Post on 10-May-2018
Advanced Hydrometallurgical Separation of Actinides and Rare Metals
in Nuclear Fuel Cycle
Masaki OZAWA1*,2, Shinichi SUZUKI1 and Kenji TAKESHITA2 1Japan Atomic Energy Agency, 2-4 Shirane, Shirakata Tokai-mura 319-1195, Japan,
2Tokyo Institute of Technology, O-okayama, Meguro, Tokyo, 152-8550, Japan
(Received March 16, 2010; Accepted March 30, 2010)
A hydrometallurgical separation technologies by novel solvent extraction (SX), ion exchange
chromatography (IXC) and electrolytic extraction techniques are reviewed as separation tools
for light PGM (Ru, Rh, Pd), Tc and f-elements in high level liquid wastes of the nuclear fuel
cycle. The SX process using N,N-dialkylamide can isolate U(VI) from fission products
without Pu(IV) valence control, and extractants with soft-hard hybrid donors (PTA and PDA)
and those containing six soft donors (TPEN) show good separation of actinides (III) from
lanthanides (III). The IXC process utilizing a tertiary pyridine resin (TPR) provides a very
high degree of separation of the f-elements in spent nuclear fuel and the recovery of pure Am
and Cm products. The catalytic electrolytic extraction (CEE) process utilizing Pdadatom or
Rhadatom can effectively separate platinum group metals (PGM), Tc and Re by means of
controlled under potential deposition (UPD). Some of the basic work on the
hydrometallurgical separation of the elements of interest has been carried out through the
strategic Advanced (Adv.-) ORIENT Cycle research in Japan. The Adv.-ORIENT Cycle process
cannot only improve the radioactive waste problem, but can also provide useful rare metals to
leading industries as from this secondary resource.
1. Introduction
Resources of natural energy (oil, gas, 235U) and most of rare metals will run out within 200 years. In
particular at fiscal year 2004, the R (resource) / P (production) ratio (year) for oil was 41 years, 67 years for
natural gas, 192 years for coal for and 85 years for uranium. Despite the rather long R/P ratio of ca.150
years for the platinum group metals (PGM), the current price increases for Ru, Rh, and Pd in the market
have been significant, and it should be noted that the production of PGM is limited to mainly those two
countries namely South Africa (75 %) and Russia (17 %) in the year 2008[1]. On the other hand, the R/P
ratio for rare earths is not so limiting, but 93 % of rare earth production is monopolized by one country,
Solvent Extraction Research and Development, Japan, Vol. 17, 19 – 34 (2010) – Reviews –
- 19 -
China. In this context, nuclear fission is said to be able to counter such a natural energy crisis issue if 238U
(239Pu) can be utilized in fast breeder reactors (FBR) in future.
Fission reaction of 235U and 239Pu currently is creating more than 40 elements and 400 nuclides as
fission products (FP) in the spent fuel, while generating enormous amounts of energy, approximately two
million times greater than that from chemical reaction per gram of fuel. Among them, 31 elements are
categorized as rare metals, and particularly Zr, Mo, Ru, Pd, Cs, Ce, Nd are highly enriched in FBR spent
fuel. Because of their individual radiochemical properties, these should be recognized as not only the
radioactive wastes but a second source of nuclear rare metals (NRM). Separation and utilization
(stock-pile) technologies should be at once developed for the next generation, and hence in the nuclear fuel
cycle, a policy change such as a Copernican Revolution is necessarily. This paper will review the state of
the art of the hydrometallurgical, e.g., solvent extraction (SX), ion exchange chromatography (IXC) and
electrolytic extraction (EE), technologies for the separation and recovery of NRM as well as actinides
present in the radioactive wastes.
2. Nuclear Rare Metals, a New Product from a Copernican Revolution of the Nuclear Fuel Cycle
Typical yields for Pd, Ru, Rh (light PGMs) and Tc will reach to around 11kg, 13kg, 4kg and 3kg,
respectively per metric ton of the reference FBR spent fuel (150 GWd/t, cooled for 5 years). The quantity
of NRM is shown in Figure 1. Since
such yields are proportional to the
degree of burn-up, those in common
light water reactors (LWR) will be
aproximately one third of FBR. It is
notable that, Mo and some heavy
lanthanides (Ln) (Dy, Er, Yb) are
already non-radioactive and
non-exothermic on reprocessing
after 5 years cooling. Also, Nd and
La are no longer radioactive beyond
the natural one’s level. Furthermore,
after cooling for more than 50 years in a stock-pile, the specific radioactivity of Ru, In, some of Ln like Pr,
Gd and Tb will be less than 0.1 Bq/g. The quality (isotopic composition) of some of NRM is shown in
Figure 2. The radiochemical properties are summarized as follows [2],
(a) After 40 years in a FP stock-pile the radioactivity of Ru will decrease to be below the exemption level
(BSS*) of 106Ru. Its isotopic abundance will become to stable Ru (99Ru, 100Ru, 101Ru, 102Ru, 104Ru) and 106Pd only. *Note BSS; International Basic Safety Standards for Protection against Ionizing Radiation
and for the Safety of Radiation Sources, Safety Series No.115, IAEA, Vienna (1996).
