Simulated High-level Nuclear Waste Research Articles

Properties and vibrational spectra of magnesium phosphate glasses for nuclear waste immobilization

The leaching behavior and structure of magnesium phosphate glasses containing 45–55 mol% MgO incorporated with simulated high level nuclear wastes (HLW) were studied. The leach rate of the waste glasses decrease with increasing of the simulated HLW content. The gross leach rate of the glass waste form containing 50 mol% MgO and 45 mass% simulated HLW is of the order of 10 −6 g/cm 2 day at 90 °C, which is small enough as compared with the corresponding release from a currently used borosilicate glass waste form. The isolated ions such as dimeric (P 2O 7) 4− and monomeric (PO 4) 3− ions increase upon as increasing the incorporating amount of the simulated HLW. The changes in properties can be attributed to the structure changes owing to the incorporation of the simulated HLW.

N,N,N′,N′‐Tetraoctyl Diglycolamide (TODGA): A Promising Extractant for Actinide‐Partitioning from High‐Level Waste (HLW)

N,N,N′,N′‐Tetraoctyl diglycolamide (TODGA) has been evaluated as an extractant for the partitioning of minor actinides from simulated high level nuclear waste solutions. Acid uptake studies suggested that TODGA is more basic (KH: 4.1) as compared to TRUEX and DIAMEX solvents viz. CMPO (KH: 2.0) and DMDBTDMA (KH: 0.32), respectively. TODGA molecules display a tendency toward aggregation in n‐dodecane at lower acidities. Effect of diluent on the distribution behavior of Am(III) was studied employing diluents of different dielectric constants. N,N‐dialkyl amides with different alkyl groups viz. dibutyl decanamide (DBDA), di(2‐ethylhexyl) acetamide (D2EHAA), di(2‐ethylhexyl) propionamide (D2EHPRA), di(2‐ethylhexyl) isobutyramide (D2EHIBA), dihexyl octanamide (DHOA) and dihexyl decanamide (DHDA) were evaluated as phase modifiers. Distribution behavior of various metal ions viz. Am(III), Pu(IV), U(VI), Eu(III), Fe(III), Sr(II) and Cs(I) was studied from pure nitric acid solution as well as from simulated high level waste solution.

Environment and oxidation state of molybdenum in simulated high level nuclear waste glass compositions

Alkali borosilicate glasses containing between 20 and 35wt% of a simulated high level nuclear waste stream with varying Li2O contents were melted under neutral (air) and reducing (nitrogen/hydrogen) conditions. XRD analysis of the as-cast glasses showed a tendency for the products to remain amorphous when melted under neutral conditions and for metallic silver to develop in the reduced melts. EXAFS analysis revealed (MoO4)2− tetrahedra in all glasses regardless of the sparge applied during melting. The glasses were heat treated to simulate an interruption to the cooling system used to prevent heat build-up in the vitrified product store. Powellite-type molybdate phases were found to develop in the heat treated samples and formed at lower waste loadings in glasses sparged with a reducing gas. A reduction in the quantity of Li2O lead to a reduction in the quantity of powellite-type molybdate phases. EDS showed the primary molybdate phase to be high in Sr and rare earth elements and TEM indicated that the presence of silver metal encouraged molybdate formation.

An Investigation into the Oxidation State of Molybdenum in Simplifies High Level Nuclear Waste Glass Compositions

The redox state of Mo in glasses containing simplified simulated high level nuclear waste (HLW) streams has been investigated using Electron Spin Resonance (ESR) and X-ray Photoelectron Spectroscopy (XPS). Melts performed in highly oxidising or neutral (air) atmospheres contained Mo6+, but our study indicates that the Mo can be at least partially reduced to Mo4+ or Mo3+ by melting in a reducing atmosphere. The implications for glasses containing a full simulated HLW stream are that the formation of detrimental crystalline phases containing Mo6+ upon heat treatment may be avoided by reducing the oxidation state of Mo during melting.

Molybdenum in Nuclear Waste Glasses - Incorporation and Redox state

The composition and structure of the mixed metal molybdates that can form in simulated high level nuclear waste (HLW) glass melts have been studied. It was found that molybdates of a tetragonal scheelite type were formed upon heat treatment of the simulated glass samples (representative of the slow cools experienced by the real vitrified product), and that these compounds are capable of incorporating the majority of the mono, di, and trivalent cations that would be present in a real HLW glass. In addition, it has been shown that altering the redox conditions prevailing upon melting can promote or suppress crystallisation in simplified model waste glasses that contain molybdenum. Experiments to investigate the effect of redox conditions during melting of simulated HLW glass on molybdate formation are also reported.

