LA-UR-97-181
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CONCEPTUAL DESIGN FOR SEPARATION OF
PLUTONIUM AND GALLIUM BY SOLVENT
EXTRACTION
Scott DeMuth
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LA-UR-97-181
Conceptual Design for
Separation of Plutonium and Gallium
by Solvent Extraction
Revised
April 4,1997
Scott F. DeMuth
Los Alamos National Laboratory
Technology & Assessment Division
Energy & Environmental Analysis Group
2
Conceptual Design for Separation of Plutonium and Gallium
by Solvent Extraction
Table of Contents
pg
1.0 INTRODUCTION....................................................................... 4
2.0 PROCESS FLOWSHEET.............................................................. 5
3.0 PROCESS SPECIFICS................................................................. 6
3.1 Oxidation of Metal...............................................................6
3.2 Dissolution of Oxide........................................................... 7
3.3 Solvent Extraction............................................................... 8
3.3.1 Distribution coefficients............................................. 8
3.3.2 Pu recovery and Ga decontamination..............................9
3.3.3 Solvent extraction equipment....................................... 15
3.4 Solvent Wash.................................................................... 16
3.5 Plutonium Product.............................................................. 17
3.6 Gallium Waste................................................................... 17
4.0 REFERENCES........................................................................... 19
List of Figures
pg
Figure 1. Solvent Extraction Block Diagram.............................................. 5
Figure 2. Solvent Extraction Process Flowsheet.......................................... 6
Figure 3. Electrolytic Dissolution of PuO2 ................................................. 7
Figure 4. Counter-current Multi-stage Solvent Extraction............................... 9
Figure 5. Distribution Coefficients for Nitric Acid/TBP from Reference 5........... 10
Figure 6. Extraction Conditions............................................................ 11
Figure 7. Stages Required for Extraction of Pu and Decontamination of Ga......... 13
Figure 8. Stripping Conditions............................................................. 14
Figure 9. Stages Required for Strip of Pu from Solvent................................ 15
Figure 10. 5.5-cm-Diameter Centrifugal Contactors..................................... 16
This report was prepared as an account of work sponsored by an agency of the United States Government.
Neither the United States Government nor any agency thereof, nor any of their employees, makes any
warranty, express or implied, or assumes any legal liability or responsibility for the accuracy,
completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that
its use would not infringe privately owned rights. Reference herein to any specific commercial product,
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imply its endorsement, recommendation, or favoring by the United States Government or any agency
thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the
United States Government or any agency thereof.
3
1.0 INTRODUCTION
A potential use for plutonium from decommissioned nuclear weapons is incorporation into
commercial UO2-based nuclear reactor fuel for power generation. This fuel, which would
be a combination of plutonium oxide with uranium oxide, is referred to as a mixed oxide
(MOX). In fabricating this fuel, it is important to maintain current commercial fuel purity
specifications. While introducing impurities from the weapons plutonium may or may not
have a detrimental affect on fuel performance, certifying the effect as insignificant could be
more costly than purification.
Weapons plutonium can have a significant amount of gallium present which will force the
MOX fuel out of purity specification limits, and could have a negative affect on fuel
performance. Consequently, numerous methods of purifying the plutonium prior to MOX
fabrication have been proposed. A solvent extraction process based on aqueous nitric acid
and organic tri-butyl phosphate (TBP) has been suggested as a fully developed method for
separating gallium from the plutonium. This document summarizes a Conceptual Design
proposed for separating plutonium from gallium by nitric acid/TBP based solvent
extraction. In addition, this process will also separate a significant amount of americium
from the plutonium. This conceptual design is based on existing separations data and offthe-shelf equipment.
An important aspect of this Conceptual Design is the deliberate lack of a scrub section,
which makes spent nuclear fuel reprocessing impractical with the existing equipment. In
addition, the inclusion of a partitioning section will permit the recycle of MOX scrap,
without the scrub section required for spent fuel reprocessing.
4
2.0 PROCESS FLOWSHEET
The principal unit operations required for separation of plutonium and gallium by most
solvent extraction processes are shown in Figure 1. The Feed Preparation and Solvent
Extraction operations are the most important for this study. The Pu-Product operation will
involve classical oxalate precipitation of liquid Pu-nitrate followed by calcination. The GaWaste operation will involve volume reduction and immobilization for permanent disposal.
