Steacute;phane Gin, CEA Valrho
Robert Jubin, Oak Ridge National Laboratory
Josef Matyaacute;s, Pacific Northwest National Laboratory
Eric Vance, Australian Nuclear Science and Technology Organization
Symposium Support Department of Energy
EE2: Capture and Immobilization of Radionuclides II
Monday PM, December 01, 2014
Hynes, Level 2, Room 204
2:30 AM - *EE2.01
Technetium Getters to Improve Cast Stone Performance
Nik Qafoku 1 Jim Neeway 1 Amanda Lawter 1 Joe Westsik 2
1Pacific Northwest National Laboratory Richland USA2Pacific Northwest National Laboratory Richland USAShow Abstract
Technetium-99 (99Tc) is one of radioactive tank waste components contributing the most to the environmental impacts associated with disposal of radioactive wastes currently stored in underground tanks at the Hanford site, WA. Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. Research is being conducted to improve the retention of Tc in the Cast Stone waste forms.
One method to improve the performance of the Cast Stone waste forms is addition of “getters” that selectively sequester Tc. Getter materials that remove Tc from solution are expected to reduce Tc(VII) to the less mobile Tc(IV), In order to determine the effectiveness of the various getter materials prior to their solidification in Cast Stone, a series of batch sorption experiments was performed. Seven getter materials were tested for Tc. Testing involved placing getter materials in contact with spiked waste solutions for periods up to 45 days with periodic solution sampling. Two different solution media, 18.2 MOmega; DI H2O and a 7.8 M Na LAW waste simulant, were used in the batch sorption tests. Each test was conducted at room temperature in an anoxic chamber containing N2 with a small amount of H2 (0.7%) to maintain anoxic conditions.
In this paper we present the results of the batch experiments conducted to determine potential Tc getter materials that will undergo continued testing, selection and subsequently incorporation into Cast Stone. Results indicate that most materials perform better in the DI H2O (18.2 MOmega;) solution than in the 7.8 M Na LAW waste simulant. We have determined that Tc sequestration may be affected by the presence of other redox sensitive elements that are present in the waste simulant, such as Cr(VI). The Tc getter materials have been examined through various solid-state characterization techniques such as SEM/EDS and XANES. The results indicate that the Tc precipitate differs depending on the getter material and that Tc is reduced in most of the getters but at different extent, from Tc(VII) to Tc(IV).
3:00 AM - EE2.02
Immobilization of Technetium and Caesium in ABO4 Compounds
Eugenia Y Kuo 1 Simon C Middleburgh 1 Meng J Qin 1 Gordon J Thorogood 1 Gregory R Lumpkin 1
1Australian Nuclear Science and Technology Organization Kirrawee DC AustraliaShow Abstract
The immobilization of 99Tc and 137Cs, two problematic nuclear waste isotopes, in an ABO4-type structure (where A is an alkali metal and B is either Tc or Ru) has been investigated using atomic scale modelling techniques. The structural stability and free energies of several ABO4 compounds, including that of CsTcO4 were computed. Most perfect compositions were calculated to be scheelite structured. To understand the potential wasteforms&’ stabilities during and after transmutation of 99Tc to 99Ru, and 137Cs to 137Ba, we also computed the structures and energies of a range of defective ABO4 compounds. Full and partial transmutation of the waste isotopes were considered, i.e., those of compounds such as A(Tc1-xRux)O4, (Cs1-xBax)TcO4 and ARuO4. We present a number of compositions that may prove to be suitable for either 99Tc waste as well as the simultaneous encapsulation of 99Tc and 137Cs.
3:15 AM - EE2.03
Simulating the Selective Adsorption of Pertechnetate to Oxyanion-SAMMS
Christopher David Williams 2 1 Karl Travis 2 Neil Burton 1 John Harding 2
1University of Manchester Manchester United Kingdom2University of Sheffield Sheffield United KingdomShow Abstract
99Tc, a radioactive fission product, is discharged in nuclear fuel reprocessing operations. In its common form in the environment (TcO4minus;) it is a major concern for the remediation of contaminated waters. The difficulty with the removal of TcO4minus; is a result of its high mobility in solution and the presence of a high concentration of competing anions such as SO42minus;. A functionalized material, developed at PNNL, known as self-assembled monolayers on mesoporous supports (SAMMS) has previously been found to selectively remove contaminant monovalent oxyanions even in the presence of the divalent SO42minus;.
