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Preparation of uranium oxide powder for nuclear fuel pelletfabrication with uranium peroxide recovered from uranium oxidescraps by using a carbonate–hydrogen peroxide solution
Kwang-Wook Kim • Jae-Won Lee • Dong-Young Chung •
Eil-Hee Lee • Kweon-Ho Kang • Kune-Woo Lee • Kee-Chan Song •
Myung-June Yoo • Geun-Il Park • Jei-Kwon Moon
Received: 31 October 2011
� Akademiai Kiado, Budapest, Hungary 2011
Abstract This work studied a way to reclaim uranium
from contaminated UO2 oxide scraps as a sinterable UO2
powder for UO2 fuel pellet fabrication, which included a
dissolution of the uranium oxide scraps in a carbonate
solution with hydrogen peroxide and a UO4 precipitation
step. Dissolution characteristics of reduced and oxidized
uranium oxides were evaluated in a carbonate solution with
hydrogen peroxide, and the UO4 precipitation were con-
firmed by acidification of uranyl peroxo–carbonate com-
plex solution. An agglomerated UO4 powder obtained by
the dissolution and precipitation of uranium in the car-
bonate solution could not be pulverized into fine UO2
powder by the OREOX process, because of submicron-
sized individual UO4 particles forming the agglomerated
UO4 precipitate. The UO2 powder prepared from the UO4
precipitate could meet the UO2 powder specifications for
UO2 fuel pellet fabrication by a series of steps such as
dehydration of UO4 precipitate, reduction, and milling. The
sinterability of the reclaimed UO2 powder for fuel pellet
fabrication was improved by adding virgin UO2 powder in
the reclaimed UO2 powder. A process to reclaim the con-
taminated uranium scraps as UO2 fuel powder using a
carbonate solution was finally suggested.
Keywords Uranium scrap � Uranium peroxide �Carbonate � Hydrogen peroxide � Fuel pellet �Sinterability � OREOX
Introduction
Increase in the utilization efficiency of uranium materials
produced from uranium ore will become more important in
the near future because of the rapidly growing demand for
uranium in the many nuclear power plants being built all
over the world to cope with energy and environmental
pressures. Several uranium scraps of UO2 and (U,Gd)O2 are
generated in the steps of the fuel fabrication process, such as
granulation, pelleting, sintering, grinding, and canning. If
the uranium scraps are green scraps, i.e., scraps with no
contamination from other impure metals, they can be
recycled through re-granulation and pelleting without fur-
ther purification. However, when the uranium is recovered
from (U,Gd)O2 scraps or from dirty UO2 scraps contami-
nated with other impurities of Fe, Ni, Al, etc., which can be
entrained into green scraps during the fabrication process, a
purification process is required to obtain the pure uranium
that meets the specification of materials for nuclear fuel.
In the conventional way to recover pure uranium from
uranium scraps, the scraps are first dissolved in a hot nitric
acid of high concentration. Then, the dissolved uranium
is extracted by a solvent extraction, and the uranium
is recovered as a precipitate of (NH4)2U2O7 or
UO4�2NH4NO3 [1, 2]. However, the conventional method
is not environmentally friendly or safe because of corrosion
problems, the generation of NOX gas, and a great deal of
secondary organic and aqueous wastes.
A carbonate-based process has been recently introduced,
where the uranium of spent nuclear fuel is selectively
K.-W. Kim (&) � J.-W. Lee � D.-Y. Chung � E.-H. Lee �K.-H. Kang � K.-W. Lee � K.-C. Song � G.-I. Park � J.-K. Moon
Korea Atomic Energy Research Institute, Daedeok-daero
989-111, Yuseong-gu, Daejeon 305-353, Republic of Korea
e-mail: [email protected]
M.-J. Yoo
KEPCO NF, 1047 Daedeok daero, Yuseong, Daejeon 305-353,
Republic of Korea
123
J Radioanal Nucl Chem (2012) 292:909–916
DOI 10.1007/s10967-011-1534-8
leached in the form of a uranyl peroxo–carbonato complex
ion of UO2(O2)x(CO3)y-z in a carbonate solution containing
H2O2, and the uranium is recovered as a precipitate of UO4
by acidifying the uranyl peroxo–carbonato complex ion
solution [3–8]. Most transient metal or their oxides are not
dissolved in the solution because of their very low solu-
bilities in the carbonate solution at high pH. Therefore,
uranium can be selectively dissolved and then recovered as
UO4 precipitate from impurity-contaminated uranium
scraps using an H2O2-carbonate solution, as shown in
Fig. 1. In the case of (U,Gd)O2 scrap, Gd is slightly co-
dissolved together with uranium at a level of less than
100 ppm in the H2O2-carbonate solution, and a small
portion of the dissolved Gd is entrained into the UO4
precipitate. The Gd-contaminated UO4 can be purified by
using a dissolution of the UO4 in about a 1 M HNO3
solution with heating and a re-precipitation of UO4 by
adding H2O2 into the dissolved uranium solution [9]. When
the Gd-contaminated UO4 is dissolved in nitric acid, the
dissolved UO2?2 ions can be precipitated again as UO4 by
adding excessive H2O2 in the uranyl solution, but the co-
dissolved Gd ions remain in the acid supernatant because
of its very high solubility under the acidic conditions, so
that pure UO4 can be obtained from the contaminated
(U,Gd)O2 scrap, as explained in Fig. 1 [9, 10].