10
100
1000
10000
100000
35 40 45 50 55 60 65
Atomic No.
Wei
ght (
g/t
HM
) .
Sr
Y
Zr Mo
Tc
Ru
Rh
Pd
Ag Cd Sn
Sb
Te
I
XeCs
BaLa
Ce
Pr
Nd
Pm
Sm
Eu Gd
InNb
RbKr
BrTbDy
10
100
1000
10000
100000
35 40 45 50 55 60 65
Atomic No.
Wei
ght (
g/t
HM
) .
Sr
Y
Zr Mo
Tc
Ru
Rh
Pd
Ag Cd Sn
Sb
Te
I
XeCs
BaLa
Ce
Pr
Nd
Pm
Sm
Eu Gd
InNb
RbKr
BrTbDy
Figure 1 Typical yields for Pd, Ru, Rh (light PGM) and Tc in the reference FBR spent fuel (150 GWd/t, cooled for 5 years)
- 20 -
(b) After 80 years in a FP stockpile the radioactivity of Rh will decrease to below the exemption level
(NRPB*) of 102Rh. Its isotopic abundance will become stable at 103Rh and 102Ru. *Note NRPB; National
Radiological Protection Board-R306 (1999).
(c)Only 107Pd only is radioactive (long-lived) in isotopic abundance in FP Pd. Its ratio is ca.16 wt %, and 107Ag will be gradually generated. The radio toxicity of FP Pd is very low, just ca. 30 times as high as 107Pd’s BSS level (105 Bq/g).
(d) 99Tc is the only major radioactive (long-lived) nuclide in isotopic abundance of FP Tc. Stable 99Ru will
be gradually generated.
From close investigation, a possible"exit strategy"can be drawn up for individual NRM with
regard to utilization. Namely, (i) Material/Chemical use; Ru, Rh, Pd, Mo, Ln (La, Nd, Dy, etc) and Tc. It
is particularly noted that the isotopic abundance of Mo in stable FP will be composed of mainly higher
order nuclides like 97Mo (22.1 wt %) and 98Mo (26.8 wt %). Such abundances might be advantageous for
the production of 99Mo and 99mTc. (ii) Radiochemical use; 137Cs (e.g., radiation source as an alternative to 60Co), 90Sr, 238Pu, 241Am and 242,244Cm, (iii) Additional nuclear fuel; 237Np, 241Am and Cm (as 238,240Pu, by
α decay of 242,244Cm), (iv) Sale of stable isotopes on the market; 99Ru, 102Ru, 103Rh, 106Pd and 107Ag. These
stable nuclides will be obtained after long-term stock-piling of FP Ru, Rh, Pd and Tc. 100Ru can be also
Figure 2 Time dependence of isotopic abundance of fission products, PMGs and Tc, after the separation (for a fast reactor, 150,000MWd/t, cooled for 5 years)
- 21 -
obtained as a transmutation product of 99Tc. The exit strategy for PGM will depend on the ability to cool
for several decades.
Prior to the industrial utilization of NRM, a radiochemically precise separation of them and the
actinides in the spent fuel is required. Such separation technologies should be integrated to be well
compatible with each other in the fuel cycle where the
reprocessing function must also be changed to meet
environment-friendly requirements.
3. Advanced Reprocessing
Spent nuclear fuel reprocessing recovers more than
95.5 % uranium and plutonium from irradiated nuclear fuels in
nuclear power plants. Solvent extraction technique is used in
current spent nuclear fuel reprocessing for the separation of
uranium and plutonium efficiently from the fission products.
This reprocessing process is called the PUREX process, and tri
butyl phosphate (TBP) used as the extractant.
Around 1950, new extractant design for reprocessing was
developed energetically, and TBP was developed through
methyl isobutyl ketone and di-butyl carbitol (Figure 3).
Uranium and plutonium, which are the separation targets in the
PUREX process, are strong acids as defined by the HSAB
principle. Furthermore, uranium and plutonium are hard donor
atoms, and thus uranium and plutonium have affinities for hard
oxygen elements. Therefore, ketones, ethers, and phosphates
containing oxygen atoms have been tried as an extractants for
the separation of uranium or plutonium.