In this work we have constructed an atomistic model of the SAMMS material and validated the model by comparison to experiment. Density functional theory (DFT) calculations were used to parameterize a classical force field that accounts for the specific interactions of the competing oxyanions with the monolayer. Potentials of mean force for oxyanion adsorption were obtained using umbrella sampling and molecular dynamics (MD) simulations in order to study the material&’s preference for binding TcO4minus; over SO42minus;. The results show that the pore structure is a key parameter governing the material&’s oxyanion selectivity. Finally, we suggest ways in which the structure of the material can be optimized in order to maximize TcO4minus; adsorption.
3:30 AM - EE2.04
A Novel Vanadosilicate with Hexadeca-Coordinated Cs+ Ions as a Highly Effective Cs+ Remover
Won Kyung Moon 1 Shuvo Jit Datta 1 Do Young Choi 1 In Chul Hwang 1 Kyung Byung Yoon 1
1Sogang University Seoul Korea (the Republic of)Show Abstract
Among various radioactive nucleotides, 137Cs is the most dangerous radioactive nucleotide because of its high fission yield (6.09 %), medium half-life (30.17 years), and very high solubility in water regardless of its counter anion. Once released into the environment, it easily spreads in nature and enters the food chain, causing enormous damage to human and animal health. In this respect, the effective removal of 137Cs+ ions from contaminated groundwater, seawater and radioactive nuclear waste solutions is crucial for public health and for the continuous operation of nuclear power plants. However, it is an extremely difficult task because 137Cs+ concentrations are usually incomparably lower than those of the co-existing competing cations (Na+, Ca2+, Mg2+, K+, and others).
Herein we report a novel microporous vanadosilicate K-SGU-45 with mixed valences of vanadium (IV and V), which shows an excellent capturing and immobilization of Cs+ from ground water, seawater and highly acidic and basic nuclear waste solutions. This material is superior to other known materials in terms of selectivity, capacity, and kinetics, in particular, at very low Cs+ concentrations, it was found to be the most effective material for the removal of radioactive Cs+.
This work will trigger the syntheses of various vanadium and other transition-metal silicates that capture various radioactive nuclides, such as 90Sr2+ ions, and other toxic heavy-metal ions. Furthermore, the discovery of unprecedented hexadeca-coordinated Cs+ centers, which corresponds to the highest coordination number ever observed in chemistry, has been described.
3:45 AM - EE2.05
New Materials for Strontium Removal from Nuclear Waste Streams
Sav Neoklis Savva 1 Joseph A. Hriljac 1
1University of Birmingham Birmingham United KingdomShow Abstract
Strontium-90 and caesium-137 are waste products produced by fission processes; both have long half-lives of 28 and 30 years respectively. Strontium in particular can have a severe biological impact as it has been shown to accumulate in bones after the intake of contaminated food or water.
Ion exchange materials, such as crystalline silicotitanite (CST, Na2Ti2O3SiO4middot;2H2O) and commercially available IONSIV, have been implemented in order to target and remove these harmful radioisotopes and have been shown to be somewhat effective. However Strontium and caesium have proven difficult to immobilise selectively in some cases as ion exchange uptake has been shown to be retarded by the presence of competing cations such as calcium or magnesium .
A range of new materials similar to CST but based on zirconium and tin silicates, such as NaKSnSi3O9.H2O pictured below, have been investigated for their potential ion exchange applications. These materials are robust against thermal, chemical and radioactive conditions which would make them ideal for use in radioactive waste streams.
A range of materials and the ion exchanged heat treated waste forms have been studied using XRD, TGA, XRF, SEM and EDX analysis in order to characterise the structures and probe the ion exchange properties.
