MA(III) co-precipitation studies for MA-bearing oxide solid solution synthesis

210 Pages·2010·8.04 MB·English

RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
Results on U(IV)/MA(III) co-precipitation studies
for MA-bearing oxide solid solution synthesis
Jean-Philippe Dancausse, Stéphane Grandjean, Nathalie Herlet, Carole Viallesoubranne,
Caroline Léorier, Benedict Arab Chapelet, Olivier Conocar
Commissariat à l’Énergie Atomique
Centre de Marcoule, France
Abstract
Within the framework of minor actinide management, the French reference way is their separation
and their transmutation in fast reactors. One of the options for the transmutation is the heterogeneous
mode, using actinide oxide on a uranium matrix located at the periphery of the core (“transmutation
blanket”). The objective of the work is to study a precipitation process leading to the formation of
U(IV)/MA(III) compound, with concentration of MA in the range 5-20%.
In the case of americium and curium solid solution, for which an actinide content of less than 20% of
U+MA is aimed, the oxalic co-precipitation of a solution U(IV)/MA(III) is studied. This process was not
previously realised on a uranium/americium and/or curium mixture but the principle of oxalic
co-precipitation of the III and IV oxidation levels has already been validated on U(IV)/Pu(III) mixtures.
The main objective of tests carried out in 2006-2007 on C11/C12 Atalante hot cell was to validate the
oxalic co-precipitation of U(IV)/Am(III) and/or Cm(III) by exacerbating the radiolysis phenomena
through the use of 10% Cm of mass U+MA. This presentation describes the qualification steps of the
process into hot cell from (U,Ce) precipitation with surrogate solution to the co-precipitation of a
U(IV)/Cm(III) solution leading to a mixed oxalate U-Cm and then, after calcination under inert
atmosphere, to the solid solution (U,Cm)O having the fluorite structure type.
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RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
Introduction
Within the framework of minor actinide management, the French reference way is their separation
and their transmutation in fast reactors. One of the options for the transmutation is the
heterogeneous mode, using actinide oxide on a uranium matrix located at the periphery of the core
(“transmutation blanket”). The objective of the work is to study a precipitation process leading to the
formation of U(IV)/MA(III) compound, with concentration of MA in the range 5-20%.
In the case of americium and curium solid solution, for which an actinide content of less than
20% of U+MA is aimed, the oxalic co-precipitation of a solution U(IV)/MA(III) is studied. This process
was not previously realised on an uranium/americium and/or curium mixture but the principle of
oxalic co-precipitation of the III and IV oxidation levels has already been validated on U(IV)/Pu(III)
mixtures.
The main objective of tests carried out in 2006-2007 on C11/C12 Atalante hot cell was to validate
the oxalic co-precipitation of U(IV)/Am(III) and/or Cm(III) by exacerbating the radiolysis phenomena
trough the use of 10% Cm of mass U+MA.
The important steps of this demonstration are the obtaining of a single phase mixed oxalate
U-Cm and after calcination under inert atmosphere the obtaining of the aimed single phase solid
solution (U,Cm)O .
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Experimental details
Reagents
Uranium(IV) and curium(III) or Ce(III) surrogate solutions were prepared respectively from purified
monometallic solution and by dissolving monometallic oxide. Hydrazinium nitrate (N H +,NO –) was
2 5 3
used as anti-nitrous agent to stabilise the +IV oxidation state of uranium. The solution characteristics
(concentration, purity, oxidation state, …) were mainly determined using ICP, radiometric and
spectroscopic methods.
Hot cell facility
Studies were realised in the C11/C12 shielded cell of the Atalante facility. It is constituted of eleven
working place behind one meter of concrete and lead glasses. The versatility of this hot cell has
permitted to achieve all the required steps to obtained (U ,Cm )O solid solution: radioactive
0.9 0.1 2
solution purification (240Pu removal coming from 244Cm decay), precipitation and calcination as well as
some on-line or in situ measurements.
Synthesis methods
The synthesis of mixed-oxide of uranium and curium was realised by a batch oxalic precipitation in a
vortex effect vessel (REV), under the following conditions (Figure 1):
• realisation of approximately 2 g of (U,Cm)O at 10% in Cm mass compared to (U+Cm);
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• co-precipitation of oxalate of U(IV)/Cm(III) by simultaneous addition of:
– nitric solution of U (IV) (stabilised with hydrazine) and Cm (III) titrating approximately
40 g/L of actinides;
– oxalic acid and hydrazinium nitrate mixture;
– in an oxalic acid and hydrazinium nitrate in nitric medium, representative of the
composition of the oxalic natural brines to balance;
• filtration of the obtained mixed oxalate;
• calcination of oxalate at 700°C under a neutral argon atmosphere.
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RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
Figure 1: Experimental conditions of uranium and curium oxalate batch co-precipitation
UU((IIVV))--CCmm((IIIIII)) CCoo--ccoonnvveerrssiioonn
CCmm//((UU++CCmm)) == 1100 %% ((aatt..//aatt..))
ooxxaalliicc aacciidd ffeeeedd ssoolluuttiioonn
AAccttiinniiddee ffeeeedd ssoolluuttiioonn HHCCOO
22 22 44
UUIIVV NNHH++
22 55
CCmmIIIIII FFllooww rraattee :: 55 mmLL..mmiinn--11
HHNNOO VVoolluummee:: 3300 mmLL
33
NNHH++
22 55
FFlloowwrraattee:: 55 mmLL..mmiinn--11 RREEVV
VVoolluummee:: 3300 mmLL VVoolluummee :: 220000 mmLL
SSttiirrrriinngg :: 445500 ttrr..mmiinn--11
IInniittiiaall ssoolluuttiioonn
HHCCOO
22 22 44
HHNNOO
33
NNHH++
22 55
VVoolluummee:: 6600 mmLL

