Environ. Sci. Technol. 2007, 41, 7087-7093
Radiation Induced Spent Nuclear Fuel Dissolution under Deep Repository Conditions MATS JONSSON,* FREDRIK NIELSEN, OLIVIA ROTH, ELLA EKEROTH, SARA NILSSON, AND MOHAMMAD MOHSIN HOSSAIN KTH Chemical Science and Engineering, Nuclear Chemistry, Royal Institute of Technology, SE - 100 44 Stockholm, Sweden
The dynamics of spent nuclear fuel dissolution in groundwater is an important part of the safety assessment of a deep geological repository for high level nuclear waste. In this paper we discuss the most important elementary processes and parameters involved in radiation induced oxidative dissolution of spent nuclear fuel. Based on these processes, we also present a new approach for simulation of spent nuclear fuel dissolution under deep repository conditions. This approach accounts for the effects of fuel age, burn up, noble metal nanoparticle contents, aqueous H2 and HCO3- concentration, water chemistry, and combinations thereof. The results clearly indicate that solutes consuming H2O2 and combined effects of noble metal nanoparticles and H2 have significant impact on the rate of spent nuclear fuel dissolution. Using data from the two possible repository sites in Sweden, we have employed the new approach to estimate the maximum rate of spent nuclear fuel dissolution. This estimate indicates that H2 produced from radiolysis of groundwater alone will be sufficient to inhibit the dissolution completely for spent nuclear fuel older than 100 years.
Introduction A political decision to build and take into use a deep geological repository for long-term storage of spent nuclear fuel will largely depend on the outcome of thorough scientifically based safety assessments. Given the very long operational time span for the repository, the safety assessments must be performed on the basis of extreme extrapolations of a number of fairly complicated processes. Consequently, the quality demands on the experimental results as well as the models underlying these extrapolations must be very high. One of the key processes here is the dissolution of the spent nuclear fuel matrix in groundwater, liberating radioactive fission products and actinides. The nuclear fuel used in water cooled reactors consists of ceramic uranium dioxide enriched in 235U to 4-5% of the total uranium content. Hence, the spent nuclear fuel consists mainly of UO2 (>95%) and the rest is fission products and actinides. In granitic groundwaters at the depth of 500-700 m the conditions are reducing and UO2 has very low solubility (1). However, the ionizing radiation emitted from the fuel will induce chemical processes in the groundwater (radiolysis) producing both oxidants (OH•, H2O2, HO2•, and O2) and reductants (e-aq, H• and H2) (2). For * Corresponding author phone: +46 8 790 9123; fax: +46 8 790 8772; e-mail:
[email protected]. 10.1021/es070832y CCC: $37.00 Published on Web 09/07/2007
2007 American Chemical Society
kinetic reasons, the radiolytically produced oxidants can oxidize U(IV) to the significantly more soluble U(VI) and thereby enhance matrix dissolution. Hence, it is of utmost importance to be able to describe the dynamics of this process, preferably on the basis of well-established rate constants for all relevant elementary reactions involved. The production of oxidants depends on the dose rate, which in turn is a function of fuel age, burnup, and distance from the fuel surface. The activity of the fuel, as well as the relative importance of R-, β-, and γ-radiation, will change considerably with time (3). For this reason, experimental studies using relatively fresh spent nuclear fuel (10-20 years) will not reflect future deep repository conditions and it is also not advisable to use experimental dissolution rates for extrapolations to longer times. In addition, the surface structure and, thereby, the reactivity of the spent nuclear fuel could change with time. It should be noted that several attempts to numerically simulate spent nuclear