Comprehensive nuclear materials 3 02 nitride fuel

3.02

Nitride Fuel

Y. Arai
Nuclear Science and Engineering Directorate, Japan Atomic Energy Agency, Ibaraki, Japan

ß 2012 Elsevier Ltd. All rights reserved.

3.02.1

Introduction

41

3.02.2
3.02.2.1
3.02.2.2
3.02.2.3
3.02.2.4
3.02.2.5
3.02.2.6
3.02.3
3.02.3.1
3.02.3.2
3.02.3.3
3.02.3.4
3.02.3.5
3.02.3.6
3.02.3.7
3.02.4

3.02.5
References

Fabrication of Nitride Fuel
Actinide Nitride Compounds
Preparation from Metal or Hydride
Carbothermic Reduction
Other Nitride Formation Processes
Nitride Pellet Fabrication
Nitride Particle Fabrication
Irradiation Behavior of Nitride Fuel
Irradiation Experience
Fuel Design
Chemical Forms of FP
Restructuring
FP Gas Release
Swelling and FCMI
Fuel–Clad Chemical Interaction
Reprocessing of Nitride Fuel
Outlook of Nitride Fuel

43
43
44
44
45
46
47
47
47

47
48
49
50
50
51
51
52
53

Abbreviations
ADS
Accelerator-driven system
CAPRA Fast reactor operated to burn rather than
breed plutonium
DP
Direct pressing
EPMA Electron probe microanalyzer
FCCI
Fuel–clad chemical interaction
FCMI
Fuel–clad mechanical interaction
FIMA
Fission per initial metal atom
FP
Fission products
HLLW High-level radioactive liquid waste
HLW
High-level radioactive waste
ITU

European Institute for Transuranium
Elements
JAEA
Japan Atomic Energy Agency
LINEX Direct synthesis of actinide nitrides in the
salt by the reaction with Li3N
LOF
Loss of flow
MA
Minor actinides
MOX
Uranium and plutonium mixed oxide
PIE
Postirradiation examination
PSI
Paul-Scherrer Institute
PUREX Plutonium uranium recovery by extraction

SPS
TD
TOP
XRD

Spark-plasma sintering
Theoretical density
Transient overpower
X-ray diffraction

3.02.1 Introduction
Nitride fuel has been proposed as an advanced fuel

for fast reactors and developed since the 1960s in
almost the whole nuclear world. In this case, nitride
fuel stands for a solid solution of uranium mononitride (UN) and plutonium mononitride (PuN),
namely (U,Pu)N, in which the Pu/(U þ Pu) molar
ratio roughly ranges from 0.15 to 0.25. In addition,
UN was developed as a potential fuel for space reactors in the United States. Although the interest in
nitride fuel subsided under a global circumstance
of slowdown of fast reactor programs in the 1980s,
the solid solution of UN, PuN, and minor actinide
(MA; Np, Am, and Cm) mononitride, (U,Pu,MA)N,
has been proposed as one of the candidate fuels
for Gen IV-type fast reactors. Furthermore, as a

41


42

Nitride Fuel

dedicated fuel for MA transmutation systems such as
an accelerator-driven system (ADS), U-free nitride
fuel, such as (Pu,MA)N diluted by ZrN, has been
studied mainly in Japan.
At the beginning of the nuclear era, the development of fast reactor fuel cycles was centered on
breeding ratio and doubling time. The reason was
that metallic fuel, the binary or certain ternary alloy
of U and Pu, was adopted in the first generation of
fast reactors. The metallic fuel, however, had disadvantages for commercial use, such as anisotropic
crystal structure, low melting temperature with

phase transformations, and high fission product (FP)
gas-induced swelling. So a solid solution of uranium
dioxide (UO2) and plutonium dioxide (PuO2),
namely (U,Pu)O2 (MOX), has been a reference fuel
for fast reactors and used in many test and prototype
reactors all over the world, although the breeding
ratio is smaller and the doubling time is longer than
those of metallic fuel.
On the other hand, nitride fuel, as well as carbide
fuel, has the advantages of both metallic fuel and
oxide fuel as shown in Table 1. It has a high thermal
conductivity and high metal atom density like metallic fuel, while it has a high melting temperature and
isotropic crystal structure like oxide fuel. These characteristics led to the motivation for developing
nitride fuel for fast reactors because the high thermal
conductivity and high melting temperature allow a
high linear power operation; alternatively, the largediameter fuel pins can be used for a given linear
power. The high metal atom density allows a low
fissile material inventory with flexible core design
and good neutron economics, leading to an improved
breeding ratio and doubling time.
However, the development of nitride fuel has
fallen behind that of carbide fuel, which has similar
physical and chemical properties. The reason includes an unexploited fuel fabrication process and

Table 1

Comparison of typical properties between oxide, metallic, and nitride fuels for fast reactors

Chemical composition
Theoretical density (TD) (g cmÀ3)

Metal atom density (g cmÀ3)
Thermal conductivity (W mÀ1 KÀ1)
at 773 K
at 1273 K
Melting temperature (K)
a

the high neutron capture cross section of 14N
(99.6% abundance in natural nitrogen) deteriorating
neutron economics. However, the fuel fabrication
process has improved since the late 1980s, and the
breeding ratio and doubling time have not been
the center of the development of fast reactor fuel
cycles. Furthermore, it was found that nitride fuel is
less hygroscopic in nature than carbide fuel, which
will be advantageous for technological development.
It was also found that nitride fuel dissolves well in
nitric acid without any formation of Pu oxalate,
which will be compatible with hydrochemical reprocessing technology such as the PUREX process. So
since the late 1980s, the global interest has moved
from carbide fuel to nitride fuel.
We can find two distinguished monographs about
nitride fuel: one is written by Matzke1 published in
1986 and the other by Blank2 in 1994. These monographs describe nitride fuel and carbide fuel as MXtype fuel (X ¼ N or C) for fast reactors in detail from
scientific and technological viewpoints. It should also
be mentioned that (U,Pu)N fuel with high Pu content, in which the Pu/(U þ Pu) molar ratio is roughly
0.45–0.55, was studied in France as a fast reactor fuel
for incineration of Pu in the 1990s. The good dissolution in nitric acid and stable crystal structure even
at high Pu content led to the potential CAPRA core
with (U,Pu)N fuel for incineration of Pu.3 Although

not being described in this chapter, another interesting aspect of nitride fuel was pointed by Lyon et al.,4
who indicated the superior safety margin in case of
hypothetical loss of flow (LOF) and transient overpower (TOP) events.
In a space reactor program called SP-100 in the
United States, UN with highly enriched 235U was
chosen as a reference fuel because it has the most
favorable properties and will show the best performance for space reactor fuels.5 An extensive work
was carried out in SP-100 program and Hayes et al.

