WorldWideScience

Sample records for salt solution waste

  1. Solubility and speciation of actinides in salt solutions and migration experiments of intermediate level waste in salt formations

    International Nuclear Information System (INIS)

    1986-01-01

    A comprehensive study into the solubility of the actinides americium and plutonium in concentrated salt solutions, the release of radionuclides from various forms of conditioned ILW and the migration behaviour of these nuclides through geological material specific to the Gorleben site in Lower Saxony is described. A detailed investigation into the characterization of four highly concentrated salt solutions in terms of their pH, Eh, inorganic carbon contents and their densities is given and a series of experiments investigating the solubility of standard americium(III) and plutonium(IV) hydroxides in these solutions is described. Transuranic mobility studies for solutions derived from the standard hydroxides through salt and sand have shown the presence of at least two types of species present of widely differing mobility; one migrating with approximately the same velocity as the solvent front and the other strongly retarded. Actinide mobility data are presented and discussed for leachates derived from the simulated ILW in cement and data are also presented for the migration of the fission products in leachates derived from real waste solidified in cement and bitumen. Relatively high plutonium mobilities were observed in the case of the former and in the case of the real waste leachates, cesium was found to be the least retarded. The sorption of ruthenium was found to be largely associated with the insoluble residues of the natural rock salt rather than the halite itself. (orig./RB)

  2. Radioactive waste and special waste disposal in salt domes - phoney waste management solutions

    International Nuclear Information System (INIS)

    Grimmel, E.

    1990-01-01

    The paper tries to make aware of the fact that an indefinite safe disposal of anthropogeneous wastes in underground repositories is impossible. Suspicion is raised that the Gorleben-Rambow salt dome has never been studied for its suitability as a repository, but that it was simply taken for granted. Safety analyses are meant only to conceal uncertainty. It is demanded to immediately opt out of the ultimate disposal technique for radioactive and special wastes in salt caverns. (DG) [de

  3. Corrosion of carbon steel in saturated high-level waste salt solutions

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Parish, W.R.

    1997-01-01

    High level waste stored as crystallized salts is to be removed from carbon steel tanks by water dissolution. Dissolution of the saltcake must be performed in a manner which will not impact the integrity of the tank. Corrosion testing was performed to determine the amount of corrosion inhibitor that must be added to the dissolution water in order to ensure that the salt solution formed would not induce corrosion degradation of the tank materials. The corrosion testing performed included controlled potential slow strain rate, coupon immersion, and potentiodynamic polarization tests. These tests were utilized to investigate the susceptibility of the cooling coil material to stress corrosion cracking in the anticipated environments. No evidence of SCC was observed in any of the tests. Based on these results, the recommended corrosion requirements were that the temperature of the salt solution be less than 50 degrees C and that the minimum hydroxide concentration be 0.4 molar. It was also recommended that the hydroxide concentration not stay below 0.4 molar for longer than 45 days

  4. Laboratory simulation of salt dissolution during waste removal

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Parish, W.R.

    1997-01-01

    Laboratory experiments were performed to support the field demonstration of improved techniques for salt dissolution in waste tanks at the Savannah River Site. The tests were designed to investigate three density driven techniques for salt dissolution: (1) Drain-Add-Sit-Remove, (2) Modified Density Gradient, and (3) Continuous Salt Mining. Salt dissolution was observed to be a very rapid process as salt solutions with densities between 1.38-1.4 were frequently removed. Slower addition and removal rates and locating the outlet line at deeper levels below the top of the saltcake provided the best contact between the dissolution water and the saltcake. It was observed that dissolution with 1 M sodium hydroxide solution resulted in salt solutions that were within the current inhibitor requirements for the prevention of stress corrosion cracking. This result was independent of the density driven technique. However, if inhibited water (0.01 M sodium hydroxide and 0.011 M sodium nitrite) was utilized, the salt solutions were frequently outside the inhibitor requirements. Corrosion testing at conditions similar to the environments expected during waste removal was recommended

  5. Criticality considerations for salt-cake disolution in DOE waste tanks

    International Nuclear Information System (INIS)

    Trumble, E.F.; Niemer, K.A.

    1995-01-01

    A large amount of high-level waste is being stored in the form of salt cake at the Savannah River site (SRS) in large (1.3 x 106 gal) underground tanks awaiting startup of the Defense Waste Processing Facility (DWPF). This salt cake will be dissolved with water, and the solution will be fed to DWPF for immobilization in borosilicate glass. Some of the waste that was transferred to the tanks contained enriched uranium and plutonium from chemical reprocessing streams. As water is added to these tanks to dissolve the salt cake, the insoluble portion of this fissile material will be left behind in the tank as the salt solution is pumped out. Because the salt acts as a diluent to the fissile material, the process of repeated water addition, salt dissolution, and salt solution removal will act as a concentrating mechanism for the undissolved fissile material that will remain in the tank. It is estimated that tank 41 H at SRS contains 20 to 120 kg of enriched uranium, varying from 10 to 70% 235 U, distributed nonuniformly throughout the tank. This paper discusses the criticality concerns associated with the dissolution of salt cake in this tank. These concerns are also applicable to other salt cake waste tanks that contain significant quantities of enriched uranium and/or plutonium

  6. Saltstone: cement-based waste form for disposal of Savannah River Plant low-level radioactive salt waste

    International Nuclear Information System (INIS)

    Langton, C.A.

    1984-01-01

    Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO 3 , NaOH, Na 2 SO 4 , and NaNO 2 . After decontamination, the salt solution is classified as low-level waste. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses. The disposal system for the DWPF salt waste includes reconstitution of the crystallized salt as a solution containing 32 wt % solids. This solution will be decontaminated to remove 137 Cs and 90 Sr and then stabilized in a cement-based waste form. Laboratory and field tests indicate that this stabilization process greatly reduces the mobility of all of the waste constitutents in the surface and near-surface environment. Engineered trenches for subsurface burial of the saltstone have been designed to ensure compatibility between the waste form and the environment. The total disposal sytem, saltstone-trench-surrounding soil, has been designed to contain radionuclides, Cr, and Hg by both physical encapsulation and chemical fixation mechanisms. Physical encapsulation of the salts is the mechanism employed for controlling N and OH releases. In this way, final disposal of the SRP low-level waste can be achieved and the quality of the groundwater at the perimeter of the disposal site meets EPA drinking water standards

  7. Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane; Eapen, Jacob

    2013-10-01

    Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt

  8. Waste salt recovery, recycle, and destruction

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1992-12-01

    Starting in 1943 and continuing into the 1970s, radioactive wastes resulting from plutonium processing at Hanford were stored underground in 149 single shell tanks. Of these tanks, 66 are known or believedto be leaking, and over a period are believed to have leaked about 750,000 gal into the surrounding soil. The bulk of the aqueous solution has been removed and transferred to double shell tanks, none of which are leaking. The waste consists of 37 million gallons of salt cake and sludge. Most of the salt cake is sodium nitrate and other sodium salts. A substantial fraction of the sludge is sodium nitrate. Small amounts of the radionuclides are present in the sludge as oxides or hydroxides. In addition, some of the tanks contain organic compounds and ferrocyanide complexes, many of which have undergone radiolytic induced chemical changes during the years of storage. As part of the Hanford site remediation effort, the tank wastes must be removed, treated, and the residuals must be immobilized and disposed of in an environmentally acceptable manner. Removal methods of the waste from the tanks fall generally into three approaches: dry removal, slurry removal, and solution removed. The latter two methods are likely to result in some additional leakage to the surrounding soil, but that may be acceptable if the tank can be emptied and remediated before the leaked material permeates deeply into the soil. This effort includes three parts: salt splitting, acid separation, and destruction, with initial emphasis on salt splitting

  9. SOLUTION MINING IN SALT DOMES OF THE GULF COAST EMBAYMENT

    Energy Technology Data Exchange (ETDEWEB)

    Griswold, G. B.

    1981-02-01

    Following a description of salt resources in the salt domes of the gulf coast embayment, mining, particularly solution mining, is described. A scenario is constructed which could lead to release of radioactive waste stored in a salt dome via inadvertent solution mining and the consequences of this scenario are analyzed.

  10. Ion exchange removal of technetium from salt solutions

    International Nuclear Information System (INIS)

    Walker, D.D.

    1983-01-01

    Ion exchange methods for removing technetium from waste salt solutions have been investigated by the Savannah River Laboratory (SRL). These experiments have shown: Commercially available anion exchange resins show high selectivity and capacity for technetium. In column runs, 150 column volumes of salt solution were passed through an ion exchange column before 50% 99 Tc breakthrough was reached. The technetium can be eluted from the resin with nitric acid. Reducing resins (containing borohydride) work well in simple hydroxide solutions, but not in simulated salt solutions. A mercarbide resin showed a very high selectivity for Tc, but did not work well in column operation

  11. Molten salt hazardous waste disposal process utilizing gas/liquid contact for salt recovery

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKenzie, D.E.

    1984-01-01

    The products of a molten salt combustion of hazardous wastes are converted into a cooled gas, which can be filtered to remove hazardous particulate material, and a dry flowable mixture of salts, which can be recycled for use in the molten salt combustion, by means of gas/liquid contact between the gaseous products of combustion of the hazardous waste and a solution produced by quenching the spent melt from such molten salt combustion. The process results in maximizing the proportion of useful materials recovered from the molten salt combustion and minimizing the volume of material which must be discarded. In a preferred embodiment a spray dryer treatment is used to achieve the desired gas/liquid contact

  12. Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jongkwon [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Um, Wooyong, E-mail: wooyong.um@pnnl.gov [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Choung, Sungwook [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of)

    2014-09-15

    The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl–KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl–KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl–KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl–KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

  13. Expected environment for waste packages in a salt repository

    International Nuclear Information System (INIS)

    Pederson, L.R.; Clark, D.E.; Hodges, F.N.; McVay, G.L.; Rai, D.

    1983-01-01

    This paper discusses results of recent efforts to define the very near-field (within approximately 2 m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and standard brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported. 38 references, 3 figures, 2 tables

  14. Treatment for hydrazine-containing waste water solution

    Science.gov (United States)

    Yade, N.

    1986-01-01

    The treatment for waste solutions containing hydrazine is presented. The invention attempts oxidation and decomposition of hydrazine in waste water in a simple and effective processing. The method adds activated charcoal to waste solutions containing hydrazine while maintaining a pH value higher than 8, and adding iron salts if necessary. Then, the solution is aerated.

  15. Mixing Modeling Analysis For SRS Salt Waste Disposition

    International Nuclear Information System (INIS)

    Lee, S.

    2011-01-01

    Nuclear waste at Savannah River Site (SRS) waste tanks consists of three different types of waste forms. They are the lighter salt solutions referred to as supernate, the precipitated salts as salt cake, and heavier fine solids as sludge. The sludge is settled on the tank floor. About half of the residual waste radioactivity is contained in the sludge, which is only about 8 percentage of the total waste volume. Mixing study to be evaluated here for the Salt Disposition Integration (SDI) project focuses on supernate preparations in waste tanks prior to transfer to the Salt Waste Processing Facility (SWPF) feed tank. The methods to mix and blend the contents of the SRS blend tanks were evalutaed to ensure that the contents are properly blended before they are transferred from the blend tank such as Tank 50H to the SWPF feed tank. The work consists of two principal objectives to investigate two different pumps. One objective is to identify a suitable pumping arrangement that will adequately blend/mix two miscible liquids to obtain a uniform composition in the tank with a minimum level of sludge solid particulate in suspension. The other is to estimate the elevation in the tank at which the transfer pump inlet should be located where the solid concentration of the entrained fluid remains below the acceptance criterion (0.09 wt% or 1200 mg/liter) during transfer operation to the SWPF. Tank 50H is a Waste Tank that will be used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work described here consists of two modeling areas. They are the mixing modeling analysis during miscible liquid blending operation, and the flow pattern analysis during transfer operation of the blended liquid. The modeling results will provide quantitative design and operation information during the mixing/blending process and the transfer operation of the blended

  16. Long-term interactions of full-scale cemented waste simulates with salt brines

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

    2016-07-01

    Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO{sub 3} solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

  17. Long-term interactions of full-scale cemented waste simulates with salt brines

    International Nuclear Information System (INIS)

    Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

    2016-01-01

    Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO 3 solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

  18. Waste salt disposal at the Savannah River Plant

    International Nuclear Information System (INIS)

    Langton, C.A.; Oblath, S.B.; Pepper, D.W.; Wilhite, E.L.

    1986-01-01

    Waste salt solution, produced during processing of high-level nuclear waste, will be incorporated in a cement matrix for emplacement in an engineered disposal facility. Wasteform characteristics and disposal facility details will be presented along with results of a field test of wasteform contaminant release and of modeling studies to predict releases. 5 refs., 11 figs., 5 tabs

  19. Molten salt oxidation of organic hazardous waste with high salt content.

    Science.gov (United States)

    Lin, Chengqian; Chi, Yong; Jin, Yuqi; Jiang, Xuguang; Buekens, Alfons; Zhang, Qi; Chen, Jian

    2018-02-01

    Organic hazardous waste often contains some salt, owing to the widespread use of alkali salts during industrial manufacturing processes. These salts cause complications during the treatment of this type of waste. Molten salt oxidation is a flameless, robust thermal process, with inherent capability of destroying the organic constituents of wastes, while retaining the inorganic ingredients in the molten salt. In the present study, molten salt oxidation is employed for treating a typical organic hazardous waste with a high content of alkali salts. The hazardous waste derives from the production of thiotriazinone. Molten salt oxidation experiments have been conducted using a lab-scale molten salt oxidation reactor, and the emissions of CO, NO, SO 2 , HCl and dioxins are studied. Impacts are investigated from the composition of the molten salts, the types of feeding tube, the temperature of molten carbonates and the air factor. Results show that the waste can be oxidised effectively in a molten salt bath. Temperature of molten carbonates plays the most important role. With the temperature rising from 600 °C to 750 °C, the oxidation efficiency increases from 91.1% to 98.3%. Compared with the temperature, air factor has but a minor effect, as well as the composition of the molten salts and the type of feeding tube. The molten carbonates retain chlorine with an efficiency higher than 99.9% and the emissions of dioxins are below 8 pg TEQ g -1 sample. The present study shows that molten salt oxidation is a promising alternative for the disposal of organic hazardous wastes containing a high salt content.

  20. Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

    International Nuclear Information System (INIS)

    Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian

    2016-01-01

    This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na_2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl- in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease the waste

  1. Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-01

    This report describes the scientific basis underlying the approach being followed to design and develop “advanced” glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl– in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease

  2. Radiological consequences associated with human intrusion into radioactive waste repositories in salt formations

    International Nuclear Information System (INIS)

    Jacquier, P.

    1989-01-01

    The assessment of the radiological impact of human intrusion scenarios is extremely important in the case of repositories located in salt formations, since salt is obviously a valuable economic resource. Salt formations also represent a suitable medium for mining storage caverns for oil and gas. The scenario considered in this report is that of solution mining in salt formations to produce salt for human consumption. It is postulated that the salt is extracted by excavating a cavern through solution-mining and that, in the course of cavern enlargement, the waste is intercepted and drops to the bottom of the cavern. We have assumed that the intrusion takes place 500 or even 2 500 years after the repository has been sealed. The cases considered involve high-level vitrified waste or cemented alpha waste. The paper describes the assumptions on which the scenario is based and uses a simplified model to assess the radiological consequences associated with the ingestion of contaminated salt. The paper also provides details of a sensitivity/uncertainty analysis which identified several areas in which experimental studies should be either initiated or continued [fr

  3. Waste package materials testing for a salt repository: 1983 status summary report

    International Nuclear Information System (INIS)

    Moak, D.P.

    1986-09-01

    The United States plans to safely dispose of nuclear waste in deep, stable geologic formations. As part of these plans, the US Department of Energy is sponsoring research on the designing and testing of waste packages and waste package materials. This fiscal year 1983 status report summarizes recent results of waste package materials testing in a salt environment. The results from these tests will be used by waste package designers and performance assessment experts. Release characteristics data are available on two waste forms (spent fuel and waste-containing glass) that were exposed to leaching tests at various radiation levels, temperatures, pH, glass surface area to solution volume ratios, and brine solutions simulating expected salt repository conditions. Candidate materials tested for corrosion resistance and other properties include iron alloys; TI-CODE 12, the most promising titanium alloy for containment; and nickel alloys. In component interaction testing, synergistic effects have not ruled out any candidate material. 21 refs., 37 figs., 15 tabs

  4. Waste treatment using molten salt oxidation

    International Nuclear Information System (INIS)

    Navratil, J.D.; Stewart, A.E.

