9. TRU and Nuclear Chemistry

概要

CO9-1

Extraction chromatography and solvent extraction of Eu using TEHDGA

K. Otsu, M. Ikeno, C. Saiga, S. Takahashi, C. Kato, T.
Matsumura1, H. Suzuki1, S. Fukutani2 and T. Fujii
Graduate School of Engineering, Osaka University
1
Nuclear Science and Engineering Center, Japan Atomic
Energy Agency
2
Institute for Integrated Radiation and Nuclear Science,
Kyoto University
INTRODUCTION: The disposal of high level radioactive liquid waste produced from nuclear power plants is
one of the main problems surrounding nuclear energy. To
solve this problem, partitioning and transmutation technologies for high-level radioactive liquid waste generated
from the reprocessing of spent nuclear fuel from nuclear
power plants is being developed. The Japan Atomic Energy Agency (JAEA) has proposed the "SELECT Process" which involves the separation and transmutation of
radioactive nuclides as a method of managing radioactive
waste [e.g. 1]. The SELECT process consists of four
steps, in which various extractants are used to separate
minor actinides (MA) and rare earth elements (REE)
from the spent nuclear fuel solution. In this study, we
focused on europium (Eu), which is a REE that has two
radioisotopes, 154Eu and 155Eu with a half-life of 8.6 years
and 4.8 years respectively [2]. TEHDGA (Tetra-2-ethylhexyldiglycolamide), is a new extractant synthesized at JAEA that has both excellent extraction and
separation performance for minor actinides and REE. In
order to understand its chemical behavior and to compare
is efficiency during different extraction methods, solvent
extraction and extraction chromatography were carried
out using TEHDGA on Eu.
EXPERIMENTS: The organic phase was prepared by
diluting the extractant TEHDGA with a 4:1 mixed solution of n-dodecane and 1-octanol to adjust the concentration to 0.01, 0.02, 0.04 and 0.10 M. 1 ppm Eu was dissolved in 1.5 M HNO3 for the aqueous phase. The phases
were stirred for 30 mins to reach equilibrium and then
separated by centrifugation. Back extraction was performed with a 0.1 M HNO3 solution. To prepare the resin
for extraction chromatography, 1 g of TEHDGA was dissolved in 50ml of ethanol (99.5% purity) to make a 0.001
M solution. 1 g of styrene divinylbenzene was added to
the solution and the ethanol was evaporated. For the extraction chromatography experiment, 1.5 M HNO3 was
added to the resin and then transferred into a column. The
column was attached under another column which introduced the feed solution to the resin filled column. A valve
was attached between the columns to regulate the flow to
ensure the resin filled column would not be saturated by
the feed solution and to not disturb the resin. 10 ml of 1.5
M Eu(NO3)2 solution was introduced to the upper column
and the eluted solution was collected in increments of 1
ml. For the batch experiment, a separate batch of resin
was prepared with the same method and was shaken for

15 mins with 1.5 M Eu(NO3)2 and the liquid was collected. Europium concentration from all experiments were
measured by Inductively Coupled Plasma Quadrupole
Mass Spectrometry (ICP-QMS) at Kyoto University.
RESULTS AND DISCUSSIONS: A slope 2 correlation
between the TEHDGA concentration and Eu concentration was observed (Fig. 1). This suggests that a
TEHDGA : Eu = 2 : 1 complex forms in this acidity, and
the following equation can be obtained.
Eu3+ + 3NO3- + 2TEHDGA ⇆ Eu(NO3)3・2TEHDGA
A breakthrough curve was obtained with the extraction
chromatography (Fig 2). After 5ml, the relative concentration of Eu maintains a flat peak, suggesting the resin
became saturated with Eu. The number of theoretical
plates was calculated as 12.2, and the height equivalent of
one theoretical plane (HETP) as 2.5 mm. From the batch
experiment, this value was proven to be reasonable.

Fig. 1. Dependence of D on TEHDGA concentration for Eu.

Fig. 2. Eu elution curve of extraction chromatography.
Concentration is normalized to the starting Eu solution.
REFERENCES:
[1] T. Matsumura, Kino Zairyo, 40 (2020) 60-71.
[2] G. Audi. et al., Nucl. Phys. A, 729 (2003) 3-128.

R4015
- 224 -

CO9-2

Electrochemical Behavior of U in NaCl-CaCl2 Melt at 823 K

T. Murakami, Y. Sakamura and K. Takamiya1
Central Research Institute of Electric Power Industry
1
Institute for Integrated Radiation and Nuclear Science,
Kyoto University
INTRODUCTION: The mixture of NaCl and CaCl2 is
one of the promising candidates as the base salt for molten
chloride salt fast reactor from the viewpoints of melting
point, solubility of actinides, applicability to reprocessing,
restraining of radioactive elements formation by irradiation and so on. Electrochemical properties of actinides in
the melt are required for constructing a reprocessing process of the spent molten chloride fuel salt. However, reports on electrochemical measurements of actinides in
NaCl-CaCl2 melt [1, 2] are very limited. Thus, this study
investigated electrochemical behavior of U to evaluate the
redox potential of U in eutectic NaCl-CaCl2 melt at 823 K.
EXPERIMENTS: All experiments were performed in a
glove box filled with purified Ar gas. Eutectic NaCl-CaCl2
melt containing UCl3 was prepared as follows. NaCl (23.4
g), CaCl2 (48.1 g) and CdCl2 (0.587 g) were loaded in an
alumina crucible and melted at 823 K. A U metal rod (36.0
g) was immersed in the melt for overnight to form UCl3 in
the melt,
2U + 3CdCl2 → 2UCl3 + 3Cd

