was confirmed by X-ray microradiographs of the particles after heating. Figure 546 compares the failure
fractions of the ZrC-TRISO-coated UO2 particles
with the ordinary TRISO-coated UO2 particles.57
Under this heating condition, most of the ordinary
TRISO-coated particles would fail, as shown in
Figure 5, where the dashed line gives a model prediction for the ordinary TRISO-coated particles with
the same dimensions as the ZrC-TRISO-coated particles in this study.46

223

Calculated (SiC)
Experimental (ZrC)
50

0
1900

2000

2100
2200
T (ЊC)

2300


2400

Figure 5 Comparison of failure fractions of the
ZrC-TRISO-coated UO2 particles with the ordinary
TRISO-coated UO2 particles. Reproduced from Ogawa, T.;
Fukuda, K.; Kashimura, S.; et al. J. Am. Ceram. Soc.
1992, 75, 2985–2990.

Figure 6 A ceramograph of the ZrC-TRISO-coated UO2
particle that survived the postirradiation heating at 2673 K.
Reproduced from Ogawa, T.; Fukuda, K.; Kashimura, S.;
et al. J. Am. Ceram. Soc. 1992, 75, 2985–2990.

The different behavior of the ZrC-TRISO-coated
UO2 particles compared with that of the ordinary
TRISO-coated UO2 particles at high temperatures
was discussed with the ceramographs of the particles
after heating. Figure 6 shows a ceramograph of the


224

Advanced Concepts in TRISO Fuel

ZrC-TRISO-coated UO2 particle that survived the
postirradiation heating at 2673 K.46 The OPyC and
ZrC layers expanded, while the IPyC layer did not.
There was a significant difference in the behavior
of the ZrC coating layer compared with that of the
SiC at high temperatures. The ZrC can sustain a very

large strain, whereas the SiC is brittle in nature. The
high plasticity is explained by the fact that resistance
of ZrC crystal lattice to the dislocation motion
becomes very weak above 2473 K.58 In these heating
tests, the particles were heated in a loose condition
without mechanical support from the surrounding
graphite matrix of the fuel compact. The presence
of the graphite matrix could offset the coating expansion and would further reinforce the integrity of the
ZrC-TRISO-coated particles.46
Isothermal postirradiation heating tests were also
performed to study the coating integrity and fission

product retention of the ZrC-TRISO-coated UO2
particles. Three tests were carried out at 1873 K
for 4500 h, at 2073 K for 3000 h, and at 2273 K for
100 h in a cold-wall furnace with a graphite
heater.51,52 For each test, 100 particles were sampled
from fuel compacts irradiated at 1173 K to 1.5%
FIMA. During all the heating tests, no throughcoating failure was detected by the 85Kr release
monitoring. The X-ray microradiography on the
coated particles after the heating tests revealed no
OPyC failure, which confirmed the results of the
gas release monitoring.
Typical polished cross-sections of the ZrCTRISO-coated fuel particles after the heating tests
are shown in Figure 7.51,52 No failure was observed
in the ZrC and OPyC coating layers of the particles
heated at 1873 K for 4500 h, as shown in Figure 7(a)
and 7(b). In some particles, the IPyC coating layers

(a)


100 µm

(b)

20 µm

(c)

100 µm (d)

20 µm

(e)

100 µm (f)

20 µm

Figure 7 Ceramographs of the ZrC-TRISO-coated UO2 particles after the postirradiation heating tests; (a) and (b) at
1873 K for 4500 h, (c) and (d) at 2073 K for 3000 h, and (e) and (f) at 2273 K for 100 h. Adapted from Minato, K.; Ogawa, T.;
Fukuda, K.; et al. J. Nucl. Mater. 1995, 224, 85–92; Minato, K.; Ogawa, T.; Fukuda, K.; et al. J. Nucl. Mater. 1997, 249, 142–149.


Advanced Concepts in TRISO Fuel

were cracked in the radial direction. No palladium
attack or thermal degradation of ZrC was observed.51
On the polished cross-sections of the particles
heated at 2073 K for 3000 h, no failure of the ZrC

and OPyC coating layers was observed, as shown in
Figure 7(c) and 7(d). However, some degradation of
the ZrC coating layer seemed to have occurred. The
inner and outer surfaces of the ZrC coating layers in
most of the particles heated at 2073 K were not
smooth. In some particles, the IPyC coating layers
were cracked in the radial direction, where about a
quarter of the thickness of the ZrC coating layers
seemed to have been attacked along the grain
boundaries.52
The ceramography on the ZrC-TRISO-coated
fuel particles after heating at 2273 K for 100 h
revealed that the ZrC coating layers as well as the
IPyC coating layers were damaged in most of the
particles observed, as shown in Figure 7(e) and 7(f ).
The ZrC coating layers were damaged through the
thickness. Based on the results of the ceramographic
examination, the electron probe microanalysis, and
the thermodynamic analysis, the observed deterioration of the ZrC-TRISO-coated fuel particles was
attributed to the reaction of ZrC with CO gas caused
by the failure of the IPyC coating layer.52
In the case of the ordinary TRISO-coated fuel
particles, fission gas release was observed in the postirradiation heating of fuel elements at 2073 K for 100
or 200 h.53,54 Although the number of the ZrCTRISO-coated particles tested was small compared
with that on the fuel element, it is probably safe to say
that the ZrC-TRISO-coated fuel particles have higher
capability of fission gas retention than the ordinary
TRISO-coated fuel particles at high temperatures.
3.08.2.4.4 Retention of fission products