At 0.1 MPa N2 pressure.

Oxide fuel

Metallic fuel

Nitride fuel

(U0.8Pu0.2)O2
11.1
9.75

U–19Pu–10Zr (wt.%)
15.9
14.3

(U0.8Pu0.2)N
14.3
13.5

4.1

2.9
3083

18
31
1330

15
18
3053a


Nitride Fuel

summarized the physical,6 mechanical,7 transport,8
and thermodynamic properties9 of UN, while FP gas
release and swelling of UN were summarized by
Storms10 and Ross et al.,11 respectively. Ross et al.12
also compiled and analyzed the thermal conductivity
data of UN. On the other hand, the diffusional and
mechanical properties were reviewed by Routbort
et al.13 previously.
Since the late 1990s, the partitioning and transmutation of MA has attracted global interest. It may
contribute to the decrease of toxicity of high-level
radioactive waste (HLW) and the mitigation of burden
for its final disposal. Several transmutation systems
and MA-containing fuels have been proposed so far.
Among them, the Japan Atomic Energy Agency
( JAEA) proposed a subcritical ADS as a transmutation system and MA nitride fuel as a dedicated fuel for
transmutation.14 Besides the thermal and neutronic

properties, the mutual solubility of actinide mononitrides in a wide range of composition and combination
becomes an advantage of the fuel with high MA content. Fabrication of MA nitride fuel and its property
measurements have been carried out in JAEA.15–18
In this chapter, fabrication of nitride fuel and its
irradiation behavior are summarized in Sections
3.02.2 and 3.02.3, respectively. A brief description
about reprocessing of spent nitride fuel is given in
Section 3.02.4, because the reprocessing technologies
are closely related with the specific issues of nitride fuel
as 14C formation from natural nitrogen and 15N enrichment. On the other hand, properties of nitride fuel are
described in Chapter 2.03, Thermodynamic and
Thermophysical Properties of the Actinide Nitrides.
In addition, an outlook of nitride fuel is briefly given
in Section 3.02.5.

43

Table 2
Crystal structures and lattice parameters of
nitrides of Th, U, Np, Pu, Am, and Cm
Compounds

Structure

Lattice parameter
(nm)

ThN
Th3N4

NaCl-type fcc
Th3P4-type
hexagonal

0.5167
a ¼ 0.3871

UN
a-U2N3 þ x
b-U2N3 À x

NaCl-type fcc
Mn2O3-type bcc
La2O3-type
hexagonal

UN2 À x
NpN
PuN
AmN
CmN

CaF2-type fcc
NaCl-type fcc
NaCl-type fcc
NaCl-type fcc
NaCl-type fcc

N

c ¼ 2.7385
0.4889
1.0685
a ¼ 0.3696
c ¼ 0.5840
0.531
0.4899
0.4905
0.4995
0.5027

U – Pu – N
1000 ЊC
α-U2N3 + (U,Pu)N + N2
β + (U,Pu)N

1 atm
α-U2N3
β-U2N3
UN

Solid
U

α + β + (U,Pu)N

PuN

α-U2N3
+ (U,Pu)N

Liquid
Pu

Figure 1 Ternary U–Pu–N phase diagram at 1273 K.
Reproduced from Matzke, Hj. Science of Advanced LMFBR
Fuels; North-Holland: Amsterdam, 1986.

3.02.2 Fabrication of Nitride Fuel
3.02.2.1

Actinide Nitride Compounds

Although nitride fuel usually stands for a mononitride or its solid solution, such as UN and (U,Pu)N,
higher nitrides other than mononitrides exist in the
Th–N and U–N binary systems. Table 2 summarizes
the crystal structures and lattice parameters of actinide nitrides reported in the Th–N, U–N, Np–N,
Pu–N, Am–N, and Cm–N binary systems. The
binary U–N and Pu–N, and ternary U–Pu–N systems were investigated and reviewed by Holleck,19,20
Tagawa,21 and Potter22 in detail.
The ternary U–Pu–N phase diagram at 1273 K
in Matzke’s monograph,1 originally calculated by

Holleck,19 is shown in Figure 1. The system is characterized by a complete solubility of UN and PuN.
It is considered that (U,Pu)N phase has a narrow
composition range of the N/(U þ Pu) molar ratio.
Although Pu2N3 does not exist in the Pu–N system,
a sesquinitride phase was identified in the U–Pu–N
system at a Pu/(U þ Pu) molar ratio of 0.15.23 As seen
in Table 2, actinide mononitrides have the same

crystal structure with similar lattice parameters except for ThN, which leads to the mutual solubility.
In a mononitride lattice with NaCl-type structure,
small nitrogen atoms are incorporated into a dense
face-centered cubic packing of metal atoms.


Nitride Fuel

3.02.2.3

Carbothermic Reduction

Carbothermic reduction is the most widely used process for preparing nitride fuel. The starting material
is a dioxide and carbon, and the general reaction is
expressed as
MO2 þ 2C þ 0:5N2 ¼ MN þ 2CO

½IŠ

where M represents an actinide element, such as
U and Pu. The mixture of dioxide and carbon is
heated in N2 gas stream, usually at 1773–1973 K. It
is considered that the carbothermic reduction could
be applied in a technological production line as well
as in a laboratory scale experiment, in contrast to the
metal or hydride route.24 Furthermore, homogeneous
products can be obtained by carbothermic reduction.
However, high amounts of oxygen, up to several
thousand parts per million, are likely to remain in the
products as impurity in case the initial carbon to dioxide mixing molar ratio, C/MO2, is 2.0. Therefore, an

excess amount of carbon is usually added to the mixture
to reduce the oxygen content and the residual carbon is
removed from the products by heating in N2–H2
stream as CH(425) or HCN26 after carbothermic reduction. The initial C/MO2 mixing ratio was chosen at
2.2–2.5 for the preparation of UN and (U,Pu)N. Besides
the two-step reaction constituted by the carbothermic
reduction in N2 stream and the following decarburization in N2–H2 stream, a one-step reaction in N2–H2 or

