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An overview is given of an International Atomic Energy Agency Coordinated Research Project (CRP) on the treatment of irradiated graphite (i-graphite) to meet acceptance criteria for waste disposal. Graphite is a unique radioactive waste stream, with some quarter-million metric tons worldwide eventually needing to be disposed of. The CRP has involved 24 organizations from 10 Member States. Innovative and conventional methods for i-graphite characterization, retrieval, treatment and conditioning technologies have been explored in the course of this work, and offer a range of options for competent authorities in individual Member States to deploy according to local requirements and regulatory conditions.
This year marks the 101st anniversary of the sinking of the “unsinkable” RMS Titanic. On April 15, 1912, the Titanic struck an iceberg in the North Atlantic Ocean on its maiden voyage from Southampton, UK, to New York City. There was no single cause for the loss of the Titanic, rather the improbable combination of errors in human design and decision combined with unforeseeable circumstance lead to the loss of over 1,500 lives. The failure appears to have occurred over a range of spatial and temporal scales – from the atomic-scale process of embrittlement of iron rivets to global-scale fluctuations in climate and ocean currents. Regardless of the specific combination of causes, this failure in design and practice led to impressive improvements in both. Disaster and tragedy are harsh teachers, but critical to improvement and progress.
The important question for the nuclear waste management community is how do we learn and improve our waste management strategies in the absence of being able to fail. A geologic repository “operates” over a very distant time frame, and today’s scientists and engineers will never have the benefit of studying a failed system. In place of a failure that is followed by improvement and progress, we can only offer a general consensus on disposal strategies supported by a wide array of evidence and risk assessments. However, it may well be that consensus leads to complacency and compromise, both of which are harbingers of disaster. With this concern in mind, this is the time to review our fundamental approach, particularly the methodologies used in risk assessments that have us calculate risk out to one million years. The structure of standards and implementing regulations, as well as the standard-of-proof for compliance, should be reexamined in order to determine whether their requirements are scientifically possible or reasonable. The demonstration of compliance must not only be compelling, but must also be able to sustain scientific scrutiny and public inquiry. We should benefit from the sobering reality of how difficult it is to anticipate future failures even over a few decades. We should be humbled by the realization that for a geologic repository we are analyzing the performance, success vs. failure, over spatial and temporal scales that stretch over tens of kilometers and out to a hundreds of thousands of years.
Various candidate waste matrices such as nuclear waste glasses, ceramic waste forms and low-specification “storage” MOX have been considered within the current UK geological disposal program for the immobilization of separated civilian plutonium, in the case this material is declared as waste. A review and evaluation of the long-term performance of potential plutonium waste forms in a deep geological repository showed that (i) the current knowledge base on the behavior and durability of plutonium waste forms under post-closure conditions is relatively limited compared to HLW-glasses from reprocessing and spent nuclear fuels, and (ii) the relevant processes and factors that govern plutonium waste form corrosion, radionuclide release and total systems behavior in the repository environment are not yet fully understood in detail on a molecular level. Bounding values for the corrosion rates of potential plutonium waste forms under repository conditions were derived from available experimental data and analogue evidence, taking into account that the current UK disposal program is in a generic stage, i.e. no preferred host rock type or disposal concept has yet been selected. The derived expected corrosion rates for potential plutonium waste forms under conditions relevant for a UK geological disposal facility are in the range of 10-4 to 10-2 g m-2 d-1 and 10-5 to 10-4 g m-2 d-1 for borosilicate glasses, and generic ceramic waste forms, respectively, and ∼5·10-6 g m-2 d-1 for storage MOX. More realistic assessments of the long-term behavior of the waste forms under post-closure conditions would require additional systematic studies regarding the corrosion and leaching behavior under more realistic post-closure conditions, to explore the safety margins of the various potential waste forms and to build confidence in long-term safety assessments for geological disposal.
Spent nuclear fuel from TVO's (Teollisuuden Voima Oy) and Fortum's nuclear power plants will be deposited deep in the crystalline bedrock in Olkiluoto, Western Finland. The bedrock needs to be well characterized to assess the risks inherent to the waste disposal at the site. If radionuclides (RN) are transported, it happens via water conducting fractures. Retardation may occur either by diffusion into stagnant pore water or by immobilization on mineral surfaces of the rock matrix.
