High-level Radioactive Liquid Waste Research Articles

Removal of ruthenium from high-level radioactive liquid waste generated during reprocessing of spent fuel

Radio-ruthenium (Ru) due to its existence in the form of complexes with varied oxidation state, larger fission yield and relatively long half life is an extremely troublesome nuclide during reprocessing and subsequent waste management. In the process of the concentration of high-level waste (HLW), containing many nitrates of fission products and nitric acid, Ru is oxidised to volatile tetroxide RuO4, which is reduced to its dioxide (RuO2) at the inner surface of the equipment and is deposited there. As a result, the radiation dose of the plant equipments keeps increasing. A process was developed for the separation of Ru from HLW stream by volatilisation using KMnO4 (potassium permanganate) and O3 (ozone) as oxidising agent and its subsequent trapping on adsorbent material polyether ether ketone pellets. Various parameters like acidity, Ru concentration, temperature, time period of reaction and type of adsorbent were studied. The sorption behaviour was examined with various isotherms like Langmuir, Freundlich and Dubinin–Raduskevich isotherms. Thermodynamics parameters were also evaluated. The results indicated that maximum volatilisation of Ru occurred in the case of KMnO4 (98%) as compared to O3 (53%) at low acid concentration (2 M).

Development of Liquid-Air-Interface Corrosion of Steel in Nitrate Solutions

Liquid-air interface (LAI) corrosion is an increasing concern in the assessment of the deterioration of the integrity of underground steel storage tanks used for high-level liquid radioactive wastes. This work is aimed at expanding the current knowledge of this form of corrosion by probing its nature to provide insights for further mechanistic studies. LAI corrosion initiated in the form of random pits at the very top of the LAI meniscus under potentiostatic polarization, which is not where attack occurs as the result of the classical oxygen concentration cell mechanism. LAI corrosion pits grew upward to form hemispherical solution drops above the meniscus and gas bubbles collected on corrosion products in the meniscus. The effects of deaeration, the presence of a crevice former, and pre-passivation of the steel on LAI corrosion were also studied. LAI corrosion occurred in nitrite solutions containing sulfate or chloride instead of nitrate. No correlation in the passive current density or pitting potential with LAI corrosion susceptibility was found in solutions with varying nitrate and nitrite concentrations. Electrochemical measurements with a carbon steel microelectrode did not indicate that an aggressive meniscus solution develops to initiate LAI corrosion.

Formulation of Specialty Glass Frit and Its Use for Decontamination of Joule Melter Employed for Vitrification of High Level Radioactive Liquid Waste

Advanced vitrification system at TWMP Tarapur was used for successful vitrification of large volume of high level radioactive liquid waste (HLW) stored in PREFRE waste tank farm. After completion of the operational life of the Joule melter, dismantling was planned. Prior to dismantling, the holdup inventory of active glass product from the melter was flushed out using specially formulated inactive glass frit to reduce the air activity buildup in the cell during dismantling operations. The properties of the specialty glass frit prepared were comparable with those of the regular product glass. More than 94% of holdup activity was flushed out from the Joule melter prior to dismantling of the melter.

Infrared and X-ray photoelectron spectroscopy studies on sodium borosilicate glass interacted with thermally oxidized aluminides formed on Alloy 690

Thermally oxidized aluminides formed on Ni–Cr–Fe based Alloy 690 substrates were subjected to interaction with sodium borosilicate melt (used as matrices for immobilization of high-level radioactive liquid waste) at 1248K for 192h. The glass samples were analyzed by Fourier transform infrared spectroscopy. The observed bands in mid-infrared spectra of reacted glass correspond to the vibrational bands of specific groups, and those in far-infrared region revealed the distinct spectral features of stretching and bending AlO vibrations. In X-ray photoelectron spectra of O1s region of reacted glass, non-bridging oxygen component disappeared due to incorporation of aluminum in the glass structure. Chemical shift towards higher energy side in B1s photoelectron spectra of reacted glass indicated the modified glass structure in consequence of inclusion of Al+3 in glass network.

