Adsorption of Uranium(VI) on the MX-80 Bentonite
Ying-Chieh Lin
Sinotech Engineering Consultant, Ltd.
171, Nanking E. Rd., Sec. 5, Taipei 10570, Taiwan

Key words; Adsorption, Bentonite, HYDROGEOCHEM, Uranium(VI)

Abstract
In present work, a coupled model of groundwater flow, thermal transport, and geochemical transport throughsaturated-unsaturated media, HYDROGEOCHEM, has been employing to simulate the uranium adsorption on
the MX-80 bentonite. The numerical experiments were performed with different sodium and hydrogenconcentrations. Uranium concentrations were kept below 10 -7 mol/L to avoid precipitation of amorphousuranium-hydroxide. Adsorption of uranium on the bentonite shows significant dependency on sodium
concentration, because of the competition between uranium and sodium ions. The numerical simulations ofuranium transport in the MX-80 bentonite have also been proposed with deep groundwater condition. The
simulation shows that the retention capacity of MX-80 bentonite and also demonstrate the capacity of
HYDROIGEOCHEM model for simulation radionuclide migration in the near-field system.

1. Introduction
Radionuclide transport simulation is one of mostimport procedure for the safety assessment of spent
nuclear fuel final disposal. How to calculate therelease, transport and dose consequences ofradionuclide for different scenarios and cases is amajor topic for the procedure of safety assessment
analysis.
The multi-barrier engineering system concept is
one of the most popular design concepts for spent
nuclear fuel disposal [3]. There are two major safetyfunctions to prevent the release of radionuclide formulti-barrier engineering system. First is hydraulicfunction and chemical function is second safetyfunction. Hydraulic function and chemical function
are reducing the groundwater contact the canister andincreasing the radionuclide retardation of theengineering barrier, respectively.
Bentonite is a material with high porosity and low
hydraulic conductivity behavior and now inwidespread use in spent fuel disposal for preventing
the groundwater flow into the disposal. Furthermore,
the bentonite also provides the capacity for retaining
the nuclide [7, 12, 13]. Sorption is an important
retention mechanism to control the concentration ofcontaminants in groundwater [12]. However, thecapacity of sorption is close related with aqueous
solution properties and characteristics of sorbing
materials [13]. The surface complexation was animportant mechanism for uranium sorption on thesmectite.[7, 29]
In present work, a hydro-geo-chemical numericalmodel (HYDROGEOCHEM) has been adapted to
study the adsorption of uranium(VI) on the MX-80
bentonite and a pseudo case is conducted todemonstrate the capacity of HYDROGEOCHEM
model.
2. Bentonite material
Bentonite is a geological term for soil materials
with a high content of a swelling mineral, whichusually is montmorillonite [24]. In present work, thereference bentonite MX-80 was conducted to study.
2.1 Mineralogy
The chemical characterization of bentonite clay
differs slightly, between different modeling studies,most due to the employed of different databases [3, 5,
8, 9, 25, 27]. These reports agree on primaryprevalence minerals are quartz and calcite, apart frommontmorillonite. The residual constituents, such as
albite, analcime, chalcedony, cristobalite, gypsum,
halite, illite, K-feldspar differs slightly but has lessinfluence on the overall chemical stability.
In present work, the mineralogical composition of
bentonite MX-80 [16] has been present in table 1.
MX-80 is mainly composed of montmorillonite-Na