Retention of the fission product cesium by the ZrC
coating layer has been demonstrated to be better than
that by the SiC coating layer though the data for the
other elements are limited compared with those for
the SiC coating layer.
The diffusion coefficients for strontium and
barium in the ZrC coating layer were obtained in
strontium soaking experiments and postactivation
annealing experiments.59 The diffusion coefficient
for Sr in the ZrC coating layer, DSr(ZrC), was estimated to be 2 Â 10À18 m2 sÀ1 at 1673 K and that for
Ba, DBa(ZrC), was estimated to be 2.9 Â 10À18 to
4.6 Â 10À18 m2 sÀ1 at 1673 K. The retention of these
elements by the ZrC coating layer was better than
that by the SiC coating layer.

225

The diffusion coefficients for silver, barium, promethium, and cerium in the ZrC coating layer were
evaluated from annealing experiments.60 Based on the
distribution of the nuclides in the ZrC measured by
removing the ZrC stepwise, the diffusion coefficient
DBa(ZrC) was estimated to be 1.3 Â 10À17 m2 sÀ1 and
DCe(ZrC) was estimated to be 6.4 Â 10À18 m2 sÀ1 at
1773 K. The ZrC coating layer showed better retention of these elements, though the characteristics of
the ZrC coating layer were not reported.
The postirradiation heating tests of the ZrCTRISO-coated UO2 particles were performed at
1873 K for 4500 h, at 2073 K for 3000 h, and at 2273 K
for 100 h, to study the release behavior of the fission
products.51,52 For each heating test, 100 of the ZrCTRISO-coated particles, which had been irradiated at
1173 K to 1.5% FIMA, were used. The furnace in a hot

cell was composed of a graphite heater, a graphite
sample holder, graphite holder disks, and carbon insulators within a stainless steel vessel. The coated fuel
particles were placed individually in the holes of
the graphite disks. Each heating test was divided into
several time steps. At the end of each time step,
the graphite components and the carbon insulators
were replaced by new ones to measure the released
metallic fission products by g-ray spectrometry. The
fission gas release monitoring during the tests and
the X-ray microradiography after the tests revealed
that no through-coating failure occurred in the tests.
The measured fractional releases of 137Cs are
shown in Figure 8 as a function of heating time.52
The calculated fractional release of 137Cs from the
ordinary TRISO-coated particles at 1873 K is also
presented in the figure for comparison. This curve
was drawn based on the effective diffusion coefficient
of 137Cs in the SiC coating layer61 and the particle
dimensions. The fractional release of 137Cs was found
to be below 1 Â 10À3 even after heating at 1873 K for
4500 h or at 2073 K for 3000 h, whereas it was more
than 1 Â 10À1 after heating at 2273 K for 100 h. The
sudden increase in the fractional release at 2273 K was
probably attributed to the degradation of the ZrC
coating layer observed in the ceramography. The
high cesium retention of the ZrC-TRISO-coated
fuel particles was confirmed to 2073 K.
Based on a diffusion model, where a fuel kernel
with a single coating layer was assumed, the effective
diffusion coefficients for 137Cs in the ZrC coating

layer, DCs(ZrC), were evaluated to be between
1 Â 10À18 and 5 Â 10À18 m2 sÀ1 at 1873 K, and
between 2 Â 10À18 and 1 Â 10À17 m2 sÀ1 at 2073 K.52
The present value for DCs(ZrC) at 1873 K was more


226

Advanced Concepts in TRISO Fuel

100

100

2273 K
10–1
Fractional release

Fractional release

10–1

10–2
SiC -1873 K
10–3

10–2
2273 K

10–3

2073 K

2073 K

1873 K

10–4

10–4
1873 K
10–5
100

101

102
103
Heating time (h)

104

105

10–5 0
10

101

102


103

104

105

Heating time (h)

Figure 8 Fractional releases of
Cs during
postirradiation heating of the ZrC-TRISO-coated UO2
particles as a function of heating time. Reproduced from
Minato, K.; Ogawa, T.; Fukuda, K.; et al. J. Nucl. Mater.
1997, 249, 142–149.

Figure 9 Fractional releases of 106Ru during
postirradiation heating of the ZrC-TRISO-coated UO2
particles as a function of heating time. Reproduced from
Minato, K.; Ogawa, T.; Fukuda, K.; et al. J. Nucl. Mater.
1997, 249, 142–149.

than one-tenth of the reported effective diffusion
coefficients of 137Cs in the SiC coating layer,
DCs(SiC),61,62 whereas at 2073 K, the value of DCs(ZrC)
was less than one-hundredth of those for DCs(SiC).
The measured fractional releases of 106Ru are shown
in Figure 9 as a function of heating time.52 The release
behavior of 106Ru was evaluated by the same diffusion
model that was used for the evaluation of 137Cs release.
For the best fit, as shown in Figure 9, the effective