It is expected a waste disposal facility will exist at the site chosen for the plutonium/gallium
separation. Because a significant amount of americium from plutonium decay will be
disposed of with the gallium, the gallium waste will likely become a transuranic (TRU)
waste requiring disposal in the Waste Isolation Pilot Plant (WIPP), or disposal in an
underground repository such as Yucca Mountain.
Pu
Product
Pits
Feed
Preparation
Solvent
Extraction
Solvent
Recycle
Ga
Waste
Figure 1. Solvent Extraction Block Diagram
The detailed process flowsheet which serves as the Conceptual Design for the plutonium
and gallium separation is shown in Figure 2. If a partition step is added to Figure 2,
between the extraction and strip banks, then MOX scrap recycle is possible. As noted in
the Introduction Section, without a scrub step, nuclear fuel reprocessing with this design is
not practical. Figure 2 will be used as the basis for determining material balances and cost
estimates for follow-on activities to this work.
5
Pu-waste
Puaqueous
waste
treatment
HNO3 /Na2CO3/NaOH
solvent
wash
30-wt% TBP/dodecane
solvent extraction
stripping
Pu-HNO 3
extraction
(Ga/Am/Pu)HNO3
Pu-HNO3
nitrate/oxide
conversion
dilute
HNO3
PuO2 solid
electrolytic
dissolution
concentrated
HNO3
PuO2 -solid
metal/oxide
conversion
waste
treatment
Ga/Am/Pu-waste
Pu-metal
Figure 2. Solvent Extraction Process Flowsheet
3.0 PROCESS SPECIFICS
3.1 Oxidation of Metal
In keeping with a fully developed solvent extraction process such as Purex, dissolution of
PuO2 is favored over direct dissolution of Pu metal. Dissolution of Pu-metal requires the
addition of very strong acids such as HF, which not only can affect extraction behavior but
also can be corrosive to the equipment. The formation of Pu-oxide from Pu-metal has been
practiced often in the past. One approach is to form a hydride powder from the bulk metal,
and then essentially execute a controlled burn of the hydride. A process similar to this, and
known as HYDOX (Reference 2), is being developed as part of the MOX program.
6
3.2 Dissolution of Oxide
In keeping with a Conceptual Design based upon existing data, it is desired to use nitric
acid alone for dissolution of the plutonium feed. Plutonium metal in most any form is not
readily dissolved in nitric acid alone. Plutonium oxide up to 25 wt % in solid solution with
uranium oxide, as is the case in spent fuel, is readily dissolved in nitric acid alone.
However, plutonium oxide as itself is not readily dissolved in nitric acid alone. Based
upon these facts, electrolytic dissolution of plutonium oxide has been selected for
preparation of the solvent extraction feed.
Electrolytic dissolution of actinide-oxides with essentially nitric acid alone, has been used
by the French since 1990 on an industrial scale (Reference 3). Essentially nitric acid alone
refers to the catalytic-type use of silver to facilitate the reduction of PuVI+ to PuIV+. Figure 3
shows the mechanisms involved in electrolytic dissolution, and outlined in Equations 1-4
(Reference 4). It is worth noting that (1) the silver does not become part of the dissolved
plutonium species and hence it’s catalytic-type behavior and (2) while PuIV+(NO3)4 is more
extractable than (PuVI+)O2(NO3)2, and will exist at a significant concentration following
dissolution, the PuVI+ species will be used as the basis for this study in order to be
conservative.
e-
HNO3 -> H+ + NO3AgNO3 -> Ag + + NO3Pu(NO 3)4 + 2H2O
PuO2(NO3)2 + 2H+
PuO22+(soluble) +2 Ag+
Ag+
PuO2(solid)
Ag2+
cathode - tantalum
anode - platinum
semi-permeable barrier
Figure 3. Electrolytic Dissolution of PuO2
7
Addition of Ag+ ion:
AgNO3 ___> Ag+ + NO3-
Equation 1
Electrolytic oxidation of Ag+:
Ag+ ___> Ag2+ + e-
Equation 2
Oxidation of solid PuO2 to soluble PuO22+ by reduction of Ag2+:
PuO2,solid + 2Ag2+ ___> PuO22+soluble + 2Ag+
Equation 3
PuVI+ nitrate complexation:
PuO22+ + 2HNO3 ___>PuO2(NO3)2 + 2H+
Equation 4
PuIV+ reduction to PuIV+, followed by nitrate complexation:
PuO22+ + 2e- + 4H+ + 4NO3- ___> Pu(NO3)4 + 2H2O
where
Equation 5
PuO22+ = [PuVI+(O2-)2]2+
Pu(NO3)4 = [PuIV+(NO3-)4]
The reduction of PuVI+ to PuIV+ can be further promoted by the addition of hydrogen
peroxide (H2O2).
