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Dissolution mechanisms for U02 and spent fuel
NUCLEAR AND CHEMICAL
WASTE MANAGEMENT,
0191-815X/82/020083-08$03.00/0 1982 Pergamon Press Ltd.
Vol. 3, pp. 83-90,1982
Printed in the USA.
DISSOLUTION MECHANISMS FOR UOz AND SPENT FUEL R. Wang Y. B. Katayama Pacific Northwest Laboratory, Richland, Washington 99352, USA
ABSTRACT. The probable dissolution mechanisms for UOz and spent fuel were studied based on accelerating oxidation and dissolutions in autoclave and electrochemical experiments. The dissolution of UO, is primarily due to oxidation of U++’ (solid) to U6 (liquid). The U+6(liquid) is initially in dissolved uranyl ions, UOf2, but subsequent hydrolysis reactions lead to the formation of solid uranyl hydrates (or related complex compounds) that deposit onto the surfaces of U02 and container walls as thin or thick coatings, depending on the temperature, pH, and time. Therefore, the dissolution rate of U02 cannot be simply determined from the uranium content of the solution. Moreover, the dissolution rate is not limited by the solubility of the uranyl ions. Electrochemical dissolution of spent fuel shows that the initial dissolution and leaching behavior is similar to that of UO, surfaces.
on accelerated oxidation and dissolution in autoclave static and electrochemical dissolution experiments. Electrochemical methods were also applied to spent fuel in the three solutions to compare with the behavior of the UO, surfaces.
INTRODUCTION Understanding the dissolution and leaching mechanisms of spent fuel is important for predicting, via mathematical models, the long-term stability of spent fuel in conditions of a geologic repository. The complex structural, microstructural, compositional, and thermal characteristics of spent fuel, along with the lack of radiation-shielded analytical tools, has necessitated a more simple, systematic approach for understanding the leaching and release of radionuclides in aqueous media. Such an approach was developed at the Pacific Northwest Laboratory (PNL) by studying, in initial experiments, a singlecrystal U02 as representative of the spent-fuel matrix (1). Oxidation and dissolution experiments for U02 were designed to accommodate ready experimental verification of the same mechanisms in spent fuel. This paper presents the complex oxidation and dissolution of U02 in three aqueous solutions: deionized water, 0.03 M sodium bicarbonate, and Waste Isolation Pilot Plant (WIPP) “B” brine solution. Dissolution mechanisms were identified based
EXPERIMENTAL
PROCEDURES
Materials The U02 samples were single-crystal UO, selected from fused U02 previously prepared by PNL in kilogram quantities. The spent-fuel samples, in the form of a fuel fragment, were removed (about February 1977) from light-water reactor spent-fuel bundles discharged from the H. B. Robinson II reactor on June 6, 1974, after an average burnup of 28,000 MWd/MTU. Autoclave Experiments Single-crystal UO, samples weighing about 3 to 6 g were selected for the dissolution experiments in titanium capsules. The UO, sample was placed on a Ti screen holder and a leachant volume (cm”) was selected based on 10 times the surface area (cm2) of the sample. Pure oxygen at different pressures was maintained to obtain dissolved oxygen levels of 200 ppm for each solution at both 75 and 150 “C. After a designated time period, the test solution was transferred so that the uranium content could be analyzed by isotopic dilution/mass spectroscopy. The Ti capsule was then refilled with fresh solution before the test
RECEIVED3/l/82; ACCEPTED3/15/82. Presented at the ORNL Conference on the Leachability of Radioactive Solids, Gatlinburg, Tennessee, December 9-12, 1980. Research supported by the Waste/Rock Interactions Technology Program being conducted by the Pacific Northwest Laboratory. The program is sponsored by the Office of Nuclear Waste Isolation, managed by Battelle Memorial Institute for the U.S. Department of Energy under Contract DE-ACO6-76RL0 1830.
83
R. WANG ANDY. B. KATAYAMA
a4
proceeded. After 11, 30, and 60 d, a small piece of UOZ (about 50 mg) previously placed with the sample in the capsule was removed for surface characterization by scanning electron microscopy (SEM), X-ray diffraction, and elemental analysis.
spent fuel, the contact was made by clamping the spent-fuel sample between the tips of a passivated titanium tweezer. Blank tests have shown that the titanium tweezer holding samples of single-crystal UOZ gives similar experimental results as the conducting epoxy.
