Special Session FJ-10
Materials Technology for Nuclear Waste Treatment and Disposal


Session FJ-10.1 - Waste Form Development, Including Glass, Ceramic, and Metallic Waste Forms

FJ-10.1:IL01  Selective Decontamination and Stable Solidification of Cs and Sr by Zeolites
H. MIMURA, Tohoku University, Sendai, Japan

Large amounts of high-activity-level water (HALW) were generated from the nuclear accident in Fukushima NPP-1 caused by the Great East Japan Earthquake. At present, the cold shutdown is completed by cooling system, while large amounts of HALW (> 260,000m3) are stored. This paper deals with the selective adsorption properties of Cs and Sr and stable ceramic solidification of secondary solid wastes by zeolites. Zeolites have three important abilities; (1) high selectivity for Cs and Sr, (2) Cs gas trapping ability and (3) self-sintering ability. As for Cs and Sr selectivity, mordenite, chabazite, ferrierite and clinoptilolite group have strong adsorbability for Cs, and A and X zeolites for Sr in seawater. The Cs immobilization ratio for the zeolites adsorbing Cs was around 100% after sintering at 1,000℃; the volatilization of Cs was markedly lowered due to the strong gas trapping ability. After high-temperature sintering, the zeolites adsorbing Cs and Sr readily converted to the stable ceramic solid forms such as CsAlSi2O6, CsAlSi5O12 and SrAl2Si2O8. Zeolites are also effective for the stable solidification of insoluble ferrocyanides adsorbing Cs as a solidification matrix; Cs adsorbed insoluble ferrocyanides tend to release all Cs at higher temperature above 1,000℃, while the immobilization ratio of Cs about 100% was maintained by mixing of zeolites. Zeolites are thus effective for the decontamination of Cs and Sr and their stable solidification.

FJ-10.1:IL02  Fabrication and Chemical Durability of Ceramic Technetium-based Pyrochlores and Perovskites as Potential Waste Forms
T. HARTMANN, I.J. ALANIZ, A.J. ALANIZ, University of Nevada - Las Vegas, Las Vegas, NV, USA

Technetium-99 is a key radioisotope from a nuclear waste perspective because of its long half-life (t1/2= 2.13x105 years) and its abundance in used nuclear fuel. As such, it is targeted in separation strategies such as UREX+, for isolation and encapsulation in solid waste forms for final disposal in a nuclear repository. We report here results regarding the incorporation of Tc-99 into ternary oxides of different structure types: pyrochlore (Nd2Tc2O7), perovskite (SrTcO3), and layered perovskite (Sr2TcO4). The objective was to determine synthesis conditions of these ceramic waste forms to immobilize Tc-99 as Tc(IV) and to harvest crystallographic, thermophysical and hydrodynamic data. The fabricated ceramic technetates exhibit good crystallinity and lattice parameters and atomic coordinates could be refined with high accuracies. Thermophysical properties of the oxides, such as the critical temperature (Tc) for superconductivity, were analyzed using AC magnetic susceptibility measurements. In our efforts to compare hydrodynamic properties of the ceramic waste forms with those to Tc-bearing borosilicate glass, we applied the standard test method ASTM C1220-10. Hereby matrix corrosion and Tc-leaching of monolithic glass and ceramic specimens were determined by static leaching experiments.

FJ-10.1:IL03  Recent Advances in Ceramics for Nuclear Waste Immobilization
A. BUKAEMSKIY, S. FINKELDEI, F. BRANDT, S. NEUMEIER, G. MODOLO, D. BOSBACH, Institute of Energy and Climate Research (IEK-6) - Nuclear Waste Management and Reactor Safety, Forschungszentrum Juelich GmbH, Jülich, Germany

Due to their inherent stability ceramic materials are considered as promising matrices for the immobilization of plutonium and the minor actinides. The basic principles of the different ceramic classifications and the main criteria for ceramic characterizations are discussed. Especially physical and chemical durability and long-term stability under repository relevant conditions will be mentioned. The main attention is paid to the single-phase ceramics in which the immobilization of actinides is realized on the atomic scale. The possible mechanism of the incorporation of actinides into these ceramics and the process of the solid solution formation has been studied experimentally. For the fabrication of ceramics with specified properties the detailed analyses of the physico-chemical properties at each step of the fabrication process (powder synthesis, powder treatment and sintering step) have been carried out using modern characterisation techniques.

