Acidic Nuclear Waste Solutions Research Articles

An Inorganic Microsphere Composite for the Selective Removal of 137Cesium from Acidic Nuclear Waste Solutions. 1: Equilibrium Capacity and Kinetic Properties of the Sorbent

A new inorganic ion exchange composite consisting of ammonium molybdophosphate, (NH4)3P(Mo3O10)4·3H2O (AMP), synthesized within hollow aluminosilicate microspheres (AMP‐C) has been developed. Two different batches of the sorbent were produced resulting in 20% and 25% AMP loading for two and three loading cycles, respectively. The selective cesium exchange capacity of this inorganic composite was evaluated using simulated sodium bearing waste solution as a surrogate for the acidic tank waste currently stored at the Idaho National Laboratory (INL). Equilibrium isotherms obtained from these experiments were very favorable for cesium uptake and indicated maximum cesium loading of approximately 9% by weight of dry AMP. Batch kinetic experiments were also performed to obtain the necessary data to estimate the effective diffusion coefficient for cesium in the sorbent particle. These experiments resulted in effective intraparticle cesium diffusivity coefficients of 4.99 × 10−8 cm2/min and 4.72 × 10−8 cm2/min for the 20% and 25% AMP‐C material, respectively.

An Inorganic Microsphere Composite for the Selective Removal of 137Cesium from Acidic Nuclear Waste Solutions. 2: Bench‐Scale Column Experiments, Modeling, and Preliminary Process Design

A new inorganic ion exchange composite for removing radioactive cesium from acidic waste streams has been developed. The new material consists of ammonium molybdophosphate, (NH4)3P(Mo3O10)4·3H2O (AMP), synthesized within hollow aluminosilicate microspheres (AMP‐C), which are produced as a by‐product from coal combustion. The selective cesium exchange capacity of this inorganic composite was evaluated in bench‐scale column tests using simulated sodium bearing waste solution as a surrogate for the acidic tank waste currently stored at the Idaho National Laboratory (INL). Total cesium loading on the columns at saturation agreed very well with equilibrium values predicted from isotherm experiments performed previously. A numerical algorithm for solving the governing partial differential equations (PDE) for cesium uptake was developed using the intraparticle mass transfer coefficient obtained from previous batch kinetic experiments. Solutions to the governing equations were generated to obtain the cesium concentration at the column effluent as a function of throughput volume using the same conditions as those used for the actual column experiments. The numerical solutions of the PDE fit the column break through data quite well for all the experimental conditions in the study. The model should therefore provide a reliable prediction of column performance at larger scales.

Titanium antimonates in various Ti∶Sb ratios: ion exchange properties for radionuclide ions

Ion exchange properties of titanium antimonates have been investigated with the main objective being selective removal of radionuclides from acidic nuclear waste solutions and in the presence of strongly interfering calcium. Five samples with Ti∶Sb ratios between 0.21–4.9 were synthesized by hydrolysis in distilled water under reflux. Products of either a mopungite, rutile or pyrochlore phase crystallized depending on the Ti∶Sb ratio, the order of metal prehydrolysis and the synthesis temperature (20 °C, reflux at ∼100 °C or hydrothermal treatment at 200 °C). The acidity of the exchanger and 85Sr selectivity in acid increases with increasing antimony content, while Sr2+/Na+ exchange is favored as the acidity decreases. The tolerance for Ca2+ in strontium uptake has a reverse trend and the 85Sr/Ca2+ selectivity coefficients (kSr/Ca) increase with the Ti∶Sb ratio from 1.2 to 130 in 0.01 M Ca(NO3)2. In addition to the Ti∶Sb ratio, the crystal phase directs the radionuclide affinity through the ion sieve effect. The pyrochlore structure favors strontium while the rutile phase seems to prefer caesium in acid.

Evaluation of a novel solid phase extraction composite for the removal of actinides from acidic nuclear waste solutions

A novel approach for preparing granular solid phase extractants (SPE) for the separation of actinides has been developed, wherein the extractant is directly immobilised in an inert matrix. This allows substantially higher extractant loading in the SPE material than for conventional extraction chromatography resins. This approach utilises polyacrylonitrile (PAN) as the inert matrix material. The well-known actinide extractant octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) has been impregnated within PAN granules at extractant loadings from 20 to 33 wt. % CMPO. The porosity of the PAN matrix allows the active material to have rapid and complete access to the solution containing the target contaminants, resulting in improved kinetics and higher sorption capacities. The new CMPO-PAN SPE material was prepared in a batch process. Equilibrium batch contact tests were then performed with this material to determine the efficacy for removing plutonium and uranium from acid solutions as well as simulated solutions of tank waste currently stored at the Idaho National Engineering and Environmental Laboratory (INEEL). The potential interference of non-target metals was also investigated. The equilibrium behaviour of the new composite material will be discussed.

