Enhanced separation of palladium from simulative high-level liquid waste using a silica-supported polymer functionalized with aminopyridine

NO2 and NO generated during a boiling and drying accident, which affects the release of volatilized radioactive Ru into the atmosphere, were examined using various samples including simulated high-level liquid waste and a thermogravimetric analyzer. NO2 and NO in the gas flowing out of the analyzer were measured separately using a NOx analyzer equipped with NO2 and NO sensors. The samples were heated to 600°C at constant heating rates of mainly 0.2 and 1°C min−1 that was adopted taking into account the decay heat of high-level liquid waste. It was found that under 180°C some nitrates in the liquid waste mainly separated their nitrate groups as HNO3 without generating NOx (a mixture of NO and NO2) and above 300°C the residual HNO3 in the waste participated in thermal decomposition generating NOx. The generation rates of NO2 and NO were obtained as a function of time using Arrhenius type equations, and the O2 rate was derived from these equations using the stoichiometry of the reactions that generate NO2, NO, and O2.

Gel sphere technology developed for reactor fuel fabrication was applied to the fixation of simulated high-level radioactive waste in crystalline ceramic form for permanent disposal. Gel spheres containing simulated alkaline defense waste sludges and ceramic matrix materials were prepared by internal gelation at waste loadings as high as 90%. The gel spheres were amenable to subsequent drying, sintering, and coating procedures to produce crystalline waste forms with extremely high leach resistances. Potential application of this technique to the processing of commercial power reactor waste was demonstrated by incorporating simulated Purex solvent extraction waste in gel spheres with up to 20% waste loading. Cesium present in the simulated waste was adsorbed on zeolite and immobilized by coating with carbon.

Bechtel National, Inc. has been contracted by the Department of Energy to design a Waste Treatment and Immobilization Plant (WTP) to stabilize liquid radioactive waste that is stored at the Hanford Site as part of the River Protection Project (RPP). Because of its experience with radioactive waste stabilization, the Savannah River National Laboratory (SRNL) of the Westinghouse Savannah River Company is working with Bechtel and Washington Group International, to help design and test certain parts of the waste treatment facility. One part of the process is the separation of radioactive solids from the liquid wastes by cross-flow ultrafiltration. To test this process a cross-flow filter was used that was prototypic in porosity, length, and diameter, along with a simulated radioactive waste slurry, made to prototypically represent the chemical and physical characteristics of a Hanford waste in tank 241-AY-102/C-106. To mimic the filtration process the waste slurry undergoes several steps, including dewatering and washing. During dewatering the concentration of undissolved solids (UDS) of the simulated AY102/C106 waste is increased from 12 wt% to at least 20 wt%. Once at the higher concentration the waste must be washed to prepare for its eventual receipt in a High Level Radioactive Waste Melter to be vitrified. This paper describes the process of washing and filtering a batch of concentrated simulated waste in two cycles, which each containing 22 washing steps that used approximately 7.7 liters of a solution of 0.01 M NaOH per step. This will be the method used by the full-scale WTP to prepare the waste for vitrification. The first washing cycle started with the simulated waste that had a solids concentration of 20 wt% UDS. This cycle began with a permeate filter flux of 0.015 gpm/ft2 (3.68 cm/hr) at 19.6 wt% UDS with a density of 1.33 kg/L, consistency of 19.1 mPa·s, and yield stress of 8.5 Pa. At the end of the 22 washing steps the permeate filter flux increased to 0.023 gpm/ft2 (5.64 cm/hr) at 20.1 wt% UDS with a density of 1.17 kg/L, consistency of 12.6 mPa·s, and yield stress of 10.4 Pa. The average permeate filter flux during the 7 hours of Cycle 1 washing was 0.018 gpm/ft2 (4.41 cm/hr). During Cycle 2 the simulated waste started at a permeate filter flux of 0.025 gpm/ft2 (6.13 cm/hr). Note that the starting flux for Cycle 2 was greater than the ending flux for Cycle 1. The period between the cycles was approximately 12 hours. While no filtering occurred during that period either solids dissolution continued and/or the filter cake was dislodged somewhat with the stopping and starting of filter operation. At the end of the second set of 22 washing steps, the permeate filter flux increased to 0.032 gpm/ft2 (7.84 cm/hr) at 20.6 wt% UDS with a density of 1.16 kg/L, consistency of 9.0 mPa·s, and yield stress of 8.2 Pa. The average permeate filter flux during the 4 hours of Cycle 2 washing was 029 gpm/ft2 (7.11 cm/hr).