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan(90%), Quartz (7.63%) and Calcite (1.04%). Others
have less influence on the overall chemical stability.
2.2 Transport parameters
The condition of bentonite is according to SKB-3V
setup [21] which surrounds the canister. The dry
densities and saturated density are 1,577 kg/m3 and
2,000
kg/m3,
respectively.
The
porositycorresponding to the saturated density 2,000 kg/m3 is
0.435 [23]. The diffusion coefficient is 1.4x10-10 m2/s
[23]. The hydrodynamic parameters are listed in table
2. In this work, we do not consider the permeabilityof bentonite and the migration is controlled by the
diffusion mechanism.
Table 1. Composition of bentonite MX-80
Mineral Phases
Volume (%)
Montmorillonite-Na
90.00
Quartz
7.63
Clacite
1.04
Phillipsite-Na
0.99
Illite0.19
Saponite-Ca
0.12
Magnetite
0.02
Table 2. Various transport parameters
Parameter
Value
Dry Density (kg/m3)
1577
Porosity (%)
43.5
Diffusion Coefficient (m2/s)
1.4x10-10
Tortuosity
1.0
2.3 Surface Complexation Reactions
For the modeling of uranium (VI) adsorb on thesmectites, the surface complexation reactions are
required. A multisite layer surface complexation
model was developed by Zachara and Mckinley[29]to describe the uranium sorption by smectites. Their
model is composed of pH independent exchange (X-)
sites and hydroxylated edge sites for SiOH and AlOH.
Samer and Broder [7] propose that sorption of
uranium on the selected smectites preferentially occur
to the edge site of aluminol and silanol, even underacidic condition. This assumption is also based on therelatively high Na+ concentration (0.01 and 0.025
mol/L) in comparison to very low uranium
concentration (100 μg⁄L) in the solution.
In this study, surface complexation reactions on thesmectites are based on the Samer and Broder [7] asshown in table 3. They used the non-linear leastsquares parameter estimation program PEST [2] inconjunction with PHAST [11] to determine theequilibrium constants of these sorption reactions.
Sodium sorption was also incorporated in table 3

since the sodium was proven to sorb onto clay
minerals [26].
The thermodynamic database thermoddem V1.10[1] was employed as the reference thermodynamic
database in the model calculations.
Table 3. Surface reactions and parameters used in themodel calculations
Surface Reactions
Log K
SiOH + H+ ⇄ SiOH2+-0.95+
SiOH ⇄ SiO + H
-6.95
2+
+
+
SiOH + UO2 ⇄ SiOUO2 + H
4.06
SiOH + UO2OH+ ⇄ SiOUO2OH + H+
1.00
SiOH + UO2(OH)3 ⇄ SiOH(UO2)(OH)3
6.90
SiOH + Na+ ⇄ SiONa + H+
-2.71
SiOH + UO22+ + 3CO32- ⇄6.65
SiO(UO2)(CO3)35- + H+
AlOH + H+ ⇄ AlOH2+
12.30+
AlOH ⇄ AlO + H-13.16
AlOH + UO22+ ⇄ AlOUO2+ + H+
7.51++
AlOH + Na ⇄ AlONa + H
15.00
2+
2AlOH + 2 UO2 + CO3 + 3H2O ⇄0.99
AlO(UO2)2(CO3)(OH)32- + 4H+
SiOH: silanol surface at edge sites, AlOH: aluminol
surface at edge sites.
3. Numerical model
The computer program HYDROGEOCHEM wasemployed to model the radionuclide transport andhydrogeochemical reactions in this study. The
HYDROGEOCHEM 4.1 and 5.1 [28] are
two-dimensional and three-dimensional coupledmodel of water flow, thermal transport, solutetransport, and geochemical kinetic/equilibrium
reactions in saturated-unsaturated porous media,
respectively.
HYDROGEOCHEM 4.1 and 5.1 iteratively solve
two- and three-dimensional fluid flow, thermaltransport, and reactive chemical transport equations.
Richards’ equation is solved for fluid flow, thermaltransport equation is solved for temperature
distribution,andadvection-dispersion-reactive
transport equations are solved for all mobile
components and kinetic-variables in a two-stepmethod or for all components and kinetic variables ina one-step method. The subsidence is calculated
based on one-dimensional linear elasticity theory.
HYDROGEOCHEM 4.1 and 5.1 is designed forgeneric applications to reactive transport problems
controlled by both kinetic and equilibrium reactions
in subsurface media. The input to the program