Electrochemical Measurements Equipment for electrochemical dissolution studies included a Princeton Applied Research 1751173 potentiostat and programmer, and a Model 350 corrosion measurement system. The electrochemical cell was made of 150-ml Pyrex flasks with a standard calomel reference electrode (SCE) and a graphite counter electrode. A water bath was used for maintaining the test temperature at 75 “C. Electrical contact for the UOZ sample was made with conducting epoxy. For
FIGURE la. oxygen.
RESULTS
Formation of Uranyl Hydrate Coatings We have identified the formation of uranyl hydrate coatings on UO, surfaces under the following conditions: ?? Autoclave dissolution at 150 “C over 11 d in deionized water and NaHCO, solution (Fig. 1);
Formation of uranyl hydrate crystals on the U02 surface after 11 d at 150 “C in deionized water containing 200 ppm of
85
DISSOLUTION MECHANISMS
FIGURE lb. Formation of uranyl hydrate crystals on the UO1 surface after 30 d at 150 “C in 0.03 M sodium bicarbonate solution containing 200 ppm of oxygen. Photos at right show bare UO, surface not covered with uranyl hydrate crystals.
Electrochemical oxidation and dissolution at 0.5 V (SCE) for more than 100 h in deionized water, NaHC03 and WIPP “B” brine solutions (Fig. 2); Dissolution at 25 “C in deionized water containing 50 and 500 ppm of Hz02 (Fig. 3). All these conditions have an abundant supply of oxidation source, such as dissolved oxygen, surface anodization, and Hz02, and a pH between 5 and 9. A thick uranyl hydrate film was not observed in the following conditions: ?? Autoclave dissolution at 75 “C up to 60 d in deionized water, NaHCOJ and WIPP “B” brine solutions. ?? Electrochemical oxidation and dissolution at 25 and 75 “C for less than 24 h.
Apparently, the formation of thick uranyl hydrate films or crystals favors a high temperature, low pH, or an incubating time.
Dissolution Rate The dissolution rates determined from analysis of the uranium content in solutions (including the suspended particulates) from autoclave tests did not reveal the same effect of temperature in the leach rates as shown in Fig. 4. While the uranium contents at 75 and 150 ‘C for each solution were nearly the same, the UOZ surfaces and the container walls exposed at 150 “C clearly, showed thick layers of uranyl hydrate crystals. However, for the 75 “C tests, the uranyl hydrate cyrstals or films were not found. The solution analysis for uranium content in all those tests
R. WANG AND Y. B. KATAYAMA
FIGURE le.
The surface of UO, after 30 d at 150 “C in WIPP “B” brine solution containing 200 ppm of oxygen.
may represent more closely the real dissolution rate of the UO, surfaces. In this case, the highest dissolution rate was for the UOZ surface in NaHC03 solution, nearly 0.5 g m-’ d-l. The dissolution rates for UOZ in deionized water was on the order of 5 x lo+ g m-* d-‘, and in WIPP “B” brine solution the rate averaged 10e3 g m-* d-‘. The dissolution rates determined from electrochemical measurements for these test solutions at both 25 and 75 “C show a high initial dissolution rate on the order of nearly 1 g m-’ d-’ assuming the dissolution current was based on one of the following electrochemical reactions:
uozUOZ + Hz0 -
U@ t 2e,
pHG4,
U02(OH)+ t
H+ t 2e, 4GpHG7,
UOZ t 2Hz0 -
UO,(OH)$ t 2H+ t 2e, pH>7.
The high initial dissolution rate for a clean U02 surface may be due to the absence of a surface film of uranyl hydrate crystals. Later as the result of the buildup of uranyl ions to solubility limits, along with an incubation period, surface films of uranyl hy-
87
FIGURE 2.