fuel dissolution have been made (47). Unfortunately, many of these attempts are based on insufficient information about the mechanism and kinetics for the surface reactions involved (i.e., reactions occurring at the spent nuclear fuel-water interface). Furthermore, the heterogeneous nature of the system is not always accounted for. In this work, we describe a new method for simulating the maximum rate of spent nuclear fuel dissolution taking recent experimental data and mechanistic models into account. The method quantitatively accounts for the experimentally observed inhibiting effect of H2 (8). Using data from the two possible repository sites in Sweden currently under investigation (9, 10), we employ the new approach to estimate the rate of spent nuclear fuel dissolution under these conditions. Mechanism and Kinetics of Radiation Induced Dissolution of UO2. The mechanism and the kinetics of oxidative spent fuel and UO2 dissolution have been studied quite extensively for several decades (11-13). Most of the studies have been focused on the release rate for U(VI), other actinides, and fission products and relatively few studies have been focused on the rate of oxidant consumption. Even fewer studies have attempted to elucidate the actual rate constants for the elementary reactions involved in the process of oxidative dissolution (11, 12). Oxidative dissolution can be described by the following reaction scheme:
UO2(s) + OX f UO22+(s) + RED
(1)
UO22+(s) f UO22+(aq)
(2)
Hence, the process can be divided into two elementary steps: (1) oxidation and (2) dissolution of oxidized UO2. The oxidation process has been shown to be kinetically limited by the first one-electron-transfer step from UO2 to the oxidant, and the rate constant for the elementary reaction also depends on the one-electron reduction potential of the oxidant, i.e., the logarithm of the rate constant is linearly dependent on the one-electron reduction potential of the oxidant (11). Among the oxidants of relevance in a deep repository, the rate constant for both OH• and CO3•- (formed in water containing HCO3-/CO32-) is limited by diffusion according to the linear relationship while H2O2 and O2 react more slowly (7.33 × 10-8 m s-1 and 3.7 × 10-10 m s-1, respectively) (11, 12). It should be noted that the diffusion limit for a heterogeneous system is significantly different (i.e., lower) from that of a homogeneous system (14). The rate of oxidant consumption is given by eq 3 while the rate of UO2 oxidation is given by eq 4. VOL. 41, NO. 20, 2007 / ENVIRONMENTAL SCIENCE & TECHNOLOGY
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-
-
d[OX] SA [OX] ) k1 dt V dnUO2 dt
( )
(3)
) k1(SA)[OX]
(4)
k1 is the rate constant for the reaction between the oxidant and the UO2 surface, SA denotes the solid surface area and V is the solution volume. Dissolution of oxidized UO2 is a fairly slow process in pure water. However, HCO3- present in the groundwater enhances dissolution due to formation of water soluble complexes (15). This increases the UO2 surface accessible to oxidants and thereby increases the rate of oxidation. At low HCO3- concentrations the rate-limiting step in oxidative dissolution of UO2 is dissolution rather than oxidation. For HCO3- concentrations below 1 mM the rate of oxidation is linearly dependent on the HCO3- concentration, whereas for concentrations above 1 mM the rate is independent of HCO3- concentration (12). For concentrations above 1 mM the rate-limiting step is oxidation. HCO3- facilitated dissolution has been found to be a diffusion controlled process. Hence, under deep repository conditions oxidative dissolution of spent nuclear fuel can be assumed to be limited by oxidation (and thereby, the accessible surface area can be considered as constant during the process). The total rate of oxidation is given by the sum of the contributions from all potential oxidants (eq 5) where ki denotes the rate constant for the reaction between the UO2 surface and the oxidant i.