diffusion coefficients of DRu(ZrC) ¼ 3 Â 10À16 m2 sÀ1
at 1873 K and DRu(ZrC) ¼ 5 Â 10À15 m2 sÀ1 at 2073 K
were obtained. The measured fractional release data at
2273 K did not fit the calculated curve, which was
drawn based on the extrapolated diffusion coefficient,
DRu(ZrC) = 5 Â 10À14 m2 sÀ1, from the data at 1873 and
2073 K. The sudden increase in the measured fractional
release and the large fractional release compared with
the calculated curve suggest the deterioration of the
ZrC coating layer.
The release of 106Ru was not reported in the
postirradiation heating tests of the ordinary
TRISO-coated fuel particles at 1873 K for 500 h and
at 2073 K for 200 h.53,54 The retention of 106Ru in the
ZrC-TRISO-coated fuel particles was not better than
that in the ordinary TRISO-coated fuel particles.
The present DRu(ZrC) obtained was almost the
same as DCs(SiC) in literature.61,62 Behavior of this

kind was also observed in the annealing tests of the
short-term irradiated ZrC-TRISO-coated UO2 particles.59 The fast diffusion of ruthenium would be
explained by the empirical rule that fast diffusion
other than ordinary substitutional diffusion may occur
when the solute has the atomic radius ratio to the solvent
below 0.85; the radius ratio of Ru to Zr is 0.836.
Besides 137Cs, 134Cs, and 106Ru, the release of
radionuclides of 144Ce, 154Eu, and 155Eu were detected
by g-ray spectrometry of the graphite components and
carbon insulators. The fractional releases of 154Eu
were larger than those of 137Cs at any temperature.

Although the release data of 137Cs and 106Ru could be
treated quantitatively, the accuracy of the measured
values for 134Cs, 144Ce, 154Eu, and 155Eu was not
enough to evaluate the effective diffusion coefficients
since the counts by g-ray spectrometry were small.52
The radionuclide 110mAg, whose half life is
250.4 d, is known as one of the most releasable radionuclides from the ordinary TRISO-coated fuel particles. The release of 110mAg was not detected in this
particular experiment. However, it cannot be concluded that the ZrC coating layer has excellent
110m
Ag retention capability. The present experiment
was carried out about 9 years after the end of irradiation of the ZrC-TRISO-coated fuel particles and
the ratio of inventories of 110mAg to 106Ru was less

137


Advanced Concepts in TRISO Fuel

than 1 Â 10À3 just before the heating tests, suggesting
that the inventory of 110mAg was too small to be
detected even if 110mAg was released.52
The retention of the fission product cesium by the
ZrC-TRISO-coated fuel particles was further studied. The particles after the postirradiation heating
tests at 1873 K for 4500 h and at 2073 K for 3000 h,
together with the as-irradiated particles, were examined individually with X-ray microradiography and
g-ray spectrometry.63
Fifteen particles each were crushed individually
to recover the fuel kernel and the coating layers
separately, and fission product inventories of the
fuel kernel and coating layers of each particle were

then measured with g-ray spectrometry. The inventory of 137Cs measured on the whole particle before
crushing it was in good agreement with the sum of
the inventories measured on the fuel kernel and the
coating layers for each particle. The fractional content of 137Cs in the fuel kernel for each as-irradiated
particle was about 98% or about 2% of 137Cs was
released from each kernel to the coating layers, as
shown in Figure 10(a).63 The 137Cs retention behavior of each of the particles during the irradiation was
similar. No anomalies were found in the X-ray microradiography. The release of 137Cs from the fuel kernel
could be attributed to the fission recoil.
The normalized 137Cs inventories in the whole
particles heated at 2073 K for 3000 h agreed with
one another and the mean value agreed well with
that for the as-irradiated particles, which was consistent with the result of the fractional release measurement. Figure 10(b) shows the fractional contents
of 137Cs in the fuel kernel and the coating layers
for each particle after postirradiation heating at
2073 K for 3000 h.63 It was found that part of the
137
Cs was released from the kernel to the coating
layers in each particle during the heating test and
that the fractional content of 137Cs in the kernel was
different from particle to particle though 137Cs was
not released from the particles practically.63
When the result of the X-ray microradiography
was combined with that of the g-ray spectrometry, a
surprising relation was found between them. The first
seven ZrC-TRISO-coated fuel particles shown in
Figure 10(b), which showed relatively good retention
of 137Cs in the fuel kernels, had radially broken IPyC
layers and deformed fuel kernels. The remaining eight
particles whose cesium retention of the fuel kernels

was relatively poor showed no anomalies in the X-ray
microradiography. The same tendency was observed
in the particles heated at 1873 K for 4500 h.63

227

It was found that the cesium retention of the fuel
kernel was enhanced when the IPyC layer failed in
the ZrC-TRISO-coated UO2 particles. The mechanism for the enhancement of cesium retention was
not yet well understood, though the interaction of
ZrC with CO gas results in reduction of the oxygen
potential of the fuel, which may influence chemical
forms of the fission products in the fuel.64 An additional ZrC coating layer on the UO2 kernel of the
coated fuel particle would enhance the cesium retention capabilities. The coated UO2 particles gettered
with ZrC, where ZrC was coated on the UO2 kernel
or dispersed throughout the buffer layer, were tested,
which is described in Section 3.08.3.
3.08.2.4.5 Behavior under oxidizing
conditions

No experimental work on the performance of the
ZrC-coated fuel particles under accidental oxidizing
conditions was found in the literature.
A thermodynamic analysis was performed on
the system of ZrC–C–(O2 or H2O)–He.65 The analysis showed that there were two kinds of oxidation
behavior of ZrC, which was similar to that of SiC:
active oxidation and passive oxidation. While a loss
in mass occurs in the active oxidation, the passive
oxidation results in a net mass increase. The activeto-passive transition of oxidation occurs depending
on temperature and initial O2 or H2O pressure.