60

61

62

63
64
2q (deg)

(2 2 2)

65

PuN
NpN

AmN

CmN

PuN

NpN

Nitride preparation methods from metal or hydride
were investigated mainly in the 1960s. They include
the nitridation of U or Pu metal in N2 or NH3 at
about 1073–1173 K, arc-melting of U or Pu metal
under N2 pressure, nitridation of fine grained U or
Pu powder formed by the decomposition of hydrides
with N2 or NH3 and direct reaction of UH3 or PuH2.7
with N2 or NH3. In the case of uranium nitrides, the
products were often U2N3, which was subsequently
decomposed to UN and N2.
These reactions are exothermic and should be
carried out slowly by temperature cycling for better
control of the products. Furthermore, these methods
necessitate a high-purity inert gas atmosphere, since
the fine-grained powders of metal, hydride, and
nitride are chemically active and likely to react with
moisture and oxygen in air even at room temperature.
So it is difficult to apply the metal or hydride route to
a technological fuel production line and these methods were restricted to a laboratory scale experiment.

(3 1 1)

Preparation from Metal or Hydride

AmN

3.02.2.2

CmN

44

66

67

Figure 2 X-ray diffraction pattern of (Np,Pu,Am,Cm)N
prepared by carbothermic reduction.81 Reprinted with
permission from OECD/NEA (2007), Actinide and Fission
Product Partitioning and Transmutation, Ninth Information
Exchange Meeting, Nıˆmes, France, Sept 25–29, 2006,
p 119, www.nea.fr.

NH3 stream can be applied although a higher initial
C/MO2 mixing ratio is necessary than that for the twostep reaction. For the preparation of UN and (U,Pu)N,
the atmosphere is changed to Ar or He from N2 or
N2–H2 at a temperature lower than about 1673 K to
prevent the formation of higher nitrides.
In the case of preparation of solid solution such
as (U,Pu)N, both the reduction of the mixture of
respective dioxides and the solid solution formation
of respective mononitrides can be applied. Figure 2
shows the X-ray diffraction (XRD) pattern of (Np,Pu,
Am,Cm)N prepared by the carbothermic reduction of
the mixture of respective dioxides, from which the
formation of quaternary mononitride solid solution
was confirmed.
Mechanism and kinetics of carbothermic reduction were investigated by several authors, such as

Muromura et al.,27–30 Lindemer,31 Greenhalgh32 and
Bardelle et al.,26 mainly by chemical and XRD analyses, and weight change measurement for UN, PuN,
and (U,Pu)N. Muromura et al. investigated the mechanism of carbothermic reduction at 1693–2023 K for
UN in detail. According to their results, the reaction
is divided into four stages: (1) formation of UN1 À xCx
from UO2, (2) decarburization of UN1 À xCx, (3) formation of UN1 À xCx with equilibrium composition,
and (4) pure UN formation. They also claimed that
the carbothermic reduction followed the first-order
rate reaction expressed as
À lnð1 À aÞ ¼ kt

½1Š

where a represents the reaction ratio, k the rate
constant, and t the time, with an activation energy


Nitride Fuel

of 347 kJ molÀ1. This value is consistent with that
reported by Greenhalgh,32 360 kJ molÀ1. On the
other hand, Muromura et al. claimed that the decarburization in N2–H2 or NH3 stream after carbothermic reduction followed the phase boundary-type rate
reaction expressed as
1 À ð1 À aÞ1=3 ¼ kt

½2Š

with activation energies of 285 kJ molÀ1 in 25%
N2–75% H2 stream and 175–185 kJ molÀ1 in NH3
stream, respectively.

Kinetics was also investigated by thermogravimetry for (U,Pu)N33 and (U,Np)N.34 The results almost
agreed with that for UN by Muromura et al.; the
carbothermic reduction in N2 stream followed
the first-order rate reaction with activation energies
of 307 kJ molÀ1 for (U,Pu)N and 344–385 kJ molÀ1
for (U,Np)N. Furthermore, the decarburization for
(U,Np)N in 92% N2–8% H2 stream followed the
phase boundary-type rate equation with an apparent
activation energy of 210 kJ molÀ1. However, it should
be pointed out that the decarburization includes
both the removal of free carbon resulting in a
decrease in weight and the replacement of carbon
by nitrogen in carbonitride resulting in an increase
in weight.
Typical impurities in nitride fuel prepared by
carbothermic reduction are oxygen and carbon.
It was found that the level of impurities could be
kept lower than 1000–2000 ppm for both oxygen and
carbon by adjusting the initial C/MO2 mixing ratio.
Carbonitrides such as UN1 À xCx and PuN1 À xCx are
characterized by complete solubility of the UN–UC
and PuN–PuC systems, while solubility limits of hypothetical UO in UN and PuO in PuN were reported at
7% and 14%, respectively.35 It was reported that the
carbon impurity content in mononitride prepared by
carbothermic reduction is related to the thermodynamic equilibrium composition of carbonitride with
free carbon under nitrogen atmosphere.17 When the
same condition of carbothermic reduction was applied
for UN, NpN, and PuN, the carbon impurity content
decreased with the increase of atomic number of
actinides. Indeed, a rather high initial C/MO2 mixing ratio was chosen for the preparation of AmN and

(Pu,Cm)N,36,37 since the monocarbides of Am or
Cm are thermodynamically unstable.
It is well known that Am-bearing species have high
vapor pressures in comparison with the other actinides. Vaporization of Am during fuel fabrication process should be kept as low as possible. In the case of