In order to reuse the UO4 recovered from uranium scrap
as a nuclear fuel, UO2, the compactability and sinterability
of UO2 powder reclaimed from UO4 for fabrication of UO2
fuel pellet are necessary to be evaluated. Accordingly, in
this work, UO2 pellet fabrication characteristics was stud-
ied with the UO4 obtained from uranium scrap by using a
H2O2-carbonate solution, and a process to reclaim con-
taminated uranium scraps as UO2 fuel powder was
suggested.
Experimental
All the reagents used in this work were of reagent grade,
and they were dissolved, as received, in demineralized
water of 18.2 MX prepared by double distillation and one
ion exchange column (Milli-Q plus). The ADU-derived
UO2 (UO2 ex-ADU) scrap obtained from a Korea nuclear
fuel company was in powder form with a diameter of about
9 lm mixed with a few metallic oxide impurities of Al, Fe,
Ni, Cr, etc. The UO2 scrap was dissolved in 0.5 M Na2CO3
solutions with H2O2 at room temperature. The uranium was
selectively dissolved in the form of uranyl peroxo–carbo-
nato complex ions of UO2(O2)x(CO3)y2-2x-2y [2, 3, 9].
After the dissolution, the solution was sampled to analyze
the concentrations of U and other elements dissolved in the
solution. The uranium carbonate solution was acidified to
pH 3 by adding HNO3 in the solution to precipitate UO4
from the uranium solution. The UO4 separated from the
solution by centrifugation was washed three times with
deionized water and dewatered by drying at room tem-
perature under vacuum.
The size and shape of the obtained UO4 particles were
irregular because of the agglomeration of individual UO4
crystallites during the UO4 precipitation and because of the
formation of lumps by agglomerates coming together in the
dewatering steps. In order to evaluate the possibility of
fragmentation and pulverization of the lumped UO4 parti-
cles by the OREOX process [11–13], the reduction and
oxidation of the UO4 powder were sequentially performed
in the following steps: (1) dehydration in Ar gas with a
flow rate of 3 L/min at 200 �C for 1 h, (2) reduction in Ar–
4% H2 gas with a flow rate of 3 L/min at 700 �C for 5 h,
and (3) oxidation in air with a flow rate of 3 L/min at
500 �C for 5 h. For the physical pulverization of the
dewatered UO4 powder, a milling of the UO4 powder was
carried out for 6 h in a jar with a constant ball-to-powder
weight ratio (8:1), zirconia balls of 5 mm diameter in the
same size, and a fixed rotational velocity of 112 rpm. A
lubricant of 0.25 wt% zinc stearate powder was added and
mixed with the milled powder in Turbula mixer for 25 min
at 25 rpm. Green pellets were prepared by a compaction of
the UO2 powder at a pressure of 300 MPa using a single
acting hydraulic press. The green pellets were sintered at
1,700 �C for 6 h in an atmosphere of Ar–4% H2 gas with a
flow rate of 1 L/min.