Under most of the accidental air or water ingress
conditions, the passive oxidation of ZrC was expected
to occur. According to the experiment on the passive
oxidation of ZrC at temperatures of 1130–2160 K,66
the protective layer of ZrO2 would not be formed on
ZrC, in contrast with the case of SiC. The preferential oxidation along grain boundaries occurred in
ZrC. Between 1130 and 1560 K, this preferential oxidation resulted in intercrystalline fracture due to
grain boundary stresses. At higher temperatures, the
stresses were apparently sufficiently released so that
the samples remained intact.66 From this point of
view, ZrC may not be a retentive layer in air or
water ingress accidents.

3.08.3 ZrC-Containing TRISOCoated Particle Fuel
3.08.3.1 Designs of ZrC-Containing
TRISO-Coated Particle Fuel
Although the TRISO-coated UO2 particle fuel has
performed very well, it has some disadvantages
that stem from the oxide fuel. When the oxide fuel


228

Advanced Concepts in TRISO Fuel

1
Coatings
Kernel

Fraction


0.8

0.6

0.4

0.2

0

A2 A15 A3 A12 A9

A5

A4 A7 A14 A1
Particle name

A8 A10 A11 A6 A13

1
Coatings
Kernel

Fraction

0.8

0.6


0.4

0.2

0

C5

C1 C12 C7

C4 C10 C9 C11 C2 C14 C15 C8
Particle name

C6 C13 C3

Figure 10 Fractional contents of 137Cs in the fuel kernel and coating layers for each ZrC-TRISO-coated UO2 particle:
(a) as-irradiated at 1173 K to 1.5% FIMA and (b) after postirradiation heating at 2073 K for 3000 h. Reproduced from
Minato, K.; Ogawa, T.; Koya, H.; et al. J. Nucl. Mater. 2000, 279, 181–188.

is used, the fission-dependent buildup of gaseous
products of CO/CO2 occurs from reactions between
released oxygen and carbon coatings. The resulting
pressure within the particle must be accommodated by increasing the thickness of the buffer
layer depending on target burnups to prevent the
pressure-vessel failure. This penalizes the reactor
design because the core fissile atom density has to
be lowered. Especially in the actinide burning, this
penalty may be large. The accumulation of CO
within the particle also leads to kernel migration
toward the hotter side of the particle when the


large thermal gradient within the particle is established. In extreme cases, the kernel can traverse
through the buffer and IPyC layers, which may
interact with the SiC layer.
The mixed UO2–UC2 fuel67–71 greatly reduces the
CO pressure via the chemical buffering or gettering
action, and kernel migration rates are also reduced in
this gettered fuel. Similar results have been obtained
by the addition of nonfissionable carbides, such as
SiC,72 to oxide kernels, and additional benefits might
be gained by placing getters of this type at the outer
boundary of the UO2 kernels.19


Advanced Concepts in TRISO Fuel

Two configurations have been proposed and
tested for the gettered TRISO-coated UO2 particle
fuel containing ZrC positioned outside the kernel.19
The purpose of the ZrC in the particle is to serve
as an oxygen getter for regulating the buildup of
gas pressure caused by oxygen release during fissioning of oxide fuel, though the work described in
Section 3.08.2 on the use of ZrC in coated fuel
particles has been directed toward the development
of a replacement for the SiC barrier layer.
Figure 11(a) and 11(b)19 shows ceramographs of
the two configurations of the gettered TRISO-coated
UO2 particle fuel containing ZrC. The first one had
an additional solid layer of ZrC deposited over the
kernel, which was protected from coating gases used

in the ZrC deposition process73 by the previous
application of a thin PyC seal coat directly over the
kernel. This substrate, consisting of a UO2 kernel, a
seal coat, and a ZrC layer, was then used as the core of
an ordinary TRISO particle. The other one had a
buffer layer in which ZrC was dispersed. The ZrC
dispersed buffer layer was coated by a codeposition
process74,75 on the kernel with a PyC seal coat, to
protect the kernel during the codeposition process.
The IPyC, SiC, and OPyC layers were then coated
on the ZrC dispersed buffer layer.
Besides the two types of TRISO-coated UO2 particles gettered with ZrC shown in Figure 11, four
fuel particle designs, using ZrC layers in combination

229

with porous and dense PyC layers, were tested76:
UC2/ZrC/buffer/PyC, (8Th,U)O2/ZrC/buffer/PyC,
UC2/buffer/ZrC/PyC, and (8Th,U)O2/buffer/ZrC/
PyC. The irradiation experiments on these coated
particles have certainly revealed that the ZrC layer
has exceptional resistance to chemical attack by fission
products and good mechanical stability under irradiation. Nonetheless, the following section focuses on the
two types of particles shown in Figure 11.
3.08.3.2 Performance of ZrC-Containing
TRISO-Coated Particle Fuel
3.08.3.2.1 Irradiation performance