For new extractant design for reprocessing, the
requirements for the extractant are as follows, 1) extraction
capacity for high-concentrated uranium, 2) high selectivity for
uranium and plutonium, 3) high solubility in non-polar solvents,
4) high stability towards hydrolysis and radiolysis, 5) small
influence of degradation products, 6) complete incineration, etc.
The difference in these requirements from non-radioactive solvent extraction is the stability against
radiation. In the FBR under present development, the degree of burn-up is much higher than for LWR fuels,
and thereby the demand for radiation stability is a very important factor. Moreover, the influence of
degradation products must be evaluated and these degradation products also need to be removed.
O
Methyl isobutyl ketone (MIBK)
OO
O
Di butyl carbitol
P
O
OOO
Tri butyl phosphate (TBP)
N
O
N,N-dihexyloctanamide (DOHA)
N
O
N,N-di(2-ethyl)hexyl-(2,2-dimethyl) propanamide (D2EHDMPA)
Figure 3 Structure of oxygen donor
ligands.
- 22 -
The development of an N,N-dialkylamide as a new, more highly efficient extractant than TBP began
in the 1960’s in the USA [3,4] and, up to the present, work is also reported from Italy [5,6], France [7-9] ,
Russia [10,11], and India [12,13]. Also in the Japan Atomic Energy Agency (JAEA), development of new
N,N-dialkyamides as alternative extractants to TBP began in
1992 [14,15]. N,N-dialkylamides have some advantages,
namely, their complete incinerability (salt-free property) and
high stability against hydrolysis and radiolysis. Their main
degradation products are carboxylic acids and secondary
amines which hardly affect the separation of U (VI) and Pu (IV)
from FP. Further, the synthesis of N,N-dialkylamides is
relatively easy, and various N,N-dialkylamides can be used for
extraction. The most important advantage of solvent extraction
technique using an extractant like TBP is the recovery of the
objective metal cations by back-extraction. Therefore, it is
difficult to develop a new extractant with moderate
complexation capability between the extractant and the metal
cations.
N,N-dialkylamides have been obtained by reacting an
alkyl carboxylic chloride with a secondary amine, according to
the reaction [1].
R1-COCl + HN R2R3 R1 CO NR2R3 + HCl [1]
R1, R2 and R3 in this reaction are alkyl radicals. Purification involves a distillation under low pressure (1
mm of Hg).
Typical extraction profiles for U (VI), Pu (IV), Np (IV, VI), and several types of simulated fission
products from nitric acid media by N,N-dioctyl hexanamide (DOHA) are shown in Figure 4. From these
results, U (VI), Pu (IV), Np (IV) and Np (VI) can be recovered by DOHA. These R1, R2, and R3
substituents played an important role with regard to the coordination properties for the metal cation and the
lipophilicity of the extractant. Furthermore, steric hindrance occurs with long alkyl chains and branched
alkyl chains, which can be exploited in the separation of U (VI). Figure 5 shows extraction profiles for U
(VI) and Pu (IV) which were obtained with the linear alkyl typed amide; DOHA and the branched alkyl
type; N,N-di-(2-ethyl)hexyl-(2,2-dimethyl)propanamide (D2EHDMPA). Bulky N,N-dialkylamides like
D2EHDMPA have selectivity only for U (VI). For this selectivity, branched N,N-dialkylamides are
promising new extractants for reprocessing in the transition period from LWR to FBR (L-F transition
periods). However, although D2EHDMPA was highly selective for U (VI), D2EHDMPA had a poor
extraction capacity for highly concentrated U (VI). Therefore, a new amide compound in which only R1
was branched was developed. The new N,N-dialkylamide maintains the selectivity for U (VI) and has a
10-4
10-3
10-2
10-1
100
101
102
103
0 2 4 6 8 10
U(VI)Pu(IV)Np(VI)Np(IV)PdRuRhZrTc
Dis
trib
utio
n ra
tio:
D
HNO3 (M)
Figure 4 Nitric acid dependency of
DM for 1.0 M (mol dm-3)
DOHA-dodecane.
- 23 -
high loading performance for highly concentrated U (VI). On
the other hand, a disadvantage of N,N-dialkylamides is the
production of a secondary amine as the degradation product
from hydrolysis and/or radiolysis. This secondary amine with a
long alkyl chain exists in the organic phase, and cannot be
removed by alkali scrubbing. Now, an evaluation of secondary
amines in the reprocessing and removal techniques is being
carried out.
