Sodium-bearing Waste Research Articles

Properties of alkali-activated slag-fly ash-metakaolin hydroceramics for immobilizing of simulated sodium-bearing waste

This work is a continuation of previous study of the alkali-activated slag-fly ash-metakaolin hydroceramics (ASFMH) waste form which was designed for solidifying simulated sodium-bearing waste (S-SBW). The thermal stability, resistances to gamma irradiation, freeze–thaw, and water immersion of ASFMH waste form were investigated systematically. The thermal stability results showed that the physical appearance of ASFMH waste forms was intact and without external cracking after calcination at 300, 500, 800 and 1000 °C. Importantly, the main crystalline phase of ASFMH waste form is still analcime after being calcined at 500 °C. The irradiation results indicated that the strong gamma irradiation could not induce the lattice damage of crystalline phases. The durability tests suggested that the compressive strength loss of ASFMH waste forms was 5.07, 19.04 and 10.51% after 500 kGy doses of gamma irradiation, 5 freeze–thaw cycles and 90 days of water immersion, respectively. It is suggested that ASFMH waste form designed for immobilizing S-SBW shows superior basic properties.

Preparation of Alkali‐Activated Slag‐Fly Ash‐Metakaolin Hydroceramics for Immobilizing Simulated Sodium‐Bearing Waste

The radioactive sodium‐bearing waste (SBW), which is mainly generated from spent nuclear fuel reprocessing activities, requires to be stabilized into a solid form for geological disposal. In this study, the alkali‐activated slag–fly ash–metakaolin hydroceramic (ASFMH) waste form for solidifying simulated SBW (S‐SBW) was prepared by hydrothermal process. The mineralogy, morphology, composition, compressive strength, and chemical durability of ASFMH waste form were investigated. It is shown that crystalline analcime phase was formed in the hydration products of ASFMH waste form after alkaline activation and hydrothermal process. The introducing of S‐SBW not only contributed to forming crystalline zeolite phase but also reduced the temperature and time for forming zeolite phase. Importantly, the cations including simulated radionuclides Sr2+ and Cs+ in S‐SBW could be immobilized as cationic components in the channels or voids of zeolitic structure during the formation of analcime phase. Moreover, when the loading content of S‐SBW in ASFMH waste form was 37.5 wt%, ASFMH waste form still exhibited superior mechanical strength, which was higher than 24 MPa. The product consistency test leaching results showed that ASFMH waste form possessed superior chemical durability after leaching test carried out for 7 d. It is suggested that ASFMH waste form is a promising candidate for application in immobilizing SBW.

Modeling Reactive Transport of Strontium‐90 in a Heterogeneous, Variably Saturated Subsurface

Sodium‐bearing waste (SBW) containing high concentration of 90Sr was accidentally released to the vadose zone at the Idaho Nuclear Technology and Engineering Center, Idaho National Laboratory, Idaho Falls, ID, in 1972. To investigate the transport and fate of the 90Sr through this 137‐m‐thick, heterogeneous, variably saturated subsurface, we conducted a two‐dimensional numerical modeling using TOUGHREACT under different assumed scenarios (low permeability of an entire interbed or just its surface) for the formation of perched water whose presence reflects the unique characteristics of the geologic materials and stratification at the study site. The results showed that different mechanisms could lead to different flow geometries. The assumption of low permeability for the entire interbed led to the largest saturated zone area and the longest water travel time (55 vs. 43 or 44 yr in other scenarios) from the SBW leakage to the groundwater table. Simulated water travel time from different locations on the land surface to the groundwater aquifer varied from <30 to >80 yr. The results also indicated that different mechanisms may lead to differences in the peak and travel time of a small mobile fraction of Sr. The effective distribution coefficient and retardation factor for Sr2+ would change more than an order of magnitude for the same material during the 200‐yr simulation period because of large changes in the concentrations of Sr2+ and competing ions. Understanding the migration rate of the mobile Sr2+ is necessary for designing long‐term monitoring programs to detect it.

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.