The two types of TRISO-coated UO2 particles gettered with ZrC were irradiated in the HFIR. Three
capsule irradiation experiments (HRB-15A, HRB15B, and HRB-16) were carried out, in which three

kinds of sample forms were used: loose particles in
the holes of graphite trays, particles bonded into
place in graphite trays with carbonaceous matrix, and
fuel rods were used. Table 3 summarizes the irradiation conditions.44,45,77 The irradiation temperatures
were about 1173 K (HRB-15B) and about 1473 K
(HRB-15A, HRB-16), and the burnups were more
than 20% FIMA. For comparison, the ordinary
TRISO-coated particles having kernels consisting of
UO2, UC2 and mixed UO2–UC2 were also included in
these capsule irradiation experiments.19

As made

Irradiated

(a)

(b)

Solid ZrC

150 µm

(c)

100 µm

(d)

Dispersed ZrC


150 µm

140 µm

Figure 11 Ceramographs of two configurations of ZrC-gettered TRISO-coated UO2 particles before and after irradiation;
(a) and (b) with a solid layer of ZrC deposited over the kernel, and (c) and (d) with ZrC dispersed throughout the buffer
layer. Reproduced from Bullock, R. E.; Kaae, J. L. J. Nucl. Mater. 1983, 115, 69–83.


230

Advanced Concepts in TRISO Fuel

Table 3

Irradiation tests of ZrC-containing TRISO-coated UO2 fuel particles

Capsule

ZrC type

Temperature (K)

Burnup
(% FIMA)

Fast neutron fluence
(mÀ2, E > 29 fJ)


Reference

HRB-15A
HRB-15A
HRB-15B
HRB-15B
HRB-16

ZrC layer on UO2
ZrC dispersed in buffer
ZrC layer on UO2
ZrC dispersed in buffer
ZrC layer on UO2

1313–1398
1333–1358
1133–1178
1188
1353–1485

22.0–29.1
23.6–24.3
23.8–26.6
25.2–26.2
19.0–27.6

(4.4–6.5) Â 1025
(4.9–5.1) Â 1025
(5.4–6.6) Â 1025
(5.4–6.1) Â 1025

(3.7–6.3) Â 1025

[44]
[44]
[77]
[77]
[45]

The irradiated particles were examined by X-ray
microradiography, ceramography, and electron probe
microanalysis. Figure 11(b) and 11(d)19 shows ceramographs of the irradiated TRISO-coated UO2
particles gettered with ZrC. Both types of the ZrCgettered TRISO-coated UO2 particles behaved very
well under irradiation. No detrimental irradiation
effects were ceramographically found, except for carbon that precipitated near the outer edge of the
kernels, which may have resulted from the gettering action of the ZrC. Oxygen released from UO2
kernels during fissioning would react predominantly
with carbon from the seal coat to form CO, which
would then be gettered through the reaction:
2CO þ ZrC ! ZrO2 þ 3C.
No failures of the three outer structural coatings
(OPyC/SiC/IPyC) on these ZrC-gettered TRISOcoated UO2 particles were observed. The solid ZrC
coating over the kernel occasionally failed under
irradiation, but an intact coating is not required for
effective gettering, and it was surprising that the
majority of the ZrC layers did remain intact.19
Kernel migration was not observed in the two
types of the TRISO-coated UO2 particles gettered
with ZrC, while the UO2 fuel kernels migrated by as
much as 35 mm in the ordinary TRISO-coated UO2
particles. The absence of kernel migration in the

ZrC-gettered TRISO-coated UO2 particles was good
evidence that this potentially detrimental phenomenon in the oxide fuels can be controlled by the addition of oxygen getters.78
In addition to demonstrating the absence of kernel
migration, ceramographic examination of the UO2
particles gettered with a solid overcoating of ZrC
also yielded information on the reduced kernel
expansions observed in X-ray microradiography. A
one-to-one particle comparison between the X-rays
and the ceramographic sections of 54 particles examined revealed that of the 42 particles with round
kernels that expanded only 1% had intact ZrC
and buffer layers. However, the other 12 particles
with 10% kernel expansions had failed ZrC and

buffer layers, and the kernels had extruded into
these coating voids to produce the irregular shapes.19
It is obvious why the expansion behaviors of the
two types of particles were so different. As long as the
ZrC layer remained intact, the growth of gas bubbles
was restricted, and consequently these kernels swelled
by only about 1% in diameter. However, large gas
bubbles invariably formed when the ZrC layer failed,
and this caused the larger kernel expansions. The
compressive restraint due to the ZrC layer would be
expected to restrict gas-bubble growth in the kernel.19
3.08.3.2.2 Retention of fission products

The fission product retention of the fuel particles
was examined by postirradiation isothermal annealing
tests. Ten particles each of the two types of TRISOcoated UO2 particles gettered with ZrC, together with
the ordinary TRISO-coated particles having kernels

consisting of UO2, UC2, and mixed UO2–UC2, from
the irradiation experiment of HRB-15B in which the
irradiation temperature was about 1173 K, were used.
The inventories of fission products in the sample
particles were measured before and after the heating
experiment and the released ones collected on sleeves
were measured at certain time intervals by g-ray spectrometer. The postirradiation heating tests were conducted at temperatures of 1473, 1623, and 1773 K for
more than 36 Ms (10 000 h).19,20
The TRISO-coated UO2 particles gettered with a
solid ZrC overcoating on the kernel did not release
any measurable fission products during postirradiation annealing at 1773 K for over 42 Ms ($12 000 h).
There was significant release of silver and europium
from the TRISO-coated UO2 particles with ZrC
dispersed in the buffer layer under the same conditions, as well as from the ordinary TRISO-coated
UO2, UC2, and mixed UO2–UC2 fuels. The X-ray
microradiographs revealed that the solid ZrC layers
on all the retentive test particles were apparently
intact and in place on perfectly round kernels before
and after annealing. An intact ZrC layer seems to be
necessary for explaining the outstanding retention