99/02759 Iron-phosphate-based chemically bonded phosphate ceramics for mixed waste stabilization

The leaching behavior and structure of magnesium phosphate glasses containing 45–55 mol% MgO incorporated with simulated high level nuclear wastes (HLW) were studied. The leach rate of the waste glasses decrease with increasing of the simulated HLW content. The gross leach rate of the glass waste form containing 50 mol% MgO and 45 mass% simulated HLW is of the order of 10−6 g/cm2 day at 90 °C, which is small enough as compared with the corresponding release from a currently used borosilicate glass waste form. The isolated ions such as dimeric (P2O7)4− and monomeric (PO4)3− ions increase upon as increasing the incorporating amount of the simulated HLW. The changes in properties can be attributed to the structure changes owing to the incorporation of the simulated HLW.

Structure of Pd–Te precipitates in a simulated high-level nuclear waste glass

Structural and bonding characteristics of simplified (Pd, Te) precipitates have been determined in a simulated nuclear French glass using extended x-ray absorption fine structure (EXAFS) and x-ray diffraction. In this sample, these precipitates have a homogeneous composition, with about 10 wt.% Te. They retain a face-centered cubic structure as in pure Pd with a cell parameter which obeys Vegard's law. Pd K-edge EXAFS shows the presence of Te in the Pd coordination shell, with (Pd–Te) distances of 2.80 Å. These distances, higher by 0.05 Å than the (Pd–Pd) distances, may result in a lower packing efficiency of the CFC lattice. The comparison with the average distances derived from x-ray diffraction shows the nonmetallic character of the Pd–Te bond in these precipitates. These bonding modifications may cause the limited solubility of Te in metallic Pd.

Apatite- and Monazite-Bearing Glass-Crystal Composites for the Immobilization of Low-Level Nuclear and Hazardous Wastes

AbstractThis study demonstrates that glass-crystal composite waste forms can be produced from waste streams containing high proportions of phosphorus, transition metals, and/or halides. The crystalline phases produced in crucible-scale melts include apatite, monazite, spinels, and a Zr-Si-Fe-Ti phase. These phases readily incorporated radionuclide and toxic metals into their crystal structures, while corrosion tests have demonstrated that glass-crystal composites can be up to 300-fold more durable than simulated high-level nuclear waste glasses, such as SRL 202U.

α‐Decay Damage Effects in Curium‐Doped Titanate Ceramic Containing Sodium‐Free High‐Level Nuclear Waste

A polyphase titanate ceramic incorporating sodium‐free simulated high‐level nuclear waste was doped with 0.91 wt% of 244Cm to accelerate the effects of long‐term self‐irradiation arising from α decays. The ceramic included three main constituent minerals: hollandite, perovskite, and zirconolite, with some minor phases. Although hollandite showed the broadening of its X‐ray diffraction lines and small lattice parameter changes during damage ingrowth, the unit cell was substantially unaltered. Perovskite and zirconolite, which are the primary hosts of curium, showed 2.7% and 2.6% expansions, respectively, of their unit cell volumes after a dose of 12 × 1017α decays.g‐1 Volume swelling due to damage ingrowth caused an exponential (almost linear) decrease in density, which reached 1.7% after a dose of 12.4 × 1017α decays.g‐1. Leach tests on samples that had incurred doses of 2.0 × 1017 and 4.5 × 1017 a decays g‐1 showed that the rates of dissolution of cesium and barium were similar to analogous leach rates from the equivalent cold ceramic, while strontium and calcium leach rates were 2–15 times higher. Although the curium, molybdenum, strontium, and calcium leach fates in the present material were similar to those in the curium‐doped sodium‐bearing titanate ceramic reported previously, the cesium leach rate was 3–8 times lower.

Immobilization of Simulated High Level Nuclear Wastes with Tholeiite-Type Basalt

The powder mixtures of tholeiite-type basalt containing 5, 10 and 20wt% of simulated high-level nuclear wastes were sintered at 1000, 1050, and 1100°C for 5, 10, 30, 50 and 100h. Compressive strength, thermal expansion coefficient, chemical durability, apparent density, porosity, leaching test into alkali and acid solution were examined. The maximum compressive strength of the sample sintered at 1100°C for 100h was slightly greater than that of commercial quartz glass-rod. Judging from the results of various tests, the optimum condition of sintering for a good waste disposal material is as follows: amount of simulated waste; 10wt%, firing temperature; 1100°C, sintering duration; 100h.