In nuclear fuel reprocessing, the development of the
isolation technology for U (VI) is a research objective. As seen
from an industrial point of view, the next step in
N,N-dialkylamides development should be tests with real spent
fuels.
4. Separation of Minor Actinides
In the past three decades, partitioning and transmutation (P&T) of long-lived nuclides has been
studied world-wide with the objective of creating an environmentally friendly nuclear fuel cycle. In this
context, the TRUEX (TRansUranium EXtraction) solvent extraction process, using a bifunctional
extractant OØD[IB]CMPO (n-octyl(phenyl)- N,N-diisobutylcarbamoylmethylphosphine oxide) with TBP in
n-dodecane as the solvent, was successfully found to be a vital method for recycling trivalent actinides
(An(III)) from high level liquid wastes (HLW) of spent fuel reprocessing [16]. The separation of trivalent
actinides and lanthanides from HLW was investigated by not only CMPO but also malonamide in the
European framework projects [17]. These extractants are in the development stage for practical use. The
process flow-sheets using these extractants have been successfully tested. One of the technological issues
was inter-group separation of trivalent Ln (III)) from An(III) for efficient actinide recycling into the FBR.
As these are reviewed in other papers [18], the most topical two technologies for minor actinide (MA) / Ln
separation and americium (Am) / curium (Cm) separation by P&T are reviewed in this section.
4.1 Separation of Minor Actinides from Lanthanides
As mentioned above, tri-valent An (III) existing in spent nuclear fuels, have been treated as high
level wastes in the nuclear cycle. At present, transmutation of MA like An (III) by fast neutrons with an
Accelerator Driven System (ADS) and/or a FBR has been studied. These transmutation technologies use a
neutron source. Therefore, if a burnable poison like Ln is present in this target, the transmutation efficiency
of Am (III) and Cm (III) is decreased. In order to achieve a high transmutation efficiency of Am (III) and
Cm (III), separation of An (III) from Ln (III) is very important. On the other hand, the 5f-electrons on An
(III) are of a largely relativistic, itinerant nature, providing a degree of covalency, which makes the
behavior of An (III) ions slightly softer than Ln (III) ions. Consequently, most separations of An (III) from
10-4
10-3
10-2
10-1
100
101
102
10-2 10-1 100 101
DOHA-U(VI)DOHA-Pu(IV)D2EHDMPA-U(VI)D2EHDMPA-Pu(IV)
Dis
trib
utio
n ra
tio :
DM
HNO3 (M)
Figure 5 Nitric acid dependency of DU and DPu by 1.0 M DOHA-dodecaneand 1.0 M D2EHDMPA-dodecane.
- 24 -
Ln (III) were achieved with ligands using soft donors like sulfur (S) or nitrogen (N) atoms.
For this purpose, new MA/Ln separation ligands were developed in several countries. These ligand
structures are shown in Figure 6.
P
S
HS
N
N
N
N N
N
Bis(2,4,4-trimethylpentyl)di 2,4,6-tri(2-pyridyl)-1,3,5-triazines
-thiophosphinic acid (CYANEX 301) (TPTZ)
NN
NN N
N
N
N N
N
N
N N
N
N
2,6-bis-(5,6-di-n-propyl- 1,2,4-triazin 6,6’-bis(5,6-diphenyl-1,2,4-triazin-3-yl)-
-3-yl)pyridine (nPrBTP) 2,2’- bipyridine (C5-BTBP)
NN
R1
NR1
O O
NN
R1
O N
N,N’-dialkyl-N,N’-diphenyl pyridine N,N’-dialkyl-1,10-phenathroline
-2,6-dicarboxyamide (PDA) -2-carboxyamide (PTA)
N
NN
N
NN
R
R
R
R
N,N,N’,N’-tetrakis((5-alkyloxypyridin-2-yl)methyl)
-ethylenediamine(TPEN)
Figure 6 Structure of sulfur and nitrogen donor ligands.
- 25 -
The purified bis(2,4,4-trimethylpentyl) di-thio phosphinic acid (CYANEX 301) gave a high
separation factor ; SF [Am (III) / Eu (III) ] = 4900 [19,20]. Process flowsheets were developed and tested in
China with real HLW and the radiolysis of CYANEX 301 was also investigated [21]. In the EUROPART
project for international cooperation in Europe, many types of new N-donors have been synthesized and
tested for the separation of An (III) from Ln (III). Typical N-donors are 2-pyridinecarboxyamide
(picolinamides) [22], 2,4,6-tri- (2-pyridyl)-1,3,5-triazines (TPTZ) [23], bis-triazinyl-1,2,4-pyridines (BTP)
[24,25], and 6,6'-bis(5,6- dialkyl-1,2,4-triazin-3-yl)-2,2'-bipyridines (C5-BTBP) [26-28].