Analysis of high-level waste for tetraphenylborate using HPLC

One waste remediation process used at the Savannah River Site was the in-tank precipitation of the beta-emitting 137Cs from high-level waste (HLW) using sodium tetraphenylborate (NaTPB) followed by processing the resulting decontaminated filtrate into grout at the Saltstone Production Facility (SPF). A simple method was developed for the monitoring of tetraphenylborate (TPB) in high-level waste (HLW) containing up to 0.38 Ci/gal of 137Cs. Separation was achieved by extraction of the high sodium-bearing waste with acetonitrile followed by analysis using reversed-phase high performance liquid chromatography (HPLC). The sample preparation method allowed for the handling of an organic extraction layer that had 94% less acitivity than the HLW sample. The subsequent HPLC analysis of the extraction layer determined the TPB concentration in HLW waste to 0.8 mg/l with a %rsd of 8.

Fluidized Bed Steam Reformed (FBSR) Mineral Waste Forms: Characterization and Durability Testing

AbstractFluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium low activity wastes (LAW) such as those existing at the Hanford site, at the Idaho National Laboratory (INL), and the Savannah River Site (SRS). The addition of clay, charcoal, and a catalyst as co-reactants with the waste denitrates the aqueous wastes and forms a granular mineral waste form that can subsequently be made into a monolith for disposal if necessary. The waste form produced is a multiphase mineral assemblage of Na-Al-Si (NAS) feldspathoid minerals with cage and ring structures and iron bearing spinel minerals. The mineralization occurs at moderate temperatures between 650-750°C in the presence of superheated steam. The cage and ring structured feldspathoid minerals atomically bond radionuclides like Tc-99 and Cs-137 and anions such as SO4, I, F, and Cl. The spinel minerals stabilize Resource Conservation and Recovery Act (RCRA) hazardous species such as Cr and Ni. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW) in pilot scale facilities at the Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The results of the SPFT testing and the activation energies for dissolution are discussed in this study.

Immobilization of Sodium-Bearing High-Level Radioactive Waste in Synroc Containing Na2Al2Ti6O16-Type Freudenbergite

Because sodium has a tendency to form less-durable phases during the fabrication of synroc, it is very difficult to immobilize. This study shows that synroc that contains Na2Al2Ti6O16-type freudenbergite can be used to incorporate sodium-bearing high-level radioactive waste. Synroc samples that contain 5.0 wt% Na2O (11.5 wt% waste oxide, accordingly) exhibit very good chemical durability and physical properties. The standard Synroc-C formulation can incorporate only 2 wt% Na2O. Therefore, this study greatly improves the immobilization ability of sodium in synroc.

Binders for radioactive waste forms made from pretreated calcined sodium bearing waste

Radioactive waste generated during the reprocessing of fuel rods by the U.S. Department of Energy (DOE) is stored in underground tanks at Hanford, Savannah River and INEEL. The liquid fraction commonly referred to as sodium bearing waste (SBW), is a highly alkaline solution containing large amounts of sodium hydroxide, sodium nitrate and sodium nitrite. It has been shown that SBW can be mixed with a reducing agent and metakaolin and then calcined at 500°–700°C to form a calcine containing sodium aluminosilicate phases such as zeolite A, hydroxysodalite and/or cancrinite. Although calcination of the pretreated SBW produces a reasonable waste form in its own right, existing regulations require that granular calcines must be solidified before they can be shipped off site. It is possible to solidify the calcine in a number of ways. The calcine can be mixed with additional metakaolin and NaOH solution followed by mild curing (90°–200°C). The solid that forms (aka hydroceramic) has both strength and suitably low leachability. The current study examines the feasibility of using a more conventional Portland cement binder to solidify the calcine. Although strength was adequate, the leachabilities of the Portland cement solidified samples were higher than those of companion samples made with metakaolin. The zeolitic phases in the calcine acted like pozzolans and reacted with the Ca(OH)2 in the Portland cement binder forming additional calcium silicate hydrate (C—S—H). Typically C—S—H is unable to host large amounts of sodium ions in its structure, thus a majority of the sodium present in the zeolites became concentrated in the pore solution present in the Portland cement binder and readily entered the leachant during PCT testing. In this instance metakaolin mixed with NaOH proved to be a superior binder for solidification purposes.

Chemically durable iron phosphate glasses for vitrifying sodium bearing waste (SBW) using conventional and cold crucible induction melting (CCIM) techniques

A simulated sodium bearing waste (SBW) was successfully vitrified in iron phosphate glasses (IPG) at a maximum waste loading of 40 wt% using conventional and cold crucible induction melting (CCIM) techniques. No sulfate segregation or crystalline phases were detectable in the IPG when examined by SEM and XRD. The IPG wasteforms containing 40 wt% SBW satisfy current DOE requirements for aqueous chemical durability as evaluated from their bulk dissolution rate ( D R), product consistency test, and vapor hydration test. The fluid IPG wasteforms can be melted at a relatively low temperature (1000 °C) and for short times (<6 h). These properties combined with a significantly higher waste loading, and the feasibility of CCIM melting offer considerable savings in time, energy, and cost for vitrifying the SBW stored at the Idaho National Engineering and Environmental Laboratory in iron phosphate glasses.

