- Email: [email protected]
Physical properties and leaching behaviour of spent fuel BISO coated particles
Progress in Nuclear Energy 57 (2012) 161e164
Contents lists available at SciVerse ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Physical properties and leaching behaviour of spent fuel BISO coated particles A. Bukaemskiy a, *, J. Fachinger b, D. Bosbach a a b
Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research - Safety Research and Reactor Technology (IEK-6), Germany Furnaces Nuclear Applications Grenoble S.A.S. Zweigniederlassung Hanau, Wilhelm-Rohn-Str. 35, D-63450 Hanau, Germany
a r t i c l e i n f o
a b s t r a c t
Article history: Received 4 July 2011 Accepted 8 September 2011
The safe disposal in a geological repository is proposed for the spent fuel elements obtained from operation of High Temperature Reactors. The behavior of the fuel particles under disposal conditions is a key question for the long-term nuclear waste disposal. In the present work, the spent fuel BISO coated particles, which have been irradiated to a burn-up of 10% FIMA, were studied. The size and morphological characteristics of the coated particles were investigated by the using of optical and SEM microscopy. The distribution of the 137Cs amount in the coated particle was studied in detail. It was shown the activity was concentrated mainly inside the kernels and in the carbon buffer layer, while the outside carbon layer contained 0.1% of the total 137Cs only. Further, the thoria-based (Th0.834U0.166)O2 kernels were mechanically isolated from the coated particles, and their solution behavior was studied using the flow through experiments. In all experiments the average flow rate was w7e8 ml/day. Dissolution of irradiated and unirradiated kernels in HCl solution with the different value of pH (from 0 to 5) was investigated at the temperatures 90, 55 and 20 C. The amounts of the radionuclide leached in solutions were determined by ACP-MS, g- und a-spectrometry. On the basis of the obtained results the important leaching characteristics such as the normalized leaching rate, the activation energy value for the release of the different radionuclides were calculated. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Nuclear reactor materials Coated particle Flow through experiment Normalized leaching rate Activation energy
1. Introduction The safe disposal in a geological repository is proposed for the spent fuel elements obtained from operation of High Temperature Reactors. The direct disposal of spent fuel in deep geological formations (salt domes, clay or granite rocks) is considered one of the most attractive and economically efficient methods of utilizing spent HTR fuel (Kirch et al., 1990; De Las Cuevas and Pueyo, 1995). This disposal strategy enables to isolate the radionuclide for a time period long enough for their decay to activity level at which the incorporation into the environment can be considered nonhazardous. The fuel elements for the German HTR design consist of small coated particles, which are embedded in graphite. In the coated particle, the kernel from uranium or mixed uranium-thorium dioxide is surrounded by the different pyrocarbon and silicon carbide layers (Nickel et al., 2002). The behaviour of an HTR disposed spent fuel in a deep repository is based on the integrity of the coated particles (Alliot et al., 2005). As long as the particles coated remain intact, no radioactivity can be leached out from the fuel material, since the coatings have an excellent long-term
* Corresponding author. Tel.: þ49 2461 614574; fax: þ49 2461 612450. E-mail address: [email protected] (A. Bukaemskiy). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2011.10.001
chemical resistance (Schenk et al., 1984). Therefore, the rate of the radionuclide release is determined by the amount of broken coated particles. The behaviour of the fuel particles under disposal conditions is a key question for the long-term nuclear waste disposal. The aim of our work consists in the detail investigations of the spent fuel particles properties, which are important in the safety disposal point of view, namely morphological characteristics and leaching behaviour in aquatic medium. Caesium is one of the main fission products in radioactive wastes. Due to the high vapor pressure, the gaseous Cs releases from the fuel, fills the pores of the fuel and of the buffer layer, forming interstitial compounds with the pyrocarbon (Schenk et al., 1984). Therefore, the main part of our work is devoted to the detailed study of the Cs distribution in the spent coated particles and its release from the fuel kernels during the leaching. 2. Experimental procedure In the present work we investigated the three spent fuel BISO Coated Particles, which have been irradiated for 182 days to a burnup of w10% FIMA at an irradiation temperature between 990 C and 1180 C. Decay time since the end of irradiation was about 26 years. The investigated fuel BISO coated particle consists of the ceramic
162
A. Bukaemskiy et al. / Progress in Nuclear Energy 57 (2012) 161e164
and each carbon layers were measured. Moreover, 137Cs content in the near-surface layer of kernel was also estimated on the base of leaching experiments. The size and morphological characteristics of the spent BISO coated particles and kernels before and after leaching were investigated using optical microscopy. The dissolution processes of the irradiated and unirradiated (Th0.834U0.166)O2 kernels as the control samples were studied by using the flow through experiments (Schenk et al., 1984; Bruno et al., 1991). The schematic diagram of one canal of the experimental set-up is shown in Fig. 2. The flow-through reactor was made from tantalum thin tube (1 mm diameter). One irradiated or ten unirradiated kernels were put into middle part of reactor. These kernels were leached with HCl different molarities (pH from 0 to 5) at temperatures 90, 55 and 20 C. An overview of the experiments carried out is presented as the Table in Fig. 2. At each experimental condition (values of pH and T), the leaching process was performed for about one month. Two or three times in week, the solution from sample collector was taken for measuring of the current concentrations of radionuclides by ICPMS, g- und a-spectrometry. The solutions for the leaching experiments were degasified prior to experiments in vacuum and then argon-saturated by bubbling. All experiments were conducted under argon atmosphere to simulate the anaerobic conditions of the final repository. The average flow rate was w7e8 mL/day.