1.R.G. Anthony, C.V. Philip, R.G. Dosch, Waste Manage, 1993,13, 503
2.T. Möller, R. Harjula, M. Pillinger, A. Dyer, J. Newton, E. Tusa,S. Amin, M. Webb and A. Araya, J. Mater. Chem.,2001, 11, 1526
EE3: Atomic Simulation and Modeling
Monday PM, December 01, 2014
Hynes, Level 2, Room 204
4:45 AM - EE3.01
Advancing the Modelling Environment for the Safety Assessment of the Swedish LILW Repository at Forsmark
Henrik von Schenck 1 Ulrik Kautsky 1 Bjoern Gylling 1 Elena Abarca 2 Jorge Molinero 2
1Swedish Nuclear Fuel and Waste Management Company Stockholm Sweden2Amphos 21 Consulting S.L. Barcelona SpainShow Abstract
An extension of the Swedish final repository for short-lived radioactive waste (SFR) is planned and a safety assessment has been performed as part of the licensing process. Within this work, steps have been taken to advance the modelling environment to better integrate its individual parts. It is desirable that an integrating modelling environment provides the framework to set up and solve a consistent hierarchy of models on different scales. As a consequence, the consistent connection between software tools and models needs to be considered, related to the full assessment domain. It should also be possible to include the associated geometry and material descriptions, minimizing simplifications to source data. The usefulness of the analysis software Comsol Multiphysics as component of an integrating modelling environment has been tested and examples of development work are presented.
Geometry handling is an important part of the modelling process and is closely related to modelling assumptions and simplifications. For the SFR, the relevant geometry includes tunnel systems and storage vaults, as well as engineered structures and barriers. CAD geometries developed during planning and design work have been successfully imported into Comsol. The landscape above the repository also constitutes relevant geometrical input for assessment modelling. Development work has allowed the import of geographic information system (ArcGIS) data into Comsol, incorporating digital elevation models as well as soil and sediment domains into model geometries.
The ability to set up and solve a consistent hierarchy of models on different scales is an important capability of an integrating modelling environment. Extracting models for repository scale hydrology from regional hydrogeology models and regional surface hydrology models are two examples. The regional hydrogeology model of the SFR site covers several square kilometres of land and reaches depths of approximately one kilometre. The repository scale model is contained within the regional model and has dimensions one order of magnitude smaller. To calculate the detailed groundwater flow through the repository requires the proper boundary conditions from the regional hydrogeology. A consistent connection was achieved by programming an interface allowing Comsol to extract the near-field boundary conditions and bedrock property fields from the regional model, set up and solved in the DarcyTools software.
The repository scale hydrology models provided a basis for further model developments focused on coupled processes. An interface between Comsol the geochemical simulator PhreeqC has been developed to support reactive transport studies. An important test case involved radionclide transport in a 3D model of a catchment area. The dynamic surface hydrology was simulated with MIKE SHE and coupled to detailed chemical processes occurring in soils and sediments.
5:00 AM - EE3.02
High Performance Computing to Simulate Cement Grout Degradation in a Deep Geological Repository
Jorge Molinero 1 Luis Manuel de Vries 1 Hedieh Ebrahimi 1 Urban Svensson 2 Peter Lichtner 3 Birgitta Kalinowski 4 Bjoern Gylling 4
1Amphos 21 Consulting Barcelona Spain2Computer-Aided Fluid Engineering AB Lyckeby Sweden3OFM Research Los Alamos USA4SKB Stockholm SwedenShow Abstract
Reactive transport modelling entails the integration of hydrogeology and geochemistry. One of the challenges for such integration is the large amount of computational resources needed due to the high non-linearity of the resulting system of equations. Taking advantage of new developments of powerful numerical tools, and based on high performance parallel computing, the solution of large-scale hydro-thermal-geochemical-mechanical models has become possible. A software solution, denoted iDP, has been developed which serves as an interface between 2 standalone simulators: DarcyTools [for groundwater flow in fractured rocks] and PFLOTRAN [for reactive solute transport]. iDP has been applied for the first time to test the new update of Mare Nostrum, the main machine at the Barcelona Supercomputing Centre, the National Supercomputing Centre in Spain. An average of 8,000 processor cores during 15 days were used to solve a large-scale (100 Mcells), long-term (20,000 years) simulation to evaluate the degradation of cement grout that will be injected in the fractures of the granitic rocks during the construction of a deep geological repository for spent nuclear fuel in Forsmark (Sweden). The simulation integrates the complex 3D groundwater flow accounting for the Discrete Fracture Network (DFN) of the site, and the complexity of the geochemical system involved in cement grout dissolution and secondary minerals precipitation within the flowing fractures. Model results allow evaluating the expected durability of the injected cement grout, as well as to evaluate the risk of hyper-alkaline groundwater development and migration towards the depositional area of the repository. This work shows that High Performance Computing of reactive solute transport is a reliable and powerful tool for decision makers involved in the planning and constructions of deep geological repositories for nuclear waste.