Application of a second order kinetic model for the preliminary design of an adsorption bed system using ammonium molybdophosphate–polyacrylonitrile for the removal of 137Cs from acidic nuclear waste solutions

Ammonium molybdophosphate (AMP) immobilized on polyacrylonitrile (PAN) is an engineered form of cesium selective sorbent material developed at the Czech Technical University in Prague. A form of a second order kinetic model is applied to experimental AMP–PAN breakthrough data. A system utilizing three fixed beds in series is described that allows for continuous nuclear waste processing at approximately 890 l/h. This process would yield a product stream that is below the NRC Class A low level waste limit for 137Cs, and would result in high activity waste volume reduction factors of approximately 2200 and 570 for liquid sodium bearing waste and dissolved calcine waste, respectively, which are currently stored at the Idaho National Engineering and Environmental Laboratory.

Evaluation of ammonium molybdophosphate-polyacrylonitrile (AMP-PAN) as a cesium selective sorbent for the removal of 137Cs from acidic nuclear waste solutions

Ammonium molybdophosphate (AMP) immobilized on polyacrylonitrile (PAN) is an engineered form of cesium selective sorbent material developed at the Czech Technical University in Prague. The selective cesium capacity of this inorganic ion exchange media was evaluated with simulated sodium bearing waste (SBW) and dissolved pilot plant calcine at the Idaho Nuclear Technology and Engineering Center (INTEC). Equilibrium isotherms were obtained at 23°C and 50°C with 48-h equilibrium contact times. The equilibrium isotherms were of the ‘favorable’ type and showed asymptotic maximum cesium loading of approximately 8.7% by weight of dry AMP. Dynamic column tests were performed with similar feeds using a 1.5-cm 3 bed in an up-flow configuration. Complete breakthrough curves were generated at feed rates up to 100 bed volumes (BV) per hour and sorbent capacities were determined to be approximately 32 g Cs/kg AMP-PAN (dry) at 100% breakthrough. Stop flow data were also obtained from the column tests. These stop flow data, in conjunction with column rate data, imply that cesium mass transfer is predominately particle diffusion limited at feed rates below 40 BV/h.

Determination of a Solid Phase Mass Transfer Coefficient for Modeling an Adsorption Bed System Using Ammonium Molybdophosphate-Polyacrylonitrile (AMP-PAN) as a Sorbent for the Removal of 137 Cs from Acidic Nuclear Waste Solutions

Ammonium molybdophosphate (AMP) immobilized on a polyacrylonitrile (PAN) support is an engineered form of cesium selective sorbent material developed at the Czech Technical University in Prague. This material is being investigated as a sorbent for removing 137Cs from Idaho National Engineering and Environmental Laboratory (INEEL) acidic sodium bearing waste (SBW) solution. As part of this study, a computer program to solve the partial differential equations (PDE's) for continuity and rate of exchange in a fixed bed system has been developed using numerical finite difference algorithms. These equations are solved iteratively in order to derive a mass transfer coefficient that agrees with the results of bench scale column experiments. This mass transfer coefficient is then applied in the PDE solutions to predict breakthrough behavior in a semi-scale column experiment. The model provided excellent agreement with the semi-scale data with a mass transfer coefficient of 0.0126 min−1.

Cobalt bis(dicarbollide) ions with covalently bonded CMPO groups as selective extraction agents for lanthanide and actinide cations from highly acidic nuclear waste solutions

A new series of boron substituted cobalt bis(dicarbollide)(1−) ion (1) derivatives of the general formula [(8-CMPO-(CH2-CH2O)2-1,2-C2B9H10)(1′,2′-C2B9H11)-3,3′-Co]− (CMPO = Ph2P(O)-CH2C(O)NR, R = C4H9 (3b), -C12H25 (4b), -CH2-C6H5 (5b)) was prepared by ring cleavage of the 8-dioxane-cobalt bis(dicarbollide) (2) bi-polar compound by the respective primary amines and by subsequent reaction of the resulting amino derivatives (3a–5a) with the nitrophenyl ester of diphenylphosphorylacetic acid. The compounds were synthesized with the aim to develop a new class of more efficient extraction agents for liquid–liquid extraction of polyvalent cations, i.e. lanthanides and actinides, from high-level activity nuclear waste. All compounds were characterized by a combination of 11B NMR, 1H high field NMR, Mass Spectrometry with Electrospray and MALDI TOF ionisation, HPLC and other techniques. The molecular structure of the supramolecular Ln3+ complex of the anion 5b was determined by single crystal X-ray diffraction analysis. Crystallographic results proved that the Ln(III) atom is bonded to three functionalised cobalt bis(dicarbollide) anions in a charge compensated complex. The cation is tightly coordinated by six oxygen atoms of the CMPO terminal groups (two of each ligand) and by three water molecules completing the metal coordination number to 9. Atoms occupying the primary coordination sphere form a tri-capped trigonal prismatic arrangement. Very high liquid–liquid extraction efficiency of all anionic species was observed. Moreover, less polar toluene can be applied as an auxiliary solvent replacing the less environmentally friendly nitro- and chlorinated solvents used in the current dicarbollide liquid–liquid extraction process. The extraction coefficients are sufficiently high for possible technological applications.

