Static Dissolution of α-doped UO2 in Boom Clay Conditions: Preliminary Results

2003 ◽  
Vol 807 ◽  
Author(s):  
C. Cachoir ◽  
K. Lemmens

ABSTRACTTo support the performance assessment (PA) calculations for the possible final disposal of spent fuel in the Boom Clay, leach experiments are performed with α-doped UO2 in clay media, simulating various near field ages. The experiments allow to measure the dissolution rate of the UO2 and to determine the assumed relationship between dissolution rate and α-activity. Tests are performed at six different α-activities, simulating various fuel ages, at 25–30°C, for durations ranging from 90 to 720 days, in a glove box with Ar/0.4%CO2 atmosphere. The solutions and solids are analyzed for U isotopes and 238Pu by radiochemical measurement and by ICP-MS. The dissolution rates of the α-doped UO2 are presented for different durations. The resulting corrosion rate is around 300 μgU.m−2.d−1. This is up to 100 times higher than found for similar conditions in the literature. In the presence of clay, there appears to be no correlation between the α-activity and the corrosion rate of the α-doped UO2.

Author(s):  
Karel Lemmens ◽  
Marc Aertsens ◽  
Ve´ra Pirlet ◽  
Norbert Maes ◽  
Hugo Moors ◽  
...  

To estimate the lifetime of vitrified high level waste (HLW-glass) in Boom Clay disposal conditions, the dissolution behaviour of waste glass has been studied with experiments performed in surface laboratories and in the HADES underground research facility of SCK·CEN since 1980. We present the main topics and first results of the SCK·CEN programme 2000–2003. This programme focuses on the following items: (1) the diffusion/sorption/precipitation of silica in Boom clay or backfill clay, (2) demonstration of glass dissolution behaviour in realistic test conditions, (3) the effect of presaturation of the clay with silica, and (4) the estimation of near field concentrations of critical isotopes. The experiments have shown so far that Si, released by the glass, is effectively immobilized by Boom Clay, but it can nevertheless diffuse into the clay without immediately precipitating. The dissolution rate of glass SON68 and SM539 is determined in Boom Clay at in situ density and at 30°C (this is the long-term temperature expected near the waste glass packages in a Boom Clay repository). The dissolution rates, based on glass mass losses, are constant during the first year, at ∼ 0.010 g.m−2.day−1 for glass SON68 and ∼ 0.012 g.m−2.day−1 for glass SM539. The addition of glass frit causes a decrease of the glass dissolution rate, both with glass SON68 and SM539, and both in Boom Clay and in FoCa-clay. In FoCa-clay at high density with glass frit, the dissolution rates, based on glass mass losses, after 8 months at 30°C are ∼ 0.001 g.m−2.day−1 (SM539) and ∼0.005 g.m−2.day−1 (SON68). Because the experiments performed in Boom Clay and FoCa-clay with glass frit simulate realistic conditions (high clay density, low temperature), they can be used to estimate the maximum glass dissolution rate in a (Boom) clay repository. The corresponding minimum lifetime of a glass canister, calculated with the SCK·CEN code for lifetime predictions, is of the order of 105 to 106 years, if we neglect the internal glass surface area (due to cracking). In more diluted clay suspensions with glass frit, the glass dissolution rate is 10−4 to 10−5 g.m−2.day−1 or even zero. This would correspond to a lifetime of >>106 years. So far, there is no indication that the addition of glass frit leads to secondary phase formation at low temperature (30–40°C). Leach experiments with doped glasses SON68 and SM539 suggest that the maximum concentrations of most critical radionuclides in near field conditions are lower than the best estimate solubilities used for performance assessment studies in Boom Clay. For Se, relatively high concentrations were measured, though. The research programme for the underground laboratory is not discussed.


1993 ◽  
Vol 333 ◽  
Author(s):  
P. A. Finn ◽  
J. K. Bates ◽  
J. C. Hoh ◽  
J. W. Emery ◽  
L D. Hafenrichter ◽  
...  

ABSTRACTPreliminary results for the composition of the leachate from unsaturated tests at 90°C with spent fuel for two successive periods of ~60 days each with pretreated J-13 groundwater are reported. The pH of the leachate solutions ranged from 4 to 7. The americium concentration was 104 to 105 greater than that reported for saturated spent fuel tests in which the leachate pH was 8. The major fraction of material in the leachate was present as colloids containing both americium and curium. The presence of actinides in a form not currently directly included in repository radionuclide transport models provides information that can be used in spent fuel reaction modeling, the performance assessment of the repository and the design of the engineering barrier system.