FJ-10.1:IL04  Behaviour of Fuel and Structural Materials in Severely Damaged Reactors
N. SATO, Institute of Multidisciplinary Research for Advanced Materials, Tohoku University, Sendai, Japan

To study the fuel debris treatment at Fukushima Daiichi NPP, information on the behaviour of fuel and structural materials in severely damaged reactors, i.e., oxides and metals of uranium and zirconium is essential. Since sea water was introduced to the reactors, situation of fuel debris became different from that for TMI case. In this paper, phase relations of uranium and zirconium oxides were analyzed by powder XRD method at high temperatures. By the heat-treatment of the mixture of UO2 and ZrO2 (U:Zr=1:1) under 10 torr air, UO2 was oxidized to U3O8 over 800 oC, The UO2 like phase appeared again at 1350 oC which may be caused by the decomposition of U3O8. The oxidation behavior of the UO2-ZrO2 system was also investigated by using solid solution sample with different U/Zr ratios under different steam and oxygen pressures. The oxidation of the UO2-ZrO2 mixture seemed to be suppressed with decreasing U/Zr ratio. The behavior of fuel materials in the presence of seawater was also discussed as well as that for other structural materials.

FJ-10.1:L05  Cost-effective and Permanently Safe Containerization of Nuclear Solid Waste Using MgAl2O4 Spinel Ceramics
A. ROKHVARGER, E. VAUGHN-FLAM, Rokon Systems, Inc., New York, NY, USA

We developed and patented microwave assisted nano-technologies of a) inexpensive and fully dense sintered large-size and thick-walled MgOxAl2O3 ceramic cylinder vessels and b) microwave induced joining container vessels with their spinel lids. They are permanent chemical, water, thermal, and radiation corrosion resistant as well as impenetrable by radon gas emanating from nuclear materials. Multi-layer onion-like construction of the Rokon dry casks ("DC") consist of a) the inner ultimate corrosion and structural resistant cylindrical ceramic container vessels of 16' length/height and 42" inner diameter and 1" wall thickness with loading capacity 14.7tn or 14.3tn of correspondingly 21 PWR or 44 BWR SNF road assemblies and, b) outer shells of the multi-layer canister package including (i) radiation shielding by aluminum-boron cylinder segments and (ii) mechanical protection by carbon steel shell. We also invented a method and technique for onsite loading, closing, assembling, disposing, and permanently secure and environmentally safe storage of millennium reliable Rokon DC at the nuclear power plant backyard.When financed, Rokon DC production plant will manufacture RC ceramic cylindrical (pipe) vessels and provide pre-assembling the inexpensive Rokon casks that shall have equal capacity and perform all functions of the U.S. DOE standards for Transportation Aging and Disposal ("TAD") Canister System or AREVA (France) dry casks.

FJ-10.1:L06  A New Matrix for Conditioning Chloride Salt Wastes from the Electrorefining of Spent Nuclear Fuel
G. DE ANGELIS, M. CAPONE, C. FEDELI, G.A. MARZO, ENEA, Centro Ricerche Casaccia, Roma, Italy; M. MARIANI, E. MACERATA, M. GIOLA, Politecnico di Milano, Milano, Italy