2,6-bis-(5,6-di-n-propyl-1,2,4- triazin-3-yl)pyridine (nPrBTP) exhibited good selectivity, with a SF[Am
(III) / Eu (III)] of > 100. This high SF was obtained from 1.0 M (mol dm-3) HNO3 acid solution with 0.04
M nPr-BTP in TPH / n-octanol (70/30 vol %). However degradation of nPr-BTP by radiolysis was very fast.
In the ACSEPT (Actinide reCycling by SEParation and Transmutation) project in EU (which runs from
2008 to 2012), the development of a new BTBP and a new N-donor is under way.
Recently, N,N’-dialkyl-N,N’-diphenylpyridine-2,6-dicarboxyamide (PDA) [29,30], N,N’-
dialkyl-1,10-phenathroline-2-carboxyamide (PTA) [31] and N,N,N’,N’-tetrakis(2-pyridylmethyl)ethylene-
diamine (TPEN) [32-34] were synthesized and investigated for MA / Ln separation. These two ligands,
PDA and PTA, have soft (N)-hard (O) hybrid donors. In particular PTA exhibited a SF [Am (III) / Eu (III)]
value of ca. 20 at an aqueous concentration of 1.0 M HNO3. However, although PDA and PTA had good
SF [An / Ln] values, the behavior of FPs with PDA and PTA was not clear. In Japan and in the ACSEPT
project in the EU, the above separation of minor actinides from lanthanides is progressing and has a strong
connection with national strategy (especially fuel fabrication).
TPEN is a hexadentate ligand with 6 nitrogen-donors and complexes with softer metal ions selectively.
Trivalent MA ions, such as Am (III) and Cm (III), which are softer than lanthanide ions, are extracted to
the organic phase containing TPEN selectively. A SF [Am (III) / Eu (III)] value of ca. 100 was observed in
the pH range > 4 [32,33]. From the viewpoint of practical use, some hydrophobic TPEN analogs wherein
long alkyloxy groups were introduced to four pyridine rings were synthesized and the effect of the
introduction of the alkyloxy groups on the separation of Am (III) and Eu (III) was tested. In an acidic
solution at pH 3, the extractability of TPEN was improved by the introduction of butoxy groups, and a SF
[Am (III) / Eu (III)] value of ca. 90 was observed at pH 3 [33].
In the meantime, the simultaneous recovering of all of the
actinides with one type of solvent, bifunctional organophosphorus
extractants dissolved in highly polar fluorinated diluents, was
studied. 0.4M OØD[iB]CMPO with 30% TBP dissolved in
metanitrobenzotrifluoride (Fluoropole-732) was found to achieve
simultaneous extraction of all of the f-elements from the dissolver
solution of spent nuclear fuel without splitting a third phases [35].
This new solvent system was named the ORGA process Figure 7 Tertiary Pyridine Resin (TPR)
- 26 -
(*abbreviation of Organophosphoryl-fluoropole solvent for Recovery of Group of Actinides) [36].
4.2 Mutual Separation of Americium and Curium
The properties of a tertiary pyridine resin (TPR), as shown in Figure 7, with a nitrogen atom in its
six-membered ring, are characterized by simultaneously having two functions as a weakly basic
anion-exchanger and a soft donor ligand. The TPR is synthesized by co-polymerization of 4-vinylpyridine
and m/p-divinylbenzene on a high porous silica carrier (φ 60μm) at room temperature. The total exchange
capacity is ca. 5 meq / g (dry) in the Cl- form as an actual measurement value [37]. The TPR was hardly
damaged by up to 3MGy irradiation from a 60Co source, with an ensuing particularly high resistance to
radiation (~10MGy) in HCl media. An original idea by T.Suzuki, et al. [38] of combining the use of
conc.HCl and conc. HNO3 media for inter- and intra-group
separation of the f-elements was based on such facts, as
shown in Figure 8, as the recognition of softer ions like Am
(III) and Cm (III) (5f-elements) was particularly improved
(e.g., higher distribution ratios) compared to Ln (III) (4f
-elements) in HCl media, and, in HNO3 media, the
distribution ratios would depend not on the difference of the
softness between 4f- and 5f-elements but just on the
difference in ionic radius. The presence of a high conc. of
methanol in acidic media, improved the separation of
intra-group f-elements by enhancing the complexation of
the dehydrated f-element ions with anions [39]. Using these
facts, a HCl / HNO3 hybrid scheme was demonstrated in the
reprocessing of a trace amount of MOX spent fuel irradiated
(143.8 GWd/t) in the fast experimental reactor “Joyo” [38].