3.3 Solvent Extraction
There are three primary steps required in the preparation of a solvent extraction conceptual
process design.
(1) Determine the extraction and stripping distribution coefficients for Pu and Ga.
(2) Determine the Pu recovery and Ga decontamination for a given process
configuration.
(3) Select the extraction equipment.
3.3.1 Distribution coefficients
Equations 6 through 8 represent in a general sense the complexation reactions required to
form the organic soluble plutonium or gallium complexant.
[PuIV+] + 4[NO3-] + 2[TBP] ___> [Pu(NO3)4.2TBP]
Equation 6
[(PuVI+)O2] + 2[NO3-] + 2[TBP] ___> [PuO2(NO3)2.2TBP]
Equation 7
[Ga3+] + 3[NO3-] + 2[TBP] ___> [Ga(NO3)3.2TBP]
Equation 8
8
Equation 9 is the equilibrium expression derived from Equation 8, which expresses the
dependency of gallium complex formation on TBP concentration. Equation 9 will be used
later in this text to estimate the gallium distribution coefficient for conditions of this study.
K = [Ga(NO3)3.2TBP]/{[Ga3+][NO3-]3[TBP]2}
Equation 9
3.3.2 Pu recovery and Ga decontamination
If both aqueous and organic feed conditions are known as shown in Figure 4, an exact
solution for the number of stages, versus plutonium extraction and gallium
decontamination, is possible given constant distribution coefficients. Based on these
assumptions, Equation 10 defines the recovery of plutonium and Equation 11 defines the
decontamination of gallium (Reference 1- Chapter 4, Section 6.2).
organic solvent
extract
Qo = Qa
Co0 = 0
Co n = ?
1
2
n
aqueous feed
Q a(l/time)
C aF(mole/l)
raffinate
Ca1 = ?
Figure 4. Counter-current Multi-stage Solvent Extraction
9
ρ = β[(βn-1)/(βn+1 -1)] + βn+1[(β-1)/(βn+1 -1)][Co0/(DCaF)]
Equation 10
where β = D(Qo/Qa)
fPu/Ga = (βPu/βGa)[(βPun -1)/ (βGan-1)][(βPun+1 -1)/(βGan+1-1)]
Equation 11
Distribution coefficients for this study are based on the work of Reference 5 and Reference
6 (Tables 13.49 & 13.51). Reference 5 was used as the basis gallium extraction (or
decontamination). Reference 6 was used as the basis for the plutonium extraction and
strip. Figure 5 shows the distribution coefficient data of Reference 5.
10+2
U IV
PuVI
Ru3+
LOG (kd )
Sr2+
Cs+
-4
10
Ga 3+
Kd = (concentration in organic) / ( concentration in aqueous)
aqueous phase = 3 N HNO 3
organic phase = 100% TBP
Figure 5. Distribution Coefficients for Nitric Acid/TBP from Reference 5
10
Common fission products along with plutonium and gallium, are shown in Figure 5 to
qualitatively indicate the ease of separating gallium from plutonium. It is species such as
ruthenium which dictate the need for a scrub section during processing with nitric acid and
tri-butyl phosphate. Scrubbing is used to further separate fission products which partially
extract with the uranium and plutonium during nuclear fuel reprocessing. For the
conditions of this study, a scrub section is not required due to the large difference between
the plutonium and gallium distribution coefficients. Hence, the proposed Conceptual
Design for this study is not capable of reprocessing spent nuclear fuels.
In order to solve Equations 10 and 11 for the extraction section of Figure 2, feed conditions
have been defined as those shown in Figure 6.
organic solvent
Q o = Qa
Co,Pu0 = 0
Co,TBP0 = 30 wt%
extract
Con = ?
1
2
raffinate
Ca1 = ?
n
aqueous feed
Q a = Qo
C a,PuF = 167 g-Pu/l
C a,GaF ~ 0.036 g-Ga/l
C a,NO3F = 3 M
Figure 6. Extraction Conditions
Equal aqueous and organic flowrates often produce optimum hydraulics in the contacting
equipment, and consequently provide a good starting point for a conceptual design.