Table 1 Radionuclide inventory of the fuel kernel. Radionuclide
Dose, Bq
137
1.9*106 510 1042 14.3 1530 469 11662 326 30
Cs 134 Cs 241 Am 236 U 233 U 232 U 238 Pu 239/240 Pu 243/244 Cm
fuel (Th0.834U0.166)O2 kernel which is surrounded by a porous carbon buffer and outer dense pyrocarbon layer. For the radionuclide inventory determination, one kernel was mechanically isolated from coatings and completely dissolved in Thorex reagent (mixture of 13 M nitric acid, 0.1 M aluminium nitrate and 0.05 M hydrofluoric acid). The content of the main radionuclide in (Th0.834U0.166)O2 kernel was measured by a - and g -spectrometry, Table 1. The analysis of the kernel radionuclide inventories has shown 137Cs to be the main contributor to the total particle radioactivity. Other radionuclides are present in kernel in lower amounts. Two rest coated particles were mechanically broken step-bystep up to the porous carbon buffer (1 step) and up to the kernel (2 step), Fig. 1a, b, c. After these procedures, 137Cs contents in kernel
Fig. 1. Optical images of: (a, d) - irradiated BISO coated particle: (b, e) - kernel with buffer: (c, f) - (Th0.834U0.166)O2.kernel.
Thermocontrol
Argon
T
o
Flow
T, C 90 55 20
pH=0 + + +
pH=1 + ---
pH=2 + + +
pH=3 + ---
pH=4 + + +
pH=5 + ---
Flow-through reactor
Feeding solution
Oven
Peristaltic pump
Fig. 2. Experimental device for flow-through leaching experiments.
Sample collector
A. Bukaemskiy et al. / Progress in Nuclear Energy 57 (2012) 161e164
163
Fig. 3. Optical images of surfaces of (Th0.834U0.166)O2 kernels before irradiation (a, d), after irradiation (b, e) and after leaching (c, f).