RN’s retardation from flowing water is linked to parameters defining porosity and microscopic rock pore structure, such as pore size distribution, connectivity, tortuosity and constrictivity, and by the mineralogy and chemical nature of the minerals and charge of the pore surfaces.
In this work, centimeter scale rock cores from Olkiluoto were investigated. The work is part of the in situ project REPRO (Experiments to investigate Rock Matrix Retention Properties) where the diffusion and sorption of RN are studied experimentally. Porosity and pore structures were characterized with the PMMA autoradiography method and polarized microscopy, which was used also to ascertain the mineralogy of the samples.
The results show that the rock from the REPRO site has low porosity with a mean value of 0.5% and a range of 0.1-1.5%. Rock heterogeneity explains the variation of porosity values. Correlation between the porosity and the mineralogy was found. Areas of high porosity correspond to areas of altered minerals, such as cordierite, biotite and plagioclase, which cover spatially between 10 and 20% of the rock volume
We have developed a “Relative Rates Method” to make bounding calculations regarding radionuclide migration due to uplift/erosion (“exhumation”) of a HLW repository. Results show that this method can apply to a wide range of different uplift rates and erosion rates. In addition, for the long time period, it was shown that the relative difference of uplift rate / erosion rate and potential hydraulic change arising from extreme uplift/erosion could affect radionuclide release and migration, thus uplift/erosion concerns should be fed back to site selection. Our method provides a credible and defensible basis for analysis and interpretation of possible uplift/erosion impacts for future volunteer sites.
To evaluate a change of chemical species of groundwater composition by the metabolism of microbes, which will be introduced to deep underground from the surface and be in a deep underground, is important for the discussion of the microbial effects on the performance assessment of the high-level radioactive waste repository. The purpose of this study is to develop of a microbial kinetics database to evaluate their activities in the deep underground environment.
Some microbial metabolism data were collected and constructed their kinetics database for aerobic, denitrifying, manganese reducing, iron reducing, sulfate reducing, methanogenic and acetogenic bacteria to evaluate above groundwater chemistry. About 1260 data were selected by literature survey for some journals and books published from 1960s and summarized in this microbial kinetics database. Some sensitivity analyses were performed for some parameter of metabolism of microbes.
Around the radioactive waste repository, the pH of the groundwater greatly changes from 8 to 13 and the groundwater contains a relatively large quantity of calcium (Ca) and sodium (Na) ions due to cementitious materials used for the construction of the geological disposal system. Under such conditions, the deposition behavior of silicic acid is one of the key factors for the migration assessment of radionuclides. The deposition and precipitation of silicic acid with the change of pH and coexisting ions may contribute to the clogging in flow paths, which is expected as the retardation effect of radionuclides. Thus, this study focused on the deposition behavior of silicic acid under the condition of relatively high Ca or Na concentration.
In the experiments, Na2SiO3 solution (250 ml, 14 mM, pH>10, 298 K) was prepared in a polyethylene vessel containing amorphous silica powder (0.5 g) as the solid phase. Then, a buffer solution (to adjust to 8 in pH), HNO3, and Ca(NO3)2 as Ca ions or NaCl as Na ions were sequentially added. Such a silicic acid solution becomes supersaturated, gradually forming colloidal silicic-acid and/or the deposit on the solid surface. In this study, the both concentrations of soluble and colloidal silicic-acid were monitored over a 40-day period. As a result, the deposition rate of silicic acid decreased with up to 5 mM in Ca ions. Besides, Na ions with up to 0.1 M slightly increased the deposition rate. Under the conditions of [Na+]>0.1 M or [Ca2+]>5 mM, the supersaturated silicic acid immediately deposited. These suggest that Na or Ca ions strongly affect the deposition behavior of supersaturated silicic-acid, depending on the surface alteration of solid phase, the change of zeta potential and the decrease of water-activity due to the addition of electrolytes (coexisting ions).
In the new DR-A in-situ diffusion experiment at Mont Terri, a perturbation (replacement of the initial synthetic porewater in the borehole with a high-salinity solution) has been induced to study the effects on solute transport and retention, and more importantly, to test the predictive capability of reactive transport codes. Reactive transport modeling is being performed by different teams (IDAEA-CSIC, PSI, Univ. Bern, Univ. British Columbia, Lawrence Berkeley Natl. Lab.). Initial modeling results using the CrunchFlow code and focusing on Cs+ behavior are reported here.