Extraction of Europium(III) and Cobalt(II) by <i>N</i>,<i>N</i>,<i>N</i>’,<i>N</i>’-Tetraoctyldiglycolamide and <i>N</i>,<i>N</i>,<i>N</i>’,<i>N</i>’-Tetrahexyldiglycolamide from Aqueous Acid Solutions

One of the main concerns related to the disposal of high– level radioactive liquid waste (HLLW) generated after reprocessing of spent nuclear fuel is related to the presence within this waste of long-lived radionuclides (LLRNs). The LLRNs in the spent nuclear fuels are mostly the so-called minor actinides (neptunium (Np), americium (Am) and curium (Cm)) and certain fission products. Transmutation of LLRNs to short-lived or stable isotopes represents a viable nuclear waste management strategy to drastically reduce the time and space requirements for a geological repository of the remaining waste. To be able to destroy LLRNs in a transmutation process they must be separated from the rest of the HLLW. Therefore, the concept of partitioning and transmutation (P&T) has been proposed for treatment of HLLW. 1-3 The P&T concept imposes highly-set demands on recovery yields and decontamination factors of the separated species. In this context, there has been considerable interest in amide extractants for the extraction of

Preparation and characterisation of glass product with modified composition for vitrification of high level radioactive waste

During the vitrification of high level radioactive liquid waste (HLW) at advanced vitrification system (AVS) in Tarapur, the glass forming chemicals are introduced into the melter in the form of preformed glass frit, known as base glass. A five component system, SiO2–B2O3–Na2O–Fe2O3–TiO2 is presently used as the base glass at AVS. Towards the step to product improvement, a modified base glass composition with seven components, SiO2–B2O3–Na2O–CaO–Fe2O3–TiO2–K2O was formulated with reduced boron content and incorporation of mixed alkali elements Na and K as compared to presently adopted base glass composition. The modification was mainly aimed at reducing boron volatilization during vitrification without affecting the present state of product quality to improve melter life and to some extent take advantage of mixed alkali effect. The modified base glass and product glass based on modified base composition were evaluated for different properties with respect to existing composition. The fusion and pouring temperatures were well with in the range of plant requirement. The glass transition temperature, coefficient of thermal expansion and thermal conductivity were similar or better. The product glass was having good homogeneity and chemical durability.

Development of ICP-AES Method for the Determination of Common Metallic and Rare Earth Elements in High-Level Radioactive Liquid Waste Solutions

Development of ICP-AES Method for the Determination of Common Metallic and Rare Earth Elements in High-Level Radioactive Liquid Waste Solutions

The Application of Nitrogen Laser on Extraction of Uranium in The Long Life of High Level Radioactive Liquid Waste Using TBP-Kerosene Solvent

The radionuclide of uranium (U) is major radionuclide contained in the long life of high level radioactive liquid waste (HLLW) generated from reprocessing of spent nuclear fuel. The radioactive waste have to be treated to ready for long term disposal. Separation of the U at high efficiencies greatly lowers the volume of the long life alpha radioactive waste to be disposed and decreases the hazard level of the waste. The technology assessment of selective separation of U was carried-out as alternative and strategy for HLLW management in the future. The selective separation technology of U from fission products at very high efficiencies was developed by the extraction process using TBP-kerosene solvent and increasing the separation by exposure of nitrogen (N 2 ) laser radiation at wavelength 337.1 nm. In the extraction process for simulation waste containing U and Zr in 5 M HNO 3 (Zr as one of fission product which difficult to be separated from U) using 30 % TBP-kerosene solvent and by exposure of nitrogen laser radiation shows that increasing of distribution coefficient of U (Kd U) can obtains 135 % and increasing of separation factor of U and Zr (SF (U/Zr) ) is 189 %. The increasing of Kd U using N 2 laser higher than using CO 2 laser (at the wave-number 944 cm -1 ) which increase of Kd U only 100 %. In Indonesia, assessment for adaptation of the separation technology by extraction for separation of U process using 30 % TBP-kerosene solvent should be carried-out as alternative for treatment the HLLW generated from 99 Mo radioisotope production and from post irradiation examination of nuclear fuel. doi: http://dx.doi.org/10.12777/ijwr.2.2.2012.6-12 [ How to cite this article : Gunandjar, G. (2012). The Application of Nitrogen Laser on Extraction of Uranium in The Long Life of High Level Radioactive Liquid Waste Using TBP-Kerosene Solvent. International Journal of Waste Resources (IJWR) , 2 (2), 6-12. doi: http://dx.doi.org/10.12777/ijwr.2.2.2012.6-12 ] Best Website to Watch Movies Online

Challenges to develop single-column MA(III) separation from HLLW using R-BTP type adsorbents