The mean particle size and particle size distribution,
specific surface area, morphology of the powders, and
composition of uranium oxide powder were evaluated with
a sieve shaker (Retsch, AS200) or a laser particle size
analyzer (Malvern Co.), a BET method (Micrometrics
3000), a scanning electron microscope (SEM, Philips, XL-
30), and XRD (MAC Science, TX J-827), respectively. The
bulk and tap density of the powder were measured using
Pure UO4xH2O
Selective uranium dissolution
Impurities (Al, Fe, Ni, etc)
Contaminated uranium scraps UO2, (U,Gd)O2
UO2(O2)x(CO3)yz-
UO4 precipitation at pH 1-3
Re-dissolution of UO4 with heating
HNO3CO2
Re-precipitation of UO4
Gd
Na2CO3 H2O2
Optional for (U,Gd)O2 scrap
HNO3
H2O2
Fig. 1 Recovery of UO4 from uranium fuel scrap with metallic
impurity by using carbonate solution with hydrogen peroxide
910 K.-W. Kim et al.
123
ASTM B329 and B527, respectively. The density of the
sintered pellet was determined using a water immersion
method (Archimedes principle). The concentrations of
uranium and impurity elements of Ni, Fe, Al, Cr, etc. in the
multi-component solution were analyzed by ICP-AES or
QMS (Inductively Coupled Plasma—Atomic Emission
Spectrometry or Quadruple Mass Spectroscopy).
Results and discussion
To recycle uranium from contaminated UO2 scraps as
nuclear fuel, UO2 scraps need to be selectively dissolved in
an environmentally friendly and economical way and to be
easily recovered in the form of simple uranium compounds.
Most metal compounds, including transient elements, are
not dissolved in alkaline solutions because of their very
low solubilities. UO2 is dissolved in aqueous solutions in
the form of uranyl ion, UO22? of ?6 oxidation state.
UO22? itself has a low solubility in alkaline solutions. In
carbonate solutions, however, UO22? ion complexes with
carbonate ion to form the uranyl tri-carbonato complex ion
of UO2(CO3)3-4, which has a high solubility, as shown in
Eq. 1. This forms the complicated uranyl peroxo–carbo-
nato complex ion, UO2(O2)x(CO3)y2-2x-2y in carbonate
solutions containing H2O2 with a further higher solubility,
as shown in Eq. 2, where H2O2 plays important roles as a
source supplying a ligand to form uranyl peroxo complex
ions and an oxidant for the dissolution of UO2 [3, 4, 6–9].
UO2 þ 1=2O2 þ 3CO2�3 þ H2O ¼ UO2ðCO3Þ�4
3 þ 2OH�
ð1Þ
UO2 þ xCO2�3 þ yH2O2 þ 2yOH�
¼ ½UO2ðO2ÞyðCO3Þx�2�2x�2y þ 2yH2Oþ 2e� ð2Þ
where y = 0, 1, 2, and x/y = 1/2, 2/1, 3/0.
The carbonate species of H2CO3, HCO3-, and CO3
2-
are interchangeable depending on the solution pH. Car-
bonic acid is easily changed into CO2 to be released out of
the solution because of its instability. When a uranyl per-
oxo–carbonato complex ion is acidified, UO4 precipitation
occurs with the carbonate species of the uranyl peroxo–
carbonato complex ions being converted into CO2 gas, as
shown in Eq. 3 [3, 14].
UO2ðO2ÞxðCO3Þ2�2x�2yy þ mHþ þ 2yH2O
! UO2ðO2Þ � 4H2Oþ yH2CO3ðCO2 "Þ ð3Þ
where m = 4, 6, 8 at y = 0, 1, 2, and x/y = 1/2, 2/1, 3/0.
In order to fast dissolve UO2 scarp in a carbonate
solution, UO2 scrap has to be pulverized for an increase in
dissolution surface area. Pellet or lumped UO2 can be
easily converted to fine U3O8 or UO2 powders by using
oxidation and reduction steps of the OREOX process
[11–13]. Figure 2 shows the uranium concentration chan-
ges with time during the dissolution of U3O8 and UO2 fine
powders in a diameter of about *9.5 lm in 0.5 M car-
bonate solutions with several H2O2 concentrations. The
finally dissolved uranium concentrations and their disso-
lution rate increased with the initial concentration of H2O2
in the carbonate solution in both cases of UO2 and U3O8.
The UO2 dissolution was very fast and was completed in a
few minutes. The U3O8 dissolution was retarded and its
rate was slower than that of UO2. The reason for the slower
dissolution of U3O8 than UO2 is considered to be because
U3O8 is thermodynamically more stable than UO2 [15]. In
this carbonate condition, dissolved concentrations of Fe,
Al, and Ni were less than 1 ppm, which was the detection
limit of the elements of ICP-AES we used, regardless of the
concentration of H2O2, although these results are not
present in this paper. The H2O2 in the solution acts as an
oxidant to oxidize uranium valances of ?4 in UO2 and
?5.3 in U3O8 to ?6 oxidation state, and as a source sup-
plying a ligand of O2-2 to form uranyl peroxo–carbonato
complex ions in carbonate solution. H2O2 self-decomposes
under alkaline condition of carbonate solution [9, 16].