Advanced Concepts in TRISO Fuel

of these particles, as opposed to its mere presence as
an oxygen getter, because UO2 particles with ZrC
dispersed in the buffer layer released significant
amounts of fission products.19
While the retention mechanism was not yet well
understood, it appeared that coated particle fuel with

a solid ZrC layer had the potential for retaining highly
diffusive silver and europium isotopes for long periods of time at temperatures as high as 1773 K.19
Both the rare-earth release and the kernel migration are largely controlled in UCO fuel in which
proper proportions of UO2 and UC2 phases are
mixed in the kernel.68,69 Gas pressures in such particles will certainly be reduced compared to those in
pure UO2 fuel, and UCO fuel kernels should retain
all the rare earths, other than Eu, well. 68 Likewise,
the addition of the ZrC oxygen getter in the UO2 fuel
types accomplishes the same purpose, while also
allowing much higher retention of Eu in the pure
oxide kernels. The TRISO-coated UCO particles
were selected over the ZrC-gettered TRISO-coated
UO2 particles as the reference fissile fuel for steamcycle HTGRs, based more on fabrication considerations and the availability of a wider database than upon
the fission product release results described earlier.20
However, the excellent performance for the ZrCgettered UO2 TRISO-coated particle has made this
fuel a strong candidate for VHTR or DB-MHR.79
To improve the high-temperature stability and the
fission product retention, ZrC-TRISO-coated UO2
particles, instead of the ordinary TRISO, gettered
with a solid ZrC overcoating on the kernel were
suggested as one of the possible design variations.80

3.08.4 SiC-Containing TRISO-Coated
Particle Fuel
3.08.4.1 Designs of SiC-Containing
TRISO-Coated Particle Fuel
The chemical interaction of the SiC layer of the
TRISO-coated particle fuel with fission products
occurs when the fission products are released from
the fuel kernel and subsequently reach the SiC layer

through the IPyC layer. Generally, two methods may
be available to prevent the corrosion of the SiC layer:
(1) to keep the fission products within the fuel kernel
and (2) to make a barrier to the diffusion of the fission
products to the SiC layer.
The corrosion of the SiC layer by lanthanide seen
in the TRISO-coated UC2 and UCxOy (O/U 1.1)
particles could be avoided, when oxide or oxycarbide

231

(O/U > 1.1) fuel kernels are used.68,69,81 Lanthanide
is kept as a stable oxide within the oxide and oxycarbide (O/U > 1.1) fuel kernels. This is a typical example of the first method mentioned earlier.
On the other hand, the compositions of the fuel
kernels of oxide, oxycarbide, and carbide showed little
effect on the corrosion of the SiC layer by palladium:
the corrosion of the SiC layer by palladium has been
observed in the TRISO-coated particle fuel with
almost all kinds of fuel compositions of UO2, UC2,
UCxOy, PuO2Àx, 3ThO2–PuO2Àx, ThO2, and (Th,U)
O2.49,50,70,82–84 A new idea was needed to keep the
fission product palladium within the fuel kernel.
Based on the out-of-reactor experiments,85 a concept of the TRISO-coated UO2 and UCO particles
gettered with SiC dispersed in the kernels has been
suggested to prevent the corrosion of the SiC layer by
the fission product palladium.86 The idea is that the
formation of a compound UPd3Si3C5 with a high
melting temperature of >2225 K would keep palladium in the kernel. Although the thermodynamic
calculations have been carried out, the fabrication
and irradiation tests of particles of this kind have

not been performed yet.
To prevent the corrosion of the SiC layer by fission
product palladium based on the second method
described above, three types of new combinations of
the coating layers have been proposed and tested.87 The
idea is to add a layer that traps palladium by chemical
reaction inside the SiC layer of the TRISO coating.
Two kinds of additional layers have been selected: an
SiC + PyC layer and an SiC layer. The SiC þ PyC
layer is composed of SiC with free carbon.88 This kind
of layer has been studied to improve the capability of
fission product retention of the dense PyC layer.89
Figure 12 shows ceramographs of three types of
advanced coatings, together with that of the TRISO
coating for comparison.87 The type-A coating has an
additional layer of SiC þ PyC adjacent to the inside
of the SiC layer. As the SiC þ PyC layer has a better
capability of fission product retention,89 the thickness
of the IPyC layer can be reduced; the IPyC layer
should act as a seal to prevent the chemical reaction
of the fuel kernel with coating gases during the coating process.
The type-B coating is similar to the type-A coating, but the dense PyC layer is present between the
SiC þ PyC and SiC layers. The expected role of
the intermediate dense PyC layer is to interrupt the
radial extension of the corrosion zone from the
SiC þ PyC layer to the SiC layer. The corrosion
zone would extend circumferentially in the SiC þ PyC


232


Advanced Concepts in TRISO Fuel

Den

Den

se

se P

yC

SiC

SiC
SiC

Den

se

+ Py

C

Den

SiC


se P

yC

Por
ous

(a)