Volatility of Simulated High‐Level Nuclear Waste Glass by Thermogravimetric Analysis

The volatilities of simulated, high‐level nuclear waste glasses have been measured using thermogravimetric analysis (TGA). These volatilities were measured in the region of the glass transition temperature (Tg) of the waste glasses, which is between 450° and 500°C. These data were obtained because the Waste Acceptance Preliminary Specifications require that no foreign materials be released into the canistered waste form upon heating of the canister to this glass transition temperature. In fact, all of the waste glass samples studied actually exhibited a net weight gain upon heating. This weight gain was shown to be due to oxygen uptake through oxidation of FeO. Powdered glass samples did show a small weight loss which was smaller in magnitude than the weight gain and was associated with water desorption. No true volatility was detected to the level of sensitivity (0.01 wt%) of the TGA instrument. This converts to a sensitivity of 330 μg/m2 of glass surface and a corresponding minimum value of 41 mg of volatiles for each Defense Waste Processing Facility conistered waste form. TGA experiments carried out at higher temperatures (800°C) revealed that organic concentrations in the waste glasses are less than 0.01 wt%. Thus, these results demonstrate that the Defense Waste Processing Facility will be able to comply with the Waste Acceptance Preliminary Specifications on the exclusion of foreign materials from the canistered waste forms, after exposure to Tg.

Separation of Cs-137, Sr-90, and Th-232 in Aqueous Solution by Using a Multistage Countercurrent Batch Contactor Ion-Exchange System

Abstract The radioisotopes Cs-137, Sr-90, and Th-232, in simulated high level nuclear liquid waste, were recovered and sequentially separated by using an 8-stage flask simulation of a multistage countercurrent batch contactor (MCBC) ion-exchange system. The solution normality ratio used for the three ions Cs:Sr:Th was 1:1:1. The solution was contacted for 2 h with resin Dowex HCR-W2 initially in the H+ form. The obtained recovery efficiency of the MCBC and the purity of the separated ions for the three ions were 77 and 60% for Cs+, 64 and 93% for Sr2+, and 99 and 65% for Th4+, respectively.

Selective Exchange of Pu4+ and Am3+ by Titanium and Tin Antimonates Cation Exchangers.

The selectivity for Pu4+ and Am3+ has been studied as a function of [HNO3] on titanium and tin antimonates cation exchangers at 30°C. A large selectivity toward Pu4+ and Am3+was discovered on tin and titanium antimonates cation exchangers, while they showed low selectivity for 4f elements such as La3+ and Nd3+. The log-log plot of the distribution coefficients vs. [HNO3] showed the linear relationship with a slope nearly equal to the valence of exchanging ions. It indicates that the adsorption predominantly proceeds through the ion exchange. The ion-exchange rate in the initial stage of adsorption was about the same as that of a commercial strong acid-type organic cation exchange resin. Highly selective uptake of Pu4+ has also been demonstrated in a simulated high level nuclear waste solution containing 0.5 M HNO3.

Self‐Irradiation Damage of a Curium‐Doped Titanate Ceramic Containing Sodium‐Rich High‐Level Nuclear Waste

The polyphase titanate ceramic containing sodium‐rich simulated high‐level nuclear waste was doped with 0.69 wt% of 244Cm to accelerate long‐term self‐irradiation due to α decays. α autoradiography showed that α emissions were almost uniformaly distributed throughout the curium‐doped samples on a >20‐μm scale although micropore surfaces and titanium oxide agglomerates were free of α‐emitting nuclides. The phase assemblage of the curium‐doped titanate ceramic included freudenbergite and loveringite in addition to the more abundant oxide phases: hollandite, perovskite, and zirconolite. Accumulation of α decays was accompanied by a gradual decrease in density. The increment of density was – 1% after an equivalent age of 5000 yr. Leach tests showed a slight trend toward higher total release of curium with equivalent age. The release of soluble nonradioactive elements (e.g., Na, Cs, Sr, and Ca) in the oldest specimens (equivalent age, 2000 yr) varied from specimen to specimen but, on average, were higher than specimens that had suffered a lower radiation dose.