45

preparation of Am-bearing nitrides by the two-step
reaction, the carbothermic reduction in N2 stream
was carried out at 1573 K, which was lower than the
cases for UN and (U,Pu)N by about 200 K.36,38 Then
the temperature was raised to 1773 K for the decarburization in N2–H2 stream. It is considered that the
intermediate product of AmCO is likely to vaporize
congruently during the carbothermic reduction. On
the other hand, the vaporization of Pu during carbothermic reduction can be neglected, which is different from the preparation of Pu-bearing carbides by
carbothermic reduction carried out in vacuum.
3.02.2.4 Other Nitride Formation
Processes
Four processes were reported for the preparation of
nitride with regard to pyrochemical reprocessing
of spent fuel. The first one is the direct dissolution
of spent nitride fuel in liquid Sn, followed by the
pressurization with N2. It was reported that UN
powder with high density sank to the bottom and
could be mechanically separated from the liquid
phase.39 The second and third processes concern
the nitridation of actinides recovered in liquid Cd
cathode by molten salt electrolysis. The second one is
the nitridation by N2 gas bubbling, in which N2 gas is
passed into liquid Cd phase at 773–823 K. Kasai et al.

reported that they succeeded in preparing UN or
U2N3 granules by the N2 gas bubbling method.40 It
was found, however, that the method was not applicable to the nitridation of Pu in liquid Cd because of
the thermodynamic stabilization of Pu in liquid Cd
phase.41 On the other hand, the third one is the
nitridation–distillation combined reaction, in which
the liquid Cd cathode-containing actinides are
heated in N2 stream at 973 K. In this method, the
nitridation of actinides and distillation of Cd proceed
simultaneously. Preparation of (U,Pu)N, PuN, and
AmN has been reported so far by the nitridation–
distillation combined method.41,42 The fourth one is
called LINEX process, in which actinides dissolved
in the chloride molten salt are converted to nitride by
the direct reaction with Li3N.43
In addition, a new process was reported by Yeamans
et al.44 They successfully synthesized UN from UO2 by
making it react first with NH3(HF)2 at ambient temperature to form (NH4)4UF8, and then with NH3 at
1073 K to UN2, followed by the decomposition to UN
at 1373 K in Ar. This method has the advantage of a
low-temperature operation in comparison with the
carbothermic reduction of dioxides.


46

Nitride Fuel

3.02.2.5

Nitride Pellet Fabrication

Nitride fuel pellets are usually prepared by a classical
powder metallurgical manner; the product of carbothermic reduction is ground to powder by use of
a ball mill, pressed into green pellets and sintered in a
furnace at 1923–2023 K. An organic binder is sometimes added to the ground powder to facilitate the
pressing. Finally, the diameter of sintered pellets is
adjusted by use of a centerless grinder. As is mentioned later, one of the characteristics of nitride fuel is
that both He- and Na-bonded pins can be applied. In
general, an He-bonded fuel pin is characterized by
low-density pellets (i.e., 80–85% of theoretical density (TD)) and a small gap width between pellets and
cladding tube, whereas a Na-bonded fuel pin is characterized by high-density pellets (i.e., >90% TD)
and a large gap width.
Actinide nitride powder has a low sinter-ability in
comparison with that of oxide or carbide powder,
which is derived from a low diffusion rate of metal
atoms in mononitrides. So a rather high sintering
temperature (i.e., T >1973 K) is necessary for preparing dense UN or (U,Pu)N pellets higher than 90%
TD.45 Although a small amount of Ni powder is an
effective sintering aid for carbide fuel, it is not applicable to nitride fuel. On the other hand, Bernard et al.,
reported that oxygen impurities tend to promote the
sintering of (U,Pu)N pellets.24 However, the increase
of oxygen impurities in UN and (U,Pu)N up to
1 wt% resulted in the decrease of density and grain
size of sintered pellets.46 Microstructures of (U,Pu)N
pellets with different oxygen impurity contents are
shown in Figure 3.
Sintering atmosphere also affects the sintered
density of nitride fuel pellets. It was reported that
sintering in high N2 partial pressure, such as in N2 or

N2–H2 stream, resulted in lower density than sintering in low N2 partial pressure, such as in Ar or Ar–H2
stream.45,47 This is an opposite tendency of the
self-diffusion coefficient of Pu in (U,Pu)N at different N2 partial pressures.48 The residual oxygen
impurity contents might affect the density of pellets
sintered in different atmospheres. On the other hand,
sintering in N2 or N2–H2 stream is indispensable
for Am-bearing nitride pellets from the viewpoint
of mitigating loss of Am by evaporation. It was
reported that the density higher than 85% TD was
attained for (Np,Am)N and (Pu,Am)N pellets by
sintering in N2–H2 stream at temperatures lower
than 1953 K.49
In addition to the classical powder metallurgical
manner, a direct pressing (DP) method was proposed
by Richter et al.50 In this method, the nitride compacts after carbothermic reduction were not ground
to powder but directly pressed into green pellets,
followed by sintering under the conventional manner.
The DP method has the advantage of avoiding dust
production and shortening preparation period. The
(U,Pu)N pellets prepared by the DP method had a
density of about 83% TD with levels of oxygen and
carbon impurities lower than 0.1 wt%.51 The open
porosity predominated in the pellets prepared by the
DP method.
An isostatic hot-pressing technique was applied to
fabrication of dense UN specimens for thermal and
mechanical property measurements. Speidel et al.
prepared UN pellets higher than 95% TD by consolidating the powder sealed in a refractory metal container under a pressure of 6.9 MPa at 1753–1813 K.52
Furthermore, a spark-plasma sintering (SPS) method

for nitride fuel has been applied to preparation in a
laboratory scale experiment recently.53 The SPS

Oxide

20 μm

(U,Pu)N pellet containing
0.21 wt% oxygen

20 μm

(U,Pu)N pellet containing
0.99 wt% oxygen

Figure 3 Microstructures of (U,Pu)N pellets with different oxygen impurity contents. Reproduced from Arai, Y.; Morihira, M.;
Ohmichi, T. J. Nucl. Mater. 1993, 202, 70–78.