A novel method proposed by Korea Atomic Energy Research Institute has been applied to the treatment of chloride salt wastes coming from electrorefining of spent fuel, which allows to separate uranium from fission products. It is based on a matrix, SAP (SiO2-Al2O3-P2O5), synthesized by a conventional sol-gel process, able to stabilize the volatile salt wastes owing to the formation of metalaluminosilicates, metalaluminophosphates and metalphosphates. With this method a higher disposal efficiency and a lower waste volume can be obtained. LiCl-KCl melt containing chlorides of alkaline, alkaline-earth metals, and lanthanides has been used to simulate the waste salt. The composite SAP has been prepared by using tetraethyl ortosilicate (TEOS), aluminum chloride (AlCl3.6H2O) and phosphoric acid (H3PO4) as sources of Si, Al, and P, respectively. All reagents were dissolved in EtOH/H2O and the mixture, tightly sealed, was placed in an electric oven at 70°C. After a gelling/ageing for 3 days, the transparent hydrogels were dried at 110°C for 2 days and then thermally treated at 600°C for 2 hours. A series of final products (SAPs) were reacted with metal chlorides at 850°C for 20 hours inside an Ar-atmosphere glove-box, after mixing them at a SAP/metal chloride mixing ratio of 2. The waste forms have been characterized by means of density measurements, stereomicroscopy, scanning electron microscopy, thermogravimetric analysis, as well as by XRD and FTIR spectra.
Financial support from the Nuclear Fission Safety Program of the European Union (project SACSESS, contract FP7-CP-2012-323282) is gratefully acknowledged.

FJ-10.1:L07  Structure Determination of (Al, Nd)-doped Zirconolite Grown from CaO-SiO2-Al2O3-TiO2-ZrO2-Nd2O3-Na2O Glass
CHANGZHONG LIAO, KAIMIN SHIH, Department of Civil Engineering, The University of Hong Kong, Hong Kong

Actinides and Pu are the major contributors to the long term radioactivity of high level nuclear waste after disposal. Zirconolite-based glass-ceramics is a promising candidate for immobilizing actinide radionuclides. Thus, it is essential to study the crystal structure of zirconolite grown from the glass matrix. In this study, Nd was used as the surrogate of actinides, and the structure of zirconolite crystals grown from the glass system was first explored. Powder X-ray diffraction (PXRD), transmission electron microscopy – energy dispersive X-ray spectroscopy (TEM-EDS) and selected area electron diffraction (SAED) were used to characterize the sample. Rietveld refinement method was employed to obtain the structural details of this zirconolite phase, and the chemical compositions of the zirconolite crystal were also determined. The PXRD and SAED results show that the zirconolite crystallized in symmetry of space group C12/c1 with a= 12.562(1) Å, b= 7.2546 (6) Å, c= 11.3616 (3) Å, and beta= 100.67º. The occupancy of Nd atoms in different Wyckoff sites was analyzed by the Rietveld refinement of PXRD data and showed that Nd had a higher preference for occupying the Ca sites.

Session FJ-10.2 - Challenging Waste Constituents, such as Actinides, Noble Metals, and Volatile Species

FJ-10.2:IL03  Adsorption Materials Development for the Separation of Actinides and Specific Fission Products from High Level Waste
YUE-ZHOU WEI, School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai, China

The long-term radiotoxicity of high level waste (HLW) generated in nuclear fuel cycle is governed by the content of the long-lived minor actinides (Np, Am and Cm) and some fission products (FPs) such as 99Tc, 135Cs. From the viewpoints of minimizing the radiological risk and facilitating the management of HLW, separation of the long-lived nuclides is much more desirable. Moreover, the separation and recovery of some potentially valuable FPs such as the PGM (Pd, Ru, Rh), 137Cs and 99Sr will result in a beneficial impact on the waste management. In recent years, to separate the MA and specific FPs from the HLW systematically, we have studied an advanced separation technology using novel adsorption materials. We prepared some silica-based composite adsorbents by immobilizing functional organic or inorganic compounds in porous silica particles with mean diameter of 60 μm and pore size of around 50-600nm. This new type of adsorbents is characterized by fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. Adsorption and separation behavior of various elements was studied experimentally and theoretically. Small scale separation tests using simulated and actual HLW solutions were carried out to verify the feasibility of the proposed technology.