(1)
where Ea is activation energy, A(pH) e empirical parameter depending on pH. For the l137Cs leaching process the value of Ea
a
pH=0 T=90C pH=1 T=90C
-2
pH=2 T=90C pH=2 T=55C
log(NLR), mol m -2day -1
All investigated coated particles before crushing had intact outer pyrocarbon surface without cracks and visible damages, Fig. 1 d. The coated particle diameter equals 760 mm. Moreover, the surface of the porous carbon buffer remained also intact after soft breaking, Fig. 1e. The kernel is black particle having a regular spherical shape, the diameter equals 430 mm. The coating thicknesses, which were determined from the optical microscopy photographs equal 90 and 75 mm for porous carbon buffer and outer dense pyrocarbon layer, respectively. The surfaces of kernels before irradiation (control sample), after irradiation and after leaching are shown in Fig. 3. All investigated kernels have a dense structure with well-formed grains and grain boundaries. The long-term irradiation led to the significant increasing of the grain boundary width, but the visible damages were not observed. After the following leaching the surface of the kernels almost remained without changes. The spent coated particle was accurately crashed in layers, and the 137Cs amount in each layer was measured. It was shown that the activity is mainly concentrated in the inner part of kernel (w77% of total caesium content), on the surface of kernel (w12%) and in the carbon buffer layer (w11%), whereas the outside carbon layer contains 0.1% of the total 137Cs only, that indicates its high protective characteristic (Schenk et al., 1984). The representative kinetic dependencies of fuel kernels leaching in hydrochloric acid are presented in Fig. 4a. The estimated normalized leaching rate (NLR) do not depend on time with the exception of the leaching at pH ¼ 0 and 90 C. The experiment at these conditions was carried out at the beginning; therefore the 137 Cs release had the highest value during first ten days. This value of 137Cs fracture release (12% of total content) can be referred to the near-surface layer. The normalized leaching rate depends on the values of pH and temperature as shown in Fig. 4b. NLR decreased significantly with increasing of pH and decreasing of temperature T. The NLR data are summarised in Table 2. In order to estimate the activation energy of the leaching process the NLR values can be plotted as function on the reciprocal temperature in so called Arrhenius coordinates (Fig. 5). The linear dependencies are observed for the fixed values of pH and can be described using the following equation:
Ea NLR ¼ AðpHÞexp RT
-3
pH=2 T=20C pH=5 T=90C
-4
-5
-6
-7
-8 0
5
10
15
20
25
30
Reaction time, days
b -4
log(NLR),mol*m -2*day -1
3. Results and discussion
90C 55C 20C
-5
-6
-7
-8 0
1
2
3
4
5
pH
Fig. 4. (a) 137Cs normalized leaching rates from (Th0.834U0.166)O2 fuel kernels in HCl solutions at different pH and temperature conditions and (b) 137Cs normalized leaching rates from (Th0.834U0.166)O2 fuel kernels in HCl solution versus pH (lines e Eq. (1)).
164
A. Bukaemskiy et al. / Progress in Nuclear Energy 57 (2012) 161e164
Table 2 137 Cs normalized leaching rates (NLR, mol*m2*d1)
Table 3 Normalized leaching rates (NLR, mol*m2*d1) and activation energy (Ea, kJ/mol).
T, C
pH ¼ 0
pH ¼ 1
pH ¼ 2
pH ¼ 3
pH ¼ 4
pH ¼ 5
T, C
137
90 55 20
1.05E-04 2.54E-05 1.27E-06
3.03E-05 e e
6.77E-06 7.97E-07 1.20E-07
4.73E-06 e e
2.46E-06 2.68E-07 4.41E-08
7.76E-07 e e
90 55 20 Ea
1.05E-04 2.54E-05 1.27E-06 57.5
Cs
238
Pu
5.1E-05 e e e
Th
U
Th*
U*
8.1E-05 1.6E-05 4.5E-06 40.0
2.5E-02 2.2E-02 1.2E-02 9.8
2.4E-06 6.3E-07 2.7E-07 30.1
2.2E-03 2.3E-03 1.1E-03 9.4
* - unirradiated kernels.
-8
component in these phases and the dissolution rate of Th is much lower. According to the obtained data the activation energy of the Th dissolution was very close to that for Cs, while the estimated value for U was four times lower.
ln(NLR), mol*m -2*day -1
pH=0
-10
pH=2 pH=4
-12
4. Conclusion -14
-16
-18 0.0027
-1
0.0029
0.0031
0.0033 1/T, K
Fig. 5. 137Cs normalized leaching rates from (Th0.834U0.166)O2 fuel kernels in HCl solution versus T1.