The formation of uranyl peroxide phases was identified as a corrosion product of spent fuel by Hanson et al . The subsequent analysis of this phase showed that metastudtite retained 241Am, 237Np and 239Pu . In this study, the retention of radionuclide Pu4+ and An3+, released from the spent fuel matrix into studtite structure, has been evaluated by the precipitation of studtite from uranyl dissolution with variable concentrations of REE (Th, Nd, Sm and Eu). Three different precipitation conditions parameters were studied: media of synthesis, time of synthesis and REE concentration. Synthesized phases were characterized by XRD and the cell parameter was calculated. The REE incorporation was determined by ICP-MS analysis. The results showed that studtite could incorporate 63% of Th in solution during its precipitation. Changes in the “a” cell parameter were identified. The results suggest that studtite coprecipitated with REE could play a role as a limiting for the REE mobility.
The migration behavior of plutonium is expected to be affected by the corrosion products of carbon steel in compacted bentonite at high-level waste repositories. Electrochemical experiments were carried out to simulate the reducing environment created by ferrous iron ions in equilibrium with anoxic corrosion products of iron. The concentration profiles of plutonium could be described by the convection -dispersion equation to obtain two migration parameters: apparent migration velocity Va and apparent dispersion coefficient Da. The apparent migration velocity was evaluated within 1 nm/s and was found to be independent of the experiment duration and the dry density of bentonite in the interval 0.8-1.4 Mg/m3. The apparent dispersion coefficient increased with the experiment duration at a dry density of 1.4 Mg/m3. The results for other dry densities also showed the same trend. These findings indicate that plutonium migration likely starts after ferrous ions reach the plutonium, in other words, the reducing environment due to ferrous ions could change the chemical form of plutonium and/or the characteristics of compacted bentonite. The apparent diffusion coefficient was estimated to be around 0.5 to 2.2 µm2/s and increased with decreasing the dry density of bentonite.
Matrix diffusion is a key process for radionuclide retention in crystalline rocks. Within the LTD project (Long-Term Diffusion), an in-situ diffusion experiment in unaltered non-fractured granite was performed at the Grimsel Test Site (www.grimsel.com, Switzerland). The tracers included 3H as HTO, 22Na+, 134Cs+ and 131I- with stable I- as carrier.
The dataset (except for 131I- because of complete decay) was analyzed with different diffusion-sorption models by different teams (NAGRA / IDAEA-CSIC, UJV-Rez, JAEA, Univ. Poitiers) using different codes, with the goal of obtaining effective diffusion coefficients (De) and porosity (ϕ) or rock capacity (α) values. A Borehole Disturbed Zone (BDZ), which was observed in the rock profile data for 22Na+ and 134Cs+, had to be taken into account to fit the experimental observations. The extension of the BDZ (1-2 mm) was about the same magnitude as the mean grain size of the quartz and feldspar grains.
De and α values for the different tracers in the BDZ are larger than the respective values in the bulk rock. Capacity factors in the bulk rock are largest for Cs+ (strong sorption) and smallest for 3H (no sorption). However, 3H seems to display large α values in the BDZ. This phenomenon will be investigated in more detail in a second test starting in 2013.
Experiments on the dissolution kinetics of natural pyrrhotite (FeS1-x-) and pyrite (FeS2) under imposed redox conditions to evaluate the oxygen uptake capacity of both minerals were carried out at 25°C and 1 bar. Experimental data indicate that in both cases, Fe(II) released from dissolution of these Fe-bearing sulphides is kinetically oxidized to Fe(III) to precipitate as Fe(III)-oxyhydroxides. While the system is pH controlled by the extent of the sulphide oxidation, Eh is controlled by the redox pair Fe2+/Fe(III)-oxyhydroxides. Pyrrhotite dissolution is faster than that of pyrite but generates less acidity. Consequently, the achieved redox value is more reducing. Experimental data show that the oxidation rates of both minerals (in mol·g-1·s-1) are equivalent under the studied conditions. This fact gives a new opportunity to quantify the reductive buffering capacity of pyrrhotite, for which no kinetic rate law has been still established.