FJ-10.2:IL04  Actinides and Actinide Surrogates Solubilities in Borosilicate Glass

Lanthanides are usually regarded as effective substitutes of actinides in non-radioactive glass considering similarities in their oxidation states, coordination environments and ionic radii. It should be representative of the glass forming ability, melt behaviour and structural features of their radioactive counterparts.
This study compares the solubility and distribution of americium and neodymium in SiO2-B2O3-Na2O-Al2O3-CaO-La2O3 glasses. It assesses the substitution of americium by neodymium from a structural point of view.
The solubility of Nd and Am-bearing melts has been determined for varying La2O3 to Nd2O3 and fixed La2O3 to Am2O3 ratios. Above solubility, apatite crystals formed. Homogeneous glass and apatite containing glass-ceramics were probed by X-ray diffraction, Raman spectroscopy, scanning electron microscopy and chemical analysis. These experiments provide key information on the effects of La2O3:Nd2O3 and La2O3:Am2O3 ratios and concentrations on the connectivity of the amorphous networks and the stoichiometry of the apatite phases.

FJ-10.2:IL05  Silver-functionalized Silica Aerogel as a Mean to Capture and Immobilize Lodine-129
J. MATYAS, Pacific Northwest National Laboratory, Richland, WA, USA

There is a possibility that used nuclear fuel will be reprocessed in the U.S. If that occurs, the release of volatile 129I from reprocessing plants and its safe storage have to be controlled to meet the EPA regulations of limited emissions and disposal restrictions. Currently, a silver-loaded zeolite (AgZ) is the reference 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. Silver-functionalized silica aerogel is being developed at PNNL 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 collapsed into a dense and leach-resistant SiO2-based waste form. It was demonstrated to have a maximum sorption capacity for I2 of 48 mass%, decontamination factors in excess of 10 000, good sorption performance after long-term exposure to dry air, and retention of more than 92% of iodine in densified product. Presentation will provide an overview of the sorption studies and methods to consolidate iodine-loaded aerogel into a dense waste form.
Session FJ-10.3 - Waste Form Modeling, Performance Testing, and Advanced Characterization Techniques

FJ-10.3:IL01  Development and Characterization of Radiation Resistant Waste Glass for Immobilization of Radionuclides
S. PEUGET, J.M. DELAYE, E.A. MAUGERI, C. MENDOZA, A.H MIR, R. CARABALLO, M. TRIBET, O BOUTY, C. JÉGOU, CEA, DEN, Laboratoire d'Étude des Matériaux et Procédés Actif, Bagnols-sur-Ceze, France

This paper presents an overview of the main results of the French research on the long-term behavior of SON68 nuclear glass towards alpha decay accumulation. The effect of the radiation damage induced by alpha decay were investigated by examining glass specimens, doped with a short-lived actinide 244Cm, irradiated by light and heavy ions and irradiated by thermal neutron flux. Atomistic simulations by molecular dynamics have provided further information on the atomic-scale effects of the macroscopic phenomena observed.
These studies have shown that some macroscopic properties vary appreciably with the accumulation of alpha decay, but then stabilize after integrating doses of the order of 4 × 1018 g-1. On the contrary the initial alteration rate of the glass is not significantly affected by the glass damage induced by alpha decays or heavy ions irradiations.
Modification of the local and medium range orders of the glass structure were also noticed with irradiation. The structural evolution induced by alpha decays would be driven by the reconstruction of the glass disorganized by displacement cascades of the recoil nuclei, freezing a glass structure with a higher fictive temperature. A "ballistic disordering fast quenching" model is proposed to explain the glass evolution.

FJ-10.3:IL02  Experimental, Simulation, and Natural Analogue Studies of Nuclear Wasteform Materials
G.R. LUMPKIN, E.Y. KUO, S.C. MIDDLEBURGH, M.J. QIN, G.J. THOROGOOD, Y. ZHANG, Z. ZHANG, D.J. GREGG, ANSTO, Kirrawee DC, NSW, Australia; M. ROBINSON, N.A. MARKS, Nanochemistry Research Institute, Curtin University, Perth, WA, Australia

Here we provide a brief summary of laboratory experiments, atomistic simulations, and natural analogue studies relevant to the development and testing of crystalline ceramic nuclear wasteforms for actinides and certain fission products. The focus of this presentation is on radiation damage effects and ultimately, how we approach the scientific challenge of wasteform performance for materials that must be stored in geological repositories for long periods of time. The available results generally set out the groups of potential actinide host phases in terms of those with intrinsic radiation tolerance due to recovery of damage on picosecond time scales (e.g., fluorite), those with favorable kinetics for longer term damage recovery (e.g., monazite), and many others that are transformed to the amorphous state. We have also conducted atomistic modelling studies of some of these materials with a view toward the understanding of fundamental defect properties, e.g., the energetics of defect formation and migration and how they influence damage recovery.