5:15 AM - EE3.03
A GoldSim Model for a Probabilistic Safety Assessment of a Trench Repository for Low-Level Waste
Youn-Myoung Lee 1 Jongtae Jeong 1
1Korea Atomic Energy Research Institute Yuseong, Daejeon Korea (the Republic of)Show Abstract
A simple and effective model for a safety assessment of a conceptual repository system, in which low-level radioactive wastes that arises from nuclear power plants and other sources has been developed using the commercial GoldSim development tool. The repository system is assumed to be planned for construction on the surface area near the seashore. The computer program based on this model, developed as a GoldSim template, is ready for a total system performance assessment (TSPA), and is able to probabilistically evaluate a nuclide release from a repository and farther transport into the geosphere and biosphere under various normal, disruptive events, and scenarios that can occur after a failure of a waste drum with associated uncertainty. To quantify the nuclide release and transport through the various pathways possible in the near- and far-fields of the repository system under a normal groundwater flow and some alternative scenarios, illustrative evaluations are made and demonstrated through this study. Even though all parameter values associated with the repository system were assumed for the time being, the illustrative results should be informative since the evaluation of such releases is very important not only in view of the safety assessment of the repository, but also for the design feedback of its performance.
The 200L storage drums for low-level waste, which amounts to a total of 125,000 drums, are to be disposed of in concrete containers and then buffered by gravel or grouted with concrete. Impervious materials and multilayered covers for preventing water infiltration and some erosion as well as nuclide release are considered to place on the roof. In GoldSim modeling, a trench and its surrounding are discretized into several compartments ready for run-off, infiltration as well as diffusive and advective transport in and among them. Several principal release pathways from the trenches are set in place: the upper, side, and base pathways, all of which simultaneously reach to the far-field transport. All releases from the trenches are then later transported along with various unsaturated and saturated pathways including surface and subsurface groundwater flow pathways into the natural far-field area.
For trench type repositories at the surface or possibly subsurface depth, normally and commonly, once leakage from a damaged radioactive waste package of a drum, and through tiny holes, happens, the nuclides will spread out to the buffer material surrounding the drum, and then into other possible regions in the trench before farther transporting into the biosphere through various pathways. In the case of transport into the rock medium under the repository, the internal fractures and the major water conducting features (MWCFs) that are assumed to exist in the far-field area of the repository could be one of the main pathways through which the nuclides finally reach the human environment by passing over the geosphere-biosphere interfaces for exposure to human bodies.
The scenario mainly considered here for a probabilistic safety assessment is a normal case, under which nuclides are released by overflow and/or groundwater that normally flows along their own preferential pathways after release from each repository. Through this study, a probabilistic behavior of nuclide releases from a low-level waste trench type repository is illustrated with varying parameters, which were selected among many others in view of their possible consequences and probabilities.
5:30 AM - EE3.04
Atomistic Simulations of Clay Minerals for Nuclear Waste Management
Marco Molinari 1 David MS Martins 2 Stephen C Parker 1 Mario A Goncalves 2
1University of Bath Bath United Kingdom2Universidade de Lisboa Lisboa PortugalShow Abstract
The safe treatment of nuclear waste poses a lasting risk to the environment and has high costs. Buried repositories represent the long term storage which is required to be stable. The stability includes many aspects such as chemical and mechanical stabilities as well as impermeability. Clay minerals are excellent candidates to maintain a long lasting seal of the nuclear waste repository due to their large adsorption capacity and swelling characteristics in aqueous suspensions. However, the interaction and transport of radionuclides in clay minerals, including organic clay minerals, still need to be fully addressed.
Atom level simulations have not yet been fully exploited to investigate these processes not least because of the complexities involved. Here we present our recent work to gain atomistic insights into the factors controlling the interaction of heavy and radioactive ions at clay mineral - water interfaces. Quantum and potential based techniques are used to explore the evolution of systems of different sizes and for different lengths of time enabling us to efficiently evaluate structural and dynamical properties of this class of geosorbents. The interaction of these ions with clay minerals is generally thought to occur on the basal plane, which dominates their morphologies and has been the focus of many investigations. However, the edge surfaces are more reactive and with a greater range of compositions and charge states can indeed provide more efficient interaction sites.