Therefore, if the concentration of H2O2 in the solution is
not sufficient, a large amount of UO2 cannot be dissolved
in the solution. When a sufficient H2O2 is used, the con-
centration of UO2 dissolved in the carbonate solution
increases.
Figure 3 shows the changes of uranium concentrations
in supernatant, when the uranyl peroxo–carbonato complex
solutions with different initial uranium concentrations,
which were prepared by dissolving UO2 scraps in the
carbonate solution with H2O2, were acidified with HNO3.
0 5 10 15 20 25 300
20
40
60
80
100Dissolved UO
2 or U
3O
8 : 100 g/L
Ura
niu
m c
on
cen
trat
ion
(g
/L)
Dissolution time (hr)
in 0.5 M Na2CO
3 with H
2O
2
UO2
1.0 M
2.0 M
4.0 M
U3O
8
H2O
2
Fig. 2 Changes of the uranium concentrations with time during
dissolution of UO2 and U3O8 powders in 0.5 M Na2CO3 solutions
with several H2O2 concentrations
Preparation of uranium oxide powder for nuclear fuel pellet fabrication 911
123
The uranium concentration rapidly dropped from about pH
6.5 and became a few ppm below pH 4, showing the same
behaviors regardless of the initial uranium concentration.
At that time, the carbonate species of uranyl peroxo–car-
bonato complex ions were converted into CO2 gas to be
released out of the solution and uranium precipitation
occurred according to Eq. 3. Figure 4 shows the XRD
result of the precipitate after washing the uranium precip-
itate with distilled water and drying at 50 �C. It was
UO4�4H2O. The solubility of the uranium peroxide is very
low, in the order of 10-3 to 10-5 M [17–21], and thus the
uranium in the carbonate solution can be almost completely
recovered as a UO4. Figure 5 shows the SEM photography
of the UO4 precipitate. Somewhat longish individual par-
ticles in diverse sizes of less than 1 lm are agglomerated,
and the agglomerated particles are irregular and various in
size. Figure 6 shows the particle size distribution of
agglomerated UO4 precipitate after drying the UO4 pre-
cipitate in Fig. 3 as received without any milling. The dried
agglomerated UO4 particles are very big with an average
size of about 500 lm.
In order to reduce the obtained agglomerated UO4 to
UO2 and to pulverize the lumped UO4 particles into sub-
micron UO2 particles for good sinterability as UO2 fuel
powder at the same time, the OREOX treatment with the
repetition of oxidation and reduction was tried. In the
OREOX process, the average particle size of the fuel
powder decreases due to the breakage of particles resulting
from a volume expansion caused by a transformation of
UO2 (TD (theoretical density): 10.96 g/cm3) into a U3O8
phase (TD: 8.4 g/cm3) during the oxidation steps, whereas
it increases due to the bonding between the particles during
the reduction step [11–13]. Before applying OREOX
1 2 3 4 5 6 7 8 9 100
20
40
60
80
Ura
niu
m c
on
cen
trat
ion
(g
/L)
pH
adding HNO3 in UO
2(O
2)
x(CO
3)y
-z
U : 77.7 g/L
U : 58.4 g/L
U : 38.2 g/L
Fig. 3 Change of the uranium concentration in the solution with pH
during acidification of uranyl peroxo–carbonato complex solution
10 20 30 40 50 60 70
Inte
nsi
ty
2θ
UO4 4H
2O
Fig. 4 XRD spectra of the UO4 precipitate dried at 50 �C
Fig. 5 SEM photograph of the UO4 precipitate
> 1000 > 500 > 250 > 125 > 75 > 45 Pan0
5
10
15
20
25
Fre
qu
ency
(w
t %
)
Particle size (μm)
Fig. 6 Particle size distribution of agglomerated UO4 particles
912 K.-W. Kim et al.
123
treatment, the UO4�4H2O powder was dehydrated at
150 �C for approximately 1 h under argon inert atmosphere
to volatilize the water of UO4�4H2O crystallization. After
that, the obtained UO4 particles were reduced to UO2 under
Ar–4% H2 atmosphere at 700 �C for 5 h, removing the
peroxo component of UO4, and were then oxidized again
under air at 500 �C for 5 h. Figure 7 shows the SEM
photographs after treatments of UO4 particles with reduc-
tion and oxidation of the OREOX process. The uranium
oxides after the reduction and oxidation of UO4 particles
were confirmed to be UO2 and U3O8, respectively, by
XRD, although their results are not present in this paper.