2006 ◽  
Vol 932 ◽  
Author(s):  
Sonia Salah ◽  
Christelle Cachoir ◽  
Karel Lemmens ◽  
Norbert Maes

ABSTRACTSince reprocessing is no longer the reference policy in Belgium, studies on the direct disposal of spent fuel in a clay formation have gained increased interest in the last years. In order to determine to what extent the clay properties and the α-activity influence the dissolution kinetics of spent fuel for the long term disposal, static dissolution tests have been performed on 5 different types of α-doped UO2, representing a PWR fuel with a burn-up of 45 or 55 GWd · tHM−1 and fuel ages ranging between 150 and 90,000 years, in different Boom Clay (BC) media at room temperature and in an anoxic atmosphere for 90 to 720 days. The uranium activity in the clay fraction over time was found to be much higher than the U-activity in the leachates, which has been mainly ascribed to the high retention capacity of the BC. The average dissolution rate between 0 and 90 days obtained for the 5 types of α-doped UO2 were all found to be high and quite similar at ~263 µg · m−2 · d−1and a “longer-term” rate (181 to 720 days) ranging between zero and 15 µg · m−2· d−1. These results suggest that the activity of the fuels does not seem to have an effect on the UO2 dissolution rates under the considered test conditions. In order to study the partition/redistribution of U during UO2dissolution, sequential extraction experiments were performed. Results of the latter have provided a better mechanistic understanding of BC/spent fuel interaction processes as well as information on the role of the different minerals controlling the U-retention/immobilization.


1992 ◽  
Vol 294 ◽  
Author(s):  
Patrik Sellin ◽  
Nils Kjellbert

ABSTRACTThe near-field radionuclide migration code Tullgarn has been developed for performance assessment purposes. As a part of the PROPER-code package it has been successfully applied in the SKB 91 safety analysis.The features and processes included in the code are:- Radioactive chain decay- Different canister failure mechanisms (copper corrosion from sulphide attack, steel corrosion, internal overpressure and initially defective canisters) - Spent fuel dissolution. The model is based on the assumption that the dissolution rate is proportional to the α-dose rate- Transport calculations are done with a resistance-network model. Tullgarn calculates the stationary release of radionuclides from a defect in the canister through the buffer and out into a fracture in the rock or up to the damaged zone under the deposition tunnel.Tullgarn can be used as a stand-alone model for near-field release calculations or as a submodel in an integrated assessment. In the SKB 91 analysis, Tullgarn gave the source term to the far-field model.


2006 ◽  
Vol 932 ◽  
Author(s):  
Antonín Vokál ◽  
Dmitrij Lukin ◽  
Dušan Vopálka

ABSTRACTCarbon steel has been chosen in the Czech disposal concept of spent fuel disposal in a granite host rock as a reference material for disposal canisters. On the basis of the results of performance assessment studies, it could be decided whether this material is suitable or whether a more corrosion resistant, and also more expensive, material should substitute it. A number of papers have convincingly shown that iron transfer constraints contribute to a significant decrease in corrosion rate, but no study hasso far been devoted to modeling this process. In this paper the effects of initial corrosion rate, corrosion product solubility and porosity and other repository parameters on the transfer of iron to the host rock are modeled using a numerical transport computer code. It was found that the critical parameter for iron transfer is the solubility of corrosion products, considerably affecting the steady state corrosion rate. The initial corrosion rate of carbon steel and the sorption properties of bentonite primarily affect the time needed to achieve a steady state of corrosion. The results of the calculations strongly suggest that the constraints on iron transfer from the canister surface will govern the corrosion rate of carbon steel canisters, whose lifetime, owing to this effect, can stretch to millions of years.


Author(s):  
Jan Marivoet ◽  
Xavier Sillen ◽  
Peter De Preter

Abstract Geological repository systems for the disposal of radioactive waste are based on a multi-barrier design. Individual barriers contribute in different ways to the overall long-term performance of the repository system, and furthermore, the contribution of each barrier can considerably change with time. In a systematic analysis of the functional requirements for achieving long-term safety a number of basic safety functions can be defined: physical confinement, retardation / slow release, dispersion / dilution and limited accessibility. In the case of the geological disposal of spent fuel in a clay formation a series of barriers are designed or chosen to contribute to the realisation of the basic safety functions. The physical confinement is realised by the watertight, high-integrity container, which prevents contact between groundwater and the confined radionuclides. In first instance the retardation / slow release function is realised by the slow dissolution of the waste matrix and by the limited solubility of many elements in the near field. However, the natural clay barrier provides the main contribution to this safety function. The migration of radionuclides through the Boom Clay is mainly due to molecular diffusion, which is an extremely slow process. Furthermore, many elements are strongly sorbed by the clay minerals what makes their migration even much slower. The dispersion / dilution function mainly occurs in the aquifer and the rivers draining the aquifer in the surroundings of the disposal system. Various performance indicators are used to quantify the contributions of each safety function and to explain the functioning of the repository system.