Individual recoveries of pure Am (III) product ((106Ru, 125Sb, 137Cs, 144Ce, 155Eu, 243Cm) / 241Am: < 10ppm) and Cm (III)
product (241Am / 243Cm: 0.78 %) were attained, respectively.
These separations will satisfy any separation degrees set up
for MA product such as Ln / MAproduct < 5 %mass.
5. Separation of Nuclear Rare Metals
In the PUREX process, SX technology was historically
been developed for the separation of An (U, Pu) from FP
while the recovery of valuable FP has been an issue in its
nuclear application. Furthermore, since their valences are
101
102
0.09 0.1
101
102
Dis
trib
utio
n co
effic
inet
, Kd
Ionic radii / nm
Am
Cm
LaCePrNdSmEuGdTb
DyHo
YEr
TmLu
Yb
LaCePr
Nd
AmCm
SmEu
GdYLu
conc.HNO3: MeOH=7:3
conc.HCl: MeOH=7:3
Figure 8 Distribution Ratios of f-elements for the Tertiary Pyridine Resin in HCl and HNO3 media (TPR)
0 2 4 6 8 10100
102
104
106
[HCl] / mol dm-3
Dis
trib
utio
n co
effic
ient
Pd(II)
Rh(III)
Ru(III)
Tc(VII)
Figure 9 Distribution Coefficients of PGM and Tc on the TPR in HCl media.
- 27 -
widely distributed from I to VII, group separation as for the f-elements seems to be impossible in the
PUREX nitric acid system. Therefore non-SX methods, like IXC, solid adsorption and EE, have been
studied for their individual separation.
5.1 Ion Exchange Chromatography of PGM and Tc
The acid dependency of the distribution coefficients of PGM and Tc on the TPR is shown in Figure 9.
The distribution coefficients were evaluated by the batch experiments carried out at room temperature. The
data suggest that all light PGM and TcO4- are strongly adsorbed by the TPR over a wide HCl concentration
range. If no adsorption of actinides (U (VI), Pu (IV), Am (III)) occurs, separation of Pd2+ and TcO4- in
highly dilute HCl conditions will be possible. These predictions have been definitely supported by the
strong adsorption of 106Ru and 125Sb in the previous hot test results.
As Figure 9 shows, it will be necessary to develop viable, salt-free, complexing reagents for efficient
elution of PGM and Tc from the resin as the next step.
On the separation of cesium and strontium, the selective separation of Cs+ by highly functional
adsorbents (silica gel loaded with ammonium molybdophosphate (AMP-SG(D)) and Sr2+ by hybrid
microcapsules containing D18C6 (D18C6-MC) are discussed in the other papers [40,41].
5.2 Catalytic Electrolytic Extraction of PGM and Tc
Catalytic electrolytic extraction (CEE) [42] can effectively separate Ru, Tc and Re (Tc simulator)
from either HNO3 or HCl solution under controlled potential deposition (UPD) with the formation of
insoluble metal (Ru, Rh, Pd) / oxide (ReO3, TcO2) solid solutions in acidic media. In the case of Re oxide,
an island type deposit was sometimes observed. In the UPD concept as shown in Figure 10, Pd or Rh may
act as a promoter at the surface of the electrode (i.e., Pdadatom, Rhadatom) and a mediator in the bulk solution
(i.e., redox ion pair).
The adatom, is a
contraction of
“adsorbed (single)
atom”, lying on
surfaces and surface
roughness, will act
as a deposition
catalyst. Pd and Rh
were easy to deposit
from the both media.
Characteristics of the
deposits are also
described in the Figure 10 Concept on Mediator and Promoter in the Under Potential Deposition
- 28 -
figure. In dilute hydrochloric acid (HCl) media in particular, such CEE occurred more readily. For
instance, the addition of Rh3+ significantly accelerated the deposition of Tc and Re as shown in Figure 11.
As typical deposition characteristics in the one to four NRM metal ion mixed solutions (corresponding to
FBR spent fuel composition), the observed deposition yields (average) were typically high in the following
order;
Pd, Rh ( > 99 %) > Re (91 %) > Ru (85 %) > Tc (69 %),
where the CEE was galvanostatically carried out with typically changing cathode current density like 2.5
mA / cm2 (1 hr) → 75 mA / cm2 (2 hr) → 100 mA / cm2 (4 hr) in 0.5M HCl at 50℃. At the initial stage,
setting a lower cathode
current density with
compulsory stirring was
necessary to form Pd
cores and the adatom
layer. This procedure was
essential to obtain higher
amounts of co-deposits
of Ru and Tc (Re) ions.