The additional cathodic (c3) and anodic (a3) currents
might be due to U adsorption and desorption, respectively.
Based on the results of the cyclic voltammetry, galvanostatic electrolysis at -10 mA was performed for 10 seconds
to deposit U metal on the W wire electrode. Then, the applied current was set at 0 mA to measure the redox potential of the U metal on the W in NaCl-CaCl20.255mol%UCl3 melt. By using the measured redox potential (E = -2.466 V), the formal standard redox potential
(E0’) was calculated to be -2.325 V according to the following Nernst equation of reaction 2,
𝐸

(4)

(1)

RESULTS: Fig. 1 shows cyclic voltammogram of W wire
electrode in the melt. A cathodic current, c1, increasing
from around -2.5 V (vs. Cl2/Cl-) was ascribed to U metal
deposition (reaction 2),
(2)

The corresponding anodic current, a1, was due to the dissolution of the deposited U (reverse reaction of reaction 2).
A broad cathodic (c2) and anodic (a2) current couple at
around -1.2 V was considered to correspond to the following reaction,
U4+ + e- = U3+

ln𝐶

R is gas constant, T is temperature in Kelvin, F is faraday
constant, CUCl3 is the concentration of UCl3 in the melt.
The obtained formal standard redox potential in NaClCaCl2 melt (-2.325 V) was higher than that in LiCl-KCl
melt (-2.466 V [4]). The difference in the formal standard
redox potential indicated the stability of U3+ could depend
on the cations of the solvent.

Gibbs energy of reaction 1 is -577 kJ at 823 K [3] suggesting that almost all of CdCl2 was consumed to dissolve U.
Thus, the resulting concentration of UCl3 in the melt was
calculated to be 0.255 mol% based on reaction 1. The remained U metal rod was recovered after the completion of
the reaction.
Electrochemical measurements were performed in the
prepared melt at 823 K. The working electrode was W wire
(1 mmφ). The Ag/AgCl reference electrode was used, of
which potential was -1.255 V (vs. Cl2/Cl-). Glassy carbon
rod (3 mmφ) was used as the counter electrode.

U3+ + 3e- → U

𝐸

Fig. 1. Cyclic voltammogram of the W wire electrode
in NaCl-CaCl2 melt containing 0.255 mol% UCl3 at
823 K. Scan rate was 50 mV･s-1.
REFERENCES:
[1] H. Zhang et al., J. Electrochem. Soc., 168 (2021)
056521.
[2] M. L. Newton et al., ECS Trans., 98 (2020) 19-25.
[3] C. W. Bale et al., FactSage Thermochemical Software
and Databases, 2010-2016, Calphad, 54, (2016)
35-53 .
[4] P. Masset et al., J. Electrochem. Soc., 152 (2005)
A1109-A1115.

(3)

R4048
- 225 -

Solid phase analysis of (Zr,Ce)O2 solid solutions in aqueous systems

T. Kobayashi, Y. Sato, T. Sasaki, S. Sekimoto1, K. Takamiya2
Graduate School of Engineering, Kyoto University
1
Institute for Integrated Radiation and Nuclear Science,
Kyoto University
INTRODUCTION: It is important to understand the
solubility behavior of (Zr,U)O2 solid solution in aqueous
systems for the management of nuclear fuel debris generated in the Fukushima Daiichi Nuclear Power Plant accident since (Zr,U)O2 solid solution is one of the possible
major components of the debris. Many studies have been
dedicated to establishing a robust thermodynamic model
to quantitatively explain the solubility of ZrO2 and UO2,
while less is known for the solubility of mixed systems. In
the present study, we focused on (Zr,Ce)O2 solid solution
as an analog of (Zr,U)O2 solid solution. In the phase diagram of (Zr,Ce)O2 solid solution at 1000 ℃, for example,
pure ZrO2 exhibits a monoclinic phase, and tetragonal
ZrO2 and cubic CeO2 phases appear with increasing Ce
molar ratio [1]. This trend is similar to that observed in the
phase diagram of (Zr,U)O2 solid solution [2]. In this study,
(Zr,Ce)O2 solid solution was prepared with the molar ratio
of Zr/Ce = 7/3 at 1000 ℃ and characterized using powder
X-ray diffraction (XRD) . Then, the solid phase was placed
into aqueous sample solutions under reducing conditions
ranging from pH 0.8 to 8.0 for several months. After the
immersion of given periods, the solid phase was separated,
dried and investigated by powder XRD to elucidate the
change in the state of solid phase and discuss their solubility behavior.
EXPERIMENTS: Acidic stock solutions of Zr(IV) nitrate and Ce(IV) nitrate were prepared and mixed with the
molar ratio of Zr/Ce = 7/3. Portions of concentrated polyvinyl alcohol (PVA) were added to the mixed solution and
heated to dryness at 200 ℃. The dried powder was then
heated at 1000 ℃ for 4 hours in a muffle furnace to synthesize (Zr,Ce)O2 solid solution. ...