equals 57.5 3.5 kJ/mol, empirical parameter can be expressed as following:
AðpHÞ ¼ expð9:74 0:98pHÞ
(2)
Using the obtained data on Ea and NLR0, the value of the normalized leaching rate can be re-calculated in the form of the lineal dependencies, which are depicted in Fig. 4b as lines. Moreover, the release of other radionuclides was investigated at pH ¼ 0 for irradiated and unradiated kernels. The values of the normalised leaching rate and of the estimated parameters, Ea and A(pH), for these radionuclides are given in Table 3. The dissolution rate of the irradiated kernels was higher in ten times approximately in comparison with the values for the unirradiated kernels. It can be noted that the normalized leaching rate for the Cs, Th and Pu are comparable with each other, while the value of NLR for U is significantly lower. The similar ratio between the normalized leaching rates for U and Th remains for the unradiated kernels too. The obtained results are in agreement with the literature data (Alliot et al., 2005). In this work it has been shown that the leaching rate of U from the thorium-rich mixed oxide matrix seems to be similar to the dissolution rate of pure UO2 even though U is a minor
In the present work, the spent fuel BISO coated particles, which have been irradiated to a burn-up of 10% FIMA, were studied. The size, morphological characteristics of the coated particles, the distribution of the 137Cs amount in the coated particle were investigated in detail. It was shown the activity was concentrated mainly inside the kernels and in the carbon buffer layer, while the outside carbon layer contained 0.1% of the total 137Cs only. The thoria-based (Th0.834U0.166)O2 kernels were mechanically isolated from the coated particles, and their solution behavior was studied using the flow through experiments. The normalized leaching rate (NLR) of 137Cs from the spent fuel kernels was determined at the different pH and temperatures. The activation energy for the 137Cs release was estimated from the experimental results plotted in Arrhenius coordinates. At present experimental conditions, the normalized leaching rate can be expressed as a linear function of the pH-value. References Alliot, C., Grambow, B., Landesman, C., 2005. Leaching behaviour of unirradiated high temperature reactor (HTR) UO2eThO2 mixed oxides fuel particles. J. Nucl. Mat. 346, 32e39. Bruno, J., Casas, I., Puigdomenech, I., 1991. The kinetics of dissolution of UO2 under reducing conditions and the influence of an oxidized surface layer (UO2þx): application of a continuous flow-through reactor. Geochim. Cosmochim. Acta 52, 647e658. De Las Cuevas, C., Pueyo, J.J., 1995. The influence of mineralogy and texture in the water content of rock salt formations. Its implication in radioactive waste disposal. Appl. Geochem. 10 (3), 317. Kirch, N., Brinkmann, H.U., Brücher, P.H., 1990. Storage and final disposal of spent HTR fuel in the Federal Republic of Germany. Nucl. Eng. Des. 121, 241e248. Nickel, H., Nabielek, H., Pott, G., Mehner, A.W., 2002. Long time experience with the development of HTR fuel elements in Germany. Nucl. Eng. Des. 217, 141e151. Schenk, W., Naoumidis, A., Nickel, H., 1984. The behaviour of spherical HTR fuel elements under accident conditions. J. Nucl. Mat. 124, 25e32.
Contents lists available at SciVerse ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Physical properties and leaching behaviour of spent fuel BISO coated particles A. Bukaemskiy a, *, J. Fachinger b, D. Bosbach a a b
Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research - Safety Research and Reactor Technology (IEK-6), Germany Furnaces Nuclear Applications Grenoble S.A.S. Zweigniederlassung Hanau, Wilhelm-Rohn-Str. 35, D-63450 Hanau, Germany
a r t i c l e i n f o
a b s t r a c t
Article history: Received 4 July 2011 Accepted 8 September 2011
The safe disposal in a geological repository is proposed for the spent fuel elements obtained from operation of High Temperature Reactors. The behavior of the fuel particles under disposal conditions is a key question for the long-term nuclear waste disposal. In the present work, the spent fuel BISO coated particles, which have been irradiated to a burn-up of 10% FIMA, were studied. The size and morphological characteristics of the coated particles were investigated by the using of optical and SEM microscopy. The distribution of the 137Cs amount in the coated particle was studied in detail. It was shown the activity was concentrated mainly inside the kernels and in the carbon buffer layer, while the outside carbon layer contained 0.1% of the total 137Cs only. Further, the thoria-based (Th0.834U0.166)O2 kernels were mechanically isolated from the coated particles, and their solution behavior was studied using the flow through experiments. In all experiments the average flow rate was w7e8 ml/day. Dissolution of irradiated and unirradiated kernels in HCl solution with the different value of pH (from 0 to 5) was investigated at the temperatures 90, 55 and 20 C. The amounts of the radionuclide leached in solutions were determined by ACP-MS, g- und a-spectrometry. On the basis of the obtained results the important leaching characteristics such as the normalized leaching rate, the activation energy value for the release of the different radionuclides were calculated. Ó 2011 Elsevier Ltd. All rights reserved.