-
dnUO2 dt
) SA
∑
9
r H2O 2 )
ki[OX]i
∫
xmax
x)0
rH2O2(R)δmax (R) + rH2O2(β)δmax (β)
(5)
i
ENVIRONMENTAL SCIENCE & TECHNOLOGY / VOL. 41, NO. 20, 2007
D˙ (x) × F × G(H2O2)dx
(6)
where D˙ (x) is the dose rate at distance x from the fuel surface, F is the density of water, and G(H2O2) is the radiation chemical yield for H2O2. The maximum rate of the reaction between H2O2 and the UO2 surface corresponds to the steady-state. At steady-state, the rate of H2O2 consumption is identical to the rate of radiolytic H2O2 production. The steady-state surface concentration is calculated from eq 7
[H2O2]s-s )
Indeed, radiolytically produced reductants should also be accounted for in some cases. In the system described above, the UO2 surface is kept free from oxidized UO2 and the fraction of the surface area available for reductants is very small. On the other hand, in systems not containing HCO3- or any other species capable of facilitating removal of the oxidized UO2, the impact of radiolytical reductants could become significant. The solvated electron and the hydrogen atom are expected to react with the oxidized surface. The rate constants for these reactions are probably diffusion limited. However, the steady-state concentrations of these species will be very low. The concentration of H2 will be many orders of magnitude higher, but the reactivity is low and the reduction by H2 often requires a catalyst. Given the complexity of the system, it is interesting to elucidate the relative importance of the different radiolytic oxidants. The impact of an oxidant in the process of oxidative dissolution depends on the rate constant as well as the concentration. Experimental studies combined with numerical simulations of irradiated aqueous solutions containing UO2 pellets and powder revealed that the molecular oxidants have the largest impact on the oxidative dissolution even in γ-irradiated systems (where production of radicals is favored) (16). Under deep repository conditions, i.e., in systems dominated by R-radiolysis, the relative impact of H2O2 is >99.9% (16). Consequently, H2O2 is the only radiolytical oxidant needed to be accounted for when exploring the rate of spent nuclear fuel dissolution. Geometrical Dose Distribution as a Function of Fuel Age and Burnup. The radiolytic production of H2O2 is given by the dose rate and the type of radiation. As mentioned above, the dose rate depends on the radionuclide inventory (a function of fuel age and burnup) and the distance from the fuel surface. Using inventory data (17), taking self-shielding and absorption in the surrounding water 7088
into account, the geometrical dose distribution has been calculated (3). Figure 1 illustrates the geometrical dose distribution for 100 year old spent fuel (burnup 38 MWd/ kg U). As can be seen, radiolysis due to radiation from the spent nuclear fuel inherently introduces a concentration gradient in the surrounding aqueous phase. Hence, diffusion must be accounted for when simulating the process. The Steady-State Approach For Maximum Dissolution Rate. The geometrical dose distribution, as well as the consumption of H2O2 in surface reactions and in homogeneous reactions in solution, complicates the situation further. However, keeping in mind the following boundary conditions, we can still reduce the complexity to some extent. The rate of spent fuel dissolution can never exceed the rate of UO2 oxidation and the rate of H2O2 consumption can never exceed the radiolytic H2O2 production rate. The radiolytic H2O2 production rate is given by eq 6
kH2O2
(7)
where jr is the average production rate in the irradiated volume, δ is the maximum range of the radiation and kH2O2 is the rate constant for the reaction between H2O2 and the fuel surface. By simulating H2O2 production using the geometrical dose distribution given above and H2O2 consumption in a surface reaction taking diffusion (one dimension) into account, we were able to show that steady-state surface concentration is approached in a very short time (minutes to hours) in view of the time span of interest with regards to a deep repository (18). Consequently, the use of the steady-state approach will simplify simulation of spent nuclear fuel dissolution significantly without loss of accuracy. Indeed, reactions between H2O2 and solutes will also affect (reduce) the steady-state concentration. Dissolution of the spent nuclear fuel will release radionuclides into the groundwater and thereby increase the dose rate in the aqueous phase. This effect is not accounted for in the approach described above. In this work we have made use of the above knowledge and methods in making fairly simple yet reliable estimates of the maximum rate of spent fuel dissolution under various conditions. The additional effect of dissolved radionuclides is also discussed. Inventory data used to calculate the dose rate, and thereby the H2O2 production rate, are taken from ref (17).