Comparisons of the UO, surfaces after electrochemical dissolution. Left: Surface film formed in NaHCO, (0.03 M) solution after 144 h at 0.4 V (SCE); Right: Surface film formed in WIPP “B” brine solution after 120 h at 0.5 V (SCE).
drates will form and the dissolution reduced.
rate will be
Effects of Surface Film on Dissolution Rate Based on polarization behavior of UOZ in three solutions at 25 and 75 “C (shown in Fig. 5), only the surface in the brine solution was passivated to some degree between -0.6 and -0.2 V (SCE). The passivation of the UOZ in brine solution correlates with a low dissolution rate of the UO, samples and the absence of a thick uranyl hydrate coating observed in the autoclave tests. This passive nature of the surface
FIGURE 3a. Formation
films in brine solution is illustrated in Fig. 2 with the U02 surface in NaHCOJ solution. Here the severely corroded surface of UOZ in NaHCO, was due to lack of protecting surface film, whereas the surface of UO, in brine solution was covered with a uniform hydrate film which contained a few cracks probably caused by dehydration after the sample was dried in the SEM viewing chamber. As shown in Fig. 6, the dissolution behavior of spent fuel was similar to that of UOZ, except the spent-fuel surface was not clearly passivated by the brine solution as was observed for the UO, surface.
of thin uranyl hydrate film on the UO, surface with 50 ppm of H,O, in deionized water.
a8
FIGURE 3b. (SCE).
R. WANG ANDY. B. KATAYAMA
Formation of thick uranyl hydrate film on the UO,1surface with 500 ppm of H,OI in deionized water after 26 h at 0.5 V
The dissolution rates of spent fuel, determined from electrochemical measurement, were nearly one order of magnitude less than those of UO1 with clean surfaces. This may be due to the surface of the spent fuel, which was covered with a film in the as-received condition. Correspondingly, the open-circuit potentials for the spent fuel were higher than those of the UO, samples, evidence of surface film, and surface oxidation effects. DISCUSWON
The dissolution mechanisms for single-crystal UO, surfaces in solutions may be considered as a rate-controlled oxidation and phase transformation of U+4
(solid) into U+6(liquid and solid) via the liquid media. The probable oxidation-dissolution steps and reactions are generalized in Fig. 7. A clean UO1 surface exposed to oxidizing species in the solution rapidly oxidized form a thin, oxide layer of UO1+, (Reaction 1). This layer, combined with dissolved oxygen or oxidizing species, rapidly dissolves to form uranyl ions (UOP) or complex ions, depending on the solution chemistry. For deionized water, Reaction 2 (l-3) is postulated; thus, the dissolution of the UO1 surface is strongly dependent on both the available oxygen and H’ ions. The rapid increase of the uranyl ion concentration near the UO1 surface region leads to hydrolysis Reactions 4 and 5 (2) which are dependent on tempera-
89
DISSOLUTION MJXXANISMS 1.5 75% SCAN
RATE
1 mV/rec WIPP
“E u z \,
100
S BRINE
l.O-
e
A . 0 . 0 .
DEiDNlZED WATER 76% DEIONIZED WATER 160% 0.03 0 NaHCO, 76°C 0.03 Y NaHCO. 1SOT WlPP “S” BRINE 76% WlPP “8” BRINE 160°C
-0.6
-
I
-1 .o
70-m
!
I
I 1 cl-s
IO-’
10-7
10.’
CURRENTDENSlNA/cm’
FIGURE Sb. Potentiodynamic crystal UO1 surfaces at 75 “C.
FIGURE 4. Leach rate of single crystal UO1 obtained from solution analysis from the autoclave tests.
ture and pH value. In this case, the concentration of the uranyl ions in solution is then determined by Reactions 4 and 5 rather than Reaction 2. The dissolution of UOz thus leads to the formation of dissolved uranyl ions plus solid uranyl hydrates or complex compounds in the form of suspended particles and surface coating. The dissolution of UO, will increase as a function of temperature because the diffusion of oxygen and H’ ions to the UO, surface will increase. However, the solubility of uranyl ions is inversely related to the temperature (3); thus, most of the dissolution products will end up on the uranyl hydrate deposits as those observed at 150 “C in autoclave experiments. As long as a supply of oxygen is available, Reaction 2 will proceed in the right direction to produce 1.5
25% SCAN
RATE
polarization behaviors for single
more uranyl ions. The excess uranyl ions over the solubility limit will be deposited into the solid uranyl hydrates. It is interesting to note that the presence of H202 will enhance the oxidation-dissolution reaction [4] as well as the hydrolysis reaction [5]. Since the formation of Hz02 is expected from radiolysis of water (6), the dissolution mechanism for UO, may be carried at the UOz-solution interface region even when the solution has reducing or neutral conditions.