Nitride Fuel

method is a kind of pressure-assisted sintering that
utilizes an electric current. The method has the advantage of obtaining dense pellets at a drastically lower
sintering temperature and a shorter sintering time
than those of the conventional methods.
3.02.2.6

Nitride Particle Fabrication

Nitride particle fabrication method was vigorously

developed in the Paul-Scherrer Institute (PSI) of
Switzerland,54,55 then followed by India56 and Japan.57
The starting material is usually a nitric solution of
actinides and this method has the advantage of avoiding
dust production and feasibility of remote operation in
comparison with the conventional powder process.
The nitride particles prepared may be directly filled
into fuel pin (sphere-pac fuel) or pressed and sintered
to fuel pellets.
The production of microspheres is carried out by
a so-called sol–gel process. The feed solution is
mixed with an aqueous solution of gelation agent,
urea, dispersed carbon black, and surfactants. Different size of microspheres can be obtained by changing
the nozzle used for microspheres production. Besides
the external gelation process using gelation agent, the
internal gelation process developed by PSI consists of
falling the droplets of feed material into hot silicon
oil for microspheres production. After washing, drying and calcining to MO2 þ C microspheres, they are
subjected to carbothermic reduction. In the case of
preparing sphere-pac fuels, the carbothermic reduction is carried out at higher temperature than the
conventional powder process to obtain dense nitride
particles by reaction sintering.
The sol–gel process is proposed for the preparation of nitride fuel for the transmutation of MA under
the double-strata fuel cycle concept.14 In this concept, MA partitioned from high-level liquid waste
(HLLW) in a reprocessing plant is converted to
nitride microspheres by the sol–gel process and carbothermic reduction, followed by mixing with diluent materials and sintering for pellet preparation.

Table 3

47

3.02.3 Irradiation Behavior of
Nitride Fuel
3.02.3.1

Irradiation Experience

The irradiation experience of nitride fuel is rather
limited in comparison with the other fuels for fast
reactors, such as oxide, metallic, and carbide fuels.
Especially, the number of (U,Pu)N fuel pins irradiated
in fast reactors so far is smaller than 200 all over the
world, which is summarized in Table 3. The highest
burnup was attained in the irradiation test in the
EBR-II fast reactor, but still lower than 10% of fission
per initial metal atom (FIMA).58 On the other hand,
high burnups, that is, >15% FIMA, were attained in
thermal reactors, such as ETR in the United States62
and HFR in the Netherlands.63 Most of them were
irradiated in instrumented capsules.
In the United States, following the capsule irradiation in ETR and EBR-II, 3 subassemblies constituted by 57 (U,Pu)N fuel pins were irradiated in
EBR-II,64 whereas in Europe, more than 10 (U,Pu)N
fuel pins were irradiated in fast test reactors, such
as DFR, RAPSODIE, and PHENIX.59,60 Besides, in
Japan, two (U,Pu)N fuel pins were irradiated in fast
test reactor JOYO.61
With regard to nitride fuel other than (U,Pu)N,
five subassemblies of 235U-enriched UN fuel were
irradiated to about 9% FIMA in BR-10 in the
1980s.65 In addition, nitride fuels for the transmutation of MA have been subjected to the irradiation

tests recently. Besides (U,Pu,Np,Am)N and (Pu,Am,
Zr)N fuels irradiated in PHENIX,66 (Pu,Zr)N fuels
were irradiated in Russia67 and Japan.68
3.02.3.2

Fuel Design

There are two typical bonding concepts of (U,Pu)N
fuel pins for fast reactors: one is Na bonding and the
other is He bonding. Since (U,Pu)N fuel is compatible with liquid Na at operating temperatures, the gap
between fuel pellets and cladding tube can be filled
with liquid Na as well as gaseous He. In a sense of

Irradiation tests of (U,Pu)N fuel carried out in fast reactors

Reactor

Bonding

Max. linear power (kW mÀ1)

Max. burnup (% FIMA)

References

EBR-II
DFR
RAPSODIE
PHENIX
JOYO

He and Na
He
Na
He
He

110
130
130
73
75

9.3
7.6
3.4
6.9
4.3

Bauer et al.58
Blank59
Blank59
Fromont et al.60
Inoue et al.61


Nitride Fuel

liquid metal, liquid Li bonding was also suggested
for UN-fueled space reactors. In a He-bonding concept, the gap is filled with He of atmospheric pressure. Besides the pellet-type fuel, vibropac (U,Pu)N

fuel pins were irradiated in DFR by use of He for
bonding gas.59
A Na-bonding concept is characterized by a large
gap width (i.e., >0.5 mm) between fuel pellets and
cladding tube and a high density of fuel pellets (i.e.,
>90% TD). This concept has the advantage of
keeping the fuel temperature relatively low due to
good thermal conductivity of liquid Na. Furthermore, the temperature of fuel pellets is considered
as quasiconstant. A shroud tube was sometimes used
in order to maintain the fuel fragments in their original geometry. On the other hand, the disadvantage
of a Na-bonding concept includes the difficulty in
fuel pin fabrication and spent fuel reprocessing. Furthermore, with regard to safety consideration, the
possibility of loss of Na in a breached pin has to be
evaluated.
At present, a He-bonding concept is considered
as the reference for (U,Pu)N fuel. A He-bonding
concept is characterized by a small gap width (i.e.,
<0.2 mm) and a low density of fuel pellets (i.e.,
80–85% TD). The temperature of fuel pellets
becomes high in comparison with the fuel with Na
bonding, especially at an early stage of irradiation.
However, the small gap is closed by free swelling of
fuel pellets at a burnup of 2–3% FIMA, which
enhances the gap conductance and lowers the fuel
temperature. A schematic change in fuel temperature for He-bonded (U,Pu)C fuel pin is illustrated
in Figure 4, which is also applicable to (U,Pu)N
fuel pin.69 The irradiation period A, as shown in
Figure 4, corresponds to the first rise of power and
lasts for one to several days, the period B the resintering of pellets center and closure of He gap, and
the period C the quasistate irradiation period in

which the fuel–clad mechanical interaction (FCMI)
starts. In order to accommodate the swelling and
mitigate the strain on the cladding tube at burnup
progressing, a rather low smear density (i.e., 75–80%
TD) is adopted for He-bonded fuel pin.
Blank proposed the ‘cold fuel concept’ for MXtype fuel, in which the maximum fuel temperature is
kept lower than one-half or one-third of the melting
temperature in Kelvin.69 If this concept is realized,
both low fission gas release and mild restructuring
and mild swelling characteristics can be compatible
in both Na-bonded and He-bonded fuels.

1600

Tcenter
Temperature (ЊC)

48

1200
ΔTfuel
Tsurface

800
A

B

C

Time scale
extented
400
0

1

2

3

4

5

6

Burnup (at.%)
Figure 4 Schematic change in temperature for MX-type
fuel pin. Reproduced from Blank, C. J. Less Common Met.
1986, 121, 583–603.