Results and Discussion As mentioned above, the steady-state approach gives the maximum dissolution rate. The maximum dissolution rate and the time required for complete dissolution of the spent nuclear fuel as a function of fuel age for two different burn ups (38 and 55 MWd/kg U) are presented in Table 1. As stated above, the release of radionuclides into the groundwater will increase the dose rate and thereby the rate of H2O2 production in the aqueous phase. If the radionuclides are not removed from the system, the oxidative
FIGURE 1. Geometrical dose distribution for spent nuclear fuel (Age 100 years and burnup 38 MWd/kg U) calculated using inventory data from ref 17.
TABLE 1. Spent Nuclear Fuel Dissolution Rates and Time for Complete Dissolution for Different Fuel Ages and Burnup dissolution rate (mol m-2 s-1) fuel age (years) burn-up 38 MWd/kg U
maximum rate and time for complete dissolution
[Fe2+] ) 1 µM
catalysis of H2 + H2O2 (1% noble metal particles)
100
4.94 × 10-10 (3.6 × 103 y)
1.04 × 10-11
1.51 × 10-10
1000
8.72 × 10-11 (2.1 × 104 y)
1.84 × 10-12
2.67 × 10-11
10 000
1.71 × 10-11 (1.1 × 105 y)
3.61 × 10-13
5.24 × 10-12
100 000
1.79 × 10-12 (1.0 × 106 y)
3.78 × 10-14
5.46 × 10-13
100
7.04 × 10-10 (2.6 × 103 y)
1.48 × 10-11
2.15 × 10-10
1000
9.68 × 10-11 (1.9 × 104 y)
2.03 × 10-12
2.96 × 10-11
10 000
1.86 × 10-11 (9.7 × 104 y)
3.91 × 10-13
5.69 × 10-12
100 000
2.41 × 10-12 (7.5 × 105 y)
5.06 × 10-14
7.37 × 10-13
burnup 55 MWd/kg U
dissolution process will be accelerated and calculations show that the time required for complete dissolution of the fuel will be reduced to 1% compared to the numbers given in Table 1. It should be kept in mind that the increase in dose rate will also result in a higher rate of H2 production. As will be discussed later, H2 inhibits the oxidative dissolution. Accelerated dissolution is normally not observed in spent nuclear fuel leaching experiments. The reaction between H2O2 and UO2 has a yield of approximately 80% with respect to oxidation of UO2 (19). The remaining 20% can probably be attributed to catalytic decomposition of H2O2. Hence, the dissolution rate should be 80% of the rate of H2O2 consumption (this has been taken into account in Table 1). Relatively recent studies show that the reactivity of irradiated UO2 (using high-energy electrons from an accelerator) toward oxidants in aqueous solution is somewhat higher than that of nonirradiated UO2. The irradiated UO2 (40 kGy) is approximately 30% more reactive
than nonirradiated UO2 (20). The loss of H2O2 concentration to solute reactions is to some extent compensated by the enhanced reactivity of the irradiated UO2. However, in a system free from reactive solutes, the radiation enhanced reactivity has no effect since the rate of oxidation is completely governed by the rate of H2O2 production (18). Additional reactions consuming H2O2, either at the spent nuclear fuel surface or in the aqueous phase, will lower the steady-state concentration and thereby the rate of oxidative dissolution. Effects of Reactive Solutes. H2 is produced both by radiolysis of water and in the process of anaerobic corrosion of steel (used in the canisters for spent nuclear fuel) (21). Although thermodynamically favorable, the reaction between H2 and H2O2 in water at low temperature is very slow (22). However, the presence of high concentrations of H2 influences the radiolytical production rate of H2O2, in part by converting hydroxyl radicals to hydrogen atoms. This will VOL. 41, NO. 20, 2007 / ENVIRONMENTAL SCIENCE & TECHNOLOGY
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FIGURE 2. Relative spent nuclear fuel dissolution rate (compared to the maximum rate) as a function of noble metal nanoparticle coverage (fraction of total surface area) taking the catalytic effect on the reaction between H2 and H2O2 into account.