During the CEE,
the reduction process for
Rh was similar to Pd, i.e.,
through direct reduction
of Rh3+ to Rh metal.
However, reduction of Ru might proceed at least by two steps via following reactions,
RuNO3+ + e → Ru 2+ + NO E0 = 0.044V (vs. Ag/AgCl)
Ru2+ + 2e → Ru(s) E0 = 0.255 V (vs. Ag/AgCl)
A H-type electrochemical cell, where a nafion-117 cation-exchange membrane separated the catholyte and
the anolyte, was employed in both the CEE experiments and for hydrogen evolution by the NRM deposited
electrodes after changing the catholyte.
Recently, preliminary experiment on EE using a simulated high level liquid waste (S-HLW; 0.5M
HNO3, containing 26 metal elements corresponding to real HLW by LWR-reprocessing) has been carried
out [39]. The CEE deposits were apparently the same as shown in Figure 10 (right picture) with fine
coagulated spherical particles as were observed in the case of the Pd2+ involved deposits. The reduction
ratio of Pd was as high as 89 % as expected, while other NRM as Ru, Re and Mo were limited to around
20 % and less than 10 % for Rh, Te and Se. Such lower deposition ratios were probably attributed to the
significant consumption of the electricity by the large amount of the corrosion product iron contained in
S-HLW through the reduction of Fe3+ to Fe2+. In addition, the PGM content itself was lower compared to
Figure 11 Acceleration effect on the deposition of Tc and Re by adding Rh3+ ;
Electrolysis Condition Electrodes: Smooth Pt , Cathode (2cm2) , Anode (8cm2) , RE:
Ag/AgCl, Catholyte: 0.5MHCl, Tcconc.: 50ppm, Reconc.: 200ppm, Temp.: 50℃
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the previous experiments with FBR-HLW, and neither Pd2+ nor Rh3+ was added in this experiment. For
higher deposition yields, the CEE utilizing UPD by Pdadatom or Rhadatom is highly recommended with prior
separation of actinides and the other elements with nobler redox potentials by IXC. The reduction behaviors
of Mo, etc, require to be investigated in details.
The CEE method is advantageous for radiochemical separation of NRM, because of minimal
irradiation with no simultaneous deposition of actinides, but also because they are recovered in the solid
state at an electrode surface. The sphere deposits were dense and mechanically more stable than dendritic
deposits, while showing electrochemically high catalytic reactivity on electrolytic hydrogen production
[43]. In particular, the catalytic reactivity of the quaternary, Pd-Ru-Rh-Re (3.5:4:1:1, corresponding to the
composition of FBR spent fuel) deposit on a Pt electrode for electrolytic production of H2 was the highest,
exceeding that of a smooth Pt electrode by ca. twice both in alkaline solution and artificial seawater. Such
higher reactivity will attribute to the higher numbers of adsorption sites on the surface of the deposit for
protons.
Ru has been confirmed as the dominant element responsible for high reactivity, while Pd behaved as
just a "starter" nucleus and a "binder" among the involved elements during the deposition. It is
noticeable that Tc seems to have a higher or the same catalytic ability as Re. The high reactivity of Tc
suggests that it may be a possible alternative to Re in the field of catalysts.
6. A New Back-end Strategy as Copernican Revolution
Aiming at simultaneous realization of the utilization of elements/nuclides and ultimate minimization
of radioactive wastes, a new fuel cycle concept, Adv.-ORIENT (Advanced Optimization by Recycling
Instructive Elements) Cycle [2, 41, 43, 44], is proposed under the following strategies as shown in Figure
12;
1/ Trinitarian research on separation, transmutation and utilization (S&T, U) of nuclides and elements,
based on FBR fuel cycle.
2/ Significant reduction of radioactive wastes and eventual ecological risks: Within a few hundred years,
achieve a radiotoxic inventory decrease to the level of natural U tons corresponding to one ton of
vitrified HLW.
3/ Cascade separation of all actinides, NRM, Cs and Sr by a multi-functional and compact reprocessing
process and plant.
4/ Challenge on isotope separation of long lived radio nuclide 135Cs from FP Cs for advanced
transmutation.
5/ Accept and separate natural radioactive materials (U, Th) to burn, on demand of the RE industry.