FJ-10.3:IL03  The Belgian Approach Towards the Investigation of the Compatibility with Geological Disposal of Eurobitum Bituminized Intermediate Level Radioactive Waste
E. VALCKE1, N. BLEYEN1, S. SMETS1, M. VASILE1, X. SILLEN2, 1W&D Expert Group, SCK.CEN, Mol, Belgium; 2ONDRAF/NIRAS, Brussels, Belgium

Eurobitum bituminized waste, containing 20 to 30 wt% of NaNO3, is an important long-lived intermediate-level radioactive waste (ILW). The current reference solution envisages its direct disposal in a geologically stable deep clay formation.
The demonstration of the safety of geological disposal of such waste is a step-wise and iterative process, requiring a systematic and multi-disciplinary investigation of all relevant processes to occur after emplacement of the waste in the disposal site. In the last decades, an enormous progress was made in the understanding of the properties and the long-term behaviour of waste and other repository compounds (EBS, EDZ, host formation), facilitating the recent decisive process regarding the structuring of the research and the selection of key processes to be studied in detail.
For Eurobitum, the key processes are:
- the possible mechanical perturbation by generation of an osmosis-induced pressure and of a gas pressure;
- the possible chemical perturbation by the release of NaNO3 and molecules with radionuclide complexing properties;
- the continuous evolution of the rheological properties of bitumen due to radio-oxidation.
The presentation will address these issues, with an overview of the most recent results and insights.

FJ-10.3:IL04  New Insight into Nuclear and Natural Glasses Dissolution Mechanisms Controlling Long-term Rate
S. GIN, Marcoule DTCD/SECM, Bagnols sur Ceze, France

Nuclear glass lifetime under geological repository conditions could be derived from long-term rate (also called residual rate) measurements and calculations; however there is a need of a better understanding of the underlying mechanisms in order to make reliable estimations. This presentation highlights the current state of knowledge and the recent progress made in this field thanks to original isotopically tagged experiments and the use of probes with sub-nanometer spatial resolution applied to the characterization of glass alteration layers formed under residual rate conditions. We also show the influence of various parameters like glass composition, temperature, pH, solution composition, etc. The results are discussed in the frame of the existing theories for glass and mineral dissolution. Finally it is shown how natural and archeological analogues can help to bridge the gap between short-term laboratory studies and long-term data arising from natural systems.

Session FJ-10.4 - Design and Operation of Waste Immobilization facilities, Repository Design, Requirements, and Licensing

FJ-10.4:IL01  Progress at ANSTO on a Synroc Plant for Intermediate-level Waste from Reactor Production of 99Mo
E.R. VANCE, S. MORICCA, M.W.A. STEWART, ANSTO, Kirrawee DC, NSW, Australia

Intermediate level waste from ANSTO's expanded 99Mo production plant will consist of ~5000L/year of 6M NaOH + 1.4 NaAlO2 + fission products. Detailed engineering is being carried out on a synroc plant to immobilise this waste in a silicophosphate glass-ceramic. The liquid waste will be mixed with precursors and dried before being calcined in a reducing atmosphere to control fission product volatility. The calcine will be transferred to 30L metal cans which will be sealed and hot isostatically pressed at 1000°C/30MPa for 2h, then cooled to room temperature and stored preparatory to final disposal. The waste form material will pass 90°C PCT tests. In addition legacy intermediate level acidic uranyl nitrate-based liquid waste from 99Mo production at ANSTO between the 1980s and 2005 via irradiation of UO2 targets will also be immobilised by the same process, although the precursor additives are different. Some examples illustrating the wide applicability of hot isostatic pressing to consolidate nuclear waste forms will be given showing the advantages for particular wastes, notably high waste loadings and the absence of off-gas in the high temperature consolidation step. The immobilisation of a variety of low-level liquid and solid wastes from 99Mo production will also be discussed.