Hydraulics affect the interfacial area and consequent mass transfer required for equilibrium
stages. It is assumed all equipment is designed for subcritical conditions independent of
plutonium concentration. The plutonium concentration exiting the electrolytic dissolover is
approximately 167 g-Pu/L (Reference 3). Classical nitric acid and TBP feed concentrations
for this process are 3 M HNO3 and 30-volume(v/o) % TBP. The maximum gallium
concentration in the feed is defined as 1 wt %.
11
The gallium distribution coefficient for 30 v/o % TBP, based on the values of 100% TBP
shown in Figure 5, was determined by rearranging Equation 9 to yield Equation 12.
CGa(NO3)3.2TBP = KGa(CGa3+)(CNO3-)3(CTBP)2
Equation 12
rearranging yields:
D(30 wt.% TBP) = (0.30)2D100%-TBP
DGa(3M HNO3, 30 % TBP) = (0.30)2(10-4) ~ 10-5
where normality (N) and molarity (M) are equal fro nitric acid.
As reported in Reference 6;
DPu(IV+)(3M HNO3, 30 % TBP) = 16
DPu(VI+)(3M HNO3, 30 % TBP) ~ 3.5
DPu(IV+)(0.1M HNO3, 30 % TBP) = 0.023
DPu(VI+)(0.1M HNO3, 30 % TBP) ~ 0.05
Figure 7 shows the results of estimating Pu extraction and Ga decontamination for 3 M
HNO3 and 30 v/o% TBP, based on Equations 10 and 11. The y-axis of Figure 7
represents the logarithm of g-Pu in gallium waste per MT-Pu feed, and ppm-Ga in the Pu
product. The gallium waste exits the extraction section, and the plutonium product exists
the strip section (see Figure 2). PuVI+ rather than PuIV+ has been used as the basis for
extraction and strip behavior in order to be conservative.
Example extraction section calculations follow:
ρ = β[(βn-1)/(βn+1 -1)] + βn+1[(β-1)/(βn+1 -1)][Co0/(DCaF)]
The Pu recovery for 10-extraction stages is:
ρPu = 3.5{[(3.5)10-1]/[(3.5)11-1]} ~ 0.9999974
where βPu = D(Qo/Qa) = 3.5(1) = 3.5
C o,Pu0= 0
the Pu loss per MT-Pu feed is;
(1-0.9999974)(1-MT)(1000 Kg/MT)(1000 g/Kg) ~ 3 g-Pu
The Ga passed to the Pu loaded solvent product for 1-extraction stage is:
ρGa = 10-5{(10-5-1)/[(10-5)2-1]} ~ 0.00001
where βGa = D(Qo/Qa) = 10-5(1) = 10-5
12
the Ga in Pu product per MT of Pu feed at 1-wt % Ga feed is;
(10-5)[0.01(1-MT)(1000 Kg/MT)(1000 g/Kg)] = 0.1 g-Ga.
LOG(g-Pu waste/MT-Pu feed) & (ppm-Ga in Pu product)
6
5
4
3
Pu
Ga
2
1
0
1
2
3
4
5
6
7
8
9
10
11
12
13
- 1
- 2
number of stages
Figure 7. Stages Required for Extraction of Pu and Decontamination of Ga
In order to solve Equations 10 and 11 for the stripping section of Figure 2, feed conditions
have been defined as those shown in Figure 8. As in the Extraction section, equal aqueous
and organic flowrates often produce optimum hydraulics in the contactors. Assuming
nearly complete Pu recovery in the extraction section, and equal flowrate and phase ratios
in the extraction and strip sections, the plutonium concentration in the solvent feed to the
strip section is identical to the aqueous feed to the extraction section, i.e. 167 g-Pu/L. The
aqueous nitric acid feed for the stripping section is 0.1 M. While the Pu distribution
coefficient decreases with decreasing nitric acid concentration, which is desirable for
stripping, some nitric acid is required in the strip section for adequate phase separation
during settling. It is assumed any Ga which extracts with the Pu in the extraction section,
will strip with the Pu in the strip section and exit with the aqueous Pu product.