Den

Por
ous

UO
2

20 µm

(b)

yC

PyC

PyC

UO
2


20 µm

Den

se P

se P

yC

SiC
se P
yC
S
Den iC
se P
yC

UO
2

(c)

yC

SiC

Den

Por

ous

PyC

+P

se

PyC

Den

PyC

Den

se P

yC

Por
ous

PyC

20 µm

UO
2


(d)

PyC

20 µm

Figure 12 Ceramographs of three types of advanced coatings, together with that of the TRISO coating for comparison;
(a) type-A coating, (b) type-B coating, (c) type-C coating, and (d) TRISO coating. Reproduced from Minato, K.; Fukuda, K.;
Ishikawa, A.; et al. J. Nucl. Mater. 1997, 246, 215–222.

layer when the intermediate dense PyC layer is present. The thickness of the IPyC layer can be reduced in
the same manner as the type-A coating.
In the type-C coating, SiC is used for an additional
layer. The role of the inner SiC layer is to react with
fission products, and the intermediate dense PyC layer
is expected to interrupt the radial extension of the
corrosion from the inner to outer SiC layers.
The increase in the thickness of the SiC layer of
the TRISO coating may be one of the solutions to the
corrosion of the SiC layer by fission products. However, the thicker coating layers will result in a smaller
amount of the fuel material contained in a unit volume. With these advanced coatings, no corrosion
of the SiC layer by fission products is expected to
occur without increasing the total thickness of the
coating layers.

3.08.4.2 Fabrication of SiC-Containing
TRISO-Coated Particle Fuel
No new technology was needed and a conventional
coating apparatus without modification could be used
to fabricate the SiC-containing TRISO-coated particle

fuel. This point is of great advantage for fuel fabrication.
Three types of advanced coatings, type-A, type-B,
and type-C, were prepared. The porous PyC layer was
deposited at 1573 K with the pyrolysis of acetylene
(C2H2) in the flowing argon, and the dense PyC layer
was deposited at 1638 Kwith the pyrolysis of propylene
(C3H6) in the flowing argon. The SiC layer was chemically vapor deposited at 1873 K using methyltrichlorosilane (CH3SiCl3; MTS) and hydrogen. The SiC þ
PyC layer was deposited at 1873 K with MTS and
argon.88 The content of free carbon in the SiC þ PyC


Advanced Concepts in TRISO Fuel

layer was about 40 wt% as examined by electron probe
microanalysis. For the irradiation experiments, three
types of the advanced coatings, together with the
TRISO coating, were deposited on the fuel kernels of
UO2 with 19.7 wt% enriched uranium.87

SiC + PyC

233

SiC

3.08.4.3 Performance of SiC-Containing
TRISO-Coated Particle Fuel
3.08.4.3.1 Irradiation performance

The SiC-containing TRISO-coated fuel particles

with the advanced coatings of the type-A, type-B,
and type-C, together with the TRISO-coated fuel
particles, were irradiated in the JRR-2. The sample
particles were put individually into the holes of
graphite disks, which were piled and then loaded in
two irradiation capsules. The burnups of the fuels
irradiated in two capsules were 3.7 and 7.0% FIMA,
respectively, and the irradiation temperature was
1603 K in both the capsules.87
The irradiated coated fuel particles were examined by visual inspection, X-ray microradiography,
ceramography, and electron probe microanalysis. No
anomaly was found by visual inspection and X-ray
microradiography. The ceramography revealed no
crack in the advanced coating layers or in the TRISO
coating layers. The SiC þ PyC layer in the advanced
coatings of the type-A and type-B showed good irradiation performance as a coating layer. The mechanical
integrity of the advanced coatings was confirmed in this
irradiation experiment.87
The behavior of the fission product palladium in
the coating layers was examined by electron probe
microanalysis. In the TRISO-coated fuel particles,
palladium was distributed along the inner surface of
the SiC layer and reacted with SiC.
On the other hand, in the type-A coating, palladium
was distributed along the inner surface of the SiC +
PyC layer, and no corrosion was found in the SiC
layer. The fission product palladium released from
the fuel kernel was trapped by the SiC þ PyC layer,
as expected. The same behavior was also found in the
type-B coating, as shown in Figure 13.87 In the type-C

coating, palladium was found along the inner surface
of the inner SiC layer and no corrosion was found in
the outer SiC layer. The inner SiC layer trapped the
fission product palladium by reacting with it.87
The irradiation experiment demonstrated that the
advanced coatings had good irradiation performance,
and the additional layers of SiC and SiC þ PyC
trapped palladium effectively to prevent the corrosion of the SiC layer.87

SEI

10 µm

Si

Pd
Figure 13 Electron probe microanalysis of the polished
cross-section of a particle fuel with type-B coating after
irradiation at 1603 K to 7.0% FIMA; (a) secondary electron
image, (b) X-ray image for silicon, and (c) X-ray image for
palladium. Reproduced from Minato, K.; Fukuda, K.;
Ishikawa, A.; et al. J. Nucl. Mater. 1997, 246, 215–222.

The effect of the intermediate dense PyC layer
between the SiC and SiC + PyC layers in the type-A
or between the inner and outer SiC layers in the typeC on the extension of the corrosion zone could not be
demonstrated in the irradiation experiment since the
corrosion depth of the coating layer by palladium was
small compared with the thickness of the layer. But
this point was examined by the out-of-reactor tests.