Table 4
Chemical forms of typical FP in the irradiated
(U,Pu)N fuel
Elementa Chemical
forms
Ba
Cs
Kr

Mo
Pd
Pr
Rh
Sm
Tc
Xe
Zr

Ba3N2
Cs, CsI, CsTe
Kr
Mo
(U,Pu)(Pd,Ru,Rh)3
PrN
(U,Pu)(Pd,Ru,Rh)3
SmN
Tc
Xe
ZrN

Element Chemical
forms
Ce
I
La
Nd
Pm
Rb
Ru

Sr
Te
Y

CeN
CsI
LaN
NdN
PmN
Rb, RbI
(U,Pu)(Pd,Ru,Rh)3
Sr3N2
Te, CsTe
YN

a
Elements with concentrations <0.08 at.% at burnup of 10% FIMA
are not shown.

3.02.3.3

Chemical Forms of FP

Chemical forms of FP in nitride fuel were evaluated
by a thermodynamic equilibrium calculation and
burnup-simulated experiments70,71 as well as postirradiation examinations (PIE).72 These results agreed
with each other in general but it is difficult to identify
the phases other than mononitride by XRD or metallographic analysis even at a burnup higher than
10% FIMA.
Table 4 shows the most probable chemical forms

of FP in the irradiated (U,Pu)N fuel. Among them,


Nitride Fuel

gaseous FP such as Xe and Kr exist as an elementary
state. Semivolatile FP such as Cs, I, and Te are likely
to exist as an elementary state or compounds such as
CsI and CsTe. Rare earth elements such as Nd, Ce,
Pr, and Y, and Zr and Nb are considered to be
dissolved in (U,Pu)N and form a mononitride solid
solution. On the other hand, Mo and Tc are considered to exist in an elementary state together, and
Ba and Sr are considered to form lower nitrides
such as Ba3N2 and Sr3N2. Platinum group elements
such as Pd, Ru, and Rh are likely to form an intermetallic compound, (U,Pu)(Pd,Ru,Rh)3, in the irradiated
nitride fuel.
The change of N/(U + Pu) ratio in mononitride
phase was evaluated and Bradbury et al.71 reported
that it increased by 2.1% at a burnup of 10% FIMA.
Furthermore, it should be mentioned that the chemical forms of FP are also influenced by the oxygen
and carbon impurity contents, that is, rare earth elements are likely to form oxide precipitates and Zr is
likely to form ZrC dissolved in mononitride phase.
It was reported that the lattice parameter of the
mononitride phase of (U,Pu)N fuel did not significantly change with burnup progressing.70 This tendency was explained by the compensation of the
increase in lattice parameter due to the dissolution
of rare earth elements and the decrease due to the
dissolution of ZrN.
3.02.3.4

Restructuring

Because of relatively low fuel temperature and temperature gradient, the restructuring of (U,Pu)N fuel
is mild in comparison with MOX fuel for fast reactors. However, in the He-bonded (U,Pu)N fuel irradiated at high linear power, a distinct restructuring
was observed, in which three structural zones shown

in Figure 5 were identified by Matzke.1 Zone I found
in the central of the fuel pellet was characterized by
very porous structure. The pores were grown to
roughly the grain size and FP gas release was high.
A small central hole was sometimes observed in
Zone I of the He-bonded (U,Pu)N fuel.59,60 The
mechanism of formation is, however, different from
that in MOX fuel; according to Coquerelle et al.72 it
results from the migration of lenticular pores up the
radial temperature gradient in MOX fuel, whereas it
is apparently created by an in-pile resintering mechanism in (U,Pu)N fuel. Zone II was characterized by
pseudocolumnar grains observed in MOX fuel. However, it was not observed in (U,Pu)N but only in
(U,Pu)C irradiated at a linear power higher than
100 kW mÀ1. Zone III was characterized by the structure accompanied with grain growth, grain boundary
bubbles, and healing of cracks. FP gas release was
relatively high and swelling of (U,Pu)N fuel was
mostly responsible in this zone. On the other hand,
Zone IV had the as-fabricated structure. In the case of
low-density pellets, slight densification occurred
because of in-pile resintering. Both FP gas release
and swelling were small in Zone IV.73
Temperature range of each zone found in (U,Pu)N
fuel irradiated in DFR to 4% FIMA was roughly
evaluated by Matzke1 as follows. Zone I appeared at
a temperature higher than 1673 K, while Zone IV

predominated at a temperature lower than 1423 K.
The intermediate temperature ranging from 400 to
500 K corresponded to Zone III. On the other hand,
under the ‘cold fuel concept’ proposed by Blank,69
most part of fuel pellets should represent the asfabricated structure seen in Zone IV characterized
by low FP gas release and mild swelling.
Richter et al.74 observed the macro- and
microstructures of unirradiated (U,Pu)N pellets

Low temperatures,
edge

Zone

IV
Structure of
as-fabricated fuel

49

High temperatures,
center

III

II

Grain growth, grain
Pseudocolumnar
boundary bubbles grain zone, elongated

grains and pores

I
Very porous
central zone

Figure 5 Schematic presentation of structural zones observed in MX-type fuels. Reproduced from Matzke, Hj. Science of
Advanced LMFBR Fuels; North-Holland: Amsterdam, 1986.


Nitride Fuel

heated in He or N2 atmosphere under temperature
gradient. From the viewpoint of structural stability,
they suggested that the operational limit of temperature for (U,Pu)N was 2000 K in normal condition,
although the structural change observed was affected
by the oxygen impurity contents.
During the power rise at an early stage of irradiation, tensile stresses are created in the outer zone and
compressive stresses at the central zone of fuel pellets. Since most of ceramics are more sensitive to
tensile than compressive stresses, crack formation
usually occurs at the periphery during start up. It
was reported that radial cracks predominated in the
Na-bonded (U,Pu)N fuel pellets irradiated in thermal test reactors, while circumferential cracks predominated in the He-bonded (U,Pu)N fuel pellets.62
On the other hand, many short radial cracks were
observed at the periphery in the He-bonded (U,Pu)N
fuel pellets irradiated in EBR-II.58 Healing of cracks
was often observed in the He-bonded (U,Pu)N fuel
after the closure of He gap by fuel swelling.
3.02.3.5