13
Distribution coefficients for stripping are the inverse of that for extraction. Therefore, the
PuVI+ distribution coefficient for stripping in 0.1 M nitric acid and 30 vol % TBP is:
D(PuVI+) = 1/0.05 = 20
Based upon the strip section feed conditions shown in Figure 8, the Pu distribution
coefficient for stripping, and Equation 10 of this text, Figure 9 has been generated
according to the example calculations which follow.
aqueous strip
Q a = Qo
Ca,Pu0 = 0
Ca,NO30 = 0.1 M
purified Pu
Can = ?
1
2
solvent recycle
n
solvent feed from extraction
Q o = Qa
Co,PuF = 167 g-Pu/l
Co,TBPF = 30 wt%
Co1 = ?
Figure 8. Stripping Conditions
Example Strip section calculations follow:
ρ = β[(βn-1)/(βn+1 -1)] + βn+1[(β-1)/(βn+1 -1)][Co0/(DCaF)]
The Pu recovery for 4 strip stages is:
ρPu = 20(204-1)/(205-1) ~ 0.9999941
based upon reverse extraction
βPu = D(Qo/Qa) = (1)(20) = 20
C o,Pu0= 0
the Pu loss per MT-Pu feed is;
(1-0.9999941)(1-MT)(1000 Kg/MT)(1000 g/Kg) ~ 6 g-Pu
14
The y-axis of Figure 9 represents the logarithm of g-Pu waste in recycled solvent per MT
Pu feed. It is assumed the plutonium remaining in the solvent will be aqueous-based waste
following the solvent wash.
6
4
LOG(g-Pu waste/MT-Pu feed)
2
0
1
2
3
4
5
6
7
8
9
10
11
12
- 2
- 4
- 6
- 8
-10
number of stages
Figure 9. Stages Required for Strip of Pu from Solvent
3.3.3 Solvent extraction equipment
Centrifugal contactors have been selected as the contacting equipment for the solvent
extraction. This is due to their (1) small compact size, (2) criticality conscious design, (3)
demonstrated high reliability, and (4) rapid achievement of steady-state operation. The
centrifugal contactor size most appropriate for the process flows of this study is the 5.5cm-diameter shown in Figure 10. The 5.5-cm-diameter centrifugal contactor design is
highly evolved with respect to (1) fabrication, (2) hydraulic performance, and (3) masstransfer performance (Reference 7). The 5.5-cm-diameter centrifugal contactors shown in
Figure 10 are in a cascade configuration of 8-strip stages, 2-scrub stages, and 4-extract
stages from left-to-right. The back row of stages in Figure 10 are difficult to see as they
15
are diagonally off-set; however, their electrical connections are visible. The 5.5-cmdiameter maximum throughput is 0.1-MT of heavy metal per day or 3-L/min total,
whichever is limiting. The 5.5-cm-diameter centrifugal contactor geometry guarantees
subcritical operation independent of plutonium inventory (or concentration).
Figure 10. 5.5-cm-Diameter Centrifugal Contactors
3.4 Solvent Wash
There are two primary purposes for the inclusion of a solvent wash operation. The first is
minimization of plutonium which returns with the recycled solvent to the extraction section.
The second is removal of hydrolyzed TBP. The affect of plutonium in the recycled solvent
is increased plutonium losses in the gallium waste from the extraction section. The affect
of TBP hydrolysis is the formation and accumulation of dibutyl and monobutyl phosphoric
acid in the solvent. These acidic species in the solvent can form plutonium complexes
which will not strip from the solvent; and hence, accumulate in the solvent upon recycle.
16
A reasonable solvent wash design is based upon a caustic aqueous wash to remove dibutyl
and monbutyl phosphoric acid, followed by very dilute nitric acid wash to further strip the
Pu and rebalance the solvent pH. The caustic wash can consist of Na2CO3 and/or NaOH.
Centrifugal contactors are an excellent choice for solvent wash equipment
3.5 Plutonium Product
Plutonium nitrate conversion to PuO2 can be achieved by several means (Reference 1 Chapter 9, Section 4.7). The most simple, which requires a very pure stream of plutonium
nitrate, is direct thermal decomposition more commonly known as calcination. Another
conversion process for less pure plutonium is precipitation as either a peroxide or oxalate,
followed by calcination. While the direct thermal decomposition (calcination) is more
simple, the precipitation followed by calcination has been used more frequently on an
industrial scale.
3.6 Gallium Waste
The gallium waste stream will consist primarily of gallium nitrate and americium nitrate,
with a very small amount of plutonium nitrate. Other elements may present in the original
Pit material at concentrations assumed to be less than gallium; and consequently, may also
be present in the waste stream. For this conceptual design it is assumed the waste activity
is due primarily to americium which is the decay product of plutonium, and a Hazardous
waste is not created.