5:45 AM - EE3.05
Understanding How Zn Improves the Durability of Nuclear Waste Glasses through Atomic Scale Simulation
Thorsten R Stechert 1 Michael J D Rushton 1 Robin W Grimes 1
1Imperial College London London United KingdomShow Abstract
Glass has been widely adopted as the first generation host material for the immobilisation of high level nuclear waste. It is intended that immobilised waste will go for permanent disposal in geological repositories and it is desirable that any wasteform should be durable under these conditions for an extended period. Atomic scale computer simulation can be used to provide a mechanistic basis for the structure and properties of glasses and as a result offers opportunities for the compositional optimisation of nuclear waste glasses.
Experimental studies have reported that zinc oxide improves the durability of nuclear glasses. Through the use of molecular dynamics, in conjunction with a simulated melt-quench procedure, atomic structures of sodium silicate glasses, have been generated with and without zinc. The structure of these glasses was studied through the use of pair distribution functions, ring size distributions and cluster analysis. Using the insights gained from these analyses the structural role of zinc oxide within silicate glass is discussed and consideration is given to reports of its differing roles as a network former and a network modifier. The effects of Zn addition on sodium ion distribution and clustering behaviour within the glasses is also reported. This is used to explain changes to intermediate-range structure and hence provide a possible explanation for the experimentally observed increase in durability obtained with the addition of Zn.
EE1: Capture and Immobilization of Radionuclides I
Monday AM, December 01, 2014
Hynes, Level 2, Room 204
9:00 AM - *EE1.01
Current Status of Immobilization Techniques for Radioactive Iodine for Geological Disposal in Japan
Kazuya Idemitsu 1 Tomofumi Sakuragi 2
1Kyushu University Fukuoka Japan2Radioactive Waste Management Funding and Research Center Tokyo JapanShow Abstract
Radioactive iodine-bearing materials, such as spent silver adsorbent, are produced in nuclear reprocessing plants in Japan. According to Japanese disposal plan radioactive wastes that contain a certain quantity of iodine-129 are classified as Transuranic Waste Group 1 (TRU 1) for spent silver adsorbent or as Group 3 for bitumen-solidified waste and they should be disposed of by burial deep underground. Because the half-life of iodine-129 is 15.7 million years, it would be difficult to prevent release of iodine-129 from the wastes into the surrounding environment over such a prolonged time. Moreover, because iodine in its ionic forms is soluble and not readily adsorbed, its migration is not retarded significantly in engineered or natural barriers. Therefore the release of iodine-129 from nuclear wastes needs to be restricted to permit reliable safety assessment; this technique is called “controlled release”. It is desirable that iodine release period will be longer than 100,000 years.
Several techniques for immobilization of iodine have been developed for this purpose. These are narrowed down to three techniques such as synthetic rock, BPI (BiPbO2I) glass and high performance cement. Iodine will be fixed as AgI in grain boundary of corundum or quartz through hot isostatic pressing (HIP) in the synthetic rock, as BPI in boron-lead based glass, or as some cement minerals such as ettringite in alumina cement. These techniques are assessed by three models such as the leaching model, the distribution equilibrium model, and the solubility-equilibrium model. In this paper current status of these techniques are described.
9:30 AM - *EE1.02
Novel Metal Sulfides to Achieve Effective Capture and Durable Consolidation of Radionuclides
Surya S Kota 1 Debajit Sarma 1 Mercouri Kanatzidis 1
1Northwestern University Evanston USAShow Abstract
To support the future expansion of nuclear energy an effective method is needed for the capture and safe storage of radioisotopes released during reprocessing of spent nuclear fuel. The Department of Energy Office of Nuclear Energy (DOE-NE) is currently investigating alternative waste forms for 129I. DOE is interested in new waste forms that can provide higher waste loadings, more efficient consolidation routes, lower costs, etc. PNNL has been developing non-oxide aerogels made with metal sulfides, termed chalcogels, for iodine immobilization and thus far, the materials do show promise as a potential replacement avenue for AgZ. These chalcogels are stable in aqueous solutions. Scientists at the university lead on this proposal who area the inventors of the chalcogel class of materials have already demonstrated selective affinity with chalcogels for metal ions in aqueous media such as Cs+, Sr2+, and Co2+. Aerogels have been studied for confinement of radioactive wastes in recent years and are under investigation as waste forms for 129I. Aerogels can act as precursors to the final glass matrix that actually immobilizes the wastes. Use of silica aerogels for the purpose has been limited by their brittleness in the presence of water that is commonly present in off-gas treatment and also due to their low permeability to nuclear waste.