TABLE 2. Relative Steady-State H2O2 Surface Concentration as a Function of Surface Rate Constant and Homogeneous Pseudo First-Order Rate Constanta homogeneous pseudo first-order rate constant (s-1) 0 surface rate constant (m s-1) 7.33 × 10-8 7.33 × 10-7 7.33 × 10-6 7.33 × 10-5 and
0.01
0.1
1
1 0.1 0.01 0.001
0.021 nd nd nd
0.0059 0.0056 nd 7.2 × 10-4
1000
0.0015 0.0015 0.0012 4.7 × 10-4
3.2 × 10-4 nd 3.0 × 10-4 1.9 × 10-4
4.8 × 10-5 nd 4.5 × 10-5 4.0 × 10-5
nd nd nd 5.3 × 10-6
is not determined.
Fe2+ + H2O2 f Fe3+ + OH• + OH-
(8)
The lifetime of the hydroxyl radical is very short and the impact of the radicals produced in this reaction on the dissolution of spent fuel will be insignificant (as has previously been shown) due to the extremely low surface concentration (16). However, the reaction will consume H2O2 in the aqueous phase leading to significantly lower surface steady-state concentrations of H2O2 and, thereby, also significantly lower rates of dissolution. Maximum dissolution rates taking this process into account, assuming [Fe2+] ) 1 µM, are presented in Table 1. As can be seen, the bulk reaction reduces the rate of dissolution by a factor of 50. 9
100
relative steady-state H2O2 surface concentration
influence the final yield of H2O2 in the system; however, the primary G-value is not affected. The magnitude of the effect depends on the type of radiation. For R-radiation, 40 bar H2 reduces the H2O2 production rate by close to 90%, while that for β-radiation is 20% (based on numerical simulation of radiation chemistry in water). Intrusion of groundwater into a canister for spent nuclear fuel implies that the canister has either been severely corroded or has been exposed to significant mechanical stress. In both cases, iron will be exposed to the groundwater and the concentration of Fe2+ will, therefore, be higher than in the groundwater. Fe2+ reduces H2O2 according to the Fenton reaction (reaction 8) producing Fe3+ and hydroxyl radicals (23).
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Micro organisms present in the repository can also influence the dissolution kinetics. Bacteria have been found to reduce many metal ions (24). This will reduce the rate of oxidative dissolution further. Effects of Surface Reactions Not Leading to Oxidation of UO2. Spent nuclear fuel contains nanometer sized metallic particles composed of the fission products Pd, Mo, Ru, Tc, and Rh (25). These so-called -particles are potential catalysts for reduction by H2. Electrochemical studies have shown that the inclusion of noble metal particles in UO2, in combination with H2, gives very low corrosion potentials (26). Recent studies of the catalytic effect of pure Pd-particles on the reaction between H2O2 and H2 show that the secondorder rate constant with respect to H2O2 and Pd is virtually diffusion controlled for H2 pressures higher than 1 bar (22). Assuming -particles have the same catalytic ability as Pd (the rationale for this is that Pd is one of the main constituents of -particles), we can calculate the oxidation rate as a function of -particle surface coverage (using simple competition kinetics). The relative dissolution rate as a function of surface coverage is presented in Figure 2 and the maximum dissolution rates taking this effect into account are presented in Table 1. As can be seen, the effect is fairly small and very high -particle surface coverage is required to reduce the rate of oxidation by a factor of 10 or more. Pd-particles also catalyze the reduction of UO22+ (aq) by H2 (27). However, this process will reduce the concentration of U(VI) in solution, but not
FIGURE 3. Elementary processes of importance in the process of oxidative dissolution. From top to bottom: (1) Consumption of H2O2 in the bulk. (2) Noble metal nanoparticle catalyzed reaction between H2 and H2O2. (3) Oxidation of U(IV) and HCO3- facilitated dissolution of U(VI). (4) Noble metal nanoparticle catalyzed solid-phase reduction of U(VI) to U(IV) by H2.