Mercury Removal from Acidic Waste Solutions Using a Thiol Functional Organo‐ceramic Adsorbent

Removal of mercury from highly acidic (1–2 M acid concentrations) waste solutions using a novel thiol‐functionalized organo‐ceramic adsorbent (SOL‐AD‐IV) has been investigated. The viability of mercury extraction was tested by employing Idaho National Engineering and Environmental Laboratory (INEEL) aluminum calcine and sodium bearing waste (SBW) surrogates. A maximum equilibrium uptake capacity of 740 mg/g in the 1–2 M acid concentrations was found, indicating a 1:1 adsorption ratio between Hg(NO3)2 0 and thiol (SH). Loss of uptake activity due to high contents of aluminum, nitrate, and hydrogen ions was not evident. Over 99.7 wt% extraction efficiency was observed up to 4 M HNO3 solutions. Rapid kinetics were observed with an extraction efficiency of >91 wt% within 5 min of reaction time. Dynamic adsorption of mercury on a fixed bed demonstrated a removal capacity of 680 mg/g and effluent concentrations as low 0.005 mg/L. Results showed applicability of this material for effective removal of mercury from highly acidic waste solutions.

Iron – Phosphate Glass (IPG) Waste Forms Produced using Induction Melter with Cold Crucible

ABSTRACTA simulated sodium bearing waste (SBW), which represented a type of high sodium and sulfate waste, was successfully vitrified in iron phosphate glasses (IPG), at a maximum waste loading of 40 wt%, using a cold crucible induction melter (CCIM). Scanning electron microscopy (SEM) and X-ray diffraction (XRD) showed that all of the IPG waste forms did not contain sulfate salt segregation or crystalline phases. The calculated composition and the average analytical composition obtained by Electron Probe Microanalysis (EPMA) were in good agreement. The major elements were uniformly distributed throughout the samples. The chemical durability of the IPG waste forms containing 40wt% SBW was evaluated by the product consistency test (PCT) and met current DOE requirements. IPG waste forms were melted at a relatively low temperature and for short times compared to borosilicate glasses. These advantages, combined with those of a significantly higher waste loading and the feasibility for CCIM melting, offer a considerable savings in time, energy, and cost for vitrifying this high sodium and sulfate waste.

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.

PARTITIONING OF MERCURY FROM ACTINIDES IN THE TRUEX PROCESS

A mercury complexant, L-cysteine hydrochloride, was tested for use in separating Hg(II) from actinides during transuranic extraction (TRUEX) processing of wastes at the Idaho National Engineering and Environmental Laboratory (INEEL). Mercury, americium, plutonium, and uranyl distributions for the TRUEX solvent were characterized over a nitric acid concentration range of 0.01 to 2 M with and without cysteine. The applicability of cysteine was evaluated for selective Hg(II) complexation in an INEEL sodium-bearing waste simulant. A test was also conducted to evaluate the applicability of cysteine to separate Hg(II) from Sr in the strontium extraction (SREX) process with Sr Resin used as a stand-in for the SREX process solvent. In all cases, the use of L-cysteine HCl retained Hg in the aqueous phase while causing no or little perturbation in the actinide and Sr distribution behavior.

Mercury Extraction By The TRUEX Process Solvent: III. Extractable Species And Stoichiometry†