Recently, we reported a new type of aerogel made with metal chalcogenides (where chalcogen is S, Se, and/or Te) and is referred as chalcogel. Non-oxide materials such as the chalcogels have different properties than oxide materials and, in this case, some of those differences are actually advantages. For example, the high polarizability of the chalcogens (over oxygen) can be used to capture iodine. We will report the exploration of chalcogels as high affinity materials for capturing iodine and the conversion of the loaded materials to glass forms. The efficiency of a chalcogel-based waste form is expected from strong complex formation based on the high chemical affinity of chalcogen atoms for iodine gas. The strong chemical affinity is due to the soft Lewis acid/soft Lewis base complex formation, according to Pearson&’s Hard/Soft Acid-Base (HSAB) principle. We also report that chalcogels can be chemically tailored to exhibit additional strong I2 capture mechanisms.
10:00 AM - EE1.03
Chalcogel Sorbents for Effective Capture and Consolidation of Radioiodine
Suryasubrahmanyam Kota 1 Debajit Sarma 1
1Northwestern University Evanston USAShow Abstract
129I is a major byproduct generated from nuclear fission of uranium fuel. Due to its adverse health effects in humans, safe removal and storage of 129I is of utmost importance across various nuclear energy plants. The sorbents for the absorption of radioiodine has to be stable during the treatment process and also it should be capable of sorbing large amounts of 129I. The most commonly used sorbents are silver-loaded zeolites and Ag-loaded silica aerogel. However, due to the poor mechanical stability of silica aerogels in an aqueous environment there is a need to develop new material with better mechanical stability. The chalcogen-based aerogels called “chalcogels” are highly porous and have showed good affinity towards heavy metal ions. Herein we report the use of chalcogels and silver functionalized analogues as host materials for capture and immobilization of 129I. Iodine capture was studied with different chalcogels (Sb4Sn4S12, Zn2Sn2S6, NiMoS4 and CoMoS4), their silver functionalized analogues, and binary metal sulfides. All the chalcogels showed high uptake reaching up to 200 mass% and the iodine chemically reacted with the sorbents to form metal -iodide complexes. We will also report the consolidation of various iodine loaded chalcogels with different glass-forming additives into a final waste form.
10:15 AM - EE1.04
Efficient Capture and Immobilization of Iodine-129 with Silver-Functionalized Silica Aerogel
Josef Matyas 1
1Pacific Northwest National Laboratory Richland USAShow Abstract
Reprocessing of spent nuclear fuel is being considered in the U.S. In that case, the release of volatile 129I from reprocessing plants and its safe storage would have to be controlled to meet the Environmental Protection Agency emissions regulations (which require capture of 99.4% of 129I) and disposal restrictions. Currently, a silver-loaded zeolite (AgZ) is the baseline material for removing 129I. However, recent studies indicate limitations in the sorption performance and long-term stability of AgZ. Also, AgZ requires addition of low-temperature glass to immobilize trapped radioiodine. To avoid these drawbacks, silver-functionalized silica aerogel is being developed for the efficient capture and immobilization of 129I. This novel sorbent has a high affinity for iodine at the low concentrations expected in the off-gas and a high sorption capacity, and, after loading with iodine, it can be consolidated into a dense and leach-resistant SiO2-based waste form. It was demonstrated to have a sorption capacity for I2 of 480 mg/g, decontamination factors in excess of 10 000, good sorption performance after long-term exposures to dry and humid air, and retention of more than 92% of iodine in the densified product. The presentation will highlight the results from a series of sorption and consolidation studies.