CONCLUSIONS The dissolution mechanisms for U02 and spent fuel can be tentatively summarized as follows: 1. The initial dissolution of UO, and spent fuel involves the oxidation of U+4 (solid) to U+6 (liquid) by forming uranyl ions. The oxidationdissolution is a rate-controlled process whose rate will increase as the temperature increases. 2.0
-
HER SPENT FUEL 2S°C, SCAN RATE = 1 mV,sec AREA-, cd
1 m”/sec
1.0-
iii
s
0.6 -
;
! 5
0.0 -
i
-0.6 -
_.-.-._.---r:~:,~_
-1.01
’
I
10-n
10.’ CURRENT
FIGURE 5a. Potentiodynamic crystal UO, surfaces at 25 “C.
I
I
IO-'
IO-'
I
I
10-a
OENB1TV Ai/cm’
polarization behaviors for single
10-a
10-s
10-7 CURRENT
FIGURE 6. Potentiodynamic fuel at 25 “C.
DENSITY
A/cm’
polarization behaviors for a spent
R. WANG AND Y. B. KATAYAMA
(1) Surface
oxidation (several A):
uoz +x/20,
IJO’;
-uo~+X,o
pH< 4, U02(0H)+ + 2H,Oc-UOa
uo z+X + 2H+ •t (1 -x)/202
-
uo 2 +X + H+ + (1 - x)/202 -
H,O+ (1 - x)/202 -
4,
UO,(OH): t
Hz 0--
UOB 2H2 0, ??
7,
Probable
pHa7.f
uo’,”
t 2H20c--UO~(OH),
t 2H+, pH<4,
7.
U02(OH)+ t
H20--U02(OH)2
t
H+, 4
(3) Transport. (4) Hydrolysis-Film Formation, 25-75 “C: 7.
7,
(5) Hydrolysis-Crystal Growth, 150 “C:
UOz (OH)27 pH>
H’,
4
UOz (OH)+, 4
FIGURE
* 2H20 t
UO;2 t H2 0, PHG
2 +x +
* 2H20 t 2H+,
< 1.
(2) Oxidation-dissolution:
uo
+ 3H20 c-uoj
Uo2W-9;
oxidation-dissolution
2. The dissolved uranyl ions will form solids of uranyl hydrates or related complex compounds by hydrolysis reaction which occur more rapidly at high temperature to form large hydrate crystals. 3. The presence of H202, probably from radiolysis of water, enhances the oxidationdissolution of UOz and also increases the formation of hydrate deposits. 4. Dissolution rates for UOz in brine solution are low because the surface was passivated. 5. The dissolution behavior of spent fuel, based on electrochemical measurement, is similar to that of the UO,, except the surface was not clearly passivated in brine solution, as was observed for the UOz surface.
-
mechanisms
UO2 (OH),,
7,
pH>7.
for UO,.
REFERENCES 1. Wang, R. Spent fuel studies progress report: Probably leaching mechanisms for oxidation and dissolution of single-crystal UO, surfaces. PNL-3566, Pacific Northwest Laboratory, Richland, WA (1980). 2. Lemire, R. J. and Tremaine, P. R. AECL-6655 Report and J. Chem. Eng. Data 25361-370 (1980). 3. Holland, H. D. and Brush, L. H. Uranium oxides in ores and spent fuels, Proceedings of the conference on high-level radioactive solid waste forms, December 19-21, 1978 Denver, CO. NUREG/CP-005 (1979); L. H. Brush, Ph.D. Thesis, Harvard University (1980). 4. Hiskey, J. B. Hydrogen peroxide leaching or uranium oxide in carbonate solutions, presented at Symposium of Hydrogen Peroxide ASM Meetings, Las Vegas, NV (1980). 5. Dawson, J. K., Wait, E., Alcock, K., and Chilton, D. R. Some aspects of the system uranium trioxide-water. J. Chem. Sot. 3531 (1956). 6. Burns, W. G. and Moore, P. B. Radiation Effects 30: 233 (1976).