    1996-01-01

    MSO technology can be characterized as a submerged oxidation process; the basic concept is to introduce air and wastes into a bed of molten salt, oxidize the organic wastes in the molten salt, use the heat of oxidation to keep the salt molten and remove the salt for disposal or processing and recycling. The molten salt (usually sodium carbonate at 900-1000 C) provides four waste management functions: providing a heat transfer medium, catalyzing the oxidation reaction, preventing the formation of acid gases by forming stable salts, and efficiently capturing ash particles and radioactive materials by the combined effects of wetting, encapsulation and dissolution. The MSO process requires no wet scrubbing system for off-gas treatment. The process has been developed through bench-scale and pilot-scale testing, with successful destruction demonstration of a wide variety of hazardous and mixed (radioactive and hazardous wastes). (author). 24 refs, 2 tabs, 2 figs

  5. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    International Nuclear Information System (INIS)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-01-01

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE's mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies

  6. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    Energy Technology Data Exchange (ETDEWEB)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-07-07

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

  7. Corrosion processes of alloyed steels in salt solutions

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Institut fuer Nukleare Entsorgung

    2018-02-15

    A summary is given of the corrosion experiments with alloyed Cr-Ni steels in salt solutions performed at Research Centre Karlsruhe (today KIT), Institute for Nuclear Waste Disposal (INE) in the period between 1980 and 2004. Alloyed steels show significantly lower general corrosion in comparison to carbon steels. However, especially in salt brines the protective Cr oxide layers on the surfaces of these steels are disturbed and localized corrosion takes place. Data on general corrosion rates, and findings of pitting, crevice and stress corrosion cracking are presented.

  8. Recovery of plutonium and americium from chloride salt wastes by solvent extraction

    International Nuclear Information System (INIS)

    Reichley-Yinger, L.; Vandegrift, G.F.

    1987-01-01

    Plutonium and americium can be recovered from aqueous waste solutions containing a mixture of HCl and chloride salt wastes by the coupling of two solvent extraction systems: tributyl phosphate (TBP) in tetrachloroethylene (TCE) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in TCE. In the flowsheet developed, the salt wastes are dissolved in HCl, the Pu(III) is oxidized to the IV state with NaClO 2 and recovered in the TBP-TCE cycle, and the Am is then removed from the resultant raffinate by the CMPO-TCE cycle. The consequences of the feed solution composition and extraction behavior of these species on the process flowsheet design, the Pu-product purity, and the decontamination of the aqueous raffinate from transuranic elements are discussed. 16 refs., 6 figs

  9. Modeling of Sulfate Double-Salt in Nuclear Wastes

    International Nuclear Information System (INIS)

    Toghiani, B.; Lindner, J.S.; Weber, C.F.; Hunt, R.D.

    2000-01-01

    The Environmental Simulation Program (ESP) continues to adequately predict the solubility of most key chemical systems in the Hanford tank waste. For example, the ESP predictions were in fair agreement with the solubility experiments for the fluoride-phosphate system, although ESP probably underestimates the aqueous amounts. Due to the importance of this system in the formation of pipeline plugs, additional experiments have been made at elevated temperatures, and improvements to the ESP database will be made. ESP encountered problems with sulfate systems because the Public database for ESP does not include anhydrous sodium sulfate in mixed solutions below 32.4 C. This limitation leads to convergence problems and to spurious predictions of solubility near the transition point with sodium sulfate decahydrate when other salts such as sodium nitrate are present. However, ESP was able to make reasonable solubility predictions with a corrected database, demonstrating the need to validate and document the various databases that can be used by ESP. Even though ESP does not include the sulfate-nitrate double salt, this omission does not appear to be a major problem. The solubility predictions with and without the sulfate-nitrate double salt are comparable. In sharp contrast, the sulfate-fluoride double salt is included, but ESP still underestimates solubility in some cases. This problem can misrepresent the ionic strength of the solution, which is an important factor in the formation of pipeline plugs. Solubility tests on the sulfate-fluoride system are planned to provide additional data at higher temperatures and in caustic solutions. These results will be used to improve the range and accuracy of ESP predictions. ESP will continue to provide important predictions for waste processing operations while being evaluated and improved. For example, ESP will be used to determine the amount of water for the saltcake dissolution efforts at Hanford. When ESP underestimates the

  10. Solubility of pllutonium in alkaline salt solutions

    International Nuclear Information System (INIS)

    Hobbs, D.T.; Edwards, T.B.

    1993-01-01

    Plutonium solubility data from several studies have been evaluated. For each data set, a predictive model has been developed where appropriate. In addition, a statistical model and corresponding prediction intervals for plutonium solubility as a quadratic function of the hydroxide concentration have been developed. Because of the wide range of solution compositions, the solubility of plutonium can vary by as much as three orders of magnitude for any given hydroxide concentration and still remain within the prediction interval. Any nuclear safety assessments that depend on the maximum amount of plutonium dissolved in alkaline salt solutions should use concentrations at least as great as the upper prediction limits developed in this study. To increase the confidence in the prediction model, it is recommended that additional solubility tests be conducted at low hydroxide concentrations and with all of the other solution components involved. To validate the model for application to actual waste solutions, it is recommended that the plutonium solubilities in actual waste solutions be determined and compared to the values predicted by the quadratic model

  11. Study on application of molten salt oxidation technology (MSO) for PVC wastes treatment

    International Nuclear Information System (INIS)

    Tran Thu Ha; Nguyen Hong Quy; Pham Quoc Ky; Nguyen Quang Long; Vuong Thu Bac; Dang Duc Nhan

    2007-01-01

    The project 'Study on application of molten salt oxidation (MSO) for PVC plastic wastes treatment' aims at three followings: 1) Installation of lab-scale MSO unit with essential compositions builds up foundation for the 2) estimation of waste destruction efficiency of the technology. 3) Based on the results of testing PVC - the chlorinated organic wastes on the lab-scale unit, the ability of the technology application at pilot-scale level will be primary estimated. The adjustment and correction of some compositions in the lab-scale unit theoretically designed during experiment overcame the shortages by design and fabrication such as heat distribution regime, feeding wastes and draining spent salt. These solutions adapt to the technical requirement of operation as well as scientific requirement of the research on MSO process. PVC waste treatment was tested on the MSO lab-scale unit in different conditions of operation temperature, superficial air velocity related to air/oxygen feeding rate, waste feeding rate. The testing results showed that destruction efficiency of chlorine in MSO technology was almost absolute. HCl and Cl 2 emission were insignificant in different operation conditions. HCl and Cl 2 emission depend on resident time and nature of molten salt. However, with inherent attributes of MSO technology emission of CO is not avoided in processing waste treatment. Therefore, finding active solutions for reduction CO emission is essential to complete the technology. The experiments also were carried in conditions of single molten salt (Na 2 CO 3 ) and molten (Na 2 CO 3 - K 2 CO 3 ) eutectic. The comparison of efficiency of these tests gives idea of using molten salt eutectic to reduce operation cost in MSO technology. Based on operation parameters and scientific verification results during experiments, the introductory procedure of waste treatment by MSO process was built up. Thereby, primary estimation of development of the technology in pilot-scale is given

  12. X-ray diffraction of slag-based sodium salt waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-30

    The attached report documents sample preparation and x-ray diffraction results for a series of cement and blended cement matrices prepared with either water or a 4.4 M Na salt solution. The objective of the study was to provide initial phase characterization for the Cementitious Barriers Partnership reference case cementitious salt waste form. This information can be used to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) potentially to design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available.

  13. Small Column Ion Exchange Analysis for Removal of Cesium from SRS Low Curie Salt Solutions Using Crystalline Silicotitanate (CST) Resin

    International Nuclear Information System (INIS)

    ALEMAN, SEBASTIAN

    2004-01-01

    Savannah River Technology Center (SRTC) researchers modeled ion exchange removal of cesium from dissolved salt waste solutions. The results assist in evaluating proposed configurations for an ion exchange process to remove residual cesium from low curie waste streams. A process for polishing (i.e., removing small amounts) of cesium may prove useful should supernate draining fail to meet the Low Curie Salt (LCS) target limit of 0.1 Ci of Cs-137 per gallon of salt solution. Cesium loading isotherms and column breakthrough curves for Low Curie dissolved salt solutions were computed to provide performance predictions for various column designs

  14. Salt disposal of heat-generating nuclear waste

    International Nuclear Information System (INIS)

    Leigh, Christi D.; Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

  15. Salt disposal of heat-generating nuclear waste.

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

  16. Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report

    International Nuclear Information System (INIS)

    Miller, C.M.; Loomis, G.G.; Prewett, S.W.

    1997-01-01

    A proof-of-concept experimental study was performed to investigate the use of Orbit Technologies polysiloxane grouting material for encapsulation of U.S. Department of Energy mixed waste salts leading to a final waste form for disposal. Evaporator pond salt residues and other salt-like material contaminated with both radioactive isotopes and hazardous components are ubiquitous in the DOE complex and may exceed 250,000,000 kg of material. Current treatment involves mixing low waste percentages (less than 10% by mass salt) with cement or costly thermal treatment followed by cementation to the ash residue. The proposed technology involves simple mixing of the granular salt material (with relatively high waste loadings-greater than 50%) in a polysiloxane-based system that polymerizes to form a silicon-based polymer material. This study involved a mixing study to determine optimum waste loadings and compressive strengths of the resultant monoliths. Following the mixing study, durability testing was performed on promising waste forms. Leaching studies including the accelerated leach test and the toxicity characteristic leaching procedure were also performed on a high nitrate salt waste form. In addition to this testing, the waste form was examined by scanning electron microscope. Preliminary cost estimates for applying this technology to the DOE complex mixed waste salt problem is also given

  17. Projected Salt Waste Production from a Commercial Pyroprocessing Facility

    Directory of Open Access Journals (Sweden)

    Michael F. Simpson

    2013-01-01

    Full Text Available Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.

  18. Cementitious Stabilization of Mixed Wastes with High Salt Loadings

    International Nuclear Information System (INIS)

    Spence, R.D.; Burgess, M.W.; Fedorov, V.V.; Downing, D.J.

    1999-01-01

    Salt loadings approaching 50 wt % were tolerated in cementitious waste forms that still met leach and strength criteria, addressing a Technology Deficiency of low salt loadings previously identified by the Mixed Waste Focus Area. A statistical design quantified the effect of different stabilizing ingredients and salt loading on performance at lower loadings, allowing selection of the more effective ingredients for studying the higher salt loadings. In general, the final waste form needed to consist of 25 wt % of the dry stabilizing ingredients to meet the criteria used and 25 wt % water to form a workable paste, leaving 50 wt % for waste solids. The salt loading depends on the salt content of the waste solids but could be as high as 50 wt % if all the waste solids are salt

  19. Process for denitrating waste solutions containing nitric acid actinides simultaneously separating the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1984-01-01

    The invention should reduce the acid and nitrate content of waste solutions containing nitric acid as much as possible, should reduce the total salt content of the waste solution, remove the actinides contained in it by precipitation and reduce the α radio-activity in the remaining solution, without having to worry about strong reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig.) [de

  20. EFFECTS OF INORGANIC SALT SOLUTION ON SOME PROPERTIES OF COMPACTED CLAY LINERS

    Directory of Open Access Journals (Sweden)

    KHALID R. MAHMOOD AL-JANABI

    2017-12-01

    Full Text Available Processed and natural clays are widely used to create impermeable liners in solid waste disposal landfills. The engineering properties of clay liners can be significantly affected by the leachate from the waste mass. In this study, the effect of inorganic salt solutions will be investigated. These solutions used at different concentrations. Two type of inorganic salt MnSO4 and FeCl3 are used at different concentration 2%,5%, 10%. Clay used in this study was the CL- clay (kaolinite. The results show that the consistency limits and unconfined compressive strength increased as the concentration of salts increased. While the permeability tends to decrease as salt concentration increased. Also, the compression index decreases as the concentration increased from 2% to 5%. The swelling index tends to increase slightly as the concentration of MnSO4 increased, while its decrease as the concentration of FeCl3. In this paper, it is aimed to investigate the performance of compacted clay liner exposed to the certain chemicals generated by the leachate and their effects on the geotechnical properties of compacted clay liner such consistency limits, permeability coefficient, compressibility characteristics and unconfined compressive strength.

  1. Hydrological methods preferentially recover cesium from nuclear waste salt cake

    International Nuclear Information System (INIS)

    Brooke, J.N.; Hamm, L.L.

    1997-01-01

    The Savannah River Site is treating high level radioactive waste in the form of insoluble solids (sludge), crystallized salt (salt cake), and salt solutions. High costs and operational concerns have prompted DOE to look for ways to improve the salt cake treatment process. A numerical model was developed to evaluate the feasibility of pump and treat technology for extracting cesium from salt cake. A modified version of the VAM3DCG code was used to first establish a steady-state flow field, then to simulate 30 days of operation. Simulation results suggest that efficient cesium extraction can be obtained with low displacement volumes. The actual extraction process will probably be less impressive because of nonuniform properties. 2 refs., 2 figs

  2. Alternatives for definse waste-salt disposal

    International Nuclear Information System (INIS)

    Benjamin, R.W.; McDonell, W.R.

    1983-01-01

    Alternatives for disposal of decontaminated high-level waste salt at Savannah River were reviewed to estimate costs and potential environmental impact for several processes. In this review, the reference process utilizing intermediate-depth burial of salt-concrete (saltcrete) monoliths was compared with alternatives including land application of the decontaminated salt as fertilizer for SRP pine stands, ocean disposal with and without containment, and terminal storage as saltcake in existing SRP waste tanks. Discounted total costs for the reference process and its modifications were in the same range as those for most of the alternative processes; uncontained ocean disposal with truck transport to Savannah River barges and storage as saltcake in SRP tanks had lower costs, but presented other difficulties. Environmental impacts could generally be maintained within acceptable limits for all processes except retention of saltcake in waste tanks, which could result in chemical contamination of surrounding areas on tank collapse. Land application would require additional salt decontamination to meet radioactive waste disposal standards, and ocean disposal without containment is not permitted in existing US practice. The reference process was judged to be the only salt disposal option studied which would meet all current requirements at an acceptable cost

  3. Organic waste processing using molten salt oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M. G., LLNL

    1998-03-01

    Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

  4. Blending Of Radioactive Salt Solutions In Million Gallon Tanks

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R.

    2012-12-10

    Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 ? 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, ?One good experiment fixes a lot of good theory?. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks.

  5. Americium Separations from High-Salt Solutions Using Anion Exchange

    International Nuclear Information System (INIS)

    Barr, Mary E.; Jarvinen, Gordon D.; Stark, Peter C.; Chamberlin, Rebecca M.; Bartsch, Richard A.; Zhang, Z.Y.; Zhao, W.

    2001-01-01

    The aging of the US nuclear stockpile presents a number of challenges, including the increasing radioactivity of plutonium residues due to the ingrowth of 241 Am from the β-decay of 241 Pu. We investigated parameters that affect the sorption of Am onto anion-exchange resins from concentrated effluents derived from nitric acid processing of plutonium residues. These postevaporator wastes are nearly saturated solutions of acidic nitrate salts, and americium removal is complicated by physical factors, such as solution viscosity and particulates, as well as by the presence of large quantities of competing metals and acid. Single- and double-contact batch distribution coefficients for americium and neodymium from simple and complex surrogate solutions are presented. Varied parameters include the nitrate salt concentration and composition and the nitric acid concentration. We find that under these extremely concentrated conditions, Am(III) removal efficiencies can surpass 50% per contact. Distribution coefficients for both neodymium and americium are insensitive to solution acidity and appear to be driven primarily by low water activities of the solutions

  6. Regularities of radium coprecipitation with barium sulfate from salt solutions

    International Nuclear Information System (INIS)

    Kudryavskij, Yu.P.; Rakhimova, O.V.

    2007-01-01

    Coprecipitation of radium with barium sulfate from highly concentrated NaCl solutions is studied, including the effects of the initial solution composition, alkaline reagent (CaO, NaOH), supporting electrolyte (NaCl) concentration, and pH. The process is promoted by high NaCl concentration in the initial solution, which is due to structural transformation and change in the sorption activity of the BaSO 4 precipitate in salt solutions. The results obtained were applied to recovery of radium from process solutions during the development and introduction of improved procedure for disinfection and decontamination of waste yielded by chlorination of loparite concentrates [ru

  7. Process for denitrating waste solutions containing nitrates and actinides with simultaneous separation of the actinides

    International Nuclear Information System (INIS)

    Gompper, K.