Electrolytic extraction of palladium from nitric acid and simulated high-level liquid waste

A pyrometallurgical partitioning process is being developed for recovering minor actinides from high-level liquid waste resulting from PUREX reprocessing. Since the high-level liquid waste consists of concentrated raffinate, concentrated alkaline waste and insoluble residues, the various elements in the waste must be converted to chlorides before they can be sent on to the pyrometallurgical partitioning process. The conversion to chlorides is done by a combination of denitration and chlorination. The mass balance of these processes was measured in the present study using simulated high-level liquid waste. The results indicate that almost all of the alkali elements and Re, substituting for Tc, and significant amounts of Se, Cr, and Mo were separated by denitration, and that Cr, Fe, Zr, Mo, and Te were separated by chlorination. The remaining noble metals, Ni, U, and alkaline-earth and rare-earth elements were efficiently converted to chlorides, which were then supplied to the reductive extraction test using a molten salt/liquid-Cd system to demonstrate that the obtained chlorides are appropriate for processing by pyrometallurgical partitioning. In further reduction, noble metals and Ni were reductively extracted into the liquid-Cd phase, and the rare-earth elements and U into the liquid-Cd phase by adding Li reductant. These elements were completely separated from the alkaline-earth elements remaining in the chloride phase.

A pyrometallurgical partitioning process is being developed for recovering minor actinides from high-level liquid waste resulting from PUREX reprocessing. Since the high-level liquid waste consists of concentrated raffinate, concentrated alkaline waste and insoluble residues, the various elements in the waste must be converted to chlorides before they can be sent on to the pyrometallurgical partitioning process. The conversion to chlorides is done by a combination of denitration and chlorination. The mass balance of these processes was measured in the present study using simulated high-level liquid waste. The results indicate that almost all of the alkali elements and Re, substituting for Tc, and significant amounts of Se, Cr, and Mo were separated by denitration, and that Cr, Fe, Zr, Mo, and Te were separated by chlorination. The remaining noble metals, Ni, U, and alkaline-earth and rare-earth elements were efficiently converted to chlorides, which were then supplied to the reductive extraction test using a molten salt/liquid-Cd system to demonstrate that the obtained chlorides are appropriate for processing by pyrometallurgical partitioning. In further reduction, noble metals and Ni were reductively extracted into the liquid-Cd phase, and the rare-earth elements and U into the liquid-Cd phase by adding Li reductant. These elements were completely separated from the alkaline-earth elements remaining in the chloride phase.

Back end of Indian nuclear fuel cycle-A road to sustainability

Static corrosion tests were performed for the glass phase of a simulated waste form of non-combustible radioactive low-level waste to study a basic aqueous corrosion behavior. The waste form, which was fabricated from simulated waste sample by use of in-can type induction-heated melting, consists of two separated phases; a glass phase and a metal phase. Tests were performed for the glass phase from two types of the waste form with different chemical composition at 35°C and S/V ratio of 2,600 m−1. The glass phase with both types showed an incongruent dissolution similar to conventional high-level radioactive waste (HLW) glasses, i.e., the normalized elemental mass loss (NLi) for soluble elements such as B and Na continued to increase after the saturation of insoluble elements such as Si, A1 and Ca. The NLi for B increased in proportion to the square root of time except for early stage, which suggests that the rate of the long-term dissolution or alteration may be controlled by a diffusion process. Potential secondary phases forming as the results of incongruent dissolution were estimated to be kaolinite and calcite by comparison of the measured solution data with the thermodynamically calculated phase stability relationships. These results suggest that the glass phase has a potential chemical durability not so different from conventional HLW glasses.

Monazite is one of the candidate ceramic matrices for the immobilization of high level radioactive waste (HLW) from the reprocessing of spent nuclear fuel. The monazite phase, Ce0.8Ca0.2PO4, can accommodate cations of different valences due to the mixed valence state (+3 and +4) of Ce in this compound, by facilitating the oxidation and reduction of the Ce3+ and Ce4+ as required by the in-coming cation. This will assist in accommodating HLW of different compositions in the monazite crystal structure even if the average valence of the HLW elements is other than 3. Therefore, the monazite phase, Ce0.8Ca0.2PO4, can be a versatile host for the immobilization of HLW. The enthalpy increment and heat capacity of this versatile monazite phase and a simulated waste form based on it with 20 mass% HLW oxides were measured by drop calorimetry in the temperature range from 373 to 873 K, and the results are compared with those measured for CePO4.