ABSTRACT Mercury extraction from acidic aqueous solutions by the TRUEX process solvent, 0.2 M n-octyl(phenyl)N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), 1.4 M tributylphosphate (TBP) in n-dodecane, has not been extensively examined, Research is currently in progress at the Idaho Chemical Processing Plant to evaluate the TRUEX process for actinide removal from several acidic waste streams, including liquid sodium-bearing waste (SBW), which contains significant quantities of mercury. Reactions for two mercury species, HgCl2 and HgCl4 −2, are reported. Classical slope analysis techniques were utilized to evaluate the stoichiometric coefficients of each Hg species independently for both CMPO and TBP. The slope analysis results indicate that even the HgCl4 −2 species extracts as HgCl2. The results also indicate that 1.9 to 2.2 moles of CMPO and 1.5 to 1.81 moles of TBP are required per mole of HgCl2 extracted. A generic equation for mercury extraction into CMPO or TBP has been shown to be: where x = 2 or 4, y = 0 or −2, E = CMPO or TBP, a = the experimentally determined CMPO or TBP stoichiometry, and over lines indicate organic species. Equilibrium constants for HgCl2 species were determined to be 295 when extracted by CMPO and 18.2 when extracted by TBP. Equilibrium constants for the HgCl4 −2 species were found to range between 5.8 and 12.0 for the CMPO reaction and 0.13 to 0.27 with TBP.

Development and Testing of Srex Flowsheets for Treatment of Idaho Chemical Processing Plant Sodium-Bearing Waste Using Centrifugal Contactors

Laboratory experimentation has indicated that the SREX process is effective for partitioning 90Sr from acidic radioactive waste solutions located at the Idaho Chemical Processing Plant. A baseline flowsheet has been proposed for the treatment of sodiumbearing waste (SBW) which includes extraction of strontium from liquid SBW into the SREX solvent (0.15 M 4′,4′ (5′)-di-(tert-butyldicyclohexo)-18-crown-6 and 1.2 M TBP in Isopar L®), a 0.01 M nitric acid strip section to back-extract components from the loaded solvent, and a 2.0 M HNO3 solvent acidification section to equilibrate the solvent with HNO3 prior to recycle to the extraction section. The flowsheet was designed to provide a decontamination factor (DF) of >103 which will reduce the 90Sr activity in the waste solution to below the NRC Class A LLW limit of 0.04 Ci 90Sr/m3. SREX flowsheet testing was performed using sixteen stages of 5.5-cm diameter centrifugal contactors. The behavior of stable Sr and other components which are potentially extracted by the SREX solvent were evaluated. Specifically, the behavior of the matrix components including Pb, K, Hg, Na, Ca, Zr, and Fe was studied. The described flowsheet achieved 99.98% Sr removal (DF=4250) with one cycle of SREX. Potassium and Zr were partially extracted into the SREX solvent with 35% and 21%, respectively, exiting in the strip product. Sodium, Ca, and Fe were essentially inextractable. Lead was determined to extract and accumulate in the SREX solvent and in the strip section. As a result, a Pb precipitate formed in the strip stages of the contactors. Mercury was also determined to extract and accumulate in the SREX solvent.

Immobilization of Sodium‐Bearing High‐Level Radioactive Waste in Synroc Containing (Na0.5Nd0.5)TiO3‐Type Perovskite

This study on the immobilization of high‐sodium‐bearing HLW in synroc indicates that (Na0.5Nd0.5)TiO3‐type perovskite can be used to incorporate a high content of sodium in synroc. Synroc samples containing 13.0 wt% waste oxide and 5.7 wt% Na2O show very well chemical durability and physical properties. The standard Synroc‐C formulation can incorporate only 2 wt% Na2O, so this study greatly improved the immobilization ability of sodium in Synroc‐related material.

Cesium Absorption from Acidic Solutions Using Ammonium Molybdophosphate on a Polyacrylonitrile Support (AMP-PAN)

Recent effort at the Idaho Chemical Processing Plant (ICPP) have included evaluation of cesium removal technologies as applied to ICPP acidic radioactive waste streams. Ammonium molybdophosphate (AMP) immobilized on a polyacrylonitrile support (AMP-PAN) has been studied as an ion exchange agent for cesium removal from acidic waste solutions. Capacities, distribution coefficients, elutability, and kinetics of cesium extraction have been evaluated. Exchange breakthrough curves using small columns have been determined from 1M HNO3 and simulated waste solutions. The theoretical capacity of AMP is 213 g Cs/Kg AMP. The average experimental capacity in batch contacts with various acidic solutions was 150 g Cs/kg AMP. The measured cesium distribution coefficients from actual waste solutions were 3287 mL/g for dissolved zirconia calcines, and 2679 mL/g for sodium-bearing waste. The cesium in the dissolved alumina calcines was analyzed for; however, the concentration was below analytical detectable limits resulting in inconclusive results. The reaction kinetics are very rapid (2-10 minutes). Cesium absorption appears to be independent of acid concentration over the range tested (0.1 M to 5 M HNO3).