WASTE MANAGEMENT,
0191-815X/82/020083-08$03.00/0 1982 Pergamon Press Ltd.
Vol. 3, pp. 83-90,1982
Printed in the USA.
DISSOLUTION MECHANISMS FOR UOz AND SPENT FUEL R. Wang Y. B. Katayama Pacific Northwest Laboratory, Richland, Washington 99352, USA
ABSTRACT. The probable dissolution mechanisms for UOz and spent fuel were studied based on accelerating oxidation and dissolutions in autoclave and electrochemical experiments. The dissolution of UO, is primarily due to oxidation of U++’ (solid) to U6 (liquid). The U+6(liquid) is initially in dissolved uranyl ions, UOf2, but subsequent hydrolysis reactions lead to the formation of solid uranyl hydrates (or related complex compounds) that deposit onto the surfaces of U02 and container walls as thin or thick coatings, depending on the temperature, pH, and time. Therefore, the dissolution rate of U02 cannot be simply determined from the uranium content of the solution. Moreover, the dissolution rate is not limited by the solubility of the uranyl ions. Electrochemical dissolution of spent fuel shows that the initial dissolution and leaching behavior is similar to that of UO, surfaces.
on accelerated oxidation and dissolution in autoclave static and electrochemical dissolution experiments. Electrochemical methods were also applied to spent fuel in the three solutions to compare with the behavior of the UO, surfaces.
INTRODUCTION Understanding the dissolution and leaching mechanisms of spent fuel is important for predicting, via mathematical models, the long-term stability of spent fuel in conditions of a geologic repository. The complex structural, microstructural, compositional, and thermal characteristics of spent fuel, along with the lack of radiation-shielded analytical tools, has necessitated a more simple, systematic approach for understanding the leaching and release of radionuclides in aqueous media. Such an approach was developed at the Pacific Northwest Laboratory (PNL) by studying, in initial experiments, a singlecrystal U02 as representative of the spent-fuel matrix (1). Oxidation and dissolution experiments for U02 were designed to accommodate ready experimental verification of the same mechanisms in spent fuel. This paper presents the complex oxidation and dissolution of U02 in three aqueous solutions: deionized water, 0.03 M sodium bicarbonate, and Waste Isolation Pilot Plant (WIPP) “B” brine solution. Dissolution mechanisms were identified based
EXPERIMENTAL
PROCEDURES
Materials The U02 samples were single-crystal UO, selected from fused U02 previously prepared by PNL in kilogram quantities. The spent-fuel samples, in the form of a fuel fragment, were removed (about February 1977) from light-water reactor spent-fuel bundles discharged from the H. B. Robinson II reactor on June 6, 1974, after an average burnup of 28,000 MWd/MTU. Autoclave Experiments Single-crystal UO, samples weighing about 3 to 6 g were selected for the dissolution experiments in titanium capsules. The UO, sample was placed on a Ti screen holder and a leachant volume (cm”) was selected based on 10 times the surface area (cm2) of the sample. Pure oxygen at different pressures was maintained to obtain dissolved oxygen levels of 200 ppm for each solution at both 75 and 150 “C. After a designated time period, the test solution was transferred so that the uranium content could be analyzed by isotopic dilution/mass spectroscopy. The Ti capsule was then refilled with fresh solution before the test
RECEIVED3/l/82; ACCEPTED3/15/82. Presented at the ORNL Conference on the Leachability of Radioactive Solids, Gatlinburg, Tennessee, December 9-12, 1980. Research supported by the Waste/Rock Interactions Technology Program being conducted by the Pacific Northwest Laboratory. The program is sponsored by the Office of Nuclear Waste Isolation, managed by Battelle Memorial Institute for the U.S. Department of Energy under Contract DE-ACO6-76RL0 1830.
83
R. WANG ANDY. B. KATAYAMA
a4
proceeded. After 11, 30, and 60 d, a small piece of UOZ (about 50 mg) previously placed with the sample in the capsule was removed for surface characterization by scanning electron microscopy (SEM), X-ray diffraction, and elemental analysis.
spent fuel, the contact was made by clamping the spent-fuel sample between the tips of a passivated titanium tweezer. Blank tests have shown that the titanium tweezer holding samples of single-crystal UOZ gives similar experimental results as the conducting epoxy.