the rate of UO2 oxidation. Given the electrochemical observation mentioned above, the main effect of the presence of -particles and H2 is probably solid-phase reduction of oxidized UO2 on the surface of the spent nuclear fuel. This process and the other redox processes of major importance are illustrated in Figure 3. Assuming the uptake of H2 to be limited by diffusion (as indicated by recent studies on the catalytic effect of Pd on the reaction between H2 and H2O2) (22), an -particle surface coverage of only 1 ppm would be sufficient to completely stop spent nuclear fuel dissolution for fuel as fresh as 100 years (40 bar H2). Long-term experiments on the dissolution of spent nuclear fuel indicate that the rate of dissolution is approaching zero when the concentration of radiolytically produced H2 is in the range of 10-5 to 10-4 mol dm-3 (corresponding to a partial pressure of 0.01-0.1 bar) (8). This corresponds to a diffusion controlled rate constant (10-6 m s-1) for the reaction between H2 and the -particles. Taking this process into account, we obtain the following expression for oxidative dissolution of spent nuclear fuel:
rdiss ) rox - kH2[H2]rel
(9)
where rdiss is the dissolution rate, rox is the oxidation rate, kH2 is the rate constant for the reaction between H2 and the -particles, [H2] is the concentration of H2, and rel is the fraction of the fuel surface area covered by -particles. Combined Effects of Reactive Solutes and Surface Reactions Not Leading to Oxidation of UO2. As mentioned
above, noble metal nanoparticles are expected to catalyze the reaction between H2 and H2O2 and thereby increase the total rate constant for consumption of H2O2 at the fuel surface. This will reduce the steady-state concentration of H2O2 and, thereby, also the rate of UO2 oxidation. Quantitative elucidation of the impact of solutes capable of reducing H2O2 in solution is not straight forward for a system with inhomogeneous production of H2O2. As mentioned above, simulations show that solutes do have a significant effect on the steady-state surface concentration. The combined effect of reactive solutes and additional surface reactions consuming H2O2 must be quantified to enable accurate simulation of a specific system. In Table 2, the relative steady-state surface concentration (with respect to the maximum steady-state concentration) as a function of surface reaction rate constant and pseudo first-order rate constant for the bulk reaction is given as a basis for sensitivity analysis. As can be seen, the steady-state surface concentration decreases with increasing pseudo first-order rate constant for the bulk reaction. The decrease is, however, not linearly related to the rate constant for the process. The main reason for the nonlinearity is the inhomogeneous production of H2O2, giving rise to concentration gradients. For higher homogeneous rate constants, the bulk concentration of H2O2 approaches zero, and the actual volume containing H2O2 decreases. As can also be seen, the steady-state surface concentration is mainly determined by the homogeneous pseudo first-order rate constant. However, for high surface reaction rate constants, the steady-state surface concentrations are significantly lower for the lowest homogeneous pseudo first-order rate constants. Hence, the system is governed by the surface reaction in these extreme cases. As mentioned above, the rate constant for the reaction between H2O2 and the spent nuclear fuel surface is expected to be less than 1 order of magnitude higher than the rate constant for the reaction between H2O2 and UO2. Consequently, even for the low homogeneous pseudo firstorder rate constant, the steady-state surface concentration and, thereby, the rate of oxidation will be governed by the bulk reaction. For this system, the relative steady-state surface / concentration ([H2O2]s-s /[H2O2]s-s) as a function of the homogeneous pseudo first-order rate constant is given by / eq 10 (28). [H2O2]s-s is the surface concentration taking the bulk reaction into account and [H2O2]s-s is the surface concentration in the absence of bulk reactions consuming H2O2.