100
Bauer et al.63
This study

Fission gas release (%/% FIMA)

50

10
Release through surfaceconnected porosity

L414
1
L413

Estimated recoil release
from geometric surface
0.1

FP Gas Release

0

Since the number of nitride fuel pins subjected to PIE
is limited, there have been no systematic results
dealing with FP gas release of nitride fuel. But it is
generally known that FP gas release of nitride fuel is
much lower than that of MOX fuel. The FP gas
release will be influenced by burnup, pellet density,
grain size, and the characteristics of porosities as well

as fuel temperature. By statistically dealing with the
data reported for 95 UN and 39 (U,Pu)N fuels.
Storms10 proposed an equation for FP gas release of
nitride fuel as a function of fuel temperature, burnup,
and density as follows:
È Â
Ã
É
R ¼ 100= exp 0:0025ð90D0:77 =Bu0:09 À TÞ þ 1
½3Š
where R is the FP gas release rate (%), D the fuel
pellets density (% TD), Bu the burnup (% FIMA),
and T the temperature of fuel (K).
Bauer et al.62 summarized the results of FP gas
release for (U,Pu)N fuel irradiated at a relatively
low fuel temperature. From the fuel porosity dependence, they suggested that the recoil of FP gas atom
from the geometric surface was responsible for the
release from pellets with a density higher than 85%
TD, and it rapidly increased with the decrease of
density lower than 82% TD since the release through
the surface connected porosity became responsible.
Figure 6 shows the results for (U,Pu)N fuel

10

20
Fuel porosity (%)

30

40

Figure 6 Porosity dependence of FP gas release rate of
(U,Pu)N fuel. Reproduced from Tanaka, K.; Maeda, K.;
Katsuyama, K.; et al. J. Nucl. Mater. 2004, 327, 77–87.

irradiated in JOYO to 4.3% FIMA75 in comparison
with the porosity dependence reported by Bauer
et al.62 Furthermore, for the (U,Pu)N fuel pellets
irradiated in JOYO, Tanaka et al. suggested that
about 80% of FP gas was still retained in the intragranular region, about 15% was in the gas bubbles,
and about 5% was released from the pellets, based on
the results of pin puncture test and electron probe
microanalysis (EPMA) of fuel pellets.
On the other hand, Coquerelle et al.72 reported
that the release rate of Xe from the central region of
(U,Pu)N pellets was about 45% and about 15% from
the outer part of the fuel pellets irradiated in DFR.
Although the burnup was low in their irradiation
campaign, FP gas release was relatively high because
of a linear power higher than 100 kW mÀ1. Therefore,
the high FP gas release could be explained by the
diffusion process as in the MOX fuel.
3.02.3.6

Swelling and FCMI

As mentioned above, FP gas release of nitride fuel is
low in general. This characteristic potentially leads to

Nitride Fuel

the increase of FP gas-induced swelling, in case FP
gas bubbles are formed and retained in the pellets.
Volumetric swelling is also caused by the accumulation of solid FP and crack formation in the pellets.
The mechanism of swelling of MX-type fuel was
studied in detail in European Institute for Transuranium Elements (ITU) and the results were summarized in the monographs of Matzke1 and Blank.2 On
the other hand, the creep rate of (U,Pu)N fuel is low
in comparison with MOX or metallic fuel at
operating temperatures due to a slow diffusion rate
of metal atoms in nitride fuel. Therefore, the FCMI is
considered the most important subject for the evaluation of (U,Pu)N fuel performance, especially at a
high burnup. Indeed, in the irradiation tests carried
out in the 1960s and 1970s, several He-bonded (U,Pu)
N fuel pins were possibly breached by FCMI.59,64
However, most of He-bonded (U,Pu)N fuel pins
breached were irradiated at a linear power higher
than 100 kW mÀ1 and had a smear density higher
than 85–90% TD. In these pin designs, the strain on
the cladding tube became severe because of the
swelling of fuel pellets, although the integrity of
fuel pins depended on the properties of cladding
tube itself. After the 1980s, the trend of He-bonded
(U,Pu)N fuel design has been changed to a mild
linear power condition (i.e., <80 kW mÀ1) and a low
smear density of fuel pins (i.e., 75–80% TD).
In contrast, the large gap between the pellets and
cladding tube will not be completely closed during the
irradiation period in Na-bonded (U,Pu)N fuel pins. In

this case, however, the relocation of the fragments of
(U,Pu)N pellets caused the ovalization of cladding tube
leading to localized FCMI, which sometimes resulted
in breaching. In the irradiation test of (U,Pu)N fuel in
EBR-II, a shroud tube was used inside the cladding
tube and the effectiveness for maintaining the fuel
fragments in the original position was confirmed.58
The ovalization was also observed in the He-bonded
(U,Pu)N fuel pin irradiated in JOYO with a relatively
large He-gap width.75
The volumetric swelling rate of (U,Pu)N fuel irradiated to 9.3% FIMA was evaluated at 1.83%/%
FIMA without constraint of cladding tube and at
1.44%/% FIMA under constraint of cladding tube
by Bauer et al.64 The similar values were also reported
by Lyon et al.76 and Tanaka et al.75 It should be noted
that FP gas-induced swelling strongly depends on
fuel temperature. Bauer et al.62 reported that the
swelling rate sharply rose at 1573–1673 K for (U,Pu)N
fuel mainly irradiated in thermal test reactors. On the
other hand, Inoue et al.61 suggested that the threshold

51

temperature for the beginning of FP gas bubble precipitation was 1200–1300 K for (U,Pu)N fuel irradiated to 4.3% FIMA in JOYO.
In contrast to FP gas-induced swelling, the
swelling rate due to the accumulation of solid FP is
considered almost independent of temperature and
burnup. Its contribution to the volumetric swelling
was evaluated at 0.5%/% FIMA by a thermodynamic
equilibrium calculation.70

Ross et al.11 proposed an equation for representing
the volumetric swelling of UN fuel as a function of
average fuel temperature, burnup, and as-fabricated
pellet density by statistically dealing with 75 reported
data in the SP-100 program,
3:12
Bu0:83 D0:5
DV =V ð%Þ ¼ 4:7 Â 10À11 Tav