Keywords: Nuclear reactor materials Coated particle Flow through experiment Normalized leaching rate Activation energy
1. Introduction The safe disposal in a geological repository is proposed for the spent fuel elements obtained from operation of High Temperature Reactors. The direct disposal of spent fuel in deep geological formations (salt domes, clay or granite rocks) is considered one of the most attractive and economically efficient methods of utilizing spent HTR fuel (Kirch et al., 1990; De Las Cuevas and Pueyo, 1995). This disposal strategy enables to isolate the radionuclide for a time period long enough for their decay to activity level at which the incorporation into the environment can be considered nonhazardous. The fuel elements for the German HTR design consist of small coated particles, which are embedded in graphite. In the coated particle, the kernel from uranium or mixed uranium-thorium dioxide is surrounded by the different pyrocarbon and silicon carbide layers (Nickel et al., 2002). The behaviour of an HTR disposed spent fuel in a deep repository is based on the integrity of the coated particles (Alliot et al., 2005). As long as the particles coated remain intact, no radioactivity can be leached out from the fuel material, since the coatings have an excellent long-term
* Corresponding author. Tel.: þ49 2461 614574; fax: þ49 2461 612450. E-mail address: [email protected] (A. Bukaemskiy). 0149-1970/$ e see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2011.10.001
chemical resistance (Schenk et al., 1984). Therefore, the rate of the radionuclide release is determined by the amount of broken coated particles. The behaviour of the fuel particles under disposal conditions is a key question for the long-term nuclear waste disposal. The aim of our work consists in the detail investigations of the spent fuel particles properties, which are important in the safety disposal point of view, namely morphological characteristics and leaching behaviour in aquatic medium. Caesium is one of the main fission products in radioactive wastes. Due to the high vapor pressure, the gaseous Cs releases from the fuel, fills the pores of the fuel and of the buffer layer, forming interstitial compounds with the pyrocarbon (Schenk et al., 1984). Therefore, the main part of our work is devoted to the detailed study of the Cs distribution in the spent coated particles and its release from the fuel kernels during the leaching. 2. Experimental procedure In the present work we investigated the three spent fuel BISO Coated Particles, which have been irradiated for 182 days to a burnup of w10% FIMA at an irradiation temperature between 990 C and 1180 C. Decay time since the end of irradiation was about 26 years. The investigated fuel BISO coated particle consists of the ceramic
162
A. Bukaemskiy et al. / Progress in Nuclear Energy 57 (2012) 161e164
and each carbon layers were measured. Moreover, 137Cs content in the near-surface layer of kernel was also estimated on the base of leaching experiments. The size and morphological characteristics of the spent BISO coated particles and kernels before and after leaching were investigated using optical microscopy. The dissolution processes of the irradiated and unirradiated (Th0.834U0.166)O2 kernels as the control samples were studied by using the flow through experiments (Schenk et al., 1984; Bruno et al., 1991). The schematic diagram of one canal of the experimental set-up is shown in Fig. 2. The flow-through reactor was made from tantalum thin tube (1 mm diameter). One irradiated or ten unirradiated kernels were put into middle part of reactor. These kernels were leached with HCl different molarities (pH from 0 to 5) at temperatures 90, 55 and 20 C. An overview of the experiments carried out is presented as the Table in Fig. 2. At each experimental condition (values of pH and T), the leaching process was performed for about one month. Two or three times in week, the solution from sample collector was taken for measuring of the current concentrations of radionuclides by ICPMS, g- und a-spectrometry. The solutions for the leaching experiments were degasified prior to experiments in vacuum and then argon-saturated by bubbling. All experiments were conducted under argon atmosphere to simulate the anaerobic conditions of the final repository. The average flow rate was w7e8 mL/day.
Table 1 Radionuclide inventory of the fuel kernel. Radionuclide
Dose, Bq
137
1.9*106 510 1042 14.3 1530 469 11662 326 30
Cs 134 Cs 241 Am 236 U 233 U 232 U 238 Pu 239/240 Pu 243/244 Cm
fuel (Th0.834U0.166)O2 kernel which is surrounded by a porous carbon buffer and outer dense pyrocarbon layer. For the radionuclide inventory determination, one kernel was mechanically isolated from coatings and completely dissolved in Thorex reagent (mixture of 13 M nitric acid, 0.1 M aluminium nitrate and 0.05 M hydrofluoric acid). The content of the main radionuclide in (Th0.834U0.166)O2 kernel was measured by a - and g -spectrometry, Table 1. The analysis of the kernel radionuclide inventories has shown 137Cs to be the main contributor to the total particle radioactivity. Other radionuclides are present in kernel in lower amounts. Two rest coated particles were mechanically broken step-bystep up to the porous carbon buffer (1 step) and up to the kernel (2 step), Fig. 1a, b, c. After these procedures, 137Cs contents in kernel
Fig. 1. Optical images of: (a, d) - irradiated BISO coated particle: (b, e) - kernel with buffer: (c, f) - (Th0.834U0.166)O2.kernel.
