8
Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps 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 Ó Akade ´miai Kiado ´, Budapest, Hungary 2011 Abstract This work studied a way to reclaim uranium from contaminated UO 2 oxide scraps as a sinterable UO 2 powder for UO 2 fuel pellet fabrication, which included a dissolution of the uranium oxide scraps in a carbonate solution with hydrogen peroxide and a UO 4 precipitation step. Dissolution characteristics of reduced and oxidized uranium oxides were evaluated in a carbonate solution with hydrogen peroxide, and the UO 4 precipitation were con- firmed by acidification of uranyl peroxo–carbonate com- plex solution. An agglomerated UO 4 powder obtained by the dissolution and precipitation of uranium in the car- bonate solution could not be pulverized into fine UO 2 powder by the OREOX process, because of submicron- sized individual UO 4 particles forming the agglomerated UO 4 precipitate. The UO 2 powder prepared from the UO 4 precipitate could meet the UO 2 powder specifications for UO 2 fuel pellet fabrication by a series of steps such as dehydration of UO 4 precipitate, reduction, and milling. The sinterability of the reclaimed UO 2 powder for fuel pellet fabrication was improved by adding virgin UO 2 powder in the reclaimed UO 2 powder. A process to reclaim the con- taminated uranium scraps as UO 2 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 UO 2 and (U,Gd)O 2 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)O 2 scraps or from dirty UO 2 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 (NH 4 ) 2 U 2 O 7 or UO 4 2NH 4 NO 3 [1, 2]. However, the conventional method is not environmentally friendly or safe because of corrosion problems, the generation of NO X 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

Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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Page 1: Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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

Page 2: Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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

Page 3: Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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

Page 4: Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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

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

Page 5: Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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

Page 6: Preparation of uranium oxide powder for nuclear fuel pellet fabrication with uranium peroxide recovered from uranium oxide scraps by using a carbonate–hydrogen peroxide solution

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

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.

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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

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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

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