½4Š

where Tav is the average fuel temperature (K), Bu the
burnup (% FIMA), and D the as-fabricated pellet
density (% TD).
3.02.3.7

Fuel–Clad Chemical Interaction

Nitride fuel has a good enough chemical compatibility
with stainless steels used for cladding tube in fast
reactors. Although the formation of chromium nitride
is possible for hyperstoichiometric nitrides under
a thermodynamic equilibrium condition, no results
have been reported for the deterioration of mechanical
properties of the cladding tube. On the other hand, a
slight carburization of inner surface of cladding tube
was found in Na-bonded (U,Pu)N fuel with high oxygen and carbon contents previously. Semivolatile FPinduced intergranular oxidation of the cladding tube
seen in MOX fuel did not occur in (U,Pu)N fuel since
the oxygen potential in the gap region is low enough.
A hypostoichiometric composition of (U,Pu)N fuel

should be avoided because of the reaction of free
metal phase and cladding tube, leading to the formation of (U,Pu)Fe2 and (U,Pu)Ni5-type compounds.1
In general, it is concluded that both He-bonded
and Na-bonded nitride fuels have no problems of
fuel–clad chemical interaction (FCCI) except for
the fuel with high impurity contents or metal phase
precipitation.

3.02.4 Reprocessing of Nitride Fuel
Reprocessing of spent nitride fuel has been investigated only in a laboratory scale experiment so
far. Both hydrochemical77,78 and pyrochemical43,79
processes were proposed for the reprocessing


52

Nitride Fuel

technology. The disposal of long-lived 14C produced
by 14N(n,p)14C reaction in case of nitride fuel with
natural nitrogen or the recovery of expensive 15N in
the case of nitride fuel with 15N-enriched nitrogen
becomes a debatable point in the reprocessing of
nitride fuel.
Two methods, the direct dissolution of spent nitride fuel in HNO3 and the voloxidation of spent
nitride fuel followed by the dissolution in HNO3,
are considered as hydrochemical processes. The
product of hydrochemical reprocessing is the nitric
solution of U þ Pu to be converted to oxide, and then
to nitride by carbothermic reduction. On the other

hand, the pyrochemical reprocessing of spent nitride
fuel has many similarities with the reprocessing of
U–Pu–Zr metallic fuel developed in the United
States. In this case, the product of reprocessing is
U þ Pu alloy recovered in cathode by molten salt
electrorefining, which is to be converted to nitride
again. In addition, a kind of partial reprocessing technique was proposed, in which spent nitride fuel was
pulverized by heating in N2 stream to sesquinitride,
accompanied by the release of FP gas and semivolatile FP.80 A part of the product is subjected to nitride
fuel fabrication without further reprocessing and the
other part is subjected to further reprocessing. By
adopting the partial reprocessing, the material
throughput will be drastically decreased in comparison with normal reprocessing.
Selection of two hydrochemical processes, direct
dissolution or voloxidation, will depend on the
enrichment of 15N in nitride fuel. In case natural
nitrogen is used in nitride fuel, the direct dissolution method will be promising. The direct dissolution
method has the advantage of a simple flowsheet and
recovery of 14C as 14CO2 in the off-gas system. The
dissolution of (U,Pu)N in HNO3 proceeds quasicongruently and any formation of Pu oxalate seen for
(U,Pu)C fuel has not been observed. The conventional hydrochemical reprocessing technology developed for oxide fuel can be applied after the
dissolution process.
In case 15N-enriched nitrogen is used in nitride
fuel, the voloxidation method will be promising for
hydrochemical process. The voloxidation method has
the advantage of recovery of 15N as 15NOx gas in the
off-gas system, which prevents 15N from being contaminated with 14N and isotropic exchange reaction
in the dissolution process.
On the other hand, similarities of electrical conductivity and free energy change of formation for
chlorides in nitride and metallic fuels make it possible

to apply the pyrochemical reprocessing technology
developed for metallic fuel in the United States. In
this case, spent nitride fuel is electrochemically dissolved at the anode with the evolution of N2 gas, and
U and U + Pu are recovered in solid and liquid cathodes, respectively, by electrorefining in a LiCl–KCl
molten salt. For the moment, Cd has been considered
as the most promising material for liquid cathode.
Then actinides are to be converted to nitride in molten
cadmium.41 In addition to the reference pyrochemical
process mentioned above, a concept called as LINEX
process was proposed by Ogawa et al.,43 in which spent
nitride fuel is dissolved in the molten salt, followed by
the nitridation of actinides by the reaction with Li3N.
Since 15N can be recovered as 15N2 gas in the off-gas
system of electrorefining process, the pyrochemical
reprocessing technology is promising for nitride fuel
with 15N-enriched nitrogen.

3.02.5 Outlook of Nitride Fuel
Research and development of nitride fuel as an
advanced fuel for Gen IV-type fast reactors and a
dedicated fuel for MA transmutation systems is still
ongoing. Its superior characteristics as nuclear fuel
make it possible to design the core with high performance and enough safety margins. For the industrialization of nitride fuel, however, several subjects
should be solved by a lot of efforts.
The irradiation performance of nitride fuel should
be demonstrated at a high burnup. The highest
burnup attained by irradiation tests is below 10%
FIMA under fast neutron circumstances. FCMI of
nitride fuel will be a limited factor at burnup progressing for the industrialization. In addition to the

modeling of fuel behavior under normal and transient conditions, the accomplishment of irradiation
tests, in which the average burnup is 15% FIMA at
the lowest, is essential. The irradiation condition
should be well defined and the fuel itself should be
well characterized.
From the viewpoint of fuel cycle, reprocessing is
important. Including the hydrochemical and pyrochemical technologies, the process for reprocessing was
tested only in a beaker-scale experiment. The material
balance of heavy metals should be evaluated through
the process and the data for designing the apparatus and
material handing are almost lacking.
Of course, in case of using the nitride fuel with
15
N-enriched nitrogen, the economical enrichment
technology should be developed. The technology


Nitride Fuel

should also be easy for enlarging the scale and
friendly to the environments. Especially, the use of
15
N-enriched nitrogen is inevitable for the dedicated
fuel for MA transmutation systems. Recycling of
15
N in fuel fabrication process will be necessary for
economic reason, while recovery of 15N in reprocessing will be an optional process in this case.

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