    1986-01-01

    The invention is intended to reduce the acid and nitrate content of nitrate waste solutions, to reduce the total salt content of the waste solution, to remove the actinides contained in it by precipitation, without any danger of violent reactions or an increase in the volume of the waste solution. The invention achieves this by mixing the waste solution with diethyl oxalate at room temperature and heating the mixture to at least 80 0 C. (orig./PW) [de

  8. Disposal of Savannah River Plant waste salt

    International Nuclear Information System (INIS)

    Dukes, M.D.

    1982-01-01

    Approximately 26-million gallons of soluble low-level waste salts will be produced during solidification of 6-million gallons of high-level defense waste in the proposed Defense Waste Processing Facility (DWPF) at the Savannah River Plant (SRP). Soluble wastes (primarily NaNO 3 , NaNO 2 , and NaOH) stored in the waste tanks will be decontaminated by ion exchange and solidified in concrete. The resulting salt-concrete mixture, saltcrete, will be placed in a landfill on the plantsite such that all applicable federal and state disposal criteria are met. Proposed NRC guidelines for the disposal of waste with the radionuclide content of SRP salt would permit shallow land burial. Federal and state rules require that potentially hazardous chemical wastes (mainly nitrate-nitrate salts in the saltcrete) be contained to the degree necessary to meet drinking water standards in the ground water beneath the landfill boundary. This paper describes the proposed saltcrete landfill and tests under way to ensure that the landfill will meet these criteria. The work includes laboratory and field tests of the saltcrete itself, a field test of a one-tenth linear scale model of the entire landfill system, and a numerical model of the system

  9. Hydrometallurgical treatment of plutonium. Bearing salt baths waste

    International Nuclear Information System (INIS)

    Bros, P.; Gozlan, J.P.; Lecomte, M.; Bourges, J.

    1993-01-01

    The salt flux issuing from the electrorefining of plutonium metal alloy in salt baths (KCI + NaCI) poses a difficult problem of the back-end alpha waste management. An alternative to the salt process promoted by Los Alamos Laboratory is to develop a hydrometallurgical treatment. A new process based on the electrochemistry technique in aqueous solution has been defined and tested successfully in the CEA. The diagram of the process exhibits two principal steps: in the head-end, a dissolution in HNO 3 medium accompanied with an electrolytic dechlorination leading to a quantitative elimination of chloride as CI 2 gas followed by its trapping one soda lime cartridge, a complete oxidative dissolution of the refractory Pu residues by electrogenerated Ag(II), in the back-end: the Pu and Am recoveries by chromatographic extractions. (authors). 10 figs., 9 refs

  10. Using Aspen simulation package to determine solubility of mixed salts in TRU waste evaporator bottoms

    Energy Technology Data Exchange (ETDEWEB)

    Hatchell, J.L.

    1998-03-01

    Nitric acid from plutonium process waste is a candidate for waste minimization by recycling. Process simulation software packages, such as Aspen, are valuable tools to estimate how effective recovery processes can be, however, constants in equations of state for many ionic components are not in their data libraries. One option is to combine single salt solubility`s in the Aspen model for mixed salt system. Single salt solubilities were regressed in Aspen within 0.82 weight percent of literature values. These were combined into a single Aspen model and used in the mixed salt studies. A simulated nitric acid waste containing mixed aluminum, calcium, iron, magnesium and sodium nitrate was tested to determine points of solubility between 25 and 100 C. Only four of the modeled experimental conditions, at 50 C and 75 C, produced a saturated solution. While experimental results indicate that sodium nitrate is the first salt to crystallize out, the Aspen computer model shows that the most insoluble salt, magnesium nitrate, the first salt to crystallize. Possible double salt formation is actually taking place under experimental conditions, which is not captured by the Aspen model.

  11. SAVANNAH RIVER SITE INCIPIENT SLUDGE MIXING IN RADIOACTIVE LIQUID WASTE STORAGE TANKS DURING SALT SOLUTION BLENDING

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, R.; Poirier, M.; Lee, S.; Steeper, T.; Fowley, M.; Parkinson, K.

    2011-01-12

    This paper is the second in a series of four publications to document ongoing pilot scale testing and computational fluid dynamics (CFD) modeling of mixing processes in 85 foot diameter, 1.3 million gallon, radioactive liquid waste, storage tanks at Savannah River Site (SRS). Homogeneous blending of salt solutions is required in waste tanks. Settled solids (i.e., sludge) are required to remain undisturbed on the bottom of waste tanks during blending. Suspension of sludge during blending may potentially release radiolytically generated hydrogen trapped in the sludge, which is a safety concern. The first paper (Leishear, et. al. [1]) presented pilot scale blending experiments of miscible fluids to provide initial design requirements for a full scale blending pump. Scaling techniques for an 8 foot diameter pilot scale tank were also justified in that work. This second paper describes the overall reasons to perform tests, and documents pilot scale experiments performed to investigate disturbance of sludge, using non-radioactive sludge simulants. A third paper will document pilot scale CFD modeling for comparison to experimental pilot scale test results for both blending tests and sludge disturbance tests. That paper will also describe full scale CFD results. The final paper will document additional blending test results for stratified layers in salt solutions, scale up techniques, final full scale pump design recommendations, and operational recommendations. Specifically, this paper documents a series of pilot scale tests, where sludge simulant disturbance due to a blending pump or transfer pump are investigated. A principle design requirement for a blending pump is UoD, where Uo is the pump discharge nozzle velocity, and D is the nozzle diameter. Pilot scale test results showed that sludge was undisturbed below UoD = 0.47 ft{sup 2}/s, and that below UoD = 0.58 ft{sup 2}/s minimal sludge disturbance was observed. If sludge is minimally disturbed, hydrogen will not be

  12. Solid waste disposal into salt mines

    International Nuclear Information System (INIS)

    Repke, W.

    1981-01-01

    The subject is discussed as follows: general introduction to disposal of radioactive waste; handling of solid nuclear waste; technology of final disposal, with specific reference to salt domes; conditioning of radioactive waste; safety barriers for radioactive waste; practice of final disposal in other countries. (U.K.)

  13. Blending of Radioactive Salt Solutions in Million Gallon Tanks - 13002

    International Nuclear Information System (INIS)

    Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R.

    2013-01-01

    Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 - 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, 'One good experiment fixes a lot of good theory'. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks. (authors)

  14. Blending of Radioactive Salt Solutions in Million Gallon Tanks - 13002

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, Robert A.; Lee, Si Y.; Fowley, Mark D.; Poirier, Michael R. [Savannah River National Laboratory, Aiken. S.C., 29808 (United States)

    2013-07-01

    Research was completed at Savannah River National Laboratory (SRNL) to investigate processes related to the blending of radioactive, liquid waste, salt solutions in 4920 cubic meter, 25.9 meter diameter storage tanks. One process was the blending of large salt solution batches (up to 1135 - 3028 cubic meters), using submerged centrifugal pumps. A second process was the disturbance of a settled layer of solids, or sludge, on the tank bottom. And a third investigated process was the settling rate of sludge solids if suspended into slurries by the blending pump. To investigate these processes, experiments, CFD models (computational fluid dynamics), and theory were applied. Experiments were performed using simulated, non-radioactive, salt solutions referred to as supernates, and a layer of settled solids referred to as sludge. Blending experiments were performed in a 2.44 meter diameter pilot scale tank, and flow rate measurements and settling tests were performed at both pilot scale and full scale. A summary of the research is presented here to demonstrate the adage that, 'One good experiment fixes a lot of good theory'. Experimental testing was required to benchmark CFD models, or the models would have been incorrectly used. In fact, CFD safety factors were established by this research to predict full-scale blending performance. CFD models were used to determine pump design requirements, predict blending times, and cut costs several million dollars by reducing the number of required blending pumps. This research contributed to DOE missions to permanently close the remaining 47 of 51 SRS waste storage tanks. (authors)

  15. Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

    International Nuclear Information System (INIS)

    Hsu, P.C.

    1997-01-01

    Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment

  16. Permanent Disposal of Nuclear Waste in Salt

    Science.gov (United States)

    Hansen, F. D.

    2016-12-01

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. Both nations are revisiting nuclear waste disposal options, accompanied by extensive collaboration on applied salt repository research, design, and operation. Salt formations provide isolation while geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Salt response over a range of stress and temperature has been characterized for decades. Research practices employ refined test techniques and controls, which improve parameter assessment for features of the constitutive models. Extraordinary computational capabilities require exacting understanding of laboratory measurements and objective interpretation of modeling results. A repository for heat-generative nuclear waste provides an engineering challenge beyond common experience. Long-term evolution of the underground setting is precluded from direct observation or measurement. Therefore, analogues and modeling predictions are necessary to establish enduring safety functions. A strong case for granular salt reconsolidation and a focused research agenda support salt repository concepts that include safety-by-design. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Author: F. D. Hansen, Sandia National Laboratories

  17. Risk analyses for disposing nonhazardous oil field wastes in salt caverns

    Energy Technology Data Exchange (ETDEWEB)

    Tomasko, D.; Elcock, D.; Veil, J.; Caudle, D.

    1997-12-01

    Salt caverns have been used for several decades to store various hydrocarbon products. In the past few years, four facilities in the US have been permitted to dispose nonhazardous oil field wastes in salt caverns. Several other disposal caverns have been permitted in Canada and Europe. This report evaluates the possibility that adverse human health effects could result from exposure to contaminants released from the caverns in domal salt formations used for nonhazardous oil field waste disposal. The evaluation assumes normal operations but considers the possibility of leaks in cavern seals and cavern walls during the post-closure phase of operation. In this assessment, several steps were followed to identify possible human health risks. At the broadest level, these steps include identifying a reasonable set of contaminants of possible concern, identifying how humans could be exposed to these contaminants, assessing the toxicities of these contaminants, estimating their intakes, and characterizing their associated human health risks. The contaminants of concern for the assessment are benzene, cadmium, arsenic, and chromium. These were selected as being components of oil field waste and having a likelihood to remain in solution for a long enough time to reach a human receptor.

  18. Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Trone, Janis R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Lawrence C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-12-01

    The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

  19. Comparison of the salt domes Asse and Gorleben with regard to their suitability for the final storage of radoactive wastes

    International Nuclear Information System (INIS)

    Deisenroth, Norbert; Kokorsch, Rudolf

    2012-01-01

    In Germany, the search for a proper solution to the issue of final disposal of radioactive wastes is complicated by political leaders. The Gorleben moratorium from October 2000 delayed the proper solution unnecessary to ten years. Asse proves that salt domes such as Gorleben do not offer a permanent partitioning of the waste over the biosphere. With this in mind, the authors of the contribution under consideration compare the two salt domes Gorleben and Asse from a mining and geological point of view based on publicly available data with regard to their suitability for the disposal of radioactive waste.

  20. Molten salt treatment to minimize and optimize waste

    International Nuclear Information System (INIS)

    Gat, U.; Crosley, S.M.; Gay, R.L.

    1993-01-01

    A combination molten salt oxidizer (MSO) and molten salt reactor (MSR) is described for treatment of waste. The MSO is proposed for contained oxidization of organic hazardous waste, for reduction of mass and volume of dilute waste by evaporation of the water. The NTSO residue is to be treated to optimize the waste in terms of its composition, chemical form, mixture, concentration, encapsulation, shape, size, and configuration. Accumulations and storage are minimized, shipments are sized for low risk. Actinides, fissile material, and long-lived isotopes are separated and completely burned or transmuted in an MSR. The MSR requires no fuel element fabrication, accepts the materials as salts in arbitrarily small quantities enhancing safety, security, and overall acceptability

  1. Waste package designs for disposal of high-level waste in salt formations

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.; Carr, J.A.

    1984-01-01

    In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

  2. Geologic disposal of nuclear wastes: salt's lead is challenged

    International Nuclear Information System (INIS)

    Kerr, R.A.

    1979-01-01

    The types of radioactive waste disposal sites available are outlined. The use of salt deposits and their advantages are discussed. The reasons for the selection of the present site for the Waste Isolation Pilot Plant are presented. The possibilities of using salt domes along the Gulf Coast and not-salt rocks as nuclear waste repositories are also discussed. The sea bed characteristics are described and advantages of this type of site selection are presented

  3. Nuclear waste solutions

    Science.gov (United States)

    Walker, Darrel D.; Ebra, Martha A.

    1987-01-01

    High efficiency removal of technetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

  4. Molten salt processing of mixed wastes with offgas condensation

    International Nuclear Information System (INIS)

    Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R.; Gay, R.L.; Stewart, A.; Yosim, S.

    1991-01-01

    We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000 degrees C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700 degrees C. 15 refs., 5 figs., 1 tab

  5. Radiological consequences of a human intrusion in a nuclear waste repository in a salt formation

    International Nuclear Information System (INIS)

    Jacquier, P.; Raimbault, P.

    1989-07-01

    The assessment of the consequences of human intrusion scenarios for a repository is very important for salt formations, since this material has an undeniable economic interest. In this work, the scenario considers the solution mining of salt for human consumption: salt is extracted from a cavern; by leaching, this cavern enlarges and uncovers the waste, which falls down into the sump. It was assumed that the intrusion takes place either 500 years or 2500 years after the closing of the repository. High-level vitrified waste or alpha cemented waste were considered. This paper displays the assumptions made and, using a simplified modelling of the phenomena, the estimation of the radiological consequences due to ingestion of contamined sals. A sensitivity/uncertainty analysis is presented which emphasizes several fields where experimental studies have to be pursued or launched [fr

  6. Recovery of fission products from acidic waste solutions thereof

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.; Dubois, D.W.

    1975-01-01

    Fission products, e.g., palladium, ruthenium and technetium, are removed from aqueous, acidic waste solutions thereof. The acidic waste solution is electrolyzed in an electrolytic cell under controlled cathodic potential conditions and technetium, ruthenium, palladium and rhodium are deposited on the cathode. Metal deposit is removed from the cathode and dissolved in acid. Acid insoluble rhodium metal is recovered, dissolved by alkali metal bisulfate fusion and purified by electrolysis. In one embodiment, the solution formed by acid dissolution of the cathode metal deposit is treated with a strong oxidizing agent and distilled to separate technetium and ruthenium (as a distillate) from palladium. Technetium is separated from ruthenium by organic solvent extraction and then recovered, e.g., as an ammonium salt. Ruthenium is disposed of as waste by-product. Palladium is recovered by electrolysis of an acid solution thereof under controlled cathodic potential conditions. Further embodiments wherein alternate metal recovery sequences are used are described. (U.S.)

  7. Selective separation of radionuclides from nuclear waste solutions with inorganic ion exchangers

    International Nuclear Information System (INIS)

    Lehto, J.; Harjula, R.

    1999-01-01

    Nuclear industry produces and stores large volumes of radioactive waste solutions. Removal of radionuclides from the solutions is an important and challenging task for two main reasons: reductions in the volumes of solidified waste, which have to be disposed of, and reductions in the radioactive discharges into the environment. Since the radioactive elements in most waste solutions are in trace concentrations and the waste solutions contain large excesses of inactive metal ions, highly selective separation methods are needed for the removal of radionuclides. A number of inorganic ion exchange materials are very selective to key radionuclides and they can play an important role in solving these problems. The spectrum of nuclear waste solutions is rather wide considering their radionuclide contents, concentrations of interfering salts and acidity/alkalinity. Therefore, several inorganic ions exchangers are needed for the removal of most harmful radionuclides from a variety of solutions. This paper discusses the use and requirements of inorganic ion exchange materials in nuclear waste management. Special attention is paid to the novel ion exchange materials developed in the Laboratory of Radiochemistry, University of Helsinki. (orig.)

  8. Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

    International Nuclear Information System (INIS)

    Dmitriev, S. A.; Adamovich, D. V.; Demkin, V. I.; Timofeev, E. M.