log
/ [H2O2]s-s
[H2O2]s-s
) -0.66log k* - 2.9
(10)
It should be noted that eq 10 is independent of average dose rate. Systems Limited by Dissolution Kinetics. In groundwater containing less than 1 mM HCO3- the process of oxidative dissolution will be limited by the rate of dissolution of oxidized UO2 rather than oxidation of UO2. The rate constants for HCO3- facilitated dissolution of oxidized UO2, as well as that for dissolution of oxidized UO2 in pure water, have previously been determined experimentally (12). In cases where the process is limited by dissolution, the surface area accessible to H2O2 will be reduced. At a given concentration of H2O2 this will reduce the rate of UO2 oxidation. However, in a system with constant rate of H2O2 production, the steadystate surface concentration will increase to compensate for the decreased reactivity of the surface. The maximum dissolution rate in such a system is given by eq 11 (12).
-
dnUO2 dt
) 1.7 × 10-5[HCO3]
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FIGURE 4. The logarithm of the dissolution rate for spent nuclear fuel (burn up 38 MWd/kg U) as a function of the logarithm of the fuel age. [ ) The maximum dissolution rate. 2 ) The effect of Fe2+ (1 µM) consuming H2O2 in the bulk. The lines corresponding to the different H2 partial pressures represent the rates of solid-phase reduction (see eq 9). It should be noted that the concentration of HCO3- is given in mol m-3 and the dissolution rate in mol m-2 s-1. In the complete absence of HCO3- the maximum dissolution rate for a fully oxidized UO2 surface is 7 ( 1 × 10-8 mol m-2 s-1 (12). Hence, the lowest maximum dissolution rate in a system limited by dissolution kinetics is still significantly higher than the maximum rate of oxidation determined from the steadystate approach. Consequently, dissolution kinetics will not limit the rate of spent fuel dissolution under deep repository conditions. Effects of Secondary Phase Formation on the Spent Nuclear Fuel Surface. In systems where the solubility of oxidized UO2 is limited, secondary phases are formed (13). Secondary phases formed on the spent nuclear fuel surface could be porous and contain a significant fraction of water. The water will, to some extent, be contained by the secondary phase, and this will reduce the diffusivity of solutes present in the contained water volume as well as the accessibility for solutes outside the contained volume. The direct consequence of this is that the steady-state surface concentration and, thereby, the rate of oxidation will decrease. The reason for this is that a smaller fraction of the radiolytically produced H2O2 will be accessible to the surface. In Figure 4, the maximum spent nuclear fuel dissolution rates and the dissolution rates in the presence of 1 µM Fe2+ as a function of fuel age are given along with limiting values corresponding to four different H2 partial pressures (assuming an -particle surface coverage of 1%). If the dissolution rate exceeds the limiting value, dissolution will occur. If not, radiation induced dissolution of spent nuclear fuel is effectively inhibited by the -particle catalyzed solid-phase reduction by H2. As can clearly be seen, a partial pressure of only 0.1 bar H2 will effectively inhibit the dissolution of the spent fuel aged 100 years or more. In the presence of 1 mM Fe2+, even 0.01 bar H2 will be sufficient to stop the dissolution. Using borehole data from the two possible repository sites in Sweden (9, 10), we can estimate the dissolution rates. The groundwater concentration of Fe2+ at a depth of 550-570 m is 40 and 36 µM in Forsmark and Oskarshamn, respectively. This will reduce the inherent rate of dissolution in the system by a factor of 400 compared to the maximum dissolution rate. Consequently, radiolytically produced H2 alone can be expected to inhibit the dissolution completely. 7092
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Acknowledgments The Swedish Nuclear Fuel and Waste Management Company (SKB) and the Swedish Nuclear Power Inspectorate (SKI) are gratefully acknowledged for financial support.
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Received for review April 9, 2007. Revised manuscript received June 18, 2007. Accepted July 28, 2007. ES070832Y
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