3.08.4.3.2 Behavior under simulated
conditions

The coated fuel particles having the three types
of advanced coatings inside out were heated with the
powder of palladium at 1773 K for 1 h in flowing argon.


234

Advanced Concepts in TRISO Fuel

In the out-of-reactor tests, palladium was designed to
diffuse into the coating layers from the outside of the
particles. The heated coated fuel particles were polished
and examined with an optical microscope.87
In the type-B coating, which has the intermediate
dense PyC layer between the SiC + PyC and SiC layers,
part of the SiC + PyC layer was completely attacked
through the coating thickness by palladium, while no
corrosion was found in the SiC layer. This interruption
of the radial extension of the corrosion zone from
the SiC + PyC layer to the SiC layer was exactly the
expected role of the intermediate dense PyC layer.
The interruption of the radial extension of the
corrosion zone from the outer to inner SiC layers
by the presence of the intermediate dense PyC layer
was also observed in the type-C coating. The intermediate dense PyC layer between two SiC layers
functioned effectively, as expected.
In the out-of-reactor experiment, the SiC + PyC

layer was proved to be effective as a barrier to the
diffusion of palladium to the SiC layer, and the
intermediate dense PyC layer was found to interrupt the radial extension of the corrosion zone
from the barrier layer to the SiC layer. However, it
should be noted that the intermediate dense PyC
layer could not interrupt the radial extension of
the corrosion zone when the amount of palladium
exceeded the capacity of the barrier layer to react
with palladium.

3.08.5 TiN-Coated Particle Fuel
3.08.5.1
Fuel

Designs of TiN-Coated Particle

The fast neutron environment of GFR requires a new
and different fuel design. The TRISO-coated particle
fuel for HTGR contains the PyC layers, but they will
not withstand the fast neutron environment. The
HTGR fuel elements, in which the coated particles
are dispersed, consist of a light element of carbon, but
structural materials made of light elements have to be
minimized in the fast spectrum reactor core.1
The chemical form of the fuel kernel should
also be considered. Among the various fuel types,
nitride fuel has good potential. The heavy-metal
density of the oxide kernel is smaller than those
of the carbide and nitride kernels. The carbide
kernel has better potential than that of the oxide,

but the fabrication of carbides of transuranium elements is more difficult due to high vapor pressures
of plutonium and americium over the carbide.90

The nitrides of transuranium elements could be
fabricated with the prevention of the vaporization
loss of the transuranium elements by keeping the
nitrogen partial pressure sufficiently high during
the fabrication processes.91
The coating has to be selected so as to be compatible
with the fast neutron environment, the nitride kernel,
and the matrix or structural materials. The use of SiC
as a coating material would have to be avoided as it
reacts with transition metals, such as Ti, Fe, Cr, and
Ni,92 unless the matrix or the structural container is
made of ceramics with good thermal properties, such
as TiN and TiSi2, which are compatible with SiC.1
Titanium nitride with enriched 15N may be a good
choice because of its refractoriness (Tm = 3223 K),
low neutron cross-sections, and compatibility with
steels. It is advantageous for the coating that TiN is
little miscible with actinide nitrides.93
Two configurations of the coated particles have
been proposed. One is the duplex coating, where the
inner layer is a low-density buffer layer and the outer
layer is of high-density fission products container.
The other configuration is a porous kernel and a
thin but strong and dense outer layer to maximize
the fraction of fuel in the particle while maintaining
the integrity of the miniature pressure vessel.1,5
Ferritic steel may be a good choice for the metal

matrix of a pebble-bed type core or the porous metal
frit of a particle bed type core. TiN is completely
stable when in contact with the ferritic steels.1
3.08.5.2 Fabrication of TiN-Coated
Particle Fuel
A series of experiments were carried out to explore
the fluidized bed chemical vapor deposition of the
TiN coating layer for GFR particle fuel.5 In the
experiments, microspheres of zirconium dioxide
(ZrO2)-coated with carbon layers were used. The
TiN layers were chemically vapor deposited on the
particles from titanium tetrachloride (TiCl4), nitrogen (N2), and hydrogen (H2).
The deposition conditions of temperature, reactant gas concentration, and reactant gas composition
were varied to study the effects on the deposited
coating properties. Coating thickness, composition,
density, crush strength, and microstructure were
examined and related to changes in deposition conditions. Coating rate, density, strength, appearance,
cracking, and crystallinity were influenced by deposition conditions. A standard set of deposition parameters was developed.5


Advanced Concepts in TRISO Fuel

3.08.6 Outlook
The recent interest in the coated particle fuel concept
includes its application outside the past experience of
HTGR. To improve the high-temperature performance of the TRISO-coated fuel particles, a new
material other than SiC is needed. Zirconium carbide
is a candidate and the ZrC-coated particles have been
tested. To improve the chemical stability of the
TRISO-coated particles, new configurations of the

coating layers have been proposed and tested. For the
application to the fast reactor fuel, TiN coating layers
have been proposed and tested instead of PyC layers.
The laboratory-scale experiments on these
advanced fuels form the basis of further discussions
of the application and development of advanced concepts in TRISO fuel. Although engineering-scale fabrication of the advanced fuels and irradiation tests are
needed for the demonstration, fundamental studies are
invaluable to develop the advanced concepts. Mechanical and thermal properties measurements of the coating layers, and thermochemical analyses of the fuel
including fission products and coating layers/additives
would be very helpful in developing the concepts and
further modeling the fuel behavior.

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