Since americium predominantly exists as a plus-three cation in 3 M nitric acid, it is assumed
to be inextractable as gallium. Therefore, as a first approximation for this Conceptual
Design, the waste activity will be based on all of the americium present in the original Pit
material, which is approximately 200 ppm maximum. The plutonium activity will be
neglected at this time since the amount will depend on the cascade configuration, which can
be selected later based on overall cost considerations.
Example activity calculations:
λ Pu-239 = ln(2)/t1/2
= ln(2)/[24,400 yr(365day/yr)(24hr/day)(3600s/hr)]
= 9.01x10-13 (disintegrations/atom)/s or (dps)
λ = activity
t1/2 = half-life
N Pu
= (6.023x1023 atom/mole)/(239g/mole)
= 2.52x1021 atom/g
(λ Pu Npu)
= (9.01x10-13 dps)(2.52x1021 atom/g)
= 2.27x109 dps/g
10 nano-Ci Pu = [10 nano-Ci(37 dps/nano-Ci)]/(2.27x109 dps/g)
= 1.63x10-7 g-Pu
17
λAm-241 = ln(2)/t1/2
= ln(2)/[458 yr(365day/yr)(24hr/day)(3600s/hr)]
= 4.8x10-11 (disintegrations/atom)/s or (dps)
NAm
= (6.023x1023 atom/mole)/(241g/mole)
= 2.50x1021 atom/g
(λAm NAm)
= (4.8x10-11 dps)(2.5x1021 atom/g)
= 1.2x1011 dps/g
Since the mole weights of Pu-239 and Am-241 are approximately equal:
(Am-activity/Pu-activity) ~ (1.2x1011 dps/2.27x109 dps) = 53
With regard to calculating final waste volumes, it has been assumed the americium
concentration is 200 ppm in the feed, or 200 g-Am/MT-Pu feed.
100 nano-Ci Am= [100 nano-Ci(37 dps/nano-Ci)]/(1.2x1011 dps/g)
= 3.08x10-8 g-Am
200 g-Am(10 nano-Ci Am/3.08x10-9 g-Am) = 6.4x1011 nano-Ci Am
Since Class-C LLW can not exceed 100 nano-Ci/gram-waste for TRUs, if less than 6400
MT of total waste is generated, the waste may be classified as TRU waste based on Am
alone.
(6.4x1011 nano-Ci Am)/(100 nano-Ci/gram-waste) = 6.4x109 g-waste (6400 MT)
The activity for plutonium is provided in the previous text for similar waste calculations
once the cascade configuration, and consequent plutonium losses, are determined.
18
4.0 References
(1) M. Benedict, T.H. Pigford and H.W. Levi, Nuclear Chemical Engineering, 2nd
Edition, McGraw-Hill Book Company, 1981.
(2) C Colmenares, Y. Zundelevich, J. Lawson and M. Bronson, Evaluation of
Hydride/Oxidation and Hydride/Nitride/Oxidation Processes to Produce PuO2 Powders
in the HYDOX Module, ARIES Review at Lawerence Livermore National Laboratory,
4 November 1996.
(3) C. Bernard, et al., Advanced Purex Process for the New French Reprocessing Plants,
Global ‘93 Conference: Future Nuclear Systems - Emerging Fuel cycles and Waste
Disposal Options, CEA Centre d’Etudes de la Vallee du Rhone, CEA-CONF-11678,
1993.
(4) X. Machuron-Mandard, Catalysed Electrolytic Metal Oxide Dissolution Processes,
Seminar for Institute National des Sciences et Techniques Nucleaires, Centre d’Etudes
de Saclay, CEA-CONF-11818, 1994.
(5) T. Ishimori and K. Watanabe, Inorganic Extraction Studies on the System of Tri-nbutyl Phosphate-Nitric Acid, Bulletin of the Chemical Society of Japan, Vol. 33, NO.
7, July 1960, pg. 1443.
(6) Plutonium Handbook, Edited by O.J. Wick, Gordon and Breach, Science Publishers,
New York, 1967.
(7) R.T. Jubin, S.F. DeMuth and S.P. Singh, Developments in Centrifugal Contactor
Technology, ORNL/TM-10768, Oak Ridge National Laboratory, Oak Ridge, TN,
September 1988.
19