The precipitation vessel and filtration are described below (Figures 3, 5 and 6). For the calcination
step, a horizontal tubular furnace was used. The oxalate powder is placed inside a quartz vessel under
an argon flow as shown in Figure 2.
Figure 2: Details of calcination set-up

Experiment realisation
Before the realisation of experiment using curium, a test under the real conditions (Figure 3) of hot
cell was performed using uranium and cerium. The criterion retained to validate this step was the
single phase structure of both oxalate and oxide after calcination. The obtained X-ray pattern is
presented on Figure 4, confirming the efficiency of the retained experimental design for hot cell test
using curium.
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RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
Figure 3: Hot cell precipitation set-up

Figure 4: Diffraction pattern of U-Ce oxide cfc single phase
UUCCeeOO
22

Demonstration on uranium and curium
All the steps described below were realised in one day.
U and Cm solution were mixed half an hour before the beginning of the precipitation leading to a
feed solution having the following characteristics:
• feed volume: 30 mL;
• [U]: 40 g/L;
• [Cm]: 3.65 g/L.
The precipitation step itself as described above in Figure 1, took about 7 min. (Figure 5) with an
average flow of 4.3 mL/min and the ageing step about 20 min. After the filtration (Figure 6), rinsing
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RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
and partial drying operations, nearly 2 g of oxalate precipitate were recovered (Figures 7 and 8). Then,
after two hours argon sweeping of calcination set-up, the oxalate was heated up to 720°C and
maintained at this temperature for three hours. The resulting oxide weighing was 0.95 g.
Figure 5: Precipitation step

Figure 6: Filtration step

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RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
Figure 7: Filter and oxalate powder recovery

Figure 8: Crucible filling with oxalate powder

Results
Chemical balance
The efficiency of the process was evaluated considering initial and final solution contents. Less than
0.03% of initial curium remains in the filtrate solution leading to a calculated atomic ratio Cm/U
slightly above 10% in the resulting solid.
Powder characteristics
X-ray diffraction
The X-ray diffraction pattern obtained on oxide shows a well-defined single phase corresponding to
the expected fluorite type structure. The calculated lattice parameter is equal to 54.49 nm.
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RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
Figure 9: Diffraction pattern of U -Cm O cfc single phase
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UU CCmm OO
00..99 00..11 22
44000000
33000000
Lin (Counts)Lin (Counts)22000000
11000000
00
99 1100 1111 1122 1133 1144 1155 1166 1177 1188 1199 2200 2211 2222 2233 2244 2255 2266 2277 2288 2299 3300
22--TThheettaa -- SSccaallee

Scanning electron microscopy observations
The SEM examinations of oxalate and oxide powders exhibit nearly the same morphology for both.
The slight difference (average particle size) is mainly due to the mass loss during thermal treatment.
Figure 10: (U,Cm) oxalate agglomerate