    2003-01-01

    The main fields of introduction and application of membrane methods for preliminary treatment and processing salt liquid radioactive waste (SLRW) can be nuclear power stations (NPP) and enterprises on atomic submarines (AS) utilization. Unlike the earlier developed technology for the liquid salt bearing radioactive waste decontamination and concentrating this report presents the new enhanced membrane technology for the liquid salt bearing radioactive waste processing based on the state-of-the-art membrane unit design, namely, the filtering units equipped with the metal-ceramic membranes of ''TruMem'' brand, as well as the electrodialysis and electroosmosis concentrators. Application of the above mentioned units in conjunction with the pulse pole changer will allow the marked increase of the radioactive waste concentrating factor and the significant reduction of the waste volume intended for conversion into monolith and disposal. Besides, the application of the electrodialysis units loaded with an ion exchange material at the end polishing stage of the radioactive waste decontamination process will allow the reagent-free radioactive waste treatment that meets the standards set for the release of the decontaminated liquid radioactive waste effluents into the natural reservoirs of fish-farming value

  9. Waste acid/metal solution reduction and recovery by vacuum distillation

    International Nuclear Information System (INIS)

    Jones, E.O.; Wilcox, W.A.; Johnson, N.T.; Bowdish, F.W.

    1995-01-01

    Processes involving distillation under reduced pressure were developed at the Pacific Northwest Laboratory several years ago to recover spent acid solutions generated during the manufacture of nuclear fuel for the N-Reactor at the Hanford site. Following construction and testing of a pilot-plant, the technology was licensed to Viatec Recovery Systems, Inc. for commercialization. The technology developed included specialized distillation and rectification of volatile acids, removal of water and/or volatile acid from sulfuric acid, and precipitation of salts. A key feature of the Waste Acid Detoxification and Reclamation (WADR) technology is the development and use of advanced thermoplastic and fluoropolymer materials of construction in all critical process equipment. The technology was then expanded to include crystallization to recover metal salts for possible reuse. Economic and environmental advantages of the procedures include recovery of acids for reuse, simplification or elimination of the disposal of waste solutions, and possible recovery of metals. Industries expected to benefit from such applications include galvanizing, electroplating, sand leaching and any where metals are cleaned in acid solutions. Currently a modular system has been assembled for recovery of several different spent acid solutions

  10. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fischer, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered

  11. Processing of radioactive waste solutions in a vacuum evaporator-crystallizer

    International Nuclear Information System (INIS)

    Petrie, J.C.; Donovan, R.I.; Van der Cook, R.E.; Christensen, W.R.

    1975-01-01

    Results of the first 18 months' operation of Hanford's vacuum evaporator-crystallizer are reported. This process reduces the volume of radioactive waste solutions and simultaneously converts the waste to a less mobile salt cake. The evaporator-crystallizer is operating at better than design production rates and has reduced the volume of radioactive wastes by more than 15 million gallons. A process description, plant performance data, mechanical difficulties, and future operating plans are discussed. Also discussed is a computer model of the evaporator-crystallizer process

  12. Hydration patterns and salting effects in sodium chloride solution.

    Science.gov (United States)

    Li, Weifeng; Mu, Yuguang

    2011-10-07

    The salting effects of 2M sodium chloride electrolyte are studied based on a series of model solutes with properties ranging from hydrophobic to hydrophilic. Generally, hydrophobic solutes will be salted out and hydrophilic solutes will be salted in by NaCl solution. The solvation free energy changes are highly correlated with Kirkwood-Buff integrals. The underlying mechanism resorts to the preferential binding of ions and water to solutes. Our results demonstrate that the salting effect not only depends on the salt's position in Hofmeister series, but also on the solutes' specifics. Taking the hydration free energies of solutes and ions as independent variables, a schematic diagram of salting effects is suggested. The resolved multifaceted salting effects rely on the sensitive balance of the tripartite interaction among solutes, ions, and water. © 2011 American Institute of Physics

  13. Thermal properties of fly ash substituted slag cement waste forms for disposal of Savannah River Plant salt waste

    International Nuclear Information System (INIS)

    Roy, D.M.; Kaushal, S.; Licastro, P.H.; Langton, C.A.

    1985-01-01

    Waste processing at the Savannah River Plant will involve reconstitution of the salts (NaNO 3 , NaNO 2 , NaOH, etc.) into a concentrated solution (32 weight percent salts) followed by solidification in a cement-based waste form for burial. The stability and mechanical durability of such a 'saltstone monolith' will depend largely on the temperature reached due to heat of hydration and the thermal properties of the waste form. Fly ash has been used as an inexpensive constituent and to moderate the hydration and setting processes so as to avoid reaching prohibitively high temperatures which could cause thermal stresses. Both high-calcium and low-calcium fly ashes have been studied for this purpose. Other constituents of these mixes include granulated blast furnace slag and finely crushed limestone. Adiabatic temperature increase and thermal conductivity of these mixes have been studied and related x-ray diffraction and scanning electron microscopy studies carried out to understand the hydration process

  14. Preliminary evaluation of solution-mining intrusion into a salt-dome repository

    International Nuclear Information System (INIS)

    1981-06-01

    This report is the product of the work of an ONWI task force to evaluate inadvertant human intrusion into a salt dome repository by solution mining. It summarizes the work in the following areas: a general review of the levels of defense that could reduce both the likelihood and potential consequences of human intrusion into a salt dome repository; evaluation of a hypothetical intrusion scenario and its consequences; recommendation for further studies. The conclusions of this task force report can be summarized as follows: (1) it is not possible at present to establish with certainty that solution mining is credible as a human-intrusion event. The likelihood of such an intrusion will depend on the effectiveness of the preventive measures; (2) an example analysis based on the realistic approach is presented in this report; it concluded that the radiological consequences are strongly dependent upon the mode of radionuclide release from the waste form, time after emplacement, package design, impurities in the host salt, the amount of a repository intercepted, the solution mining cavity form, the length of time over which solution mining occurs, the proportion of contaminated salt source for human consumption compared to other sources, and the method of salt purification for culinary purposes; (3) worst case scenarios done by other studies suggest considerable potential for exposures to man while preliminary evaluations of more realistic cases suggest significantly reduced potential consequences. Mathematical model applications to process systems, guided by more advanced assumptions about human intrusion into geomedia, will shed more light on the potential for concerns and the degree to which mitigative measures will be required

  15. Treatment of waste salts by oxygen sparging and vacuum distillation

    International Nuclear Information System (INIS)

    Cho, Y.J.; Yang, H.C.; Kim, E.H.; Kin, I.T.; Eun, H.C.

    2007-01-01

    Full text of publication follows. During the electrorefining process of the oxide spent fuel from LWR, amounts of waste salts containing some metal chloride species such as rare earths and actinide chlorides are generated, where the reuse of the waste salts is very important from the standpoint of an economical as well as an environmental aspect. In order to reuse the waste salts, a salt vacuum distillation method can be used. For the best separation by a vacuum distillation, the metal chloride species involved in the waste salts must be converted into their oxide(or oxychloride) forms due to the their low volatility compared to that of LiCl-KCl. In this study, an oxygen sparging process was adopted for the oxidation (or precipitation) of rare earth chlorides. The effects of oxygen flow rate and molten salt temperature on the conversion of rare earth chlorides to the precipitate phase (i.e. oxide or oxychloride) were investigated. In addition, distillation characteristics of LiCl-KCl molten salt with system pressure and temperature were studied. (authors)

  16. Salt removal from tanks containing high-level radioactive waste

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    At the Savannah River Plant (SRP), there are 23 waste storage tanks containing high-level radioactive wastes that are to be retired. These tanks contain about 23 million liters of salt and about 10 million liters of sludge, that are to be relocated to new Type III, fully stress-relieved tanks with complete secondary containment. About 19 million liters of salt cake are to be dissolved. Steam jet circulators were originally proposed for the salt dissolution program. However, use of steam jet circulators raised the temperature of the tank contents and caused operating problems. These included increased corrosion risk and required long cooldown periods prior to transfer. Alternative dissolution concepts were investigated. Examination of mechanisms affecting salt dissolution showed that the ability of fresh water to contact the cake surface was the most significant factor influencing dissolution rate. Density driven and mechanical agitation techniques were developed on a bench scale and then were demonstrated in an actual waste tank. Actual waste tank demonstrations were in good agreement with bench-scale experiments at 1/85 scale. The density driven method utilizes simple equipment, but leaves a cake heel in the tank and is hindered by the presence of sludge or Zeolite in the salt cake. Mechanical agitation overcomes the problems found with both steam jet circulators and the density driven technique and is the best method for future waste tank salt removal

  17. Risk assessment of nonhazardous oil-field waste disposal in salt caverns.

    Energy Technology Data Exchange (ETDEWEB)

    Elcock, D.

    1998-03-10

    Salt caverns can be formed in underground salt formations incidentally as a result of mining or intentionally to create underground chambers for product storage or waste disposal. For more than 50 years, salt caverns have been used to store hydrocarbon products. Recently, concerns over the costs and environmental effects of land disposal and incineration have sparked interest in using salt caverns for waste disposal. Countries using or considering using salt caverns for waste disposal include Canada (oil-production wastes), Mexico (purged sulfates from salt evaporators), Germany (contaminated soils and ashes), the United Kingdom (organic residues), and the Netherlands (brine purification wastes). In the US, industry and the regulatory community are pursuing the use of salt caverns for disposal of oil-field wastes. In 1988, the US Environmental Protection Agency (EPA) issued a regulatory determination exempting wastes generated during oil and gas exploration and production (oil-field wastes) from federal hazardous waste regulations--even though such wastes may contain hazardous constituents. At the same time, EPA urged states to tighten their oil-field waste management regulations. The resulting restrictions have generated industry interest in the use of salt caverns for potentially economical and environmentally safe oil-field waste disposal. Before the practice can be implemented commercially, however, regulators need assurance that disposing of oil-field wastes in salt caverns is technically and legally feasible and that potential health effects associated with the practice are acceptable. In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. It investigated regulatory issues; the types of oil-field wastes suitable for cavern disposal; cavern design and location considerations; and disposal operations, closure and remediation issues. It determined

  18. Risk assessment of nonhazardous oil-field waste disposal in salt caverns

    International Nuclear Information System (INIS)

    Elcock, D.

    1998-01-01

    Salt caverns can be formed in underground salt formations incidentally as a result of mining or intentionally to create underground chambers for product storage or waste disposal. For more than 50 years, salt caverns have been used to store hydrocarbon products. Recently, concerns over the costs and environmental effects of land disposal and incineration have sparked interest in using salt caverns for waste disposal. Countries using or considering using salt caverns for waste disposal include Canada (oil-production wastes), Mexico (purged sulfates from salt evaporators), Germany (contaminated soils and ashes), the United Kingdom (organic residues), and the Netherlands (brine purification wastes). In the US, industry and the regulatory community are pursuing the use of salt caverns for disposal of oil-field wastes. In 1988, the US Environmental Protection Agency (EPA) issued a regulatory determination exempting wastes generated during oil and gas exploration and production (oil-field wastes) from federal hazardous waste regulations--even though such wastes may contain hazardous constituents. At the same time, EPA urged states to tighten their oil-field waste management regulations. The resulting restrictions have generated industry interest in the use of salt caverns for potentially economical and environmentally safe oil-field waste disposal. Before the practice can be implemented commercially, however, regulators need assurance that disposing of oil-field wastes in salt caverns is technically and legally feasible and that potential health effects associated with the practice are acceptable. In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. It investigated regulatory issues; the types of oil-field wastes suitable for cavern disposal; cavern design and location considerations; and disposal operations, closure and remediation issues. It determined

  19. High-temperature vacuum distillation separation of plutonium waste salts

    International Nuclear Information System (INIS)

    Garcia, E.

    1996-01-01

    In this task, high-temperature vacuum distillation separation is being developed for residue sodium chloride-potassium chloride salts resulting from past pyrochemical processing of plutonium. This process has the potential of providing clean separation of the salt and the actinides with minimal amounts of secondary waste generation. The process could produce chloride salt that could be discarded as low-level waste (LLW) or low actinide content transuranic (TRU) waste, and a concentrated actinide oxide powder that would meet long-term storage standards (DOE-DTD-3013-94) until a final disposition option for all surplus plutonium is chosen

  20. Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

    International Nuclear Information System (INIS)

    1999-01-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

  1. ICP-MS nebulizer performance for analysis of SRS high salt simulated radioactive waste tank solutions (number-sign 3053)

    International Nuclear Information System (INIS)

    Jones, V.D.

    1997-01-01

    High Level Radioactive Waste Tanks at the Savannah River Site are high in salt content. The cross-flow nebulizer provided the most stable signal for all salt matrices with the smallest signal loss/suppression due to this matrix. The DIN exhibited a serious lack of tolerance for TDS; possibly due to physical de-tuning of the nebulizer efficiency

  2. Treatment of waste salt from the advanced spent fuel conditioning process (I): characterization of Zeolite A in Molten LiCl Salt

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    The oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) and the long-lived radioactive nuclides partitioning process based on electro-refining process, which are being developed ay the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as LiCl salt and LiCl-KCl eutectic salt, respectively. These waste salts must meet some criteria for disposal. A conditioning process for LiCl salt waste from ACP has been developed using zeolite A. This treatment process of waste salt using zeolite A was first developed by US ANL (Argonne National Laboratory) for LiCl-KCl eutectic salt waste from an electro-refining process of EBR (Experimental Breeder Reactor)-II spent fuel. This process has been developed recently, and a ceramic waste form (CWF) is produced in demonstration-scale V-mixer (50 kg/batch). However, ANL process is different from KAERI treatment process in waste salt, the former is LiCl-KCl eutectic salt and the latter is LiCl salt. Because of melting point, the immobilization of eutectic salt is carried out at about 770 K, whereas LiCl salt at around 920 K. Such difference has an effect on properties of immobilization media, zeolite A. Here, zeolite A in high-temperature (923 K) molten LiCl salt was characterized by XRD, Ion-exchange, etc., and evaluated if a promising media or not

  3. Expected brine movement at potential nuclear waste repository salt sites

    International Nuclear Information System (INIS)

    McCauley, V.S.; Raines, G.E.

    1987-08-01

    The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m 3 brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs

  4. Release rates from waste packages in a salt repository

    International Nuclear Information System (INIS)

    Chambre, P.L.; Hwang, Y.; Lee, W.W.L.; Pigford, T.H.

    1987-06-01

    In this report we present estimates of radionuclide release rates from waste packages into salt. This conservative and bounding analysis shows that release rates from waste packages in salt are well below the US Nuclear Regulatory Commission's performance objectives for the engineered barrier system. 2 refs., 2 figs

  5. The waste isolation pilot plant. Permanent isolation of defense transuranic waste in deep geologic salt. A national solution and international model

    International Nuclear Information System (INIS)

    Franco, Jose; Van Luik, Abraham

    2015-01-01

    The Waste Isolation Pilot Plant is located about 42 kilometers from the city of Carlsbad, New Mexico. It is an operating deep geologic repository in bedded salt 657 meters below the surface of the Chihuahuan desert. Since its opening in March of 1999, it has received about 12,000 shipments totaling about 91,000 cubic meters of defense related transuranic (TRU) wastes. Twenty-two sites have been cleaned up of their defense-legacy TRU waste. The WIPP's shipping program has an untarnished safety record and its trucks and trailers have safely traveled the equivalent of about 60 round-trips to the Moon. WIPP received, and deserved, a variety of safety accolades over its nearly 15 year working life. In February of 2014, however, two incidents resulted in a major operational suspension and reevaluation of its safety systems, processes and equipment. The first incident was an underground mining truck fire, followed nine days later by an airborne radiation release incident. Accident Investigation Board (AIB) reports on both incidents point to failures of plans, procedures and persons. The AIB recommendations for recovery from both these incidents are numerous and are being carefully implemented. One major recommendation is to no longer have different maintenance and safety requirements for nuclear handling equipment and mining equipment. Maintenance and cleanliness of mining equipment was cited as a contributing cause to the underground fire, and the idea that there can be lesser rigor in taking care of mining equipment, when it is being operated in the same underground space as the waste handling equipment, is not tenable. At some point in the future, the changes made in response to these two incidents will be seen as a valuable lesson learned on behalf of future repository programs. WIPP will once again be seen as a ''pilot'' in the nautical sense, in terms of 'showing the way' - the way to a national and international radioactive waste

  6. The waste isolation pilot plant. Permanent isolation of defense transuranic waste in deep geologic salt. A national solution and international model

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Jose; Van Luik, Abraham [US Department of Energy, Carlsbad, NM (United States). Carlsbad Field Office

    2015-07-01

    The Waste Isolation Pilot Plant is located about 42 kilometers from the city of Carlsbad, New Mexico. It is an operating deep geologic repository in bedded salt 657 meters below the surface of the Chihuahuan desert. Since its opening in March of 1999, it has received about 12,000 shipments totaling about 91,000 cubic meters of defense related transuranic (TRU) wastes. Twenty-two sites have been cleaned up of their defense-legacy TRU waste. The WIPP's shipping program has an untarnished safety record and its trucks and trailers have safely traveled the equivalent of about 60 round-trips to the Moon. WIPP received, and deserved, a variety of safety accolades over its nearly 15 year working life. In February of 2014, however, two incidents resulted in a major operational suspension and reevaluation of its safety systems, processes and equipment. The first incident was an underground mining truck fire, followed nine days later by an airborne radiation release incident. Accident Investigation Board (AIB) reports on both incidents point to failures of plans, procedures and persons. The AIB recommendations for recovery from both these incidents are numerous and are being carefully implemented. One major recommendation is to no longer have different maintenance and safety requirements for nuclear handling equipment and mining equipment. Maintenance and cleanliness of mining equipment was cited as a contributing cause to the underground fire, and the idea that there can be lesser rigor in taking care of mining equipment, when it is being operated in the same underground space as the waste handling equipment, is not tenable. At some point in the future, the changes made in response to these two incidents will be seen as a valuable lesson learned on behalf of future repository programs. WIPP will once again be seen as a ''pilot'' in the nautical sense, in terms of 'showing the way' - the way to a national and international radioactive waste

  7. Review of information on the radiation chemistry of materials around waste canisters in salt and assessment of the need for additional experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Jenks, G.H.; Baes, C.F. Jr.