The most important policy change is that NRM shall not be just the waste constituents but be the main
product in the fuel cycle. Actinides will no longer be the products, but will just be the material burned in
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the reactors. To realize this concept at both scientific and industrial levels, several separation tactics are
proposed as follows;
1/ Higher purity of NRM for utilization, while lower decontaminated actinides are permitted for burning in
FBR. Separation factors of 90-99.9 % will be chosen for the individual impact of radio nuclides.
2/ Adopt soft hydrometallurgical separation processes with salt-free reagents to reduce the secondary
radioactive wastes. Ultimately, low greenhouse gas emission technology is required.
3/ A high degree of separation of all actinides into 3-4 groups, U, Pu/U/Np, Am and Cm, directly from the
spent fuel by an IXC method.
4/ As non-SX methods, CEE and adsorptive separation methods for NRM are chosen to alleviate the
radiation effect. Solid state will be preferable as the end product for utilization and/or stockpiling.
5/ HCl media is allowed in combination with HNO3 media to improve the separability.
6/ Identification of anti-corrosive materials in both conc. HCl and HNO3 are indispensable from an
industrialization point of view. Verification of thermo- and radio-chemical stability of the novel IX resin
is also required.
7/ Preliminary separation of actinides from the NRM is advantageous because Zr, Mo, Pd, 99Tc, 106Ru and 125Sb, etc will disturb the operation of both reprocessing and vitrification of HLW.
Figure 12 Adv.-ORIENT Cycle Concept
LLFP Removalby Inorganic
Ion Exchanger
Spent fuel
Main AnSeparation/ Purification
ProcessbyIXC
NRM Recovery by CEE
Am, CmMutual
Separation
90Sr -Zeolite
Cs,Se,Sn
Sr
Other FPs(Near Surface and/or Deep Repository)
Ln
Am, Cm Am
Pu (←Cm)
U, Pu, Np
Reuse of Filter
Pd, Ru, Rh,Tc, Sb, etc
Ru,Rh,Pd,Tc,etc
Tc
Ap
plicatio
n for C
atalysts fo
r H
ydro
gen G
ene
ration a
nd/orF
uel C
ell
Application for Heat (Sr battery)and/or Radiation Source
Application forHeat and/or Radiation Source
Application as long-lived TcbatteryCatalyst after Transmutation99Tc(n,β-)100Ru
Isotope SeparationElution
133Cs
137Cs
Pd
Pd
107Pd135Cs
(Transmutation)
MoApplication for99mTc Source
Isotope
Sep
aration
Suzuki/Ozawa
NRM Separationby IXC Filter
Electro-Catalyst
137Cs-Zeolite
StrategicMaterials
Cm
Adv.-ORIENT Cycle since 2006
Multi-functionalReprocessing
Advanced Optimization by Recycling Instructive Elements Cycle
Fast BreederReactor
Fuel Fab.
EX-Cycle
IN-Cycle
Natural RE
U,Th
- 31 -
The time to reduce the radio-toxicity Sv (Sievert) of 1 ton of vitrified HLLW below the level of equivalent
tons of natural raw uranium is one of the indexes for environmental impact. In the Adv.-ORIENT Cycle, by
putting the separation factors at 99.9 % for all actinides, 99 % for 137Cs, 90Sr and the other NRM, and 90 %
for Ln, such a period can dramatically be reduced to around 102 years.
7. Conclusion
The isotopic composition and radiochemical properties of nuclear rare metals have been reviewed.
Hydrometallurgical separation technologies using solvent extraction (SX), ion exchange chromatography
(IXC) and catalytic electrolytic extraction (CEE) techniques were developed as vital separation tools for
light PGM (Ru, Rh, Pd), Tc and f-elements present in high level liquid wastes of the nuclear fuel cycle. The
SX process using N,N-dialkylamide can isolate U(VI) from fission products without Pu(IV) valence
control, and extractants utilizing soft-hard hybrid donors (PDA and PTA) and those containing six soft
donor atoms (TPEN), which have been developed mainly in Japan, show good separation of An(III) from
Ln(III). The IXC process utilizing a tertiary pyridine resin (TPR) gives a high separation of the
f-elements in spent nuclear fuel and produce pure Am and Cm products. The CEE process utilizing Pdadatom
or Rhadatom can effectively separate PGM, Tc and Re by utilizing under potential deposition (UPD)
phenomena.
Some of the basic work on hydrometallurgical separation of concerned elements is being progressed
in the flame of strategic Advanced (Adv.-) ORIENT Cycle research program in Japan. The Adv.-ORIENT
Cycle process can not only improve the rad. waste problem, but also offer useful rare metals to leading
industries from this secondary resource.
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