Thermocontrol
Argon
T
o
Flow
T, C 90 55 20
pH=0 + + +
pH=1 + ---
pH=2 + + +
pH=3 + ---
pH=4 + + +
pH=5 + ---
Flow-through reactor
Feeding solution
Oven
Peristaltic pump
Fig. 2. Experimental device for flow-through leaching experiments.
Sample collector
A. Bukaemskiy et al. / Progress in Nuclear Energy 57 (2012) 161e164
163
Fig. 3. Optical images of surfaces of (Th0.834U0.166)O2 kernels before irradiation (a, d), after irradiation (b, e) and after leaching (c, f).
(1)
where Ea is activation energy, A(pH) e empirical parameter depending on pH. For the l137Cs leaching process the value of Ea
a
pH=0 T=90C pH=1 T=90C
-2
pH=2 T=90C pH=2 T=55C
log(NLR), mol m -2day -1
All investigated coated particles before crushing had intact outer pyrocarbon surface without cracks and visible damages, Fig. 1 d. The coated particle diameter equals 760 mm. Moreover, the surface of the porous carbon buffer remained also intact after soft breaking, Fig. 1e. The kernel is black particle having a regular spherical shape, the diameter equals 430 mm. The coating thicknesses, which were determined from the optical microscopy photographs equal 90 and 75 mm for porous carbon buffer and outer dense pyrocarbon layer, respectively. The surfaces of kernels before irradiation (control sample), after irradiation and after leaching are shown in Fig. 3. All investigated kernels have a dense structure with well-formed grains and grain boundaries. The long-term irradiation led to the significant increasing of the grain boundary width, but the visible damages were not observed. After the following leaching the surface of the kernels almost remained without changes. The spent coated particle was accurately crashed in layers, and the 137Cs amount in each layer was measured. It was shown that the activity is mainly concentrated in the inner part of kernel (w77% of total caesium content), on the surface of kernel (w12%) and in the carbon buffer layer (w11%), whereas the outside carbon layer contains 0.1% of the total 137Cs only, that indicates its high protective characteristic (Schenk et al., 1984). The representative kinetic dependencies of fuel kernels leaching in hydrochloric acid are presented in Fig. 4a. The estimated normalized leaching rate (NLR) do not depend on time with the exception of the leaching at pH ¼ 0 and 90 C. The experiment at these conditions was carried out at the beginning; therefore the 137 Cs release had the highest value during first ten days. This value of 137Cs fracture release (12% of total content) can be referred to the near-surface layer. The normalized leaching rate depends on the values of pH and temperature as shown in Fig. 4b. NLR decreased significantly with increasing of pH and decreasing of temperature T. The NLR data are summarised in Table 2. In order to estimate the activation energy of the leaching process the NLR values can be plotted as function on the reciprocal temperature in so called Arrhenius coordinates (Fig. 5). The linear dependencies are observed for the fixed values of pH and can be described using the following equation:
Ea NLR ¼ AðpHÞexp RT
-3
pH=2 T=20C pH=5 T=90C
-4
-5
-6
-7
-8 0
5
10
15
20
25
30
Reaction time, days
b -4
log(NLR),mol*m -2*day -1
3. Results and discussion
90C 55C 20C
-5
-6
-7
-8 0
1
2
3
4
5
pH
Fig. 4. (a) 137Cs normalized leaching rates from (Th0.834U0.166)O2 fuel kernels in HCl solutions at different pH and temperature conditions and (b) 137Cs normalized leaching rates from (Th0.834U0.166)O2 fuel kernels in HCl solution versus pH (lines e Eq. (1)).
164
A. Bukaemskiy et al. / Progress in Nuclear Energy 57 (2012) 161e164
Table 2 137 Cs normalized leaching rates (NLR, mol*m2*d1)
Table 3 Normalized leaching rates (NLR, mol*m2*d1) and activation energy (Ea, kJ/mol).