In order to directly separate trivalent minor actinides (MA: Am, Cm) from fission products (FP) containing rare earths (RE) in high level radioactive liquid waste (HLLW), the authors have challenged to develop a simplified MA separation process by extraction chromatography using a single column. Attention has been paid to a new type of nitrogen-donor ligands, R-BTP (2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl) pyridine, R: alkyl group) as an extractant because it shows high extraction selectivity for Am(III) over RE(III). It is known that the R-BTP ligands show different properties such as adsorbability and stability by having different alkyl groups. Therefore, some novel adsorbents were prepared by impregnating different types of R-BTP ligands (isohexyl-, isoheptyl- and cyheptyl-BTP) and a similar ligand to the R-BTP, ATP (2,6-bis(1-aryl-1H-tetrazol-5-yl)pyridines), into the porous silica/polymer support (SiO2-P particles). This work deals with comparison in adsorption and desorption properties of Am and some FP in HNO3 solution onto such R-BTP type adsorbents, as well as chemical and radiolytic stability of the adsorbents. Then the possibility of a single-column separation of MA from main FP was pursued by evaluating the results of column experiments using the most promising adsorbent (isohexyl-BTP/SiO2-P) under temperature control. In addition, elution behaviors of U and Pd were also estimated.

Numerical modeling on the discharged fluid flow from a glass melter by a Lagrangian approach

Abstract In Japan, it is planned to use vitrification to immobilize high level radioactive liquid waste (HLLW) obtained from the nuclear reprocessing process. In the vitrification process, the HLLW will be mixed with molten glass in a joule-heated glass melter. The molten glass is then poured into a canister through a nozzle at the bottom of the melter. When some abnormalities occurred, feature of the discharged flow become different from that in normal conditions. We cannot observe the state of the molten glass inside the melter directly because of the severe environment. Accordingly, we cannot investigate the cause when the abnormalities occurred. Hence, numerical simulation of the discharged flow becomes important to understand and recover from operational problems or abnormalities with the glass melter. However, existing numerical models of the glass melter do not really simulate the discharged glass flow. In the present study, we introduce a new algorithm into an existing Lagrangian approach to efficiently simulate the discharged glass flow. To verify the model, 3D simulations were performed on a real scale system and got compared with the results of experimental validation tests. The simulation results show good agreement with the experimental results. Consequently, the new Lagrangian approach accurately simulates the molten glass flow.

Radiolytic Stability of N,N,N′,N′-Tetraoctyl Diglycolamide (TODGA) in the Presence of Phase Modifiers Dissolved in n-Dodecane

The radiolytic stability of a promising extractant for actinide partitioning from high-level radioactive liquid waste, namely N,N,N',N'-tetraoctyl diglycolamide (TODGA) was investigated in the presence of several phase modifiers, viz. N,N-dihexyloctanamide (DHOA), tri-n-butyl phosphate (TBP), 1-decanol, and iso-decanol dissolved in n-dodecane. The distribution ratio of Am(III) decreased with increased radiation dose studied up to 1000 kGy. Nevertheless, all the compositions of extractants showed fairly high extraction of Am(III) up to 500 kGy (DAm: ≥ 50), beyond which significant decrease was observed. However, the DAm values were sufficiently high for process applications for the chosen compositions even at an absorbed dose of 1000 kGy. The stripping behavior of Am(III) with 0.2 M HNO3 was found to be favorable with increased absorbed dose by the solvent up to 1000 kGy. With an increased absorbed dose, the loading of Nd(III) in the organic phase decreased due to depletion of ligand/extractant concentration (TODGA) in the organic phase. There was marginal variation in the hydrodynamic parameters such as density, viscosity, and interfacial tension (IFT) of the irradiated solvents vis-a-vis fresh/unirradiated solvent.

Development of a simplified separation process of trivalent minor actinides from fission products using novel R-BTP/SiO2-P adsorbents

From a viewpoint of direct separation of trivalent minor actinides (MA: Am, Cm etc.) from fission products (FP) including rare earths (RE) in high level radioactive liquid waste, the authors have developed a simplified separation process using a single column packed with novel extraction adsorbents. Attention was paid to a new type of nitrogen-donor ligand, R-BTP (2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl)pyridine, R: alkyl group) as an extractant because it has higher extraction selectivity for Am(III) than RE(III). Since the R-BTP ligands show different properties such as adsorbability and stability when they have different alkyl groups, several R-BTP extraction adsorbents were prepared by impregnating the R-BTP ligands with different alkyl groups (isohexyl-, isoheptyl- and cyheptyl-BTP) into a porous silica/polymer composite support (SiO2-P particles). This work investigated: (1) fundamental properties of the synthesized R-BTP/SiO2-P adsorbents, (2) adsorption and desorption properties of Am and FP in nitric acid solution and water using the adsorbents in a batch experiment, (3) radiolytic and chemical stabilities of the adsorbents, and (4) the possibility for developing a simplified separation process of MA using the most promising adsorbent (isohexyl-BTP/SiO2-P) under temperature control between 25 and 50°C.