Electrochemical Measurements Equipment for electrochemical dissolution studies included a Princeton Applied Research 1751173 potentiostat and programmer, and a Model 350 corrosion measurement system. The electrochemical cell was made of 150-ml Pyrex flasks with a standard calomel reference electrode (SCE) and a graphite counter electrode. A water bath was used for maintaining the test temperature at 75 “C. Electrical contact for the UOZ sample was made with conducting epoxy. For
FIGURE la. oxygen.
RESULTS
Formation of Uranyl Hydrate Coatings We have identified the formation of uranyl hydrate coatings on UO, surfaces under the following conditions: ?? Autoclave dissolution at 150 “C over 11 d in deionized water and NaHCO, solution (Fig. 1);
Formation of uranyl hydrate crystals on the U02 surface after 11 d at 150 “C in deionized water containing 200 ppm of
85
DISSOLUTION MECHANISMS
FIGURE lb. Formation of uranyl hydrate crystals on the UO1 surface after 30 d at 150 “C in 0.03 M sodium bicarbonate solution containing 200 ppm of oxygen. Photos at right show bare UO, surface not covered with uranyl hydrate crystals.
Electrochemical oxidation and dissolution at 0.5 V (SCE) for more than 100 h in deionized water, NaHC03 and WIPP “B” brine solutions (Fig. 2); Dissolution at 25 “C in deionized water containing 50 and 500 ppm of Hz02 (Fig. 3). All these conditions have an abundant supply of oxidation source, such as dissolved oxygen, surface anodization, and Hz02, and a pH between 5 and 9. A thick uranyl hydrate film was not observed in the following conditions: ?? Autoclave dissolution at 75 “C up to 60 d in deionized water, NaHCOJ and WIPP “B” brine solutions. ?? Electrochemical oxidation and dissolution at 25 and 75 “C for less than 24 h.
Apparently, the formation of thick uranyl hydrate films or crystals favors a high temperature, low pH, or an incubating time.
Dissolution Rate The dissolution rates determined from analysis of the uranium content in solutions (including the suspended particulates) from autoclave tests did not reveal the same effect of temperature in the leach rates as shown in Fig. 4. While the uranium contents at 75 and 150 ‘C for each solution were nearly the same, the UOZ surfaces and the container walls exposed at 150 “C clearly, showed thick layers of uranyl hydrate crystals. However, for the 75 “C tests, the uranyl hydrate cyrstals or films were not found. The solution analysis for uranium content in all those tests
R. WANG AND Y. B. KATAYAMA
FIGURE le.
The surface of UO, after 30 d at 150 “C in WIPP “B” brine solution containing 200 ppm of oxygen.
may represent more closely the real dissolution rate of the UO, surfaces. In this case, the highest dissolution rate was for the UOZ surface in NaHC03 solution, nearly 0.5 g m-’ d-l. The dissolution rates for UOZ in deionized water was on the order of 5 x lo+ g m-* d-‘, and in WIPP “B” brine solution the rate averaged 10e3 g m-* d-‘. The dissolution rates determined from electrochemical measurements for these test solutions at both 25 and 75 “C show a high initial dissolution rate on the order of nearly 1 g m-’ d-’ assuming the dissolution current was based on one of the following electrochemical reactions:
uozUOZ + Hz0 -
U@ t 2e,
pHG4,
U02(OH)+ t
H+ t 2e, 4GpHG7,
UOZ t 2Hz0 -
UO,(OH)$ t 2H+ t 2e, pH>7.
The high initial dissolution rate for a clean U02 surface may be due to the absence of a surface film of uranyl hydrate crystals. Later as the result of the buildup of uranyl ions to solubility limits, along with an incubation period, surface films of uranyl hy-
87
FIGURE 2.