T, C
pH ¼ 0
pH ¼ 1
pH ¼ 2
pH ¼ 3
pH ¼ 4
pH ¼ 5
T, C
137
90 55 20
1.05E-04 2.54E-05 1.27E-06
3.03E-05 e e
6.77E-06 7.97E-07 1.20E-07
4.73E-06 e e
2.46E-06 2.68E-07 4.41E-08
7.76E-07 e e
90 55 20 Ea
1.05E-04 2.54E-05 1.27E-06 57.5
Cs
238
Pu
5.1E-05 e e e
Th
U
Th*
U*
8.1E-05 1.6E-05 4.5E-06 40.0
2.5E-02 2.2E-02 1.2E-02 9.8
2.4E-06 6.3E-07 2.7E-07 30.1
2.2E-03 2.3E-03 1.1E-03 9.4
* - unirradiated kernels.
-8
component in these phases and the dissolution rate of Th is much lower. According to the obtained data the activation energy of the Th dissolution was very close to that for Cs, while the estimated value for U was four times lower.
ln(NLR), mol*m -2*day -1
pH=0
-10
pH=2 pH=4
-12
4. Conclusion -14
-16
-18 0.0027
-1
0.0029
0.0031
0.0033 1/T, K
Fig. 5. 137Cs normalized leaching rates from (Th0.834U0.166)O2 fuel kernels in HCl solution versus T1.
equals 57.5 3.5 kJ/mol, empirical parameter can be expressed as following:
AðpHÞ ¼ expð9:74 0:98pHÞ
(2)
Using the obtained data on Ea and NLR0, the value of the normalized leaching rate can be re-calculated in the form of the lineal dependencies, which are depicted in Fig. 4b as lines. Moreover, the release of other radionuclides was investigated at pH ¼ 0 for irradiated and unradiated kernels. The values of the normalised leaching rate and of the estimated parameters, Ea and A(pH), for these radionuclides are given in Table 3. The dissolution rate of the irradiated kernels was higher in ten times approximately in comparison with the values for the unirradiated kernels. It can be noted that the normalized leaching rate for the Cs, Th and Pu are comparable with each other, while the value of NLR for U is significantly lower. The similar ratio between the normalized leaching rates for U and Th remains for the unradiated kernels too. The obtained results are in agreement with the literature data (Alliot et al., 2005). In this work it has been shown that the leaching rate of U from the thorium-rich mixed oxide matrix seems to be similar to the dissolution rate of pure UO2 even though U is a minor
In the present work, the spent fuel BISO coated particles, which have been irradiated to a burn-up of 10% FIMA, were studied. The size, morphological characteristics of the coated particles, the distribution of the 137Cs amount in the coated particle were investigated in detail. It was shown the activity was concentrated mainly inside the kernels and in the carbon buffer layer, while the outside carbon layer contained 0.1% of the total 137Cs only. The thoria-based (Th0.834U0.166)O2 kernels were mechanically isolated from the coated particles, and their solution behavior was studied using the flow through experiments. The normalized leaching rate (NLR) of 137Cs from the spent fuel kernels was determined at the different pH and temperatures. The activation energy for the 137Cs release was estimated from the experimental results plotted in Arrhenius coordinates. At present experimental conditions, the normalized leaching rate can be expressed as a linear function of the pH-value. References Alliot, C., Grambow, B., Landesman, C., 2005. Leaching behaviour of unirradiated high temperature reactor (HTR) UO2eThO2 mixed oxides fuel particles. J. Nucl. Mat. 346, 32e39. Bruno, J., Casas, I., Puigdomenech, I., 1991. The kinetics of dissolution of UO2 under reducing conditions and the influence of an oxidized surface layer (UO2þx): application of a continuous flow-through reactor. Geochim. Cosmochim. Acta 52, 647e658. De Las Cuevas, C., Pueyo, J.J., 1995. The influence of mineralogy and texture in the water content of rock salt formations. Its implication in radioactive waste disposal. Appl. Geochem. 10 (3), 317. Kirch, N., Brinkmann, H.U., Brücher, P.H., 1990. Storage and final disposal of spent HTR fuel in the Federal Republic of Germany. Nucl. Eng. Des. 121, 241e248. Nickel, H., Nabielek, H., Pott, G., Mehner, A.W., 2002. Long time experience with the development of HTR fuel elements in Germany. Nucl. Eng. Des. 217, 141e151. Schenk, W., Naoumidis, A., Nickel, H., 1984. The behaviour of spherical HTR fuel elements under accident conditions. J. Nucl. Mat. 124, 25e32.