Stable Waste Form Research Articles

Copper tellurite glass for radioactive iodine immobilization

Radioactive iodine-129 can easily diffuse into the environment and possesses a half-life of 1.57 × 107 years. Therefore, durable and stable waste forms are required for the treatment I-129. In this study, CuI·Cu2O·TeO2 glasses were fabricated by incorporating CuI into a copper tellurite glass (0.3Cu2O·0.7TeO2) matrix at different concentrations ranging from 0 to 30 mol%. The X-ray diffraction analysis confirmed the amorphous phase of all prepared samples, and X-ray fluorescence analysis verified iodine volatilization during the melting process. With the addition of 30 mol% CuI, the maximum iodine loading was approximately 13.79 wt%. X-ray photoelectron spectroscopy and Raman spectroscopy were used to understand the structure of the CuI·Cu2O·TeO2 glass waste forms. As the CuI content increased in the glass matrix, there was an increase of +2-valent Cu and OBO. It was further confirmed that there were no significant changes in the structure of Te, which serves as the glass network former. Furthermore, all samples satisfied the US regulation of 2 g/m2 for the normalized release of all elements, as confirmed through the product consistency test. The proposed glass waste forms have cost advantages over previously developed silver-based glasses, making them a promising avenue for further research.

Effect of solvents on the properties of Ag0-containing aluminosilicate sorbents for iodine capture

Radioiodine, which is generated during the reprocessing of nuclear fuel, must be immobilized on sorbents to convert it into a stable waste form. In this study, to capture iodine, Ag0-containing aluminosilicate sorbents (Ag0ASSs) were synthesized via sol–gel method using Na, Al, and Si alkoxides; co-solvent exchange; Ag/Na ion exchange; solvent exchange with solvents having different surface tensions; and ambient pressure drying. Three different Na+AlSi-OH sols were synthesized using sodium methoxide (NaOMe): aluminum tri-sec-butoxide [Al(OsBu)3]: tetraethyl orthosilicate (TEOS) with molar ratios of 1.05:1:1, 1.3:1.1:1, and 1.5:1.3:1. The solvent effect on textural properties such as surface area and pore size distribution, Ag0 particle size, and total iodine capture capacity of Ag0ASSs was investigated using two types of solvents: 1) a solvent with high surface tension (water) and 2) solvents with low surface tension (isopropanol and n-heptane). The textural properties and particle size distribution of Ag0 in the Ag0ASSs were dependent on the type of solvents used for exchange. The average pore diameter and cumulative pore volume of the Ag0ASSs, except the Brunauer–Emmett–Teller (BET) surface area, significantly increased with decreasing surface tension of solvent and increasing Al/Si atomic ratios. Although the increase in pore diameter and volume resulted in larger particle sizes of Ag0 in Ag0ASSs, the total and chemisorbed iodine-capture capacities of the Ag0ASSs increased. Therefore, the total and chemisorbed iodine-capture capacities depends on the pore size distributions of the Ag0ASSs. These Ag0ASSs have the potential to immobilize radioiodine, which could be further processed into stable compounds.

Developments and experiences of the CHANCE, MICADO and PREDIS projects in radioactive waste characterization

Characterization is a very important step in dealing with materials and waste streams generated during the operational and decommissioning phases of nuclear installations, including nuclear power plants. Characterization allows differentiation between materials that can be released from regulatory control and those that require further treatment and conditioning to become a stable waste form suitable for future storage and final disposal according to its waste classification. Characterization is also needed in the pre-disposal stages of radioactive waste management to demonstrate compliance with the waste acceptance criteria of the facilities that will accept the different waste forms. This work will present the strategies developed and implemented by the three projects for in-depth and accurate waste characterization and investigation of the different radioactive waste packages considered. CHANCE, MICADO, and PREDIS will present their goals, the methods developed, the technologies used and the (preliminary) results contributing to the improvement of the safety and the data and information quality of the waste packages analyzed at the different stages of the waste management process. Special emphasis will also be given to complementary approaches highlighting the usability of the technologies, the accessibility of the data, and the problem-solving of the three projects within the European panorama.