FJ-10.4:IL02  Vitrification of UK Higher Activity Wastes and the Dissolution Mechanisms of Simulant Waste Glasses in Hyperalklaline Conditions Relevant to a Cementitious Geological Disposal Facility
N. HYATT, Department of Materials Science & Engineering, The University of Sheffield, Sheffield, UK

The current baseline for conditioning of UK intermediate-level radioactive waste (ILW) is encapsulation using an ordinary Portland cement composite. However, vitrification of some UK ILW waste streams is a promising alternative, achieving passivation of reactive wastes and enhanced reduction in packaged volume, compared to the baseline encapsulation technology. This presentation will summarise the current state of the art in developing vitrified wasteforms for UK intermediate level wastes, including product performance characteristics.  Successful case studies concerning the design, development and demonstration of vitrified ILW wasteforms for plutonium contaminated materials, Magnox sludge and reactor wastes will be discussed.  In particular, the opportunity to achieve high volume reduction and enhanced waste passivity, relative to raw wastes and encapsulated products, will be discussed in the context of the overall strategy and cost of geological disposal strategy.  One option for the disposal of the resulting vitrified ILW would be emplacement in a geological disposal facility in a high-pH environment with cemented ILW and a cement-based backfill. Progress in understanding the dissolution behaviour of simulant ILW glasses in high pH (∼12.5) calcium-rich solutions will be presented.  The influence of glass composition on glass dissolution mechanism will be highlighted, in particular, the role of boron which is thought to delay initial hydration of the borosilicate glasses, through the formation of calcium borates, in a similar manner to the retardation of cement hydration by soluble borates. More generally, the formation of calcium- and magnesium-containing precipitates on the surface of the vitrified wastes, and agglomeration of the powder, appeared to act to reduce the dissolution rate. Overall these results suggest that calcium has an important role in the long-term durability of vitrified wastes at high pH.  The results of recent investigations of UK HLW glass dissolution in high pH (∼12.5) calcium-rich solutions, relevant to a co-located HLW / ILW GDF, will also be discussed.  In this case the formation of a magnesium-containing smectite clay (likely saponite) in addition to CSH(II), a jennite-like CSH phase, were identifiedas important in controlling in both the long and shorter-term durability of these materials.

FJ-10.4:IL03  Review of the Development of the Proposed Yucca Mountain Geologic Repository
C.E. SANDERS, University of Nevada, Las Vegas, Las Vegas, NV, USA

It can be said that the nuclear community neglected the issue of final storage of nuclear waste in the first era of nuclear power production, with many nations not looking at this topic until some years into its program. This is a matter that must not be neglected now during the "renaissance" of nuclear if nuclear energy is to have a part on the stage of the play in world energy supplies. In 1982, the United States (U.S.) Congress passed the Nuclear Waste Policy Act (NWPA) which outlines the screening process for selecting a national site for used nuclear fuel and high level radioactive waste storage. This paper describes the U.S. nuclear waste policy dilemma and its impact on the selection and development of the nation's first long-term geologic repository for over 70,000 metric tons of used nuclear fuel and high-level radioactive waste. In 1987, the U.S. Congress designated Yucca Mountain, which can be described as the "most studied real estate on the planet", as the repository site to be characterized. However, due to political pressures, the fate and realization of the Yucca Mountain repository is uncertain.

FJ-10.4:IL04  RADON Operational Experience in High-temperature Treatment of Radioactive Wastes