Some kinds of material in the environment due to the accident at the Fukushima Nuclear Power Plant have been contaminated by radioactive cesium (134Cs and 137Cs), which are represented by dehydrated sludge, surface soil and disaster wastes generated by the Great East Japan Earthquake. Treatment (transportation, temporary storage and incineration) and disposal of the contaminated materials should be carried out while ensuring the safety of radiation for the workers and the public. In this study, in order to provide the technical information for making the criteria, the dose estimation for scenarios on the treatment and disposal is conducted, based on the method used for driving the clearance levels in Japan. Minimum radioactive cesium concentration in contaminated material, that is, limiting activity concentration which is practicable for ordinary treatment and/or disposal, is calculated from the dose results, corresponding to the effective dose criteria indicated by the Nuclear Safety Commission of Japan. From the calculation result, it is suggested that it is necessary to forbid reusing the disposal site as construction, resident and agriculture in which the calculated doses for the public are higher than those in the other exposure pathways. Limiting concentration of radioactive cesium (134Cs and 137Cs) is derived to be 8,900Bq/kg for the external exposure pathway in landfill work under the condition of limited reuse of the site. In the case of the concentration below 8,900Bq/kg, the calculated dose of the resident due to direct and sky-shine radiation scattered in the air and ground from the interim storage place is less than 1mSv/y, irrespective of the distance from the storage place.
A set of computer programs has been developed to draw chemical-equilibrium diagrams. This new software is the Java-language equivalent to the Medusa/Hydra software (developed some time ago in Visual basic at the Royal Institute of Technology, Stockholm, Sweden). The main program, now named “Spana” calls Java programs based on the HaltaFall algorithm. The equilibrium constants that are needed for the calculations may be retrieved from a database included in the software package (“Database” program). This new software is intended for undergraduate students as well as researchers and professionals.
The “Spana” code can be easily applied to perform radionuclide speciation and solubility calculations of minerals, including solubility calculations relevant for the performance assessment of a nuclear waste repository. In order to handle ionic strength corrections in such calculations several approaches can be applied. The “Spana” code is able to perform calculations based on three models: the Davies equation; an approximation to the model by Helgeson et al. (HKF); and the Specific Ion-Interaction Theory (SIT). Default SIT-coefficients may be used, which widens the applicability of SIT significantly.
A comparison is made here among the different ionic strength approaches used by “Spana” (Davies, HKF, SIT) when modelling the chemistry of radionuclides and minerals of interest under the conditions of a geological repository for nuclear waste. For this purpose, amorphous hydrous Thorium(IV) oxide (ThO2(am)), Gypsum (CaSO4·2H2O) and Portlandite (Ca(OH)2) solubility at high ionic strengths have been modelled and compared to experimental data from the literature. Results show a good fitting between the calculated values and the experimental data especially for the SIT approach in a wide range of ionic strengths (0-4 M).
Cs-137 was accidentally spilled in an industrial waste repository located in a salt marsh in southern Spain, and a permeable reactive barrier was proposed to retain it. Cs adsorption properties of different natural clayey materials were analyzed. The salt marsh waters show high salinity and high chemical variability, therefore Cs adsorption was also analyzed in the presence of competitive ions, especially K+ and NH4+.
Cs adsorption was non-linear in all the analyzed materials, indicating more than one adsorption sites with different selectivity. It was shown that in mixed clay systems with illite, montmorillonite and kaolinite, the presence of illite favors Cs retention at low and medium Cs loadings and montmorillonite at high Cs loadings. In the presence of illite and montmorillonite, kaolinite plays almost no role in Cs retention. The presence of K+ and NH4+ significantly hinders cesium adsorption.
The integrated sorption and diffusion (ISD) model has been developed to quantify radionuclide transport in compacted bentonite. The current ISD model, based on averaged pore aperture and the Gouy-Chapman electric double layer (EDL) theory can quantitatively account for diffusion of monovalent cations and anions under a wide range of conditions (e.g., salinity, bentonite density). To improve the applicability of the current ISD model for multivalent ions and complex species, the excluded volume effect and the dielectric saturation effect were incorporated into the current model, and the modified Poisson-Boltzmann equations were numerically solved. These modified models had little effect on the calculation of effective diffusivity of Sr2+/Cs+/I−. On the other hand, the model, modified considering the effective electric charge of hydrated ions, calculated using the Gibbs free energy of hydration, agreed well with the diffusion data including those of Sr2+.