    1980-03-01

    The brines, vapors, and salts precipitated from the brines will be exposed to gamma rays and to elevated temperatures in the regions close to a waste package in the salt. Accordingly, they will be subject to changes in composition brought about by reactions induced by the radiations and heat. This report reviews the status of information on the radiation chemistry of brines, gases, and solids which might be present around a waste package in salt and to assess the need for additional laboratory investigations on the radiation chemistry of these materials. The basic aspects of the radiation chemistry of water and aqueous solutions, including concentrated salt solutions, were reviewed briefly and found to be substantially unchanged from those presented in Jenks's 1972 review of radiolysis and hydrolysis in salt-mine brines. Some additional information pertaining to the radiolytic yields and reactions in brine solutions has become available since the previous review, and this information will be useful in the eventual, complete elucidation of the radiation chemistry of the salt-mine brines. 53 references.

  8. The safe disposal of radioactive wastes in geologic salt formations

    International Nuclear Information System (INIS)

    Kuehn, K.; Proske, R.

    Geologic salt formations appear to be particularly suitable for final storage. Their existance alone - the salt formations in Northern Germany are more than 200 million years old - is proof of their stability and of their isolation from biological cycles. In 1967 the storage of LAW and later, in 1972, of MAW was started in the experimental storage area Asse, south-east of Braunschweig, after the necessary technical preparations had been made. In more than ten years of operation approx. 114,000 drums of slightly active and 1,298 drums of medium-active wastes were deposited without incident. Methods have been developed for filling the available caverns with wastes and salt to ensure the security of long term disposal without supervision. Tests with electric heaters for simulation of heat-generating highly active wastes confirm the good suitability of salt formations for storing these wastes. Safety analyses for the operating time as well as for the long term phase after closure of the final storage area, which among others also comprise the improbable ''greatest expected accident'', namely break through of water, are carried out and confirm the safety of ultimate storage of radioactive wastes in geological salt formations. (orig./HP) [de

  9. Alternative methods of salt disposal at the seven salt sites for a nuclear waste repository

    International Nuclear Information System (INIS)

    1987-02-01

    This study discusses the various alternative salt management techniques for the disposal of excess mined salt at seven potentially acceptable nuclear waste repository sites: Deaf Smith and Swisher Counties, Texas; Richton and Cypress Creek Domes, Mississippi; Vacherie Dome, Louisiana; and Davis and Lavender Canyons, Utah. Because the repository development involves the underground excavation of corridors and waste emplacement rooms, in either bedded or domed salt formations, excess salt will be mined and must be disposed of offsite. The salt disposal alternatives examined for all the sites include commercial use, ocean disposal, deep well injection, landfill disposal, and underground mine disposal. These alternatives (and other site-specific disposal methods) are reviewed, using estimated amounts of excavated, backfilled, and excess salt. Methods of transporting the excess salt are discussed, along with possible impacts of each disposal method and potential regulatory requirements. A preferred method of disposal is recommended for each potentially acceptable repository site. 14 refs., 5 tabs

  10. Waste management analysis for the nuclear fuel cycle. I. Actinide recovery from aqueous salt wastes

    International Nuclear Information System (INIS)

    Martella, L.L.; Navratil, J.D.

    1979-01-01

    A preliminary feasibility study of solvent extraction methods has been completed for removing actinides from selected salt wastes likely to be produced during reactor fuel fabrication and reprocessing. The use of a two-step solvent extraction system, tributyl phosphate (TBP) followed by a bidentate organophosphorus extractant (DHDECMP), appears most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and >99.99% of the uranium. The second step, using DHDECMP, would remove >99.91% of the americium, the remaining plutonium (>99.98%), and other actinides from the acidified salt waste

  11. Defense waste salt disposal at the Savannah River Plant

    International Nuclear Information System (INIS)

    Langton, C.A.; Dukes, M.D.

    1984-01-01

    A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. The disposal process includes emplacing the saltstone in engineered trenches above the water table but below grade at SRP. Design of the waste form and disposal system limits the concentration of salts and radionuclides in the groundwater so that EPA drinking water standards will not be exceeded at the perimeter of the disposal site. 10 references, 4 figures, 3 tables

  12. Expedited demonstration of molten salt mixed waste treatment technology. Final report

    International Nuclear Information System (INIS)

    1995-01-01

    This final report discusses the molten salt mixed waste project in terms of the various subtasks established. Subtask 1: Carbon monoxide emissions; Establish a salt recycle schedule and/or a strategy for off-gas control for MWMF that keeps carbon monoxide emission below 100 ppm on an hourly averaged basis. Subtask 2: Salt melt viscosity; Experiments are conducted to determine salt viscosity as a function of ash composition, ash concentration, temperature, and time. Subtask 3: Determine that the amount of sodium carbonate entrained in the off-gas is minimal, and that any deposited salt can easily be removed form the piping using a soot blower or other means. Subtask 4: The provision of at least one final waste form that meets the waste acceptance criteria of a landfill that will take the waste. This report discusses the progress made in each of these areas

  13. Salt Repository Project Waste Package Program Plan: Draft

    International Nuclear Information System (INIS)

    Carr, J.A.; Cunnane, J.C.

    1986-01-01

    Under the direction of the Office of Civilian Radioactive Waste Management (OCRWM) created within the DOE by direction of the Nuclear Waste Policy Act of 1982 (NWPA), the mission of the Salt Repository Project (SRP) is to provide for the development of a candidate salt repository for disposal of high-level radioactive waste (HLW) and spent reactor fuel in a manner that fully protects the health and safety of the public and the quality of the environment. In consideration of the program needs and requirements discussed above, the SRP has decided to develop and issue this SRP Waste Package Program Plan. This document is intended to outline how the SRP plans to develop the waste package design and to show, with reasonable assurance, that the developed design will satisfy applicable requirements/performance objectives. 44 refs., 16 figs., 16 tabs

  14. Salt splitting of sodium-dominated radioactive waste using ceramic membranes

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Carlson, C.D.; Virkar, A.; Joshi, A.

    1994-08-01

    The potential for salt splitting of sodium dominated radioactive wastes by use of a ceramic membrane is reviewed. The technical basis for considering this processing technology is derived from the technology developed for battery and chlor-alkali chemical industry. Specific comparisons are made with the commercial organic membranes which are the standard in nonradioactive salt splitting. Two features of ceramic membranes are expected to be especially attractive: high tolerance to gamma irradiation and high selectivity between sodium and other ions. The objective of the salt splitting process is to separate nonradioactive sodium from contaminated sodium salts prior to other pretreatment processes in order to: (1) concentrate the waste in order to reduce the volume of subsequent additives and capacity of equipment, (2) decrease the pH of the waste in preparation for further processing, and (3) provide sodium with very low radioactivity levels for caustic washing of sludge or low level and mixed waste vitrification

  15. Molten salt destruction process for mixed wastes

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

    1993-04-01

    We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

  16. R and D activities on the management of waste chloride salts in KAERI

    International Nuclear Information System (INIS)

    In-Tae, Kim; Hwan-Seo, Park; Jeong-Gook, Kim; Hee-Chul, Yang; Yong-Joon, Cho; Eung-Ho Kim

    2007-01-01

    Full text of publication follows. Electrochemical treatment of spent oxide fuels has been intensively studied in KAERI to reduce the volume, heat load and radiotoxicity of high-level wastes. It consists of an electrolytic reduction process to convert the oxide fuel into a metallic form and an electro-refining process to separate TRU elements from the electro-reduced metal ingot. Two types of waste salts are expected to generate from the electrochemical pyro-processes, that is, LiCl salt from the reduction process and LiCl+KCl eutectic salt form the refining process. The R and D strategy of the waste salt management in KAERI can be categorized into two parts: 1) enhancement of safety by the stabilisation/solidification of waste salt that is to be finally disposed of and 2) reduction of the waste generation by the regeneration/recycle of the spent salt after removal of radionuclides in it. A sol-gel technique and a zeolite occlusion technique are under development to stabilize the waste salt. The LiCl salt is stabilised by a low-temperature sol-gel process and then the gel product is solidified into a ceramic-like waste form with an addition of glass frit. Another method uses Zeolite-4A to occlude the LiCl salt into its cage and adsorption site to immobilize the radionuclides. The product, salt-occluded zeolite, is fabricated into another type of a ceramic waste form. For the regeneration and recycle of the spent salt, the radionuclides in the salt are removed by a zeolite process for the LiCl salt and by an oxidation/distillation process for the eutectic salt. The target nuclides to be removed in each process are Cs/Sr and rare earth (RE) elements, respectively. In the oxidation/ distillation process, the rare earth chloride nuclides are oxidised by an oxygen sparging method, and the products are precipitated in the form of oxide or oxychloride REs. After separation of the RE elements from the precipitates by distillation, the refined spent salt with a low content

  17. Containment of solidified liquid hazardous waste in domal salt

    International Nuclear Information System (INIS)

    Domenico, P.A.; Lerman, A.

    1992-01-01

    In recent years, the solidification of hazardous liquid waste has become a viable option in waste management. The solidification process results in an increased volume but more stable waste form that must be disposed of or stored in a dry environment. An environment of choice in south central Texas is domal salt. The salt dome currently under investigation has a water content of 0.002 percent by weight and a permeability less than one nanodarcy. A question that must be addressed is whether a salt dome has a particular set of attributes that will prevent the release of contaminants to the environment. From a regulatory perspective, a ''no migration'' petition must be approved by the U.S.E.P.A. for the containment facility. By ''no migration'' it is implied that the waste must be contained for 10,000 years. A demonstration that this condition will be met will require model calculations and such models must be based on the physical and chemical characteristics of the waste form and the geologic environment. In particular, the models must address the rate of brine infiltration into the caverns, providing information on how fast an immobile solid waste form could convert to a more mobile liquid state. Additionally, the potential for migration by both diffusion and advection is of concern. Lastly, given a partially saturated cavern, the question of how far gaseous waste will be transported over the 10,000 year containment period must also be addressed. Results indicate that the containment capabilities of domal salt are exceptional. A nominal volume of brine will seep into the cavern and most voids between the injected solidified waste pellets will remain unsaturated. Very small quantities of hazardous constituents will be leached from the waste pellets

  18. Extraction, scrub, and strip test results for the solvent transfer to salt waste processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-07

    The Savannah River National Laboratory (SRNL) prepared approximately 240 gallons of Caustic-Side Solvent Extraction (CSSX) solvent for use at the Salt Waste Processing Facility (SWPF). An Extraction, Scrub, and Strip (ESS) test was performed on a sample of the prepared solvent using a salt solution prepared by Parsons to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams. This data will be used by Parsons to help qualify the solvent for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 15.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  19. Characteristics of solidified products containing radioactive molten salt waste.

    Science.gov (United States)

    Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Kim, Joon-Hyung

    2007-11-01

    The molten salt waste from a pyroprocess to recover uranium and transuranic elements is one of the problematic radioactive wastes to be solidified into a durable wasteform for its final disposal. By using a novel method, named as the GRSS (gel-route stabilization/solidification) method, a molten salt waste was treated to produce a unique wasteform. A borosilicate glass as a chemical binder dissolves the silicate compounds in the gel products to produce one amorphous phase while most of the phosphates are encapsulated by the vitrified phase. Also, Cs in the gel product is preferentially situated in the silicate phase, and it is vitrified into a glassy phase after a heat treatment. The Sr-containing phase is mainly phosphate compounds and encapsulated by the glassy phase. These phenomena could be identified by the static and dynamic leaching test that revealed a high leach resistance of radionuclides. The leach rates were about 10(-3) - 10(-2) g/m2 x day for Cs and 10(-4) - 10(-3) g/m2 x day for Sr, and the leached fractions of them were predicted to be 0.89% and 0.39% at 900 days, respectively. This paper describes the characteristics of a unique wasteform containing a molten salt waste and provides important information on a newly developed immobilization technology for salt wastes, the GRSS method.

  20. Problems of the final storage of radioactive waste in salt formations

    International Nuclear Information System (INIS)

    Hofrichter, E.

    1977-01-01

    The geological conditions for the final storage of radioactive waste, the occurrence of salt formations, and the tectonics of salt domes are discussed. The safety of salt rocks, the impermeability of the rocks, and the thermal problems in the storage of high-activity waste are dealt with. Possibilities and preconditions of final storage in West Germany are discussed. (HPH) [de

  1. Thermal denitration of high concentration nitrate salts waste water

    International Nuclear Information System (INIS)

    Hwang, D. S.; Oh, J. H.; Choi, Y. D.; Hwang, S. T.; Park, J. H.; Latge, C.

    2003-01-01

    This study investigated the thermodynamic and the thermal decomposition properties of high concentration nitrate salts waste water for the lagoon sludge treatment. The thermodynamic property was carried out by COACH and GEMINI II based on the composition of nitrate salts waste water. The thermal decomposition property was carried out by TG-DTA and XRD. Ammonium nitrate and sodium nitrate were decomposed at 250 .deg. C and 730 . deg. C, respectively. Sodium nitrate could be decomposed at 450 .deg. C in the case of adding alumina for converting unstable Na 2 O into stable Na 2 O.Al 2 O 3 . The flow sheet for nitrate salts waste water treatment was proposed based on the these properties data. These will be used by the basic data of the process simulation

  2. Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

    International Nuclear Information System (INIS)

    Koyama, Tadafumi.

    1994-01-01

    A method is described for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities

  3. Waste segregation analysis for salt well pumping in the 200 W Area -- Task 3.4

    International Nuclear Information System (INIS)

    Reynolds, D.A.

    1995-01-01

    There is an estimated 7 million liters (1.9 million gallons) of potentially complexed waste that need to be pumped from single-shell tanks (SST) in the 200 West Area. This represents up to 40% of the salt well liquor that needs to be pumped in the 200 West Area. There are three double-shell (DST) tanks in the 241-SY tank farm in the 200 West Area. Tank 241-SY-101 is full and not usable. Tank 241-SY-102 has a transuranic (TRU) sludge in the bottom. Current rules prohibit mixing complexed waste with TRU waste. Tank 241-SY-103 has three major problems. First, 241-SY-103 is on the Flammable Watch list. Second, adding waste to tank 241-SY-103 has the potential for an episodic release of hydrogen gas. Third, 241-SY-103 will not hold all of the potentially complexed waste from the SSTs. This document looks at more details regarding the salt well pumping of the 200 West Area tank farm. Some options are considered but it is beyond the scope of this document to provide an in-depth study necessary to provide a defensible solution to the complexed waste problem

  4. Leaching due to hygroscopic water uptake in cemented waste containing soluble salts

    DEFF Research Database (Denmark)

    Brodersen, K.