10:30 AM - *EE1.05
French Studies on the Development of Potential Conditioning Matrices for Iodine 129
Lionel Campayo 7 Fabienne Audubert 6 Jean-Eric Lartigue 6 Eglantine Courtois-Manara 5 Sophie Le Gallet 1 Frederic Bernard 1 Thomas Lemesle 3 Francois O. Mear 2 Lionel Montagne 2 Antoine Coulon 7 Danielle Laurencin 4 Agnes Grandjean 7
1Universitamp;#233; de Bourgogne Dijon France2Universitamp;#233; de Lille 1 Lille France3Washington State University Pullman USA4CNRS Montpellier France5Karlsruhe Institute of Technology Karlsruhe Germany6CEA Cadarache Saint Paul Lez Durance France7CEA Marcoule Bagnols sur Ceze FranceShow Abstract
Since 1991, the potential of several specific inorganic host matrices was studied at CEA to ensure a durable immobilization of iodine 129 in the frame of a possible disposal in a deep geological repository.
Due to evidence of retention of xenon 129, decay product of iodine 129, over geological time scales in apatites, these phases were the first materials to be considered. Specifically, a lead-bearing apatite with a good chemical durability was initially developed. Its composition can be written as Pb10(VO4)4.8(PO4)1.2I2. At 90°C, in pure water, its leach rate is of 2.28 10-3 g.m2.j#8209;1 on the basis of iodine release and this rate decreases with time as the progressive replacement of iodide ions by hydroxyl groups along the channels of the crystalline structure occurs. This replacement obeys to a diffusive law and the transformation of iodoapatite grains into hydroxyapatite can be qualified as being pseudomorphic. Current studies are devoted to the shaping of such an iodoapatite in order to get a dense monolith. In so doing, it was found that a reactive sintering by spark plasma sintering at 400°C under a pressure range of 40-70 MPa could offer a clear benefit over sintering techniques in sealed environment (e.g., HIP) of which the use could be seen as more complicated for a reprocessing plant. This allows pellets of more than 92% of the theoretical density to be obtained. However, this process also appears to be very sensitive to scaling effects and it requires a subsequent optimization.
Other apatite compositions were also studied to avoid the use of toxic elements like lead. These apatites were developed on a phospho-calcic basis. They have the noticeable ability of incorporating iodine under its iodate form. Depending on phases constitutive of the geological barrier around the repository site, iodate ions could be less mobile in comparison with iodide which could delays the return of iodine to the biosphere. It was demonstrated that the incorporation mechanism of iodate into such an apatite relies on a substitution of hydroxyls groups. The chemical durability of this apatite is currently evaluated.
Together with ceramics, some glass compositions were also considered. They belong to the AgI-Ag2O-P2O5-Al2O3 system. Close compositions were already proposed by Japanese teams for a similar goal. Here, the idea was to improve their properties by addition of a cross-linking reagent of the phosphate network like alumina. These glasses have an intrinsic compatibility with silver iodide which is the form adopted by iodine in most of the iodine capture processes on solid filters. They can incorporate high iodine amounts and their leaching behavior depends on phosphate chain length, iodine amount and alumina content.
Beyond the development of each matrix, the desired goal would be to have a correct opinion on the strengths and drawbacks of these solutions to face with future industrial and regulatory needs.
11:30 AM - EE1.06
Apatite-Based Ceramic Waste Forms by High Energy Ball Milling and Spark Plasma Sintering for Iodine Confinement
Tiankai Yao 1 Jie Lian 1
1Rensselaer Polytechnic Institute Troy USAShow Abstract
Apatite structure type, with a typical chemical composition of A10(BO4)6C2 (e.g. , A=Ca, Na, Pb, rare earth, fission product, actinides; B=P or V; C=F, Cl, I.) shows tremendous potentials as advanced waste forms for effective nuclear waste management. A wide range of radionuclides can be incorporated into its crystal structure by coupled substitutions at both cation and anion sublattices. Of particular importance, iodine-bearing apatite with chemical composition of Pb10(VO4)6I2 is proposed to confine extremely mobile and highly volatile I-129, a fission product of uranium fission. However, iodine-bearing apatite are typically synthesized and densified at elevated temperatures, resulting in evitable iodine loss. In this work, Pb10(VO4)6I2 powder samples are synthesized by solid state reaction at room temperature by using High energy ball milling (HEBM) followed by thermal annealing at 200 oC to control the crystallinity. Dense iodine-loaded apatite ceramic pellets were consolidated by state-of-art Spark plasma sintering (SPS) at various temperature (350 oC to 700 oC) and very short durations (0 ~20 mins). Iodine retention and the microstructure tenability, especially g