Compacted bentonite barrier in radioactive waste repositories is expected to prevent radionuclide migration, due to its high sorption capability for many radionuclides. This study analyses whether the addition of Al2O3 nanoparticles (NPs) enhances the sorption properties of bentonite. The study was carried out with 109Cd, highly pollutant heavy metal and divalent fission product. Sorption experiments were conducted in NaClO4 at different ionic strengths (5·10-4 to 10-1 M) and pH (2 to 10), using mixtures of sodium homoionised bentonite and Al2O3 in different proportions.
It has been probed that addition of Al2O3 NPs to bentonite enhances Cd sorption at pH higher than 6. The effect of Al2O3 NPs addition on the surface properties of bentonite colloids was also analyzed by measuring particle size and surface charge in all studied systems.
In order to analyze the C-14 inventory and leaching rate for safety evaluation of transuranic waste disposal, it is necessary to establish an analytical method that can measure C-14 with sufficient precision . Oxidative decomposition of organic compounds containing C-14 is carried out to absorb carbon dioxide (CO2) in an alkaline solution, which is mixed with a liquid scintillation cocktail, and the amount C-14 is quantified by measuring a beta ray spectrum with a liquid scintillation counter. It has been difficult to completely decompose carbon compounds in a sample, even to CO2, by using conventional oxidizing agents. In the work described here, we improved the method of oxidative decomposition used to completely decompose carbon compounds using peroxydisulfuric acid (K2S2O8). When C-14 in the form of CO2 was absorbed in a sodium hydroxide (NaOH) aqueous solution, only 80% of the actually used quantity was detected. Total organic carbon measurements showed that the entire quantity of CO2 was absorbed by NaOH. When NaOH aqueous solution was used, it was found that only the analytical value was 80%. The entire quantity of the actually used carbon could be measured by absorbing the CO2 in Carbo-Sorb®. An anion form and a neutral molecule exist in the organic compound released from activated metals. In order to identify organic compounds efficiently, fractionation into an anion and a neutral molecule and separation by high performance liquid chromatography (HPLC) are necessary. Here, we propose the combined use of an ion exchange resin and HPLC as an improved technique for identification of the chemical species.
The use of generic sorption data in PA requires the transfer of the data to the PA-specific conditions. A site-specific Kd setting approach for PA calculations was tested, comparing two data transfer procedures. First transfer of sorption data can be done through semi-quantitative estimation procedures, by considering differences between experimental and PA geochemical conditions (sorption capacity, radionuclide speciation, competitive reactions, etc.). On the other hand, thermodynamic sorption models allow to estimate Kd variations directly, based on quasi-mechanistic understanding. The present paper focuses on illustrating example calculations regarding the derivation of Kd values, and their uncertainties, of Cs, Ni, Am and Th, for the mineralogical and geochemical conditions of the mudstone system at the Horonobe URL. Clay minerals (illite and smectite) were considered as sorption-relevant minerals in all cases. The Kd setting results were compared with Kd measured for Horonobe mudstone by batch experiments. The results indicate that Kd can be quantitatively evaluated from generic sorption data when adequate data and models are available. The careful evaluation and conjunctive use of calculated and measured Kd values can enhance the reliability of Kd setting and uncertainty assessments.
Selenium (Se) is an important element for assessing the safety of high-level waste disposal. Se is redox-sensitive, and its oxidation state varies from -2 to 6 depending on the redox conditions and pH of the solution. Large quantities of ferrous ions formed in bentonite due to corrosion of carbon steel overpack after the closure of a repository are expected to maintain a reducing environment near the repository. Therefore, the migration behavior of Se in the presence of Fe in bentonite was investigated by electrochemical experiments. Na2SeO3 solution was used as tracer solution. Dry density range of bentonite was from 0.8 to 1.4 ×103 kg/m3.
Results indicated that Se was strongly retained by the processes such as precipitation reaction with ferrous ions in bentonite. Se K-edge X-ray absorption near-edge structure (XANES) measurements were performed at the BL-11 beamline at SAGA Light Source, and the results revealed that the oxidation state of Se in the bentonite remained Se(IV).