Federal State Unitary Enterprise (FSUE) RADON is a leading company in the Russian Federation in fields of management of low- and intermediate-level radioactive wastes (LILW) and spent radiation sources, radioecological monitoring of the city of Moscow and Moscow region, and decontamination and restoration of radioactively-contaminated areas. One of the most important fields of RADON activity is treatment of LILW to reduce their volume before long-term storage in repository and disposal into geological sites. Among the volume reduction methods high-temperature technologies seem to be the most perspective and cost effective due to highest volume reduction factors and product quality. Application of high temperatures (>1000 °C) provides for production of glassy and glass-crystalline waste forms with high chemical and radiation resistance, and strong mechanical integrity suitable for safe long-term storage and disposal in both interim repositories and underground sites.
The full-scale LILW vitrification facility energized from 1.76 MHz/160 kW generators and equipped with three 418 mm inner diameter cold crucibles produces alkali borosilicate glass with ~30-35 wt.% waste loading. Treatment of LILW is two-phase process. Liquid waste is concentrated to salt content of ~1000 g/L. The concentrate is intermixed with glass formers preparing a paste with a water content of 22-25 wt.% which is fed into the cold crucible. The facility is equipped with automated control system of the CCIM process. Molten gla ss is periodically poured into 20 L canisters. The glass blocks obtained are annealed in the tunnel furnace to remove thermal stresses in the blocks. The mix of gases, steam and aerosol formed during melting is cleaned step by step in the off-gas system.
The plasma unit is meant for joint treatment of solid organic waste and spent heat-insulators, concrete, glass breakage, construction garbage, and other high-fusible materials. The shaft is heated by arc plasmatrones. Waste in the shaft is step-by-step subjected to drying, gasification, burning, slag formation and melting. Molten slag is accumulated and homogenized in the bottom part of the shaft, and poured through a stopper unit into containers. Off-gas is purified in an off-gas system, which involves high-temperature post-combustor, chemical and catalytic toxic gases neutralization units, and two-step radioactive aerosols trapping unit. Maximum waste capacity of the unit is up to 250 kg/hr.
The bench-scale unit for cold crucible melting of ceramic and glass-ceramic waste forms is energized from a 1.76 MHz/60 kW generator and equipped with a 130 mm inner cold crucible. Numerous melted ceramics (zirconolite-, pyrochlore-, murataite-, pyrochlore-, garnet-based and complex Synroc-type) and glass-ceramics (perrierite/chevkinite-, sphene-, britholite-based and multicomponent) were produced and characterized in details using XRD, SEM/EDS, vibrational and X-ray spectroscopic techniques.
One of the promising methods for LILW treatment is application of thermochemical reactions (SHS) with high energy release. A powdered metal fuel (PMF) composed of metallic powder, stabilizer, and surfactant is intermixed with solid LILW and matrix constituents and set it on fire following by propagation of the burning front. This method was applied to treatment of organic wastes such as spent ion-exchange resins, polymers, biological wastes, irradiated graphite, and some contaminated soils. The materials pro duced have been proven to satisfy to environmental standards for conditioned LILW.

Poster Presentations

FJ-10:P02  Leaching Behaviour of Salt Wastes Conditioned with Sodalite Blended with two Different Glass Powders
M. CAPONE, G. DE ANGELIS, C. FEDELI, ENEA, Centro Ricerche Casaccia, Roma, Italy: F. GIACOBBO, M. DA ROS, E. MACERATA, M. MARIANI, Politecnico di Milano, Milano, Italy

Two different glass powders (a commercially available glass frit and a borosilicate glass) have been used as blending agents for sodalite, an aluminosilicate mineral able to condition chloride salt wastes from pyrometallurgical processes. The synthesis of the mineral phase has been made through a process recently proposed by Idaho National Laboratory in USA, starting from a homogeneous powder of nepheline, chloride salts and glass. The mix, put into an alumina crucible, was introduced in a furnace inside an argon-atmosphere glove-box. The furnace temperature was then raised at 10°C/min to 500°C where it was held for about one hour, in order to allow any residual moisture to evaporate. The temperature was then raised to 925°C for 7 hours.
Leach tests under static conditions, according to ASTM C1285-02 (reapproved 2008), have then been carried out on the final waste forms for six contact times (1, 7, 15, 30, 90, 150 days) at room temperature (23°C) and in an oven at 90°C. SEM investigations have also been made before and after leach tests, in order to check the status of the powders. In particular the effect of the leaching process on the surface of the sodalite grains at 90 °C from 30 to 150 days has been evidenced. The results obtained in the present study have been usefully compared to those from a similar test on a sodalite added with a glass frit by Idaho National Laboratory.
Financial support from the Nuclear Fission Safety Program of the European Union is gratefully acknowledged (project ACSEPT, contract FP7-CP-2007-211 267).


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