Superior Iodine Uptake Capacity Enabled by an Open Metal-Sulfide Framework Composed of Three Types of Active Sites

Superior Iodine Uptake Capacity Enabled by an Open Metal-Sulfide Framework Composed of Three Types of Active Sites

Effects of additives on the thermal stability of silver tellurite glass system

129I is notable for its long half-life of 1.57 × 107 y and high mobility in the environment. It should be immobilized in a geological disposal environment through a stable waste form. In this study, additives including Al, Bi, Pb, V, Mo, and W were added to silver tellurite glass, and the thermal properties of this matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method, including the melting of the glass precursor mixtures at 800 °C for 1 h. The loading of iodine in the matrix was approximately 11–15% by weight, excluding the oxygen element. Differential scanning calorimetry was performed to evaluate the thermal properties of the glass samples. The glass transition temperature (Tg), crystallization temperature (Tc), and glass stability were investigated. The addition of MoO3 or WO3 in the glass system increases Tg and Tc while maintaining the glass stability.

Vitrification of Nuclear Contaminated Lead-Boron Polyethylene

Lead-Boron Polyethylene (PbBPE) composite formed by diffusing lead and boron carbide powders in the polyethylene is a typical shielding material for radiation protection in nuclear facilities. Cement solidification of contaminated PbBPE could result in the waste volume increase and the radionuclides leaching risk. This study aimed to apply the thermal treatment to reduce the volume of PbBPE and immobilize radionuclides into the stable waste form by turning PbBPE into glass. PE could be pyrolyzed at ~450 oC, while the residual lead and boron were converted into an undurable lead borate glass at temperature above 900 oC. The volume of the waste decreased more than twice after thermal treatment, and the durability of the lead borate glass was enhanced by addition of SiO2. The 7-day Product Consistency Test normalized releases for elements (Pb, B, Co, Sr and Cs) were <1 g/m2, when 10 wt% SiO2 was added. Glass structure analysis indicated that the enhanced durability resulted from the increased silica tetrahedron and bridging oxygens in the glass network.

Formation of multicomponent alloy particles in doped ceria under I2+ ion irradiation and thermal annealing

Multicomponent alloys resembling the ε-phase particles in nuclear fuel have the potential to immobilize 99Tc in a stable waste form. Formation of such multicomponent alloys under extreme conditions is studied in a non-activated model system: CeO2 doped with 2 wt.% Mo, 1.5 wt.% Ru, 0.75 wt.% Pd, 0.5 wt.% Re and 0.25 wt.% Rh. The doped CeO2 films were irradiated with I2+ ions (610°C, 1016 and 5×1016 I2+/cm2). For selected samples post-irradiation heat treatment was conducted (900°C, 1100°C). Analytical transmission electron microscopy revealed the formation of 5-20 nm precipitates containing Pd, Mo, and Re sometimes associated with cavities. Iodine was found to segregate to grain boundaries and cracks. After heat treatment at 1100°C, the CeO2 matrix had recrystallized. Precipitates containing Pd, Mo, Re, and Ru were observed near the interface with the polycrystalline yttria-stabilized zirconia substrate. It follows that Pd is the most mobile element, followed by Mo and Re. While irradiation promotes precipitation, high-temperature effects are more significant.

Selective immobilization of iodide onto a novel bismuth-impregnated layered mixed metal oxide: Batch and EXAFS studies

A novel Mg/Al/Bi ternary mixed metal oxide (MMO) was successfully prepared by a simple coprecipitation and calcination process. The Mg/Al/Bi MMO (molar ratio = 3:0.4:0.6) presented synergistically enhanced adsorption efficiency for iodide in water with porous structure. Kinetic adsorption results indicated good agreement with pseudo first-order-kinetic model (R = 0.998) and the maximum adsorption capacity calculated by Langmuir isotherm of Mg/Al/Bi MMO was 202.2 mg/g, which is higher than that of Mg/Al layered double oxide (LDO) (161.0 mg/g) and bismuth oxide (57.5 mg/g). In addition, the Mg/Al/Bi MMO showed good adsorption efficiency under neutral condition (pH 7), by which the material exhibits relatively strong resistance to the presence of co-existing anions in water due to the ion selectivity. Extended X-ray absorption fine structure (EXAFS) spectra revealed that the enhanced adsorption of iodide by Mg/Al/Bi MMO can be induced by selective Bi-O-IBformation, which is stable waste form for its treatment. Consequently, it is clear that the novel Mg/Al/Bi MMO is a potentially attractive material in terms of selective uptake of iodide with forming stable chemical complex in radioactive contaminants.