    1992-01-01

    conditions, condensation of water vapour will result in generation of a certain amount of liquid in the form of a strong salt solution. The volume of liquid may well exceed the storage capacity of the pore system in the cemented material and in the release of a limited amount of free contaminated solution......Considerable amounts of easily soluble salts such as sodium nitrate, sulphate, or carbonate are introduced into certain types of cemented waste. When such materials are stored in atmospheres with high relative humidity or disposed or by shallow land burial under unsaturated, but still humid....... A model of the quantitative aspects for the equilibrium situation is presented. Experiments with hygroscopic water uptake support the model and give indications about the rate of the process. The release mechanism is only thought to be important for radionuclides which are not fixed in a low...

  5. Salt-specific effects in lysozyme solutions

    Directory of Open Access Journals (Sweden)

    T. Janc

    2016-03-01

    Full Text Available The effects of additions of low-molecular-mass salts on the properties of aqueous lysozyme solutions are examined by using the cloud-point temperature, T_{cloud}, measurements. Mixtures of protein, buffer, and simple salt in water are studied at pH=6.8 (phosphate buffer and pH=4.6 (acetate buffer. We show that an addition of buffer in the amount above I_{buffer} = 0.6 mol dm^{-3} does not affect the T_{cloud} values. However, by replacing a certain amount of the buffer electrolyte by another salt, keeping the total ionic strength constant, we can significantly change the cloud-point temperature. All the salts de-stabilize the solution and the magnitude of the effect depends on the nature of the salt. Experimental results are analyzed within the framework of the one-component model, which treats the protein-protein interaction as highly directional and of short-range. We use this approach to predict the second virial coefficients, and liquid-liquid phase diagrams under conditions, where T_{cloud} is determined experimentally.

  6. Assessment of crushed salt consolidation and fracture healing processes in a nuclear waste repository in salt

    International Nuclear Information System (INIS)

    1984-11-01

    For a nuclear waste repository in salt, two aspects of salt behavior are expected to contribute to favorable conditions for waste isolation. First, consolidation of crushed salt backfill due to creep closure of the underground openings may result in a backfill barrier with low permeability. Second, fractures created in the salt by excavation may heal under the influence of stress and temperature following sealing. This report reviews the status of knowledge regarding crushed salt consolidation and fracture healing, provides analyses which predict the rates at which the processes will occur under repository conditions, and develops requirements for future study. Analyses of the rate at which crushed salt will consolidate are found to be uncertain because of unexplained wide variation in the creep properties of crushed salt obtained from laboratory testing, and because of uncertainties in predictions of long term closure rates of openings in salt. This uncertainty could be resolved to a large degree by additional laboratory testing of crushed salt. Similarly, additional testing of fracture healing processes is required to confirm that healing will be effective under repository conditions. Extensive references, 27 figures, 5 tables

  7. Systems costs for disposal of Savannah River high-level waste sludge and salt

    International Nuclear Information System (INIS)

    McDonell, W.R.; Goodlett, C.B.

    1984-01-01

    A systems cost model has been developed to support disposal of defense high-level waste sludge and salt generated at the Savannah River Plant. Waste processing activities covered by the model include decontamination of the salt by a precipitation process in the waste storage tanks, incorporation of the sludge and radionuclides removed from the salt into glass in the Defense Waste Processing Facility (DWPF), and, after interim storage, final disposal of the DWPF glass waste canisters in a federal geologic repository. Total costs for processing of waste generated to the year 2000 are estimated to be about $2.9 billion (1984 dollars); incremental unit costs for DWPF and repository disposal activities range from $120,000 to $170,000 per canister depending on DWPF processing schedules. In a representative evaluation of process alternatives, the model is used to demonstrate cost effectiveness of adjustments in the frit content of the waste glass to reduce impacts of wastes generated by the salt decontamination operations. 13 references, 8 tables

  8. Numerical analysis of impurity separation from waste salt by investigating the change of concentration at the interface during zone refining process

    Science.gov (United States)

    Choi, Ho-Gil; Shim, Moonsoo; Lee, Jong-Hyeon; Yi, Kyung-Woo

    2017-09-01

    The waste salt treatment process is required for the reuse of purified salts, and for the disposal of the fission products contained in waste salt during pyroprocessing. As an alternative to existing fission product separation methods, the horizontal zone refining process is used in this study for the purification of waste salt. In order to evaluate the purification ability of the process, three-dimensional simulation is conducted, considering heat transfer, melt flow, and mass transfer. Impurity distributions and decontamination factors are calculated as a function of the heater traverse rate, by applying a subroutine and the equilibrium segregation coefficient derived from the effective segregation coefficients. For multipass cases, 1d solutions and the effective segregation coefficient obtained from three-dimensional simulation are used. In the present study, the topic is not dealing with crystal growth, but the numerical technique used is nearly the same since the zone refining technique was just introduced in the treatment of waste salt from nuclear power industry because of its merit of simplicity and refining ability. So this study can show a new application of single crystal growth techniques to other fields, by taking advantage of the zone refining multipass possibility. The final goal is to achieve the same high degree of decontamination in the waste salt as in zone freezing (or reverse Bridgman) method.

  9. Generic aspects of salt repositories

    International Nuclear Information System (INIS)

    Laughon, R.B.

    1979-01-01

    The history of geological disposal of radioactive wastes in salt is presented from 1957 when a panel of the National Academy of Sciences-National Research Council recommended burial in bedded salt deposits. Early work began in the Kansas, portion of the Permian Basin where simulated wastes were placed in an abandoned salt mine at Lyons, Kansas, in the late 1960's. This project was terminated when the potential effect of nearby solution mining activities could not be resolved. Evaluation of bedded salts resumed a few years later in the Permian Basin in southeastern New Mexico, and search for suitable sites in the 1970's resulted in the formation of the National Waste Terminal Storage Program in 1976. Evaluation of salt deposits in many regions of the United States has been virtually completed and has shown that deposits having the greatest potential for radioactive waste disposal are those of the largest depositional basins and salt domes of the Gulf Coast region

  10. Investigation of Various LiCl Waste Salt Purification Technologies

    International Nuclear Information System (INIS)

    Yung-Zun Cho; Hee-Chul Yang; Han-Soo Lee; In-Tae Kim

    2008-01-01

    Various purification research of LiCl waste molten salt generated from electroreduction process were tested. The purification of the LiCl waste salt very important in a various aspects, where the purification means separation of cesium and strontium form LiCl salt melts. In this study, for the separation of cesium and strontium from LiCl salt melts, precipitant agent addition techniques such as sulfate and carbonate addition method and, as a new attempt, zone freezing technique for concentration of cesium and strontium elements was investigated. As a results of this research, only strontium was carbonated by reaction with Li 2 CO 3 (cesium did not react with Li 2 CO 3 ). In case of sulfate addition method, both cesium and strontium were converted into their sulfate that is Cs 2 S 2 O 6 and SrSO 4 and maximum sulfate efficiency of cesium and strontium were about 72% and 95%, respectively. Cesium and strontium involved in LiCl molten salt could be concentrated in the molten salt by using zone freezing method. (authors)

  11. Americium separations from high salt solutions

    International Nuclear Information System (INIS)

    Barr, Mary E.; Jarvinen, Gordon D.; Schulte, Louis D.; Stark, Peter C.; Chamberlin, Rebecca M.; Abney, Kent D.; Ricketts, Thomas E.; Valdez, Yvette E.; Bartsch, Richard A.

    2000-01-01

    Americium (III) exhibits an unexpectedly high affinity for anion-exchange material from the high-salt evaporator bottoms solutions--an effect which has not been duplicated using simple salt solutions. Similar behavior is observed for its lanthanide homologue, Nd(III), in complex evaporator bottoms surrogate solutions. There appears to be no single controlling factor--acid concentration, total nitrate concentration or solution ionic strength--which accounts for the approximately 2-fold increase in retention of the trivalent ions from complex solutions relative to simple solutions. Calculation of species activities (i.e., water, proton and nitrate) in such concentrated mixed salt solutions is difficult and of questionable accuracy, but it is likely that the answer to forcing formation of anionic nitrate complexes of americium lies in the relative activities of water and nitrate. From a practical viewpoint, the modest americium removal needs (ca. 50--75%) from nitric acid evaporator bottoms allow sufficient latitude for the use of non-optimized conditions such as running existing columns filled with older, well-used Reillex HPQ. Newer materials, such as HPQ-100 and the experimental bifunctional resins, which exhibit higher distribution coefficients, would allow for either increased Am removal or the use of smaller columns. It is also of interest that one of the experimental neutral-donor solid-support extractants, DHDECMP, exhibits a similarly high level of americium (total alpha) removal from EV bottoms and is much less sensitive to total acid content than commercially-available material

  12. Phase Equilibrium Studies of Savannah River Tanks and Feed Streams for the Salt Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.

    2001-06-19

    A chemical equilibrium model is developed and used to evaluate supersaturation of tanks and proposed feed streams to the Salt Waste Processing Facility. The model uses Pitzer's model for activity coefficients and is validated by comparison with a variety of thermodynamic data. The model assesses the supersaturation of 13 tanks at the Savannah River Site (SRS), indicating that small amounts of gibbsite and or aluminosilicate may form. The model is also used to evaluate proposed feed streams to the Salt Waste Processing Facility for 13 years of operation. Results indicate that dilutions using 3-4 M NaOH (about 0.3-0.4 L caustic per kg feed solution) should avoid precipitation and reduce the Na{sup +} ion concentration to 5.6 M.

  13. Recovery of soluble chloride salts from the wastewater generated during the washing process of municipal solid wastes incineration fly ash.

    Science.gov (United States)

    Tang, Hailong; Erzat, Aris; Liu, Yangsheng

    2014-01-01

    Water washing is widely used as the pretreatment method to treat municipal solid waste incineration fly ash, which facilitates the further solidification/stabilization treatment or resource recovery of the fly ash. The wastewater generated during the washing process is a kind of hydrosaline solution, usually containing high concentrations of alkali chlorides and sulphates, which cause serious pollution to environment. However, these salts can be recycled as resources instead of discharge. This paper explored an effective and practical recovery method to separate sodium chloride, potassium chloride, and calcium chloride salts individually from the hydrosaline water. In laboratory experiments, a simulating hydrosaline solution was prepared according to composition of the waste washing water. First, in the three-step evaporation-crystallization process, pure sodium chloride and solid mixture of sodium and potassium chlorides were obtained separately, and the remaining solution contained potassium and calcium chlorides (solution A). And then, the solid mixture was fully dissolved into water (solution B obtained). Finally, ethanol was added into solutions A and B to change the solubility of sodium, potassium, and calcium chlorides within the mixed solvent of water and ethanol. During the ethanol-adding precipitation process, each salt was separated individually, and the purity of the raw production in laboratory experiments reached about 90%. The ethanol can be recycled by distillation and reused as the solvent. Therefore, this technology may bring both environmental and economic benefits.

  14. Testing of Air Pulse Agitators to Support Design of Savannah River Site Highly Radioactive Processing at the Salt Waste Processing Facility

    International Nuclear Information System (INIS)

    Gallego, R.M.; Stephens, A.B.; Wilkinson, R.H.; Dev, H.; Suggs, P.C.

    2006-01-01

    The Salt Waste Processing Facility (SWPF) is intended to concentrate the highly radioactive constituents from waste salt solutions at the Savannah River Site (SRS). Air Pulse Agitators (APAs) were selected for process mixing in high-radiation locations at the SWPF. This technology has the advantage of no moving parts within the hot cell, eliminating potential failure modes and the need for maintenance within the high-radiation environment. This paper describes the results of APA tests performed to gain operational and performance data for the SWPF design. (authors)

  15. Summary Report of Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-12

    The inadvertent creation of transuranic waste carrying hazardous waste codes D001 and D002 requires the treatment of the material to eliminate the hazardous characteristics and allow its eventual shipment and disposal at the Waste Isolation Pilot Plant (WIPP). This report documents the effectiveness of two treatment methods proposed to stabilize both the unremediated and remediated nitrate salt waste streams (UNS and RNS, respectively). The two technologies include the addition of zeolite (with and without the addition of water as a processing aid) and cementation. Surrogates were developed to evaluate both the solid and liquid fractions expected from parent waste containers, and both the solid and liquid fractions were tested. Both technologies are shown to be effective at eliminating the characteristic of ignitability (D001), and the addition of zeolite was determined to be effective at eliminating corrosivity (D002), with the preferred option1 of zeolite addition currently planned for implementation at the Waste Characterization, Reduction, and Repackaging Facility. During the course of this work, we established the need to evaluate and demonstrate the effectiveness of the proposed remedy for debris material, if required. The evaluation determined that Wypalls absorbed with saturated nitrate salt solutions exhibit the ignitability characteristic (all other expected debris is not classified as ignitable). Follow-on studies will be developed to demonstrate the effectiveness of stabilization for ignitable Wypall debris. Finally, liquid surrogates containing saturated nitrate salts did not exhibit the characteristic of ignitability in their pure form (those neutralized with Kolorsafe and mixed with sWheat did exhibit D001). As a result, additional nitrate salt solutions (those exhibiting the oxidizer characteristic) will be tested to demonstrate the effectiveness of the remedy.

  16. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-01-01

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for 238 Pu contaminated waste. Combustible low-level waste material contaminated with 238 Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble 2328 Pu in the spent salt. The valuable 238 Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of 238 Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered 238 Pu is considered

  17. ELECTRODIALYSIS IN THE CONVERSION STEP OF THE CONCENTRATED SALT SOLUTIONS IN THE PROCESS OF BATTERY SCRAP

    Directory of Open Access Journals (Sweden)

    S. I. Niftaliev

    2014-01-01

    Full Text Available Summary. The concentrated sodium sulfate solution is formed during the processing of waste battery scrap. A promising way to further treatment of the concentrated salt solution could be the conversion of these salts into acid and bases by electrodialysis, that can be reused in the same technical process cycle. For carrying out the process of conversion of salts into the corresponding acid and base several cells schemes with different combinations of cation, anion and bipolar membranes are used. At this article a comparative analysis of these cells is carried out. In the cells there were used the membranes МC-40, МА-41 and МB-2I. Acid and base solutions with higher concentration may be obtained during the process of electrodialysis in the circulation mode, when a predetermined amount of salt in the closed loop is run through a set of membranes to obtain the desired concentration of the product. The disadvantages of this method are the high cost of buffer tanks and the need to work with small volumes of treated solutions. In industrial applications it is advisable to use continuous electrodialysis with bipolar membranes, since this configuration allows to increase the number of repeating sections, which is necessary to reduce the energy costs. The increase of the removal rate of salts can be achieved by increasing the process steps, and to produce a more concentrated products after the conversion step can be applied electrodialysis-concentrator or evaporator.

  18. Deep geologic disposal of mixed waste in bedded salt: The Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Rempe, N.T.

    1993-01-01

    Mixed waste (i.e., waste that contains both chemically hazardous and radioactive components) poses a moral, political, and technical challenge to present and future generations. But an international consensus is emerging that harmful byproducts and residues can be permanently isolated from the biosphere in a safe and environmentally responsible manner by deep geologic disposal. To investigate and demonstrate such disposal for transuranic mixed waste, derived from defense-related activities, the US Department of Energy has prepared the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. This research and development facility was excavated approximately at the center of a 600 m thick sequence of salt (halite) beds, 655 m below the surface. Proof of the long-term tectonic and hydrological stability of the region is supplied by the fact that these salt beds have remained essentially undisturbed since they were deposited during the Late Permian age, approximately 225 million years ago. Plutonium-239, the main radioactive component of transuranic mixed waste, has a half-life of 24,500 years. Even ten half-lives of this isotope - amounting to about a quarter million years, the time during which its activity will decline to background level represent only 0.11 percent of the history of the repository medium. Therefore, deep geologic disposal of transuranic mixed waste in Permian bedded salt appears eminently feasible

  19. Engineering Options Assessment Report. Nitrate Salt Waste Stream Processing

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-13

    This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 above-ground UNS, and 79 candidate below-ground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

  20. Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-18

    This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 aboveground UNS, and 79 candidate belowground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

  1. A novel method of non-violent dissolution of sodium metal in a concentrated aqueous solution of Epsom salt

    International Nuclear Information System (INIS)

    Lakshmanan, A.R.; Prasad, M.V.R.; Ponraju, D.; Krishnan, H.