Figure 11: (U,Cm)O agglomerate
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RESULTS ON U(IV)/MA(III) CO-PRECIPITATION STUDIES FOR MA-BEARING OXIDE SOLID SOLUTION SYNTHESIS
Conclusion
The co-conversion process applied to U(IV)-Cm(III) solution with a ratio 90/10 led as expected to a
fluorite single-phase oxide. This experiment will be continued by studying the effects of 244Cm alpha
irradiation on structure and lattice parameters of this oxide as a function of time.
Further experiments will be oriented to study the synthesis of fluorite structure solid solutions
containing different ratios and mixtures of U, Pu, Np, Am and Cm in several grams scales in the
framework of minor actinide transmutation studies.
References
Grandjean, S., et al., World Patent, WO 2005/119699 (2005).
Grandjean, S., et al., “Structure of Mixed U(IV)-An(III) Precursors Synthesized by Co-conversion
Methods”, presented at Plutonium Futures 2008 and forthcoming in Journal of Nuclear Materials.
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SAFETY RESEARCH FOR MULTI-FUNCTIONAL REPROCESSING PROCESS BASED ON ION-EXCHANGE METHOD
Safety research for multi-functional reprocessing
process based on ion-exchange method*
S. Koyama,1 M. Ozawa,1,4 K. Okada,2 Y. Sato,2
K. Kurosawa,3 K. Tatenuma,3 T. Suzuki,4 Y. Fujii4
1Japan Atomic Energy Agency
2National Institute of Advanced Industrial Science and Technology,
3KAKEN Co., Ltd.
4Tokyo Institute of Technology
Abstract
The group separation of lanthanide (Ln) and actinide (An), and the mutual separation of americium
(Am) and curium (Cm) are still unfinished assignment as an optional separation scheme concerned
with current aqueous reprocessing process. We proposed a multi-functional separation process on the
basis of an ion-exchange method using tertiary pyridine type resin. Hot experiment of total separation
process, starting from fuel dissolution by using a fragment of a spent mixed-oxide (MOX) fuel which
had been highly irradiated in the experimental fast reactor JOYO, was demonstrated at the
Alpha-Gamma Facility (AGF) of Japan Atomic Energy Agency (JAEA). We could achieve separations of
the five groups, i.e. platinum group elements, Ln fission products, fuel elements (U, Pu and Np), Am
and Cm. The process is a key technology for Advanced ORIENT Cycle concept implemented by JAEA as
a fundamental research programme on separation, transmutation and utilisation of nuclides in the
nuclear spent fuel. In order to apply this process to engineering scale, two important subjects should
be solved so as to prove the availability. One is explication of the reactive safety between ion exchange
resin (IER) and solvent (conc. HNO – MeOH and HCl). The other is engineering aspect for the use of
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conc. HCl solution, because of its corrosive property to the material.
Thermal hazards of the pyridine-type IER/MeOH – HNO eluent and HCl system were examined from
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the viewpoints of fire and explosion safety. First, a fundamental analysis of IER was conducted, and
was analysed the chemical reactions of the IER/MeOH – HNO eluent system. Second, we applied a
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differential scanning calorimeter (DSC) to evaluate thermal hazards of the IER/MeOH – HNO eluent and
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HCl system. Third, gram-scale heat experiment was performed to confirm the actual thermal hazard.
It is well known that HCl is corrosive solvent towards structural materials made of stainless steel, and
therefore, optimised structural metal should be used for components and equipment. Four metals,
tantalum (Ta), zirconium (Zr), niobium (Nb) and hastelloy which is a nickel-based alloy (Ni-28Mo),
were selected as candidate structural materials. The conventional austenitic stainless steel SUS316L
was used as a reference. Corrosion experiment at room temperature and 90°C using liquid HCl was
performed to observe the corrosion rate and soluble elements from the matrix and corroded structure.
And moreover, we made electrochemical measurements to evaluate corrosion mechanism. In this
meeting, these results will be reported.

* The full paper being unavailable at the time of preparation of this CD-ROM, only the abstract is included.
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RECOVERY OF ACTINIDES FROM A LIQUID CATHODE BY A CADMIUM DISTILLATION
Recovery of actinides from a liquid cathode by a cadmium distillation
S-W. Kwon, Y-J. You, S-W. Paek, K-R. Kim, S-H. Kim,
J-B. Shim, H. Chung, D-H. Ahn, H-S. Lee, E-H. Kim
Korea Atomic Energy Research Institute
Daejeon, Korea
Abstract
In Korea, the studies on a partitioning have been focused on the development of pyroprocessing based
on an electrorefining of actinides because it is a kind of proliferation-resistive technology, where all the
transuranic metals are separated together as a mixture. Electrorefining is generally composed of two
recovery steps – deposit of uranium onto a solid cathode and the recovery of actinide elements by a
liquid cadmium cathode. The actinides in the liquid cadmium cathode are generally collected as an
ingot by an evaporation of cadmium at a reduced pressure. In this study, cadmium distillation
experiments were carried out to examine the behaviour of a cadmium distillation for the development
of an actinide recovery process from a liquid cadmium cathode. The experimental set-up is composed
of an evaporator, condenser, vacuum pump, control unit, and an off gas treatment system. The
evaporation temperature was varied from 400 to 700°C. Cadmium was successfully distilled and
separated from the surrogate actinide metals. The evaporation rate were measured and compared
with the values calculated by a relation based on the kinetics of gases. The theoretical value of the
evaporation rate calculated by the Hertz-Langmuir relation is higher than the experimental value.
This deviation was compensated by an evaporation coefficient obtained empirically. The evaporation
coefficient was a function of the temperature.
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