Study of the effect of Kaolin in the mortar of cement matrices by confinement of ion exchange resins

Radioactive waste arising as a result of nuclear activities should be safely managed from its generation to final disposal in an appropriate conditioned form to reduce the risk of radiation exposure of technical personnel and of the public and to limit contamination of the environment. The immobilization of low and intermediate level radioactive wastes in cementitious matrices is the most commonly used technique to produce inexpensive waste matrix that complies with regulatory requirements in order to protect humans and the environment against nuisance caused by ionizing radiation. Cement based materials are used in radioactive waste management to produce stable waste forms. This matrix constitutes the first build engineering barrier in disposal facilities. In this work, the kaolin is used to enhance the mechanical performance of the matrix of confinement of ion exchange resins by gradually replacing the sand in mortar with kaolin clay. The Kaolin clay sample was a special pure product, sourced from a foreign country. The maximum quantity of resins that can be incorporated into the mortar formulation without the packages losing their strength is 13.915% which results in a better mechanical strength at 6.7686 MPA compression with kaolin.

Dilute condition corrosion behavior of glass-ceramic waste form

Borosilicate glass-ceramics are being developed to immobilize high-level waste generated by aqueous reprocessing into a stable waste form. The corrosion behavior of this multiphase waste form is expected to be complicated by multiple phases and crystal-glass interfaces. A modified single-pass flow-through test was performed on polished monolithic coupons at a neutral pH (25 °C) and 90 °C for 33 d. The measured glass corrosion rates by micro analysis in the samples ranged from 0.019 to 0.29 g m−2 d−1 at a flow rate per surface area = 1.73 × 10−6 m s−1. The crystal phases (oxyapatite and Ca-rich powellite) corroded below quantifiable rates, by micro analysis. While, Ba-rich powellite corroded considerably in O10 sample. The corrosion rates of C1 and its replicate C20 were elevated an order of magnitude by mechanical stresses at crystal-glass interface caused by thermal expansion mismatch during cooling and unique morphology (oxyapatite clustering).

Cold crucible induction melter studies for making glass ceramic waste forms: A feasibility assessment

Glass ceramics are being developed to immobilize fission products, separated from used nuclear fuel by aqueous reprocessing, into a stable waste form suitable for disposal in a geological repository. This work documents the glass ceramic formulation at bench scale and for a scaled melter test performed in a pilot-scale (∼1/4 scale) cold crucible induction melter (CCIM). Melt viscosity, electrical conductivity, and crystallization behavior upon cooling were measured on a small set of compositions to select a formulation for melter testing. Property measurements also identified a temperature range for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form. Bench scale and melter run results successfully demonstrate the processability of the glass ceramic using the CCIM melter technology.

99Tc and Re Incorporated into Metal Oxide Polyoxometalates: Oxidation State Stability Elucidated by Electrochemistry and Theory

The radioactive element technetium-99 ((99)Tc, half-life = 2.1 × 10(5) years, β(-) of 253 keV), is a major byproduct of (235)U fission in the nuclear fuel cycle. (99)Tc is also found in radioactive waste tanks and in the environment at National Lab sites and fuel reprocessing centers. Separation and storage of the long-lived (99)Tc in an appropriate and stable waste-form is an important issue that needs to be addressed. Considering metal oxide solid-state materials as potential storage matrixes for Tc, we are examining the redox speciation of Tc on the molecular level using polyoxometalates (POMs) as models. In this study we investigate the electrochemistry of Tc complexes of the monovacant Wells-Dawson isomers, α(1)-P(2)W(17)O(61)(10-) (α1) and α(2)-P(2)W(17)O(61)(10-) (α2) to identify features of metal oxide materials that can stabilize the immobile Tc(IV) oxidation state accessed from the synthesized Tc(V)O species and to interrogate other possible oxidation states available to Tc within these materials. The experimental results are consistent with density functional theory (DFT) calculations. Electrochemistry of K(7-n)H(n)[Tc(V)O(α(1)-P(2)W(17)O(61))] (Tc(V)O-α1), K(7-n)H(n)[Tc(V)O(α(2)-P(2)W(17)O(61))] (Tc(V)O-α2) and their rhenium analogues as a function of pH show that the Tc-containing derivatives are always more readily reduced than their Re analogues. Both Tc and Re are reduced more readily in the lacunary α1 site as compared to the α2 site. The DFT calculations elucidate that the highest oxidation state attainable for Re is VII while, under the same electrochemistry conditions, the highest oxidation state for Tc is VI. The M(V)→ M(IV) reduction processes for Tc(V)O-α1 are not pH dependent or only slightly pH dependent suggesting that protonation does not accompany reduction of this species unlike the M(V)O-α2 (M = (99)Tc, Re) and Re(V)O-α1 where M(V/IV) reduction process must occur hand in hand with protonation of the terminal M═O to make the π*(M═O) orbitals accessible to the addition of electrons. This result is consistent with previous extended X-ray absorption fine structure (EXAFS) and X-ray absorption near edge structure (XANES) data that reveal that the Tc(V) is "pulled" into the -α1 framework and that may facilitate the reduction of Tc(V)O-α1 and stabilize lower Tc oxidation states. This study highlights the inequivalency of the two sites, and their impact on the chemical properties of the Tc substituted in these positions.