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, JapanHYDROGEOCHEM 4.1 and 5.1 are thecomprehensive simulators of coupled fluid flow,hydrologic transport, thermal transfer, andbiogeochemical transport under variably saturated
conditions in two dimensions and three dimensions,
respectively.
.
4. Results and discussion
4.1Effect of pH and Na+ concentration on
Uranium Sorption
The purpose of this section is to understand thesorption behavior of the present MX-80 bentonite
reaction model under different pH and Na+concentration condition. Here, the experiment
procedure [7] was used for the comparison. The
numerical experiments were carried out by reacting
1.67g of MX-80 bentonite and 1 liter of 10-7 mol/Luranium(VI) solution with 0, 10-3, 10-2, 10-1 mol/L
NaCl and the pH value is range from 4 to 9. The
uranium concentration 10-7 mol/L was chosen to
avoid
the
precipitationofamorphousuranium-hydroxides. Here we define an equation todescribe the capability of sorption on the MX-80
bentonite. The percentage of the sorbed uranium wascalculated from the difference between the initial andfinal uranium concentration. The equation is asfollowing,
C −Cf

Sorption(%) = ( iCi) × 100%(Eq. 1)

Where Ci and Cf are the uranium concentrationsin the initial and final solutions, respectively.
The effect of pH and Na+ concentration on theuranium sorption behavior onto the MX-80 bentonite
is illustrated in Figure 1. The high sorption ability
was observed at pH 4 even though the high Na+concentration that has competition with adsorption
edge sites. In the pH range 6 to 9, a distinct reduced
uranium sorption on MX-80 bentonite at higher pH.
It shows that the uranium sorption decreases rapidly
with increasing Na+ concentration. When pH larger
than 8, the uranium sorption gets close to 45% ~ 50%
gradually. These results do agree with the research bySamer and Broder [7]. They also found high sorption
capability at pH 4 and reduced uranium sorption on
smectites in the pH range 6 to 9. Higher pH is related
to the low affinity of uranium carbonate complexes tothe surface sites of clay minerals or to competition
between carbonate ions and uranium species.
10080

Uranium Sorption (%)

includes the geometry of the system, the spatial
distribution of finite elements and nodes, theproperties of the media, reaction network, and theinitial and boundary conditions. Output includes thespatial distribution of pressure head, total head,velocity fields, moisture contents, temperature, andchemical concentrations as a function of time as wellas the distribution of all chemical species atuser-specified nodes. Subsidence for the overall
region is also calculated.

60
0 M NaCl
40

0.001 M NaCl
0.01 M NaCl
0.1 M NaCl

20

0
4

5

6

7

8

9

pH

Figure 1. Comparison between different pH and Na +concentration data for uranium(VI) sorption onto the
MX-80 bentonite.
4.2 Near-field transport simulation
In the framework of the spent fuel disposal safetyassessment, a conceptual and numerical modeling ofradionuclide reactive transport in the near-surface
system [17, 18, 20] and the long-term performance ofengineering barrier [14, 15, 16, 25] have been
carrying out. These reports use the geochemical
program PHREEQC [10] and PHAST [11] to solver
reactive chemistry and reactive transport problems,
respectively. In this section, a three-dimensionalhydro-geo-thermo-chemical
couplemodel,HYDROGEOCHEM, has used for solving theradionuclide transport in the near-field engineeringbarrier base on the SKB-3V concept [19] as shown infigure 2(a). In this concept, canister and buffer are
two major components in this design. The canister issurrounding by the buffer material, here the buffermaterial is MX-80 bentonite. The detailed geometricparameters are presented in table 4.
The scenario is this single canister failure due tothe shear force applies on the canister and makes
damage. We assume an incline fracture across thesystem and conduct a 1.0 cm thickness crack on thecanister and only diffusion process is considered in
the present simulation according to the low
permeability property of MX-80 bentonite. Becauseof the diffusion mechanism dominate theradionuclide migration, the computational domain islimited with 2.0 m of height for better computational
efficiency. The three dimension computation domain
is illustrated in figure 2(b).
The dissolution rate and solubility limit of uranium
in spent fuel are 10-7 1/yr and 9.5x10-10 mol/L [22],6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japanrespectively. Besides, the inventory of uranium inmole per canister is 7,200 mol/canister. The selected
groundwater [20] in present work was deep
groundwater (underground 300 m) which is the
Na-Cl type with moderate salinity. The groundwatercomposition is reported in table 5.
The spatial discretization of the computational
domain was shown in figure 3(a). There are total
73,359 tetrahedron elements and 16,622 nodes to
describe the computational domain. In order to get agood approximation, the crack of the canister has
been used fine grids to describe the geometry (seefigure 3(b)).