Evaluation of Alternative Filter Media for the Rotary Microfilter

The Savannah River Site is currently developing and testing several processes to treat high level radioactive liquid waste. Each of these processes has a solid-liquid separation process that limits its throughput. Savannah River National Laboratory researchers identified and tested the rotary microfilter as a technology to increase solid-liquid separation throughput. The authors believe the rotary microfilter throughput can be improved by using a better filter membrane. Previous testing showed that asymmetric filters composed of a ceramic membrane on top of a stainless steel support produced higher filter flux than 100% stainless steel symmetric filters in crossflow filter tests. Savannah River National Laboratory and Oak Ridge National Laboratory are working together to develop asymmetric ceramic–stainless steel composite filters and asymmetric 100% stainless steel filters to improve the throughput of the rotary microfilter.

Multiplier Effect on Separation of Am and Cm with Hydrophilic and Lipophilic Diamides

Mutual separation of trivalent lanthanides (Ln) and actinides (An) is a challenging study in order to dispose effectively these metals in high-level radioactive liquid waste. Both extractant and masking agent are used simultaneously in an extraction system. The strong extractants, diglycolamide (DGA) and dioxaoctane-diamide (DOODA), and several kinds of the masking agents, were employed. The four novel-synthesized aminopolyacetamide compounds, namely, nitrilotriacetamide (NTAamide), hydroxypropane-triacetamide (Citricamide), ethylenediamine-tetracetamide (EDTAamide), and diethylenetriamine-pentacetamide (DTPAamide), were also examined. It can be seen that these compounds has more affirmative to heavy Ln than light ones and mutual separation factors by addition of these compounds are slightly high.

Treatment of low level radioactive liquid waste containing appreciable concentration of TBP degraded products

The acidic and alkaline low level radioactive liquid waste (LLW) generated during the concentration of high level radioactive liquid waste (HLW) prior to vitrification and ion exchange treatment of intermediate level radioactive liquid waste (ILW), respectively are decontaminated by chemical co-precipitation before discharge to the environment. LLW stream generated from the ion exchange treatment of ILW contained high concentrations of carbonates, tributyl phosphate (TBP) degraded products and problematic radio nuclides like 106Ru and 99Tc. Presence of TBP degraded products was interfering with the co-precipitation process. In view of this a modified chemical treatment scheme was formulated for the treatment of this waste stream. By mixing the acidic LLW and alkaline LLW, the carbonates in the alkaline LLW were destroyed and the TBP degraded products got separated as a layer at the top of the vessel. By making use of the modified co-precipitation process the effluent stream (1–2 μCi/L) became dischargeable to the environment after appropriate dilution. Based on the lab scale studies about 250 m 3 of LLW was treated in the plant. The higher activity of the TBP degraded products separated was due to short lived 90Y isotope. The cement waste product prepared using the TBP degraded product was having good chemical durability and compressive strength.

Effective separation of palladium from simulated high level radioactive waste

Aiming the selective recovery of palladium from high level radioactive liquid waste (HLW), a chelating thiamide type sorbent, CWP–TU, was prepared by the modification of Japanese cedar wood powder (CWP). Convection oven and microwave heating were separately used for modification purpose and found that microwave heating is more effective over oven heating. CWP–TU was extensively studied for the adsorption of Pd(II) from nitric acid medium. The batch test showed that nitric acid concentration of 3 M or higher is favorable for Pd(II) loading. Consistent adsorption of Pd(II) under gamma irradiation condition demonstrated the feasibility of using CWP–TU in real HLW. Also, Pd(II) only adsorption from simulated HLW solution verified the palladium only selectivity of the sorbent as well as the lack of influence of coexisting metal ions on its affinity toward Pd(II). CWP–TU holds maximum Pd(II) loading capacities of 0.98 mol/kg at 30 °C and 1.04 mol/kg under gamma irradiation. A comparative study using some ion exchange resins revealed that the resins are either ineffective in nitrate medium or lack stability under irradiation.