Synthesis, Structure Elucidation, and Redox Properties of99Tc Complexes of Lacunary Wells−Dawson Polyoxometalates: Insights into Molecular99Tc−Metal Oxide Interactions

The isotope (99)Tc (β(max), 293.7; half-life, 2.1 × 10(5) years) is an abundant product of uranium-235 fission in nuclear reactors and is present throughout the radioactive waste stored in underground tanks at the Hanford and Savannah River sites. Understanding and controlling the extensive redox chemistry of (99)Tc is important in identifying tunable strategies to separate (99)Tc from spent fuel and from waste tanks and, once separated, to identify and develop an appropriately stable waste form for (99)Tc. Polyoxometalates (POMs), nanometer-sized models for metal oxide solid-state materials, are used in this study to provide a molecular level understanding of the speciation and redox chemistry of incorporated (99)Tc. In this study, (99)Tc complexes of the (α(2)-P(2)W(17)O(61))(10-) and (α(1)-P(2)W(17)O(61))(10-) isomers were prepared. Ethylene glycol was used as a "transfer ligand" to minimize the formation of TcO(2)·xH(2)O. The solution structures, formulations, and purity of Tc(V)O(α(1)/α(2)-P(2)W(17)O(61))(7-) were determined by multinuclear NMR. X-ray absorption spectroscopy of the complexes is in agreement with the formulation and structures determined from (31)P and (183)W NMR. Preliminary electrochemistry results are consistent with the EXAFS results, showing a facile reduction of the Tc(V)O(α(1)-P(2)W(17)O(61))(7-) species compared to the Tc(V)O(α(2)-P(2)W(17)O(61))(7-) analog. The α(1) defect is unique in that a basic oxygen atom is positioned toward the α(1) site, and the Tc(V)O center appears to form a dative metal-metal bond with a framework W site. These attributes may lead to the assistance of protonation events that facilitate reduction. Electrochemistry comparison shows that the Re(V) analogs are about 200 mV more difficult to reduce in accordance with periodic trends.

Physical properties of an alumino-silicate waste form for cesium and strontium

Nuclear fuel reprocessing will be required to sustain nuclear power as a baseload energy supplier for the world. New reprocessing schemes offer an opportunity to develop a better strategy for recycling elements in the fuel and preparing stable waste forms. Advanced strategies could create a waste stream of cesium, strontium, rubidium, and barium. Some physical properties of a waste form containing these elements sintered into bentonite clay were evaluated. We prepared samples loaded to 27% by mass to a density of approximately 3 g/cm 3. Sintering temperatures of up to 1000 °C did not result in volatility of cesium. Instead, the crystallinity noticeably increased in the waste form as temperatures increased from 600 to 1000 °C. Assemblages of silicates were formed. Significant water evolved at approximately 600 °C but no other gases were generated at higher temperatures.

Composition Changes and Future Challenges for the Sellafield Waste Vitrification Plant

AbstractThe Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquors produced during reprocessing of spent nuclear fuel by bonding the fission products as metal oxides into a borosilicate glass matrix. This provides a stable and durable waste form suitable for safe long term storage and ultimate disposal.WVP was commissioned with feed from reprocessing of Magnox reactor fuel. This material is relatively low in fission product content per tonne of fuel, but contains significant Al and Mg from fuel cladding. WVP also routinely treats a blended feed made from a mixture of Magnox and Oxide reprocessing products. The Oxide fuel from Light Water Reactor (LWR) and Advanced Gas Cooled (AGR) power stations is of higher burnup and contains more fission products per tonne of fuel, also Gd and other process additives. Blending allows 25% incorporation of waste oxides by weight in glass to be achieved routinely.Recent programmes of development work in WVP have been aimed at increasing incorporation rates for these feeds, to reduce the number of waste containers produced for disposal. Work has also focussed on increasing the throughput of WVP, to more rapidly treat current stocks of liquid reprocessing waste, both by increasing the feed rate and by improving the lifetime of key components to improve plant availability.Future challenges for WVP include flowsheet changes to treat historic stocks of reprocessing wastes containing high U, Fe and Cr. Washout of solids from the base of waste storage tanks in preparation for decommissioning is also likely to give high Mo feeds. Development of flowsheet and glass formulation to accept these changes in feed composition will be a key objective of future work.