The particle shapes themselves in Fig. 7 did not have any
apparent change and any breakage in individual particles of
the agglomerated U3O8 particles after the oxidation fol-
lowing the reduction, compared with Fig. 5 of the initial
UO4 precipitate. The reason for this is considered to be
because the stress of volume change due to phase trans-
formation by the reduction–oxidation was absorbed in the
submicron sized-individual particles of the uranium oxide.
It is reported that submicron-sized particles of UO2 appear
to be sufficiently small to withstand expansion without
cracking, because they normally generate only one U3O8
nucleus per prior UO2 grain, or because the tensile strains
are proportional to the grain size [22]. The size of the
agglomerated UO4 particles are big, but the individual UO4
particles of the agglomerated UO4 particles are fine in the
sub-micrometer level, as seen in Fig. 5, so that no spalling
and cracking occured in the U3O8 particles prepared after
the reduction of the UO4 precipitate.
Accordingly, to make UO2 fuel powder with good fuel
characteristics of compactability and sinterability from the
agglomerated UO4 powder, the agglomerated UO4 powder
is required to be pulverized by physical milling instead of
the OREOX method. The agglomerated UO4 after
dehydration was first reduced to UO2 state, then pulverized
to UO2 fine powder by milling. The usual reduction of
uranium oxide is performed at 600–800 �C under hydrogen
reducing atmosphere. If the treatment is performed at a
temperature of less than 600 �C, the reduction reaction rate
is low, and if the temperature exceeds 800 �C, a UO2
powder with low compactability and sinterability is pro-
duced due to the aggregation of powder particles. Figure 8
shows the XRD result after reduction of agglomerated UO4
powder at 700 �C under Ar–4% H2 atmosphere. The UO4
was confirmed to be clearly converted into UO2. The UO2
powder after the reduction is usually passivated in 1–3%
O2–Ar atmosphere at 75–85 �C to form a protective oxide
film of UO2?x (0 \ x B 0.17) on the powder. In this work,
the passivation was carried out in 2% O2–Ar atmosphere at
80 �C to get UO2?x (x = 0.08). After that, the UO2?x
powder with a protective oxide film was ball-milled for
6 h. Figures 9 and 10 show the SEM photograph and
particle size distribution after milling the agglomerated
UO2 powder, respectively. The agglomerated lumps or
crystalline shape of individual UO4 or UO2 particles shown
in Fig. 5 and Fig. 7 disappeared and were fragmented in
fine powder form. The milled UO2 powder had an average
particle size of 0.54 lm, specific surface area of 3.1 m2/g,
bulk density of 1.77 g/cm3, and tap density of 2.74 g/cm3.
The x in UO2?x increased from 0.08 to 0.14 by oxidation
during milling. To evaluate the sinterability of the finally
prepared UO2?x powder for nuclear fuel pellet, a green
pellet was fabricated by a compaction of the prepared UO2
powder at a pressure of 300 MPa. The green pellet had a
density of 5.63 g/m3 (51.4% TD). Then it was finally sin-
tered at 1,700 �C for 6 h under hydrogen atmosphere. The
sintered pellet had a density of 10.42 g/m3 (95.1% TD)
which satisfied the density specification of nuclear fuel
pellet. The sintered density suitable for utilization as a fuel
Fig. 7 SEM photographs after treatment of UO4 particles with reduction and oxidation of OREOX process
Preparation of uranium oxide powder for nuclear fuel pellet fabrication 913
123
in nuclear reactor is usually above 95% of the theoretical
density. The density of sintered pellet can be further
increased by using higher compaction pressure, or by using
higher sintering temperature and longer sintering time.