    2004-01-01

    A new technique of non-violent and fast dissolution of sodium metal in a concentrated aqueous solution of Epsom salt (MgSO 4 .7H 2 O) at room temperature (RT) has been developed. The dissolution process is mildly exothermic but could be carried out even in a glass beaker in air under swift stirring condition. The reaction products consist of mixed salts of MgSO 4 and Na 2 SO 4 as well as Mg(OH) 2 which are only mildly alkaline and hence are non-corrosive and non-hazardous unlike NaOH. A 50 mL solution having Epsom salt concentration of 2 M was found to give the optimal composition for disposal of 1 g of sodium. Supersaturated (>2.7 M), as well as dilute (<1.1 M) solutions, however, cause violent reactions and hence should be avoided. Repeated sodium dissolution in Epsom solution produced a solid waste of 4.7 g per g of sodium dissolved which is comparable with the waste (4 g) produced in 8 M NaOH solution. A 1.4 M Epsom solution sprayed with a high-pressure jet cleaner at RT in air easily removed the sodium blocked inside a metal pipe made of mild steel. The above jet also dissolved peacefully residual sodium collected on the metal tray after a sodium fire experiment. No sodium fire or explosion was observed during this campaign. The Epsom solution spray effectively neutralized the minor quantity of sodium aerosol produced during this campaign. This novel technique would hence be quite useful for draining sodium from fast breeder reactor components and bulk processing of sodium as well as for sodium fire fighting

  2. Long-term effects of waste solutions on concrete and reinforcing steel

    International Nuclear Information System (INIS)

    Daniel, J.I.; Stark, D.C.; Kaar, P.H.

    1982-04-01

    This report has been prepared for the In Situ Waste Disposal Program Tank Assessment Task (WG-11) as part of an investigation to evaluate the long-term performance of waste storage tanks at the Hanford Site. This report, prepared by the Portland Cement Association, presents the results of four years of concrete degradation studies which exposed concrete and reinforcing steel, under load and at 180 0 F, to simulated double-shell slurry, simulated salt cake solution, and a control solution. Exposure length varied from 3 months to 36 months. In all cases, examination of the concrete and reinforcing steel at the end of the exposure indicated there was no attack, i.e., no evidence of rusting, cracking, disruption of mill scale or loss of strength

  3. Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

    International Nuclear Information System (INIS)

    Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G.

    2005-08-01

    Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research

  4. Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G

    2005-08-01

    Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research.

  5. Radiolysis salt phenomenology: application to storage of high level radioactive waste

    International Nuclear Information System (INIS)

    Akram, Najib

    1993-01-01

    In France, rock salt is a candidate repository for highly radioactive waste. Rock salt contains water and adsorbed gases which can be released in boreholes after heating due to vitrified wastes. In addition, waste-induced irradiation in near-field conditions induce radiolytic reactions which also contribute to gas release. The aim of this work is to understand and evaluate the effects of heat and irradiation produced by waste containers in a deep disposal, primarily concerning gas production. This is justified by the impact of gases on long-term safety: toxicity, explosibility, chemical reactivity, pressure build-up. We have evidenced the influence of integrated dose, filling gases, temperature and grain size on an homogeneous medium (Asse Mine rock salt). We have then studied heterogeneous samples, which allowed to determine the influence of the chemical and mineralogical composition of rock salt (bedded rock salt from the Mine de Potasse d'Alsace). The role played by organic matter on gas production is important, leading for instance to high consumption rates of oxygen. Through this study, we have also considered the behaviour of clay-rich materials under irradiation. Our results constitute important bases for the future modelling of the phenomena which will take place in the near-field of a rock salt-type repository, especially concerning its long-term safety. (author) [fr

  6. Salt Effect on Osmotic Pressure of Polyelectrolyte Solutions: Simulation Study

    Directory of Open Access Journals (Sweden)

    Jan-Michael Y. Carrillo

    2014-07-01

    Full Text Available Abstract: We present results of the hybrid Monte Carlo/molecular dynamics simulations of the osmotic pressure of salt solutions of polyelectrolytes. In our simulations, we used a coarse-grained representation of polyelectrolyte chains, counterions and salt ions. During simulation runs, we alternate Monte Carlo and molecular dynamics simulation steps. Monte Carlo steps were used to perform small ion exchange between simulation box containing salt ions (salt reservoir and simulation box with polyelectrolyte chains, counterions and salt ions (polyelectrolyte solution. This allowed us to model Donnan equilibrium and partitioning of salt and counterions across membrane impermeable to polyelectrolyte chains. Our simulations have shown that the main contribution to the system osmotic pressure is due to salt ions and osmotically active counterions. The fraction of the condensed (osmotically inactive counterions first increases with decreases in the solution ionic strength then it saturates. The reduced value of the system osmotic coefficient is a universal function of the ratio of the concentration of osmotically active counterions and salt concentration in salt reservoir. Simulation results are in a very good agreement with osmotic pressure measurements in sodium polystyrene sulfonate, DNA, polyacrylic acid, sodium polyanetholesulfonic acid, polyvinylbenzoic acid, and polydiallyldimethylammonium chloride solutions.

  7. Results for the Fourth Quarter Calendar Year 2015 Tank 50H Salt Solution Sample

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-01-11

    In this memorandum, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2015 (CY15) sample of Tank 50H salt solution are presented in tabulated form. The Fourth Quarter CY15 Tank 50H samples were obtained on October 29, 2015 and received at Savannah River National Laboratory (SRNL) on October 30, 2015. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter Calendar Year 2015 (CY15) sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan.

  8. Salt decontamination demonstration test results

    International Nuclear Information System (INIS)

    Snell, E.B.; Heng, C.J.

    1983-06-01

    The Salt Decontamination Demonstration confirmed that the precipitation process could be used for large-scale decontamination of radioactive waste sale solution. Although a number of refinements are necessary to safely process the long-term requirement of 5 million gallons of waste salt solution per year, there were no observations to suggest that any fundamentals of the process require re-evaluation. Major accomplishments were: (1) 518,000 gallons of decontaminated filtrate were produced from 427,000 gallons of waste salt solution from tank 24H. The demonstration goal was to produce a minimum of 200,000 gallons of decontaminated salt solution; (2) cesium activity in the filtrate was reduced by a factor of 43,000 below the cesium activity in the tank 24 solution. This decontamination factor (DF) exceeded the demonstration goal of a DF greater than 10,000; (3) average strontium-90 activity in the filtrate was reduced by a factor of 26 to less than 10 3 d/m/ml versus a goal of less than 10 4 d/m/ml; and (4) the concentrated precipitate was washed to a final sodium ion concentration of 0.15 M, well below the 0.225 M upper limit for DWPF feed. These accomplishments were achieved on schedule and without incident. Total radiation exposure to personnel was less than 350 mrem and resulted primarily from sampling precipitate slurry inside tank 48. 3 references, 6 figures, 2 tables

  9. Experimental results on salt concrete for barrier elements made of salt concrete in a repository for radioactive waste in a salt mine

    International Nuclear Information System (INIS)

    Gutsch, Alex-W.; Preuss, Juergen; Mauke, Ralf

    2012-01-01

    The Bartensleben rock salt mine in Germany was used as a repository for low and intermediate level radioactive waste from 1971 to 1991 and from 1994 to 1998. The repository with an overall volume of about 6 million m 3 has to be closed. Salt concrete is used for the refill of the voids of the repository. The concrete mixtures contain crushed salt instead of natural aggregates as the void filling material should be as similar to the salt rock as possible. Very high requirements regarding low heat development and little or even no cracking during concrete hardening had to be fulfilled even for the barrier elements made from salt concrete which separate the radioactive waste from the environment. Requirements for the salt concrete were set up with regard to the fluidity of the fresh concrete during the hardening process and its durability. In the view of a comprehensive numerical calculations of the temperature development and thermal stresses in the massive salt concrete elements of the backfill of the voids, experimental results for material properties of the salt concrete are presented: mixture of the salt concrete, thermodynamic properties (adiabatic heat release, thermal dilatation, thermal conductivity and heat capacity), mechanical short term properties, creep (under tension, under compression), autogenous shrinkage

  10. Potentiometric titration curves of aluminium salt solutions and its ...

    African Journals Online (AJOL)

    Potentiometric titration curves of aluminium salt solutions and its species conversion ... of aluminium salt solutions under the moderate slow rate of base injection. ... silicate radical, and organic acid radical on the titration curves and its critical ...

  11. Salt splitting with ceramic membranes

    International Nuclear Information System (INIS)

    Kurath, D.

    1996-01-01

    The purpose of this task is to develop ceramic membrane technologies for salt splitting of radioactively contaminated sodium salt solutions. This technology has the potential to reduce the low-level waste (LLW) disposal volume, the pH and sodium hydroxide content for subsequent processing steps, the sodium content of interstitial liquid in high-level waste (HLW) sludges, and provide sodium hydroxide free of aluminum for recycle within processing plants at the DOE complex. Potential deployment sites include Hanford, Savannah River, and Idaho National Engineering Laboratory (INEL). The technical approach consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON). As the name implies, sodium ions are transported rapidly through these ceramic crystals even at room temperatures

  12. Waste isolation facility description: bedded salt

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria. (LK)

  13. Waste isolation facility description: bedded salt

    International Nuclear Information System (INIS)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria

  14. Different Methods for Conditioning Chloride Salt Wastes

    International Nuclear Information System (INIS)

    De Angelis, G.; Fedeli, C.; Capone, M.; Marzo, G.A.; Mariani, M.; Da Ros, M.; Giacobbo, F.; Macerata, E.; Giola, M.

    2015-01-01

    Three different methods have been used to condition chloride salt wastes coming from pyro-processes. Two of them allow to synthesise sodalite, a naturally occurring mineral containing chlorine: the former, starting from Zeolite 4A, which transforms the zeolite into sodalite; the latter, which starts from kaolinite, giving sodalite as well. In addition, a new matrix, termed SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ), has been synthesised. It is able to form different mineral phases which occlude fission metals. The products from the different processes have been fully characterised. In particular the chemical durability of the final waste forms has been determined using the standard product consistency test. According to the results obtained, SAP seems to be a promising matrix for the incorporation of chloride salt wastes from pyro-processes. Financial support from the Nuclear Fission Safety Programme of the European Union (projects ACSEPT, contract FP7-CP-2007- 211 267, and SACSESS, Collaborative Project 323282), as well as from Italian Ministry for Economic Development (Accordo di Programma: Piano Annuale di Realizzazione 2008-2009) is gratefully acknowledged. (authors)

  15. Concepts and Technologies for Radioactive Waste Disposal in Rock Salt

    Directory of Open Access Journals (Sweden)

    Wernt Brewitz

    2007-01-01

    Full Text Available In Germany, rock salt was selected to host a repository for radioactive waste because of its excellent mechanical properties. During 12 years of practical disposal operation in the Asse mine and 25 years of disposal in the disused former salt mine Morsleben, it was demonstrated that low-level wastes (LLW and intermediate-level wastes (ILW can be safely handled and economically disposed of in salt repositories without a great technical effort. LLW drums were stacked in old mining chambers by loading vehicles or emplaced by means of the dumping technique. Generally, the remaining voids were backfilled by crushed salt or brown coal filter ash. ILW were lowered into inaccessible chambers through a borehole from a loading station above using a remote control.Additionally, an in-situ solidification of liquid LLW was applied in the Morsleben mine. Concepts and techniques for the disposal of heat generating high-level waste (HLW are advanced as well. The feasibility of both borehole and drift disposal concepts have been proved by about 30 years of testing in the Asse mine. Since 1980s, several full-scale in-situ tests were conducted for simulating the borehole emplacement of vitrified HLW canisters and the drift emplacement of spent fuel in Pollux casks. Since 1979, the Gorleben salt dome has been investigated to prove its suitability to host the national final repository for all types of radioactive waste. The “Concept Repository Gorleben” disposal concepts and techniques for LLW and ILW are widely based on the successful test operations performed at Asse. Full-scale experiments including the development and testing of adequate transport and emplacement systems for HLW, however, are still pending. General discussions on the retrievability and the reversibility are going on.

  16. Pyrolytic conversion of plastic and rubber waste to hydrocarbons with basic salt catalysts

    Science.gov (United States)

    Wingfield, Jr., Robert C.; Braslaw, Jacob; Gealer, Roy L.

    1985-01-01

    The invention relates to a process for improving the pyrolytic conversion of waste selected from rubber and plastic to low molecular weight olefinic materials by employing basis salt catalysts in the waste mixture. The salts comprise alkali or alkaline earth compounds, particularly sodium carbonate, in an amount of greater than about 1 weight percent based on the waste feed.

  17. Potential for creation of a salt dome following disposal of radioactive waste in a salt layer

    International Nuclear Information System (INIS)

    Fries, G.

    1987-01-01

    The study aims at quantifying the possibility of creation of a salt dome from a salt layer in which heat-emitting radioactive waste would be buried. Volume 1 describes the results of numerical computer simulations, and of laboratory-scale models in centrifuges. Volume 2 envisages, in a geological perspective, the origin of salt domes, the mechanisms of thei formation, and the associated parameters [fr

  18. Potential for creation of a salt dome following disposal of radioactive waste in a salt layer

    International Nuclear Information System (INIS)

    Charo, L.; Habib, P.

    1987-01-01

    The study aims at quantifying the possibility of creation of a salt dome from a salt layer in which heat-emitting radioactive waste would be buried. Volume 1 describes the results of numerical computer simulations, and of laboratory-scale models in centrifuges. Volume 2 envisages, in a geological perspective, the origin of salt domes, the mechanisms of their formation, and the associated parameters [fr

  19. [Determination of Chloride Salt Solution by NIR Spectroscopy].

    Science.gov (United States)

    Zhang, Bin; Chen, Jian-hong; Jiao, Ming-xing

    2015-07-01

    Determination of chloride salt solution by near infrared spectrum plays a very important role in Biomedicine. The near infrared spectrum analysis of Sodium chloride, potassium chloride, calcium chloride aqueous solution shows that the concentration change of chloride salt can affect hydrogen bond, resulting in the variation of near infrared spectrum of water. The temperature influence on NIR spectrum has been decreased by choosing reasonable wavelength range and the wavelength where the temperature effects are zero (isosbestic point). Chlorine salt prediction model was established based on partial least squares method and used for predicting the concentration of the chlorine ion. The impact on near infrared spectrum of the cation ionic radius, the number of ionic charge, the complex effect of ionic in water has also discussed in this article and the reason of every factor are analysed. Experimental results show that the temperature and concentration will affect the near-infrared spectrum of the solution, It is found that the effect of temperature plays the dominant role at low concentrations of chlorine salt; rather, the ionic dominates at high concentration. Chloride complexes are formed in aqueous solution, It has an effect on hydrogen bond of water combining with the cations in chlorine salt solution, Comparing different chloride solutions at the same concentration, the destruction effects of chloride complexes and catnions on the hydrogen bond of water increases in the sequences: CaCl2 >NaCl>KC. The modeling result shows that the determination coefficients (R2) = 99.97%, the root mean square error of cross validation (RM- SECV) = 4.51, and the residual prediction deviation (RPD) = 62.7, it meets the daily requirements of biochemical detection accuracy.

  20. Waste package for a repository located in salt

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.

    1983-01-01

    This paper describes the current status of the waste package designs for salt repositories. The status of the supporting studies of environment definition, corrosion of containment materials, and leaching of waste forms is also presented. Emphasis is on the results obtained in FY 83 and the planned effort in FY 84. 8 references, 3 figures, 1 table

  1. Tank Waste Transport Stability: Summaries of Hanford Slurry and Salt-Solution Studies in FY 2000

    Energy Technology Data Exchange (ETDEWEB)

    Welch, T.D.

    2002-07-08

    This report is a collection of summary articles on FY 2000 studies of slurry transport and salt-well pumping related to Hanford tank waste transfers. These studies are concerned with the stability (steady, uninterrupted flow) of tank waste transfers, a subset of the Department of Energy (DOE) Tanks Focus Area Tank (TFA) Waste Chemistry effort. This work is a collaborative effort of AEA Technology plc, the Diagnostic Instrumentation and Analysis Laboratory at Mississippi State University (DIAL-MSU), the Hemispheric Center for Environmental Technology at Florida International University (HCET-FIU), Numatec Hanford Corporation (NHC), and the Oak Ridge National Laboratory (ORNL). The purpose of this report is to provide, in a single document, an overview of these studies to help the reader identify contacts and resources for obtaining more detailed information and to help promote useful interchanges between researchers and users. Despite over 50 years of experience in transporting radioactive tank wastes to and from equipment and tanks at the Department of Energy's Hanford, Savannah River, and Oak Ridge sites, waste slurry transfer pipelines and process piping become plugged on occasion. At Hanford, several tank farm pipelines are no longer in service because of plugs. At Savannah River, solid deposits in the outlet line of the 2H evaporator have resulted in an unplanned extended downtime. Although waste transfer criteria and guidelines intended to prevent pipeline plugging are in place, they are not always adequate. To avoid pipeline plugging in the future, other factors that are not currently embodied in the transfer criteria may need to be considered. The work summarized here is being conducted to develop a better understanding of the chemical and waste flow dynamics during waste transfer. The goal is to eliminate pipeline plugs by improving analysis and engineering tools in the field that incorporate this understanding.