Transuranic waste remediation in Idaho: Pit 9, failed contracts, and lawsuits

AbstractThe U.S. Department of Energy (US DOE) remediation responsibilities include its Idaho National Laboratory. In 1989, the U.S. Environmental Protection Agency placed the Idaho site on its National Priority List for environmental cleanup. The site's contamination legacy from operations included inactive reactors and other structures, spent nuclear fuel, high‐level liquid radioactive wastes, calcined radioactive wastes, and transuranic wastes. Documents governing cleanup include a 1995 Settlement Agreement between the US DOE and the US Navy as responsible parties, and the State of Idaho. The Subsurface Disposal Area contains buried transuranic wastes, lies above the East Snake River Plain Aquifer, and could be the “site's most nettlesome cleanup issue,” according to an outside observer. This article describes the technical and legal difficulties that have been encountered in remediating this area. © 2010 Wiley Periodicals, Inc.

Isomorphism of actinides and REE in synthetic ferrite garnets

The reprocessing of spent nuclear fuel (SNF) is accompanied by the formation of liquid high-level radioactive waste (HLW). To increase the safety of handling HLW, it is proposed to extract actinide isotopes (An) and REE from them. These elements may be incorporated into crystalline matrices, e.g., based on ferrites with garnet structure, and then disposed in a geologic repository. The actinide-REE fraction is characterized by a complex composition. In addition to major components (An and REE), Al, Si, Na, and Sn occur therein in small amounts (a few wt %). Possible incorporation of the admixtures into ferrite garnets, as well as their effect on the phase composition of matrices and Th, Ce, Gd, and La contents were studied. It was shown that admixtures enter into garnet by means of isomorphic replacement. The properties of samples change only when admixtures are added in amounts exceeding their concentrations in HLW. The ability of ferrite garnets to accumulate significant amounts of An, REE, and admixture elements makes them suitable for use as matrices in immobilizing actinide-REE HLW of complex composition.

Adsorption of Molybdenum Ion in Nitric Acid Solution by Using a Pb-Fe Based Adsorbent

In order to remove molybdenum (Mo) from the high-level radioactive liquid waste (HLLW), Pb-Fe based adsorbents were synthesized and their adsorption capacities were investigated. The adsorbent precursors were prepared with precipitation at various pH values by adding sodium hydroxide into a mixture of 1 mol/L lead nitrate and 1 mol/L ferric nitrate aqueous solutions, and each precipitate was then filtered, dried and ground to obtain a fine powder precursor. The precursors were also calcinated at various temperatures to obtain the Pb-Fe based adsorbents.Two main parameters were considered when synthesizing the adsorbents, i.e. the pH value of precipitation of adsorbent precursor as well as the calcination temperature of the Pb-Fe based adsorbent. The experimental results indicated that the calcination temperature has a great influence on adsorption capacity of the Pb-Fe based adsorbent for Mo. It was found that when the precursor was precipitated at pH 9 and then calcinated at 500 °C, the sorbed amount of Mo by the adsorbent was the highest. In addition, the equilibrium sorption isotherm of Mo onto the adsorbent was estimated by the Langmuir, Freundlich and Redlich-Peterson sorption models. We found that the overall adsorption process was described well by Redlich-Peterson sorption isotherm equation.

Sulfur incorporation in high level nuclear waste glass: A S K-edge XAFS investigation

We perform X-ray absorption fine structure (XAFS) spectroscopy measurements at the sulfur K-edge to elucidate the electronic and geometric bonding of sulfur atoms in borosilicate glass used for the vitrification of high level radioactive liquid waste. The sulfur is incorporated as sulfate, most probably as sodium sulfate, which can be deduced from the X-ray absorption near edge structure (XANES) by fingerprint comparison with reference compounds. This finding is backed up by Raman spectroscopy investigation. In the extended XAFS data, no second shell beyond the first oxygen layer is visible. We argue that this is due to the sulfate being present as small clusters located into voids of the borosilicate network. Hence, destructive interference of the variable surrounding prohibits the presence of higher shell signals. The knowledge of the sulfur bonding characteristics is essential for further optimization of the glass composition and to balance the requirements of the process and glass quality parameters, viscosity and electrical resistivity on one side, waste loading and sulfur uptake on the other side.