Since amount of the uranium oxide scraps generated during
nuclear fuel fabrication process is not much, it is desirable
to mix reclaimed UO2 powder and virgin UO2 powder to
improve the sintering characteristics of the UO2 powder
reclaimed from uranium oxide scraps via UO4 precipita-
tion, rather than directly recycling the reclaimed UO2
powder for the nuclear fuel pellet fabrication, which can
reduce the sintering temperature and time. Also, using the
mixture of the reclaimed UO2 powder and virgin UO2
powder can make UO2 pellet with higher sintered density,
which can exclude the possibility of fabrication of nuclear
fuel pellet below the required specification, because the
density of the sintered pellet fabricated with the reclaimed
UO2 powder, 95.1% TD, is not high enough, even though it
meets the requirement of exceeding 95% TD. Accordingly,
a series of experiments to evaluate the change of sintered
densities of pellets fabricated with mixtures of reclaimed
UO2 powder and virgin UO2 powder, where the amount of
virgin UO2 was varied by 10–50 wt%, were carried out at
the same condition as one used for the fabrication of sin-
tered pellet with only reclaimed UO2 powder, and their
results are given in Table 1. The specific surface area of
UO2 powder mixture and final density of sintered
pellet almost linearly increased with the mixing ratio of
virgin UO2 powder to the reclaimed UO2 powder. The
density of sintered pellet increased from 10.47 to 10.62 g/
cm3 when the ratio of added virgin UO2 powder was
changed from 10 to 50%, which was in the range of
95.5–96.9% TD. Therefore, it was confirmed that the
density specification of nuclear fuel pellet using the
reclaimed UO2 powder and virgin UO2 powder could be
fully satisfied. Based on all the above results, a process to
10 20 30 40 50 60 70 80
Inte
nsi
ty
2θ
UO2
Fig. 8 XRD spectra after reduction of agglomerated UO4 powder
Fig. 9 SEM photographs after milling UO2 powder obtained by
reduction of the agglomerated UO4
0 2 4 6 8 100
1
2
3
4
5
6
Fre
qu
ency
(%
)
Particle Size Distribution (μm)
Fig. 10 Particle size distribution of UO2 powder after milling
Table 1 Specific surface areas and sintered densities of pellets pre-
pared with mixtures of reclaimed and virgin UO2 powders
Mixing ratio (wt%) Specific
surface
area (m2/g)
Sintered
density
(g/cm3)
Theoretical
density (%)Reclaimed
UO2
Virgin
UO2
Run 1 100 0 3.10 10.42 95.1
Run 2 90 10 3.32 10.47 95.5
Run 3 80 20 3.54 10.53 96.1
Run 4 75 25 3.65 10.55 96.3
Run 5 70 30 3.75 10.57 96.4
Run 6 60 40 3.97 10.60 96.7
Run 7 50 50 4.19 10.62 96.9
914 K.-W. Kim et al.
123
fabricate UO2 fuel with uranium recovered from uranium
oxide scrap in an environmentally friendly and economical
way using a carbonate solution with hydrogen peroxide can
be finally suggested on the basis of Fig. 1, as shown in
Fig. 11.
Conclusions
The uranium of metallic oxide-contaminated uranium oxide
scraps generated from nuclear fuel fabrication processes
was selectively dissolved in the form of a uranyl peroxo–
carbonato complex in a carbonate solution with H2O2, while
the metallic impurity was undissolved. The uranium car-
bonate complex ion could be recovered as UO4 crystalli-
zation by acidifying the uranium carbonate complex
solution at pH 2–4. Because the individual particles of UO4
precipitates were in the submicron size, the agglomerated
UO4 powder could not be pulverized into fine UO2 powder
for nuclear fuel pellet fabrication by the OREOX process.
The UO2 powder reclaimed from uranium oxide scraps via
UO4 precipitation could meet the UO2 powder specification
for UO2 fuel pellet fabrication by a series of steps including
dehydration of UO4 precipitate, reduction, and milling. The
UO2 powder mixture of the reclaimed UO2 powder and
virgin UO2 powder improved the sinterability of the
reclaimed UO2 powder for fuel pellet fabrication.
Acknowledgments This work was supported by the Ministry of
Education, Science, and Technology (MEST) of the Republic of
Korea under the nuclear R&D Project.
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Dehydration at 150oC in Ar
Reduction at 700oCin Ar-4%H2
Passivation at 80oC in 2% O2-Ar
Milling
Homogenization
Adding virgin UO2
Pelletizing at 300 MPa
Sintering at 1700oC in Ar
Pure UO4 xH2O
Selective uranium dissolution
Contaminated uranium scraps UO2, (U,Gd)O2
UO4 precipitation at pH 1-3
Re-dissolution of UO4 in acid with
heating
Re-precipitation of UO4 with H2O2
Na2CO3 H2O2
Optional for (U,Gd)O2 scrap
Fig. 11 Flow diagram of a
process to fabricate UO2 fuel
with uranium recovered from
uranium oxide scrap
Preparation of uranium oxide powder for nuclear fuel pellet fabrication 915
123
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