Leach Rates and Thermal Properties of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate (LIP) glasses loaded with a simulated high-level nuclear waste were studied on their leach rates and thermal properties. The obtained results showed that the phosphate glass matrix consisting of lead monoxide, phosphorus pentoxide and ferric oxide of 56:35:9w/0 is able to vitrify the waste, pretreated with formic acid to remove Zr, to about 15 w/0 at 950°C. The leach rate of the vitrified waste glass was in the order of 10−7 g/cm2.d at 110°C, which is low compared with that of the borosilicate glass waste form. Increasing the phosphorus pentoxide content of the matrix to higher than 35% enabled it to produce the glass form with the waste near 20 w/0 at 950°C, but this increase rendered the glass waste form more soluble than the former. Thermal properties such as thermal expansion coefficient, critical cooling rate for vitrification and temperatures of glass transition, softening and maximum rate of crystallization were measured and discussed. Removing Na ions from wastes improves consi...

Titanate Ceramics for the Immobilization of Sodium‐Bearing High‐Level Nuclear Waste

The phase chemistries and microstructures of titanate‐based ceramics containing simulated high‐level nuclear waste with varying sodium contents were compared. Incorporation of relatively low sodium levels (∼0.9 wt% Na2O) resulted in more complex phase assemblages. The principal hosts for sodium were hibonite and freudenbergite, and, when iron and sodium were present in combination, loveringite was also stabilized. During fabrication, oxygen potential was controlled by Ti‐TiO2 or TiH2‐Ti3O5 solid‐state buffers. These metal and hydride oxygen getters behaved similarly, neither disturbing the phase assemblages nor significantly altering the partitioning of waste elements between radiophases. It is believed that the hydrothermal stability of the sodium‐bearing ceramics (containing up to 2.7 wt% Na2O) will be comparable to sodium‐free material, since less durable sodium‐rich phases are encapsulated in a resistant matrix. Extensive formation of glassy phases may cause embrittlement at higher sodium loadings.

Leach Rates of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate glasses loaded with simulated high-level nuclear wastes at temperatures between 900 and 1,100°C were studied on their soaking behavior in distilled water by means of leachate solution analysis. The obtained results showed that the leach rates of the glass waste forms were at least 10 to 100 times lower than that of the currently investigated borosilicate glass, even though the selective release of Na ion from the forms was observed. Zirconium of the waste led the glass to partial crystallization at 900°C, but was able to be incorporated in the glass at near 1,100°C. The liquid chromatographic analysis of poly-phosphate ions in the leachate solution revealed that the low leachability of the glass forms was brought about by a certain degree of depoly-merization of long poly-phosphate chains of lead metaphosphate caused by the addition of ferric oxide.

Leach rates of lead-iron phosphate glass waste forms.

Lead-iron phosphate glasses loaded with simulated high-level nuclear wastes at temperatures between 900 and 1,100°C were studied on their soaking behavior in distilled water by means of leachate solution analysis. The obtained results showed that the leach rates of the glass waste forms were at least 10 to 100 times lower than that of the currently investigated borosilicate glass, even though the selective release of Na ion from the forms was observed. Zirconium of the waste led the glass to partial crystallization at 900°C, but was able to be incorporated in the glass at near 1,100°C. The liquid chromatographic analysis of poly-phosphate ions in the leachate solution revealed that the low leachability of the glass forms was brought about by a certain degree of depoly-merization of long poly-phosphate chains of lead metaphosphate caused by the addition of ferric oxide.

Features of fracture faces in synroc C

Fracture faces and polished surfaces of Synroc C specimens with 10% and 20% simulated high-level nuclear waste (HLW) have been studied with X-ray photoelectron spectroscopy. Elemental abundances and chemical environments for all matrix and most HLW atomic species have been investigated in the surface and near-surface layers before and after chemical attack. These observations suggest that the fracture mechanism is mainly intergranular, that particular phases are not preferentially exposed in the fracture faces, that some HLW species (caesium and molybdenum) are found in the grain boundaries, that intergranular regions are thin amorphous films or larger “glassy” triple points and that the intergranular films exhibit low chemical durability in an aqueous environment. These effects are independent of waste loading up to 20% and are consistent with the very good chemical durability reported for large monoliths.

Lead phosphate glass as a stable medium for the immobilization and disposal of high-level nuclear waste

Investigations of the characteristics of lead phosphate glasses pertinent to the use of these materials as a storage medium for high-level nuclear waste have been carried out. By combining lead phosphate glass with various types of simulated high-level nuclear waste, it is possible to form a highly corrosion resistant, homogeneous, easily processed glass. For corroding solutions at 90° C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. Lead phosphate nuclear waste glasses can be prepared at temperatures as low as 800°C, since they exhibit very low melt viscosities in the 800–1050°C temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550°C and are not adversely affected by large doses of gamma radiation in H 2O at 135°C. It appears that lead phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear waste forms.