(a)(b)
Figure 3. The computational grid for radionuclide
migration simulation. (a) the full domain, (b) zoom inview at the crack position of canister.
The figure 4 and figure 5 show the uranium(VI)concentration in the solution and on the sorption siteat the simulation time 500 year, respectively. The
high uranium(VI) concentration is observed near thecrack position of the canister, then the concentration
decreases rapidly due to the sorption reaction. Thisphenomenon also observed in figure 5. Higher
uranium(VI) concentration in solution will result inhigh uranium(VI) concentration sorbed on the solid
surface. Because of the migration is dominated by the
diffusion mechanism, therefore the uranium(VI)migrate uniformly in space.

(a)(b)
Figure 2. Sketch of simulation case.
Table 4. The geometric parameters.
Parameter
Height of the canister (CH)
Diameter of the canister (CD)Height of disposition hole (H)Diameter of disposition hole (D)Void inside canister
Thickness of canister wallValue
4.835 m1.05 m
6.835 m1.75 m
1.0 m30.05 mTable 5. Chemistry of the selected groundwater.
SpeciesValue (mol/L)Na+
6.13x10-2+
K
7.98x10-4Ca++
1.82x10-2++
Mg
4.73x10-3
HCO3
4.72x10-3Cl
1.04x10-1-SO4
2.21x10-3
H4SiO45.63x10-4Fe
5.78x10-5
pH
7.6

In the previous section, the Eq. 1 has been used for
calculating the percentage of sorption. Here, thesolid/liquid partition coefficient (KD) is proposed todefine the concentration of uranium(VI) in the solid
phase over its concentration in pore water. Since theconcentration of uranium(VI) in figure 4 and figure 5
are reported as mole per liter water, but theconcentration of uranium(VI) on the sorption siteneed to be converted to mole per kilogram of rock inorder to determine the KD value. For this calculation,
we need the porosity and the density of bulk rock.
According to table 2, the porosity of MX-80
bentonite is 0.435 and the density is 1577 kg/m3. The
calculation of KD is shown as following,[Sorbed]mol⁄L

KD = [Solute]mol⁄ ×L

0.435
1.557kg⁄L(Eq.

2)
The distribution of KD at the simulation time 500
year is illustrated in figure 6. The result shows wellretention capability of MX-80 bentonite for
uranium(VI). We can observe that the KD value near
the crack position is lower than the average. Becauseof the uranium is continued release from the canisterand the free edge site in the bentonite will decrease
gradually. Therefore, in the beginning, the KD value
can be approached 1.2x105 L/kg and reducing to 104
L/kg at the simulation time 500 year.

6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan

5. Conclusion
A geochemical numerical model of MX-80
bentonite with uranium surface complexation
reaction has been proposed in present work.
Furthermore, batch numerical experiments withdifferent NaCl concentration and a simplify KBS-3V
type repository case implement in the present
numerical model.
Batch experiments clearly showed that the muchhigher uranium sorption capability in the pH range 4
to 6 and rapidly reduce in the pH range 6 to 9. Higher
Na+ concentration will result in lower sorption
capability.
Figure 4. The distribution concentration contour of
uranium in solution at x=0 plane.

The established reactive model also implement to aKBS-3V type case. In this case, a shear force causes
the failure of the canister, and only diffusion
mechanism is assumed to control the migration of
uranium(VI). The results demonstrated that the HGCmodel has the capacity to modeling the uranium
reactive transport in the near-field and calculate theretention capacity of the buffer system.
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Figure 5. The distribution concentration contour of
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Figure 6. The distribution contour of KD value for
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6th East Asia Forum on Radwaste Management Conference
November 27-29, 2017, Osaka, Japan