2000 ◽  
Vol 663 ◽  
Author(s):  
Esther Cera ◽  
Juan Merino ◽  
Jordi Bruno

ABSTRACTIn the framework of the Enresa 2000 PA exercise and as a continuation of the developments made during SR 97, we have developed a conceptual and numerical model to calculate the release of radionuclides from spent fuel under repository conditions. The model includes both thermodynamic and kinetic considerations. Hence, although certain radionuclides are solubility controlled, for other radionuclides their release is governed by kinetic processes such as radiolytically promoted oxidative dissolution of the matrix and the associated water turnover inthe gap. The fluxes of selected radionuclides are calculated as an indication of the relative importance of the various processes considered to define source term concentrations in the performance assessment of the spent fuel repository.


2006 ◽  
Vol 985 ◽  
Author(s):  
Andreas Loida ◽  
Volker Metz ◽  
Bernhard Kienzler

AbstractRecent studies have shown that in the presence of H2 overpressure, which forms due to the corrosion of the Fe based container, the dissolution rate of the spent fuel matrix is slowed down by a factor of about 10, associated with a distinct decrease of concentrations of important ra-dionuclides. However, in a natural salt environment as well as in geological formations with chloride rich groundwater the presence of radiation chemically active impurities such as bro-mide must be taken in consideration. Bromide is known to react with β/γ radiolysis products, thus counteracting the protective H2 effect. In the present experiments using high burnup spent fuel it is observed that during 212 days the matrix dissolution rate was enhanced by a factor of about10 in the presence of up to 10-3 M bromide and 3.2 bar H2 overpressure. However, concen-trations of matrix bound actinides were found at the same level or below as found under identical conditions, but in the absence of bromide. In the long-term it is expected that the effect of bro-mide becomes less important, because the decrease of β/γ-activity results in a decrease of oxida-tive radicals, which react with bromide, while α activity will dominate the radiation field.


2002 ◽  
Vol 713 ◽  
Author(s):  
Jan Marivoet ◽  
Xavier Sillen ◽  
Dirk Mallants ◽  
Peter De Preter

ABSTRACTIn Belgium the possibilities to dispose of high-level waste in the plastic Boom Clay formation has been studied since 1975. Consequently many results of the site characterisation programme are already available. Various performance assessments have been carried out dealing with the disposal of high-level waste arising from reprocessing and with direct disposal of spent fuel. The performance assessment consists of two main steps: scenario development and consequence analyses. The scenario analysis is based on a catalogue of features, events and processes (FEPs) having the potential to influence the behaviour of the repository system. Two main groups of scenarios are distinguished. The normal evolution scenario, including a number of variants, treats the FEPs that are fairly sure to take place. Altered evolution scenarios focus on FEPs having a probability of occurrence lower than one but that might influence the performance of the repository system. For the impact analyses, a robust concept approach was introduced, which focused the analyses on a limited number of well-characterised barriers and processes. The impact analyses are complemented with sensitivity and uncertainty analyses based on deterministic and probabilistic approaches.


2006 ◽  
Vol 932 ◽  
Author(s):  
Sonia Salah ◽  
Christelle Cachoir ◽  
Karel Lemmens ◽  
Norbert Maes

ABSTRACTSince reprocessing is no longer the reference policy in Belgium, studies on the direct disposal of spent fuel in a clay formation have gained increased interest in the last years. In order to determine to what extent the clay properties and the α-activity influence the dissolution kinetics of spent fuel for the long term disposal, static dissolution tests have been performed on 5 different types of α-doped UO2, representing a PWR fuel with a burn-up of 45 or 55 GWd · tHM−1 and fuel ages ranging between 150 and 90,000 years, in different Boom Clay (BC) media at room temperature and in an anoxic atmosphere for 90 to 720 days. The uranium activity in the clay fraction over time was found to be much higher than the U-activity in the leachates, which has been mainly ascribed to the high retention capacity of the BC. The average dissolution rate between 0 and 90 days obtained for the 5 types of α-doped UO2 were all found to be high and quite similar at ∼263 µg · m−2· d−1and a “longer-term” rate (181 to 720 days) ranging between zero and 15 µg · m−2· d−1. These results suggest that the activity of the fuels does not seem to have an effect on the UO2 dissolution rates under the considered test conditions. In order to study the partition/redistribution of U during UO2dissolution, sequential extraction experiments were performed. Results of the latter have provided a better mechanistic understanding of BC/spent fuel interaction processes as well as information on the role of the different minerals controlling the U-retention/immobilization.


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