Physico-chemical and ion exchange characteristics of ammonium molybdophosphate irradiated with electrons

Physico-chemical changes accompanying the irradiation of ammonium molybdophosphate (AMP) with electrons up to a dose of 200 MGy have been studied by X-ray diffraction, SEM, EDX, thermal analyses, EPR and NMR measurements. The ion exchange characteristics of the control and irradiated AMP with respect to Cs+ have been studied using simulated acidic nuclear waste solutions in a column. Apparent equilibrium constants for the Cs+–H+ exchange have been calculated from breakthrough curves. Structural breakdown occurred on irradiating AMP to a dose of 100–200 MGy and was partially converted to MoO3 and ammonium molybdate. Although the useful exchange capacities of the irradiated AMP were reduced, the apparent ion exchange selectivity for Cs+ (with respect to H+) improved as compared to the control AMP.

Evaluation of the SREX Solvent Extraction Process for the Removal Of 90Sr and Hazardous Metals from Acidic Nuclear Waste Solutions Containing High Concentrations of Interfering Alkali Metal Ions

The SREX process, which was developed at Argonne National Laboratories, has been evaluated for its effectiveness for the decontamination of radioactive liquid waste at the Idaho Chemical Processing Plant located at the Idaho National Engineering Laboratory. The extraction solvent consists of 0.15 M 4′,4′,(5′)-di-(t-butyl-dicyclohexo)-18-crown-6 in either 1.2 M TBP/Isopar L®. Suppressed extraction due to matrix interferences limits DSr values to the range of 3-4 in batch extraction experiments. The SREX solvent has been shown to be effective in the removal of non-radioactive Pb from liquid wastes, in addition to the extraction of 90Sr. The information from this study has been used to develop a proposed flowsheet for the treatment of liquid waste in centrifugal contactors.

3-Phenyl-4-benzoyl-5-isoxazolone: A promising chelate extractant for actinide separation from acidic nuclear waste solutions

In the presence of tri-n-octyl-phosphine (TOPO) oxide, 3-phenyl-4-benzoyl-5-isoxazolone (PBI) has been found to be a promising chelate extractant for the partitioning of actinides from acidic nuclear waste solutions. Quantitative extraction of Pu and U is possible in the nitric acid concentration range 1–6 M, whereas Am can be extracted only from solutions with an acidity $ ̌ 2 M . Extraction studies of Am and U under varying loading conditions are also carried out and conditions for quantitative stripping are arrived at.

The application of novel extraction chromatographic materials to the characterization of radioactive waste solutions

A simple method for the separation and preconcentration of radiostrontium from acidic nuclear waste solutions for subsequent determination is described. The method involves passage of the waste solution, acidified to at least 2M with nitric acid, through an extraction chromatographic column consisting of a 1M solution of bis-4,4′(5′)[(t-butyl)cyclohexano]-18-crown-6 in 1-octanol sorbed on an inert polymeric substrate, which preferentially retains strontium. The strontium may then be stripped from the column with a small volume of either dilute (≤0.05M) nitric acid or water. Actinides present are removed quantitatively prior to strontium separation by passage of the sample through an actinide-specific extraction chromatographic column.

Solvent Extraction Recovery of Byproduct137Cs and90Sr from HNO3Solutions—A Technology Review and Assessment

Abstract A principal purpose of this paper is to call out what is perceived to be a timely need for increased research to develop improved and new solvent extraction technology for recovery of 137Cs and 90Sr from strong (≥ 0.5M HNO3) acid media, technology which can be quickly scaled up to routine plant-scale operation with highly radioactive waste solutions when needed. The present and forseeable future (2020) U.S. inventory of 137Cs and 90sr is listed; much of this inventory is expected to be available only in strongly acidic nuclear waste solutions. Comparison of available methods-precipitation and ion exchange as well as solvent extraction—for recovering 137 Cs and 90Sr from alkaline and acid media demonstrates, as expected, that very limited technology is available for use with strongly acidic solutions. Solvent extraction procedures which employ either macrocyclic polyethers (crown ethers) or dicarbolide. H+{[π-(3)-1, 2-B9C2H11Cl2±2Co}−, to extract 137Cs and 90Sr from ≥ 0.5M HNO3 solutions have rece...