Development and characterization of solidified high-level waste forms are directed at determining both process properties and long-term behaviors of various solidified high-level waste forms in aqueous, thermal, and radiation environments. Waste glass properties measured as a function of composition were melt viscosity, melt electrical conductivity, devitrification, and chemical durability. The alkali metals were found to have the greatest effect upon glass properties. Titanium caused a slight decrease in viscosity and a significant increase in chemical durability in acidic solutions (pH-4). Aluminum, nickel and iron were all found to increase the formation of nickel-ferrite spinel crystals in the glass. Four multibarrier advanced waste forms were produced on a one-liter scale with simulated waste and characterized. Glass marbles encapsulated in a vacuum-cast lead alloy provided improved inertness with a minimal increase in technological complexity. Supercalcine spheres exhibited excellent inertness when coated with pyrolytic carbon and alumina and put in a metal matrix, but the processing requirements are quite complex. Tests on simulated and actual high-level waste glasses continue to suggest that thermal devitrification has a relatively small effect upon mechanical and chemical durabilities. Tests on the effects radiation has upon waste forms also continue to show changes to be relatively insignificant. Effects caused by decay of actinides can be estimated to saturate at near 10/sup 19/ alpha-events/cm/sup 3/ in homogeneous solids. Actually, in solidified waste forms the effects are usually observed around certain crystals as radiation causes amorphization and swelling of th crystals.

The co-extraction performance of Sr and Cs from simulated high-level liquid waste (HLLW) was studied. The extraction solvent consists of 0.1 mol/L dicyclohexano-18-crown-6 (DCH18C-6) and 0.025 mol/L 25,27-bis (isopropoxy) calix[4]-26,28-crown-6 (iPr-C[4]C-6) in n-octanol as a diluent. Testing included the extraction performance of Sr and Cs in nitric acid and in the simulated HLLW medium, and the countercurrent cascade tests. The countercurrent cascade tests included 10 stages for Sr and Cs co-extraction, 2 stages for scrubbing and 8 stages for Sr and Cs co-stripping, or 2 stages for the supplementary extraction of Cs, 4 stages for stripping Sr and 8 stages for stripping Cs were carried out on a miniature centrifugal contactor set. The removal efficiencies of Sr and Cs in the simulated HLLW were 99.0% and 99.9%, respectively, and Sr and Cs could be co-stripped together completely or individually stripped by two stripping sections. Thus, the above extractants could be used to achieve the efficiency required for co-extracting Sr and Cs from HLLW. This process is simpler than the original extracting processes of Sr and Cs.

Currently, at the Idaho Chemical Processing Plant (ICPP) there are about 6800 m3 of liquid sodium-bearing and liquid high-level wastes (HLW), and 3800 m3 of solid calcined HLW. One of the waste processing options under consideration includes separation of the HLW into high activity and low activity (LAW) wastes, followed by immobilization. Preliminary glasses were synthesized for the sodium-bearing, alumina-bearing, and the zirconia-bearing LAW fractions after radionuclide separations. The glasses were formed by crucible melting of a mixture of reagent chemicals representative of the LAW waste streams and frit additives at 1200 °C for 5 hours, followed by overnight annealing at 550 °C and furnace cooling of the melt. These glasses were characterized for density, elastic property, viscosity, chemical durability, structural parameters, and glass phase separation. The results are compared with that of the Hanford's standard glass ARM-i, Savannah River's benchmark glass EA, and the ICPP's grout waste form prepared using the simulated non-radioactive sodium-bearing waste fraction.

Research on calcination thermal decomposition process of simulated high-level liquid waste based on two-step vitrification

The ceramics process flow diagram has been simplified and upgraded to utilize only two major processing steps - fluid-bed calcination and hot isostatic press consolidating. Full-scale fluid-bed calcination has been used at INEL to calcine high-level waste for 18 y; and a second-generation calciner, a fully remotely operated and maintained calciner that meets ALARA guidelines, started calcining high-level waste in 1982. Full-scale hot isostatic consolidation has been used by DOE and commercial enterprises to consolidate radioactive components and to encapsulate spent fuel elements for several years. With further development aimed at process integration and parametric optimization, the operating knowledge of full-scale demonstration of the key process steps should be rapidly adaptable to scale-up of the ceramic process to full plant size. Process flowsheets used to prepare ceramic and glass waste forms from defense and commercial high-level liquid waste are described. Preliminary layouts of process flow diagrams in a high-level processing canyon were prepared and used to estimate the preliminary cost of the plant to fabricate both waste forms. The estimated costs for using both options were compared for total waste management costs of SRP high-level liquid waste. Using our design, for both the ceramic and glass plant, capital and operating costs are essentially the same for both defense and commercial wastes, but total waste management costs are calculated to be significantly less for defense wastes using the ceramic option. It is concluded from this and other studies that the ceramic form may offer important advantages over glass in leach resistance, waste loading, density, and process flexibility. Preliminary economic calculations indicate that ceramics must be considered a leading candidate for the form to immobilize high-level wastes.

Extraction and aggregation behaviour of Zr(IV) in diglycolamide solvents during the treatment of high-level liquid waste solution arising from metallic fuel reprocessing