Comparisons of the UO, surfaces after electrochemical dissolution. Left: Surface film formed in NaHCO, (0.03 M) solution after 144 h at 0.4 V (SCE); Right: Surface film formed in WIPP “B” brine solution after 120 h at 0.5 V (SCE).
drates will form and the dissolution reduced.
rate will be
Effects of Surface Film on Dissolution Rate Based on polarization behavior of UOZ in three solutions at 25 and 75 “C (shown in Fig. 5), only the surface in the brine solution was passivated to some degree between -0.6 and -0.2 V (SCE). The passivation of the UOZ in brine solution correlates with a low dissolution rate of the UO, samples and the absence of a thick uranyl hydrate coating observed in the autoclave tests. This passive nature of the surface
FIGURE 3a. Formation
films in brine solution is illustrated in Fig. 2 with the U02 surface in NaHCOJ solution. Here the severely corroded surface of UOZ in NaHCO, was due to lack of protecting surface film, whereas the surface of UO, in brine solution was covered with a uniform hydrate film which contained a few cracks probably caused by dehydration after the sample was dried in the SEM viewing chamber. As shown in Fig. 6, the dissolution behavior of spent fuel was similar to that of UOZ, except the spent-fuel surface was not clearly passivated by the brine solution as was observed for the UO, surface.
of thin uranyl hydrate film on the UO, surface with 50 ppm of H,O, in deionized water.
a8
FIGURE 3b. (SCE).
R. WANG ANDY. B. KATAYAMA
Formation of thick uranyl hydrate film on the UO,1surface with 500 ppm of H,OI in deionized water after 26 h at 0.5 V
The dissolution rates of spent fuel, determined from electrochemical measurement, were nearly one order of magnitude less than those of UO1 with clean surfaces. This may be due to the surface of the spent fuel, which was covered with a film in the as-received condition. Correspondingly, the open-circuit potentials for the spent fuel were higher than those of the UO, samples, evidence of surface film, and surface oxidation effects. DISCUSWON
The dissolution mechanisms for single-crystal UO, surfaces in solutions may be considered as a rate-controlled oxidation and phase transformation of U+4
(solid) into U+6(liquid and solid) via the liquid media. The probable oxidation-dissolution steps and reactions are generalized in Fig. 7. A clean UO1 surface exposed to oxidizing species in the solution rapidly oxidized form a thin, oxide layer of UO1+, (Reaction 1). This layer, combined with dissolved oxygen or oxidizing species, rapidly dissolves to form uranyl ions (UOP) or complex ions, depending on the solution chemistry. For deionized water, Reaction 2 (l-3) is postulated; thus, the dissolution of the UO1 surface is strongly dependent on both the available oxygen and H’ ions. The rapid increase of the uranyl ion concentration near the UO1 surface region leads to hydrolysis Reactions 4 and 5 (2) which are dependent on tempera-
89
DISSOLUTION MJXXANISMS 1.5 75% SCAN
RATE
1 mV/rec WIPP
“E u z \,
100
S BRINE
l.O-
e
A . 0 . 0 .
DEiDNlZED WATER 76% DEIONIZED WATER 160% 0.03 0 NaHCO, 76°C 0.03 Y NaHCO. 1SOT WlPP “S” BRINE 76% WlPP “8” BRINE 160°C
-0.6
-
I
-1 .o
70-m
!
I
I 1 cl-s
IO-’
10-7
10.’
CURRENTDENSlNA/cm’
FIGURE Sb. Potentiodynamic crystal UO1 surfaces at 75 “C.
FIGURE 4. Leach rate of single crystal UO1 obtained from solution analysis from the autoclave tests.
ture and pH value. In this case, the concentration of the uranyl ions in solution is then determined by Reactions 4 and 5 rather than Reaction 2. The dissolution of UOz thus leads to the formation of dissolved uranyl ions plus solid uranyl hydrates or complex compounds in the form of suspended particles and surface coating. The dissolution of UO, will increase as a function of temperature because the diffusion of oxygen and H’ ions to the UO, surface will increase. However, the solubility of uranyl ions is inversely related to the temperature (3); thus, most of the dissolution products will end up on the uranyl hydrate deposits as those observed at 150 “C in autoclave experiments. As long as a supply of oxygen is available, Reaction 2 will proceed in the right direction to produce 1.5
25% SCAN
RATE
polarization behaviors for single
more uranyl ions. The excess uranyl ions over the solubility limit will be deposited into the solid uranyl hydrates. It is interesting to note that the presence of H202 will enhance the oxidation-dissolution reaction [4] as well as the hydrolysis reaction [5]. Since the formation of Hz02 is expected from radiolysis of water (6), the dissolution mechanism for UO, may be carried at the UOz-solution interface region even when the solution has reducing or neutral conditions.