Durability of Borosilicate Glass Compositions for the Immobilisation of the UK's Separated Plutonium Stocks

AbstractSeveral glass compositions are currently under investigation for immobilisation of the separated PuO2 that has been produced as a result of civil nuclear fuel reprocessing in the UK. Whilst a final decision on the fate of what ultimately will be over 100 tonnes of plutonium has yet to be made, all options for the disposition of this material are currently being investigated by Nexia Solutions Ltd on behalf of the Nuclear Decommissioning Authority (NDA).As one of the immobilisation options, vitrification in borosilicate glass could potentially provide a criticality-safe and stable waste form with durability suitable for long term storage and subsequent repository disposal. From an initial experimental survey of potential candidates, three borosilicate compositions were selected for a more detailed study of the waste loading and chemical durability: lanthanide borosilicate (LaBS), alkali tin silicate (ATS) and high-lanthanide alkali borosilicate (modified-MW). In these inactive tests, hafnium was used as the surrogate for plutonium. This paper describes a range of static leach tests that were undertaken in order to understand the overall durability of the waste forms, as well as the release rates of the Pusurrogate when compared to any neutrons poisons present in the glass. For the LaBS compositions it was found that the release rate of gadolinium was potentially slightly higher than that of hafnium, although both were as low as 10-5 to 10-6 g m2 day-1. The potential implications for long-term repository behaviour are discussed.

The Behavior of ThO2-Based Fuel Rods during Normal Operation and Transient Events in LWRs

The thermal, mechanical, and chemical behavior of both thorium and uranium dioxide (ThO2-UO2) and thorium and plutonium dioxide (ThO2-PuO2)-based fuels during in-service and hypothetical accident conditions in light water reactors (LWRs) is described. These fuels offer the possibility for increased proliferation resistance and a reduction in the stockpile of weapons-grade and reactor-grade PuO2 as well as being a more stable waste form. The behavior is described for three different designs of ThO2-based fuels: a homogeneous mixture of ThO2-UO2, a microheterogeneous arrangement of the ThO2 and UO2, and a homogeneous mixture of ThO2-PuO2. The behavior was calculated with widely known LWR analysis tools extended for ThO2-based fuels: (a) MATPRO for calculating material properties, (b) FRAPCON-3 for calculating in-service fuel temperature and fission-gas release, (c) VIPRE-01 for calculating the possibility for departure from nucleate boiling, (d) HEATING7 for calculating in-service two-dimensional temperature distributions in microheterogeneous fuel, (e) SCDAP/RELAP5-3D for calculating the transient reactor system behavior and fuel behavior during loss-of-coolant accidents, and (f) FRAP-T6 for calculating the vulnerability of the cladding to cracking due to swelling of the fuel during hypothetical reactivity-initiated accidents.The analytical tools accounted for the following differences in ThO2-based fuels relative to 100% UO2 fuel: (a) higher thermal conductivity, lower density and volumetric heat capacity, less thermal expansion, and higher melting point; (b) higher fission-gas production for 233U fission than 235U fission, but a lower gas diffusion coefficient in the ThO2 than in the UO2; (c) less plutonium accumulation at the rim of the fuel pellets; (d) greater decay heat; (e) microheterogeneous arrangement of fuel; and (f) more-negative moderator temperature and Doppler coefficients and a smaller delayed-neutron fraction. The newly developed models for ThO2 were checked against data from the light water breeder reactor program. Calculations by these analytical tools indicate that the in-service and transient performance of homogeneous ThO2-UO2-based fuels with respect to safety is generally equal to or better than that of 100% UO2 fuel. The in-service and transient temperatures in the most promising neutronic design of microheterogeneous ThO2–UO2–based fuel are greater than the temperatures in 100% UO2 fuel but are still within normal LWR safety limits. The reactor kinetics parameters for ThO2-PuO2–based fuel cause a higher transient reactor power for some postulated accidents, but in general, the margin of safety for ThO2-PuO2 fuels is equal to or greater than that in 100% UO2 fuels.