CONCLUSIONS The dissolution mechanisms for U02 and spent fuel can be tentatively summarized as follows: 1. The initial dissolution of UO, and spent fuel involves the oxidation of U+4 (solid) to U+6 (liquid) by forming uranyl ions. The oxidationdissolution is a rate-controlled process whose rate will increase as the temperature increases. 2.0
-
HER SPENT FUEL 2S°C, SCAN RATE = 1 mV,sec AREA-, cd
1 m”/sec
1.0-
iii
s
0.6 -
;
! 5
0.0 -
i
-0.6 -
_.-.-._.---r:~:,~_
-1.01
’
I
10-n
10.’ CURRENT
FIGURE 5a. Potentiodynamic crystal UO, surfaces at 25 “C.
I
I
IO-'
IO-'
I
I
10-a
OENB1TV Ai/cm’
polarization behaviors for single
10-a
10-s
10-7 CURRENT
FIGURE 6. Potentiodynamic fuel at 25 “C.
DENSITY
A/cm’
polarization behaviors for a spent
R. WANG AND Y. B. KATAYAMA
(1) Surface
oxidation (several A):
uoz +x/20,
IJO’;
-uo~+X,o
pH< 4, U02(0H)+ + 2H,Oc-UOa
uo z+X + 2H+ •t (1 -x)/202
-
uo 2 +X + H+ + (1 - x)/202 -
H,O+ (1 - x)/202 -
4,
UO,(OH): t
Hz 0--
UOB 2H2 0, ??
7,
Probable
pHa7.f
uo’,”
t 2H20c--UO~(OH),
t 2H+, pH<4,
7.
U02(OH)+ t
H20--U02(OH)2
t
H+, 4
(3) Transport. (4) Hydrolysis-Film Formation, 25-75 “C: 7.
7,
(5) Hydrolysis-Crystal Growth, 150 “C:
UOz (OH)27 pH>
H’,
4
UOz (OH)+, 4
FIGURE
* 2H20 t
UO;2 t H2 0, PHG
2 +x +
* 2H20 t 2H+,
< 1.
(2) Oxidation-dissolution:
uo
+ 3H20 c-uoj
Uo2W-9;
oxidation-dissolution
2. The dissolved uranyl ions will form solids of uranyl hydrates or related complex compounds by hydrolysis reaction which occur more rapidly at high temperature to form large hydrate crystals. 3. The presence of H202, probably from radiolysis of water, enhances the oxidationdissolution of UOz and also increases the formation of hydrate deposits. 4. Dissolution rates for UOz in brine solution are low because the surface was passivated. 5. The dissolution behavior of spent fuel, based on electrochemical measurement, is similar to that of the UO,, except the surface was not clearly passivated in brine solution, as was observed for the UOz surface.
-
mechanisms
UO2 (OH),,
7,
pH>7.
for UO,.
REFERENCES 1. Wang, R. Spent fuel studies progress report: Probably leaching mechanisms for oxidation and dissolution of single-crystal UO, surfaces. PNL-3566, Pacific Northwest Laboratory, Richland, WA (1980). 2. Lemire, R. J. and Tremaine, P. R. AECL-6655 Report and J. Chem. Eng. Data 25361-370 (1980). 3. Holland, H. D. and Brush, L. H. Uranium oxides in ores and spent fuels, Proceedings of the conference on high-level radioactive solid waste forms, December 19-21, 1978 Denver, CO. NUREG/CP-005 (1979); L. H. Brush, Ph.D. Thesis, Harvard University (1980). 4. Hiskey, J. B. Hydrogen peroxide leaching or uranium oxide in carbonate solutions, presented at Symposium of Hydrogen Peroxide ASM Meetings, Las Vegas, NV (1980). 5. Dawson, J. K., Wait, E., Alcock, K., and Chilton, D. R. Some aspects of the system uranium trioxide-water. J. Chem. Sot. 3531 (1956). 6. Burns, W. G. and Moore, P. B. Radiation Effects 30: 233 (1976).