Use of Thoria-Urania Fuels in PWRs: A General Review of a NERI Project to Assess Feasible Core Designs, Economics, Fabrication Methods, In-Pile Thermal/Mechanical Behavior, and Waste Form Characteristics

This paper provides an introduction to and a summary of the remaining papers in this issue of Nuclear Technology. The papers in this issue present the important results from a U.S. Department of Energy–sponsored Nuclear Engineering Research Initiative (NERI) project to study the efficacy of the thorium-uranium dioxide (ThO2-UO2) once-through fuel cycle in current light water reactors. The project addressed fuel cycle neutronics and economics; ThO2-UO2 fuel manufacturing; the in-pile thermal/mechanical behavior of ThO2-UO2 fuel during normal, off-normal, and accident conditions; and the long-term stability of ThO2-UO2 waste. Results from this work show that a small-scale separation of the uranium and thorium will enhance the fuel reactivity and achievable burnup from uranium-thorium dioxide fuels. Under conditions that meet the thermal requirements in present pressurized water reactors (PWRs), a properly designed microheterogeneous fuel will have more reactivity than all-uranium fuel, and the overall production of plutonium is significantly reduced. The use of thorium as a host for actinide fuels when PWRs are used for actinide transmutation was also explored. It was also determined that there were no fundamental obstacles to converting the current plants that manufacture uranium oxide–only fuel to a mixed ThO2-UO2 fuel. Also, the in-service and transient thermal and mechanical performance of homogeneous ThO2-UO2–based fuels with respect to safety is generally equal to or better than that of all-uranium fuel. Furthermore, a mixed thorium-uranium dioxide spent fuel appears to be a much more stable waste form than uranium oxide spent fuel.

Influence of technetium on the microstructure of a stainless steel–zirconium alloy

Stainless steel–zirconium alloys are being developed for the disposal of metallic waste generated during the electrometallurgical treatment of spent Experimental Breeder Reactor (EBR-II) fuel. The metallic waste contains the fission product technetium, which must be incorporated into a stable waste form matrix to prevent its release into the environment. The baseline waste form for metallic waste from EBR-II fuels is a stainless steel–15 wt% zirconium (SS–15Zr) alloy. The microstructure of SS–15Zr alloys containing 2 wt% technetium was characterized using a combination of microscopy, spectroscopy, diffraction, and chemical analysis techniques. Peaks corresponding to the iron solid solutions ferrite and austenite, ZrFe 2-type Laves polytypes C36 and C15, and an Fe 23Zr 6-type intermetallic were identified in diffraction patterns of the alloy. Discrete technetium-rich phases were not observed either in diffraction patterns or in the microstructure; the element partitioned into various phases of the SS–15Zr alloy. Technetium favors ferrite and austenite over the Zr-based intermetallics. The lattice parameters of the Zr-based intermetallics are smaller than those in an alloy without technetium, which appears to substitute at the zirconium sites of the intermetallic lattice.

Graphite Processing With Carbon Retention in a Waste Form

Conversion of waste graphite into a stable waste form acceptable for long term storage and disposal was considered both theoretically and experimentally. A self-sustaining transformation process of graphite composited with suitable precursors was studied. The powdered precursors that were used were used were: Al+SiO 2 (1), Al+TiO 2 (2) and Ti+SiO 2 (3). Numeric thermodynamic simulation was performed. Equilibrium temperatures and chemical compositions of reaction products were determined for a wide range of component ratios in the source mixtures. The highest temperatures (up to 2300 K) were observed for precursor type (2). Precursor type (3) demonstrated a minimal rise of temperature of up to 1900 K. Regions of compositions with complete binding of all chemical elements as well as production of stable final products were found to be rather narrow. About 10 – 13 wt.% of carbon can be processed in composition with given precursors. The gas phase reaction products were studied to minimize carry over of radionuclides. Carbon monoxide was shown to be the main component of the gas phase. The self-sustaining synthesis process was conducted in ceramic crucibles at ambient pressure in an air atmosphere. Batch masses ranged between 0.1 – 1 kg. Best results were obtained for processing of graphite composited with Al and TiO 2 . XRD analysis has identified titanium carbide and corundum in the waste form produced. These experiments confirmed that carbon can be converted completely into a stable waste form.