  2. Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

  3. Summary Report of Comprehensive Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-04

    The inadvertent creation of transuranic waste carrying hazardous waste codes D001 and D002 requires the treatment of the material to eliminate the hazardous characteristics and allow its eventual shipment and disposal at the Waste Isolation Pilot Plant (WIPP). This report documents the effectiveness of two treatment methods proposed to stabilize both the unremediated and remediated nitrate salt waste streams (UNS and RNS, respectively) at Los Alamos National Laboratory (LANL). The two technologies include the addition of zeolite (with and without the addition of water as a processing aid) and cementation. Surrogates were developed to evaluate both the solid and liquid fractions expected from parent waste containers, and both the solid and liquid fractions were tested. Both technologies are shown to be effective at eliminating the characteristic of ignitability (D001), and the addition of zeolite was determined to be effective at eliminating corrosivity (D002), with the preferred option1 of adding zeolite currently planned for implementation at LANL’s Waste Characterization, Reduction, and Repackaging Facility (WCRRF). The course of this work verified the need to evaluate and demonstrate the effectiveness of the proposed remedy for debris material, if required. The evaluation determined that WypAlls, cheesecloth, and Celotex absorbed with saturated nitrate salt solutions exhibit the ignitability characteristic (all other expected debris is not classified as ignitable). Finally, liquid surrogates containing saturated nitrate salts did not exhibit the characteristic of ignitability in their pure form (those neutralized with Kolorsafe and mixed with sWheat did exhibit D001). Sensitivity testing and an analysis were conducted to evaluate the waste form for reactivity. Tests included subjecting surrogate material to mechanical impact, friction, electrostatic discharge and thermal insults. The testing confirmed that the waste does not exhibit the characteristic of

  4. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fisher, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes from the fuel cycle of an integral fast reactor (IFR). The IFR is a sodium-cooled fast reactor with metal fuel. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500 degrees C. This cell has a cadmium anode and a liquid salt electrolyte. The salt will be a low-melting mixture of alkaline and alkaline earth chlorides. This paper discusses one method being considered for immobilizing this treated salt, to disperse it in a portland cement-base motar, which would then be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canisters where it will solidify into a strong, leach-resistant material

  5. Influence of complexing on physicochemical properties of polymer-salt solutions

    International Nuclear Information System (INIS)

    Ostroushko, A.A.; Yushkova, S.M.; Koridze, N.V.; Skobkoreva, N.V.; Zhuravleva, L.I.; Palitskaya, T.A.; Antropova, S.V.; Ostroushko, I.P.; AN SSSR, Moscow

    1993-01-01

    Using the methods of spectrophotometry, viscosimetry, conductometry the influence of salt-polymer complexing processes on physicochemical prperties of aqueous solutions of yttrium, barium, copper nitrates and formates with polyvinyl alcohol was studied. Change of dynamic viscosity, specific electric conductivity of solutions in the process of complexing was shown. Thermal effects of salt-polymer interaction were measured. It is shown that decrease of transition temperature of polymer to plastic state in films, temperature and effective activation energy of salt decomposition is also connected with complexing. Effective values of surface tension on the boundary with air are measured. Coefficients of cation diffusion in polymer-salt solutions are estimated

  6. Cerebral salt wasting following tuberculous meningoencephalitis in an infant

    Directory of Open Access Journals (Sweden)

    Syed Ahmed Zaki

    2012-01-01

    Full Text Available In patients with central nervous system disease, life-threatening hyponatremia can result from either the syndrome of inappropriate secretion of antidiuretic hormone or cerebral salt wasting. Clinical manifestations of the two conditions may be similar, but their pathogeneses and management protocols are different. Cerebral salt wasting syndrome is a disorder in which excessive natriuresis and hyponatremia occurs in patients with intracranial diseases. We report a 6-month-old girl with CSWS associated with tuberculous meningoencephalitis. She was diagnosed as having CSWS on the basis of hypovolemia, polyuria, natriuresis, and the relatively high level of fractional excretion of uric acid. Aggressive replacement of urine salt and water losses using 0.9% or 3% sodium chloride was done. Fludrocortisone was started at 0.1 mg twice daily on the seventh day of admission and was continued for 17 days.

  7. Preliminary study on lithium-salt aqueous solution blanket

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Naruse, Yuji; Yamaoka, Mitsuaki; Ohara, Atsushi; Ono, Kiyoshi; Kobayashi, Shigetada.

    1992-06-01

    Aqueous solution blanket using lithium salts such as LiNO 3 and LiOH have been studied in the US-TIBER program and ITER conceptual design activity. In the JAERI/LANL collaboration program for the joint operation of TSTA (Tritium Systems Test Assembly), preliminary design work of blanket tritium system for lithium ceramic blanket, aqueous solution blanket and liquid metal blanket, have been performed to investigate technical feasibility of tritium demonstration tests using the TSTA. Detail study of the aqueous solution blanket concept have not been performed in the Japanese fusion program, so that this study was carried out to investigate features of its concept and to evaluated its technical problems. The following are the major items studied in the present work: (i) Neutronics of tritium breeding ratio and shielding performance Lithium concentration, Li-60 enrichment, beryllium or lead, composition of structural material/beryllium/solution, heavy water, different lithium-salts (ii) Physicochemical properties of salts Solubility, corrosion characteristics and compatibility with structural materials, radiolysis (iii) Estimation of radiolysis in ITER aqueous solution blanket. (author)

  8. Volume reduction of waste contaminated by fission product elements and plutonium using molten salt combustion

    International Nuclear Information System (INIS)

    McKenzie, D.E.; Grantham, L.F.; Paulson, R.B.

    1979-01-01

    In the Molten Salt Combustion Process, transuranic or β-γ organic waste and air are continuously introduced beneath the surface of a sodium carbonate-containing melt at a temperature of about 800 0 C. Complete combustion of the organic material to carbon dioxide and steam occurs without the conversion of nitrogen to nitrogen oxides. The noxious gases formed by combustion of the chloride, sulfur or phosphorus content of the waste instantly react with the melt to form the corresponding sodium compounds. These compounds as well as the ash and radionuclides are retained in the molten salt. The spent salt is either fused cast into an engineered disposal container or processed to recover salt and plutonium. Molten salt combustion reduces the waste to about 2% of its original volume. Many reactor or reprocessing wastes which cannot be incinerated without difficulty are readily combusted in the molten salt. A 50 kg/hr molten salt combustion system is being designed for the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. Construction of the combustor started during 1977, and combustor startup was scheduled for the spring of 1978

  9. Geochemical processes in marine salt deposits: Their significance and their implications in connection with disposal of radioactive waste within salt domes

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, A G [Goettingen Univ. (Germany, F.R.). Geochemisches Inst.

    1980-01-01

    Attempts to effect permanent disposal of radioactive wastes in marine evaporites should do nothing to disturb, either in the short or the long term, the present relative stability of such bodies of rock. It is necessary to take account of all of the geochemical and physico-chemical reactions known to have been involved in the processes which formed the evaporites before proceeding to an acceptable strategy for disposal of radionucleides. These processes can be represented as three kinds of metamorphism: 1. solution metamorphism, 2. thermal metamorphism, 3. dynamic metamorphism. In all of the evaporite occurrences in Germany such processes have been influential in altering, on occasion significantly, the primary mineralogical composition and have also promoted a considerable degree of transposition of material. Given similar geochemical and physico-chemical premises, these metamorphic processes could become effective now or in the future. It is therefore necessary to discuss the following criteria when examining salt domes as permanent repositories of highly radioactive substances: (1) Temperatures <= 90/sup 0/ +- 10/sup 0/C at the contact between waste containers and rock salt; (2) Temperatures <= 75/sup 0/C within zones of carnallite rocks; (3) Immobilisation of high-level waste in crystalline forms whenever possible; (4) Systems of additional safety barriers around the waste containers or the unreprocessed spent fuel elements. The geochemical and physical effectiveness of the barriers within an evaporite environment must be guaranteed. For example: Ni-Ti-alloys, corundum, ceramic, anhydrite.

  10. Destruction of high explosives and wastes containing high explosives using the molten salt destruction process

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Brummond, W.A.; Pruneda, C.O.

    1992-01-01

    This paper reports the Molten Salt Destruction (MSD) Process which has been demonstrated for the destruction of HE and HE-containing wastes. MSD has been used by Rockwell International and by Anti-Pollution Systems to destroy hazardous wastes. MSD converts the organic constituents (including the HE) of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. In the case of HE-containing mixed wastes, any actinides in the waste are retained in the molten salt, thus converting the mixed wastes into low-level wastes. (Even though the MSD process is applicable to mixed wastes, this paper will emphasize HE-treatment.) The destruction of HE is accomplished by introducing it, together with oxidant gases, into a crucible containing a molten salt, such as sodium carbonate, or a suitable mixture of the carbonates of sodium, potassium, lithium and calcium. The temperature of the molten salt can be between 400 to 900 degrees C. The combustible organic components of the waste react with oxygen to produce carbon dioxide, nitrogen and steam

  11. Salt-occluded zeolite waste forms: Crystal structures and transformability

    International Nuclear Information System (INIS)

    Richardson, J.W. Jr.

    1996-01-01

    Neutron diffraction studies of salt-occluded zeolite and zeolite/glass composite samples, simulating nuclear waste forms loaded with fission products, have revealed complex structures, with cations assuming the dual roles of charge compensation and occlusion (cluster formation). These clusters roughly fill the 6--8 angstrom diameter pores of the zeolites. Samples are prepared by equilibrating zeolite-A with complex molten Li, K, Cs, Sr, Ba, Y chloride salts, with compositions representative of anticipated waste systems. Samples prepared using zeolite 4A (which contains exclusively sodium cations) as starting material are observed to transform to sodalite, a denser aluminosilicate framework structure, while those prepared using zeolite 5A (sodium and calcium ions) more readily retain the zeolite-A structure. Because the sodalite framework pores are much smaller than those of zeolite-A, clusters are smaller and more rigorously confined, with a correspondingly lower capacity for waste containment. Details of the sodalite structures resulting from transformation of zeolite-A depend upon the precise composition of the original mixture. The enhanced resistance of salt-occluded zeolites prepared from zeolite 5A to sodalite transformation is thought to be related to differences in the complex chloride clusters present in these zeolite mixtures. Data relating processing conditions to resulting zeolite composition and structure can be used in the selection of processing parameters which lead to optimal waste forms

  12. Problems of evaluating isotope analysis of concentrated salt solutions in potash mines

    International Nuclear Information System (INIS)

    Schmiedl, H.D.

    1980-01-01

    Three problems of quantitative evaluation of analytic D and 18 O isotope data of concentrated salt solutions are discussed: (1) Consideration of the influence of admixtures of hydrated salts in determining meteoric or marine water fractions in a concentrated salt solution, (2) analytic accuracy and detection limits in determining meteoric water in salt solutions, and (3) processes of isotopic exchange with reservoir rock and sample matrix

  13. Neptunium sorption and co-precipitation of strontium in simulated DWPF salt solution

    International Nuclear Information System (INIS)

    McIntyre, P.F.; Orebaugh, E.G.; King, C.M.

    1988-01-01

    Batch experiments performed using crushed slag saltstone (∼40 mesh) removed >80% of 237 Np from simulated Defense Waste Processing Facility (DWPF) salt solution. The concentration of 237 Np (110 pCi/ml) used was 1000x greater than levels in actual DWPF solutions. Neptunium-239 was used as a tracer and was formed by neutron activation of uranyl nitrate. Results showed that small amounts of crushed saltstone (as little as 0.05 grams), removed >80% of neptunium from 15 ml of simulated DWPF solution after several hours equilibration. The neptunium is sorbed on insoluble carbonates formed in and on the saltstone matrix. Further testing showed that addition of 0.01 and 0.10 ml of 1 molar Ca +2 (ie. Ca (NO 3 ) 2 , CaCl 2 ) into 15 ml of simulated DWPF solution yielded a white carbonate precipitate which also removed >80% of the neptunium after 1 hour equilibration. Further experiments were performed to determine the effectiveness of this procedure to co-precipitate strontium

  14. Hydrogen Generation in Microbial Reverse-Electrodialysis Electrolysis Cells Using a Heat-Regenerated Salt Solution

    KAUST Repository

    Nam, Joo-Youn

    2012-05-01

    Hydrogen gas can be electrochemically produced in microbial reverse-electrodialysis electrolysis cells (MRECs) using current derived from organic matter and salinity-gradient energy such as river water and seawater solutions. Here, it is shown that ammonium bicarbonate salts, which can be regenerated using low-temperature waste heat, can also produce sufficient voltage for hydrogen gas generation in an MREC. The maximum hydrogen production rate was 1.6 m3 H2/m3·d, with a hydrogen yield of 3.4 mol H2/mol acetate at a salinity ratio of infinite. Energy recovery was 10% based on total energy applied with an energy efficiency of 22% based on the consumed energy in the reactor. The cathode overpotential was dependent on the catholyte (sodium bicarbonate) concentration, but not the salinity ratio, indicating high catholyte conductivity was essential for maximizing hydrogen production rates. The direction of the HC and LC flows (co- or counter-current) did not affect performance in terms of hydrogen gas volume, production rates, or stack voltages. These results show that the MREC can be successfully operated using ammonium bicarbonate salts that can be regenerated using conventional distillation technologies and waste heat making the MREC a useful method for hydrogen gas production from wastes. © 2012 American Chemical Society.

  15. Dechlorination and Stabilization of Molten Salt Waste by Using xSiO2-yAl2O3- zP2O5 at Melting Temperature

    International Nuclear Information System (INIS)

    Park, Hwanseo; Kim, Intae; Kim, Hwanyoung; Kim, Joonhyung

    2007-01-01

    Molten salt waste, which is generated from the pyroprocess to separate uranium and trans-uranium elements from spent nuclear fuel, has been interested to researchers in the radioactive waste management. For its final disposal, direct immobilization into a suitable host matrix or indirect solidification by other chemical routes requires the control of chlorides and its volatility since molten salt wastes mainly consist of volatile metal chlorides. Glass-bonded sodalite (Na 6 M 2 Al 6 Si 6 O 24 Cl 2 , 1-5) suggested by Argonne National Laboratory (ANL), to the present, could be a practical solution to the immobilization of this waste, where waste form can be fabricated at about 915 .deg., lower than the melting temperature of many borosilicate glasses ( -1150 .deg.). A wet dechlorination to oxides or a thermal conversion into borate glass was suggested to remove Cl from salt waste (6-7) and it seemed that the preference of radionuclides for the intended chemical conversions or immobilizations described above could be hardly accomplished or failed, except the phosphate precipitation method suggested by Volkovich and his co-workers (8). Our research group suggested a novel method to treat molten salt waste, named GRSS (Gel-Route Stabilization/Solidification) using Si-P-Al system as a gel-forming system. This showed little vaporization during high temperature process and good leach resistance on Cs and Sr. As another method, this study suggested a method to stabilize molten salt wastes by using xSiO 2 -yAl 2 O 3 - zP 2 O 5 material. GRSS method is considered as a 'reaction system' to completely convert salt waste into stable product while the inorganic material used in this study is a stabilizer for salt wastes. Using this material, this study investigated the reactivity on different metal chlorides, thermal stability, leach-resistance and etc

  16. Nuclear waste in sea or salt? No, wrong

    International Nuclear Information System (INIS)

    Damveld, H.; Van Duin, S.; Bannink, D.

    1994-04-01

    Eighteen years of successful action against ocean dumping and storage of nuclear waste in salt domes are reviewed for the Dutch situation. The aim of this book is to hand some support to those who want to act against trial borings, in particular the people living close to the most important salt domes in the Netherlands: Ternaard, Zuidwending, Pieterburen, Onstwedde, Winschoten, Schoonlo and Gasselte-Drouwen. In 1976 the Interdepartmental Commission on Nuclear Energy with its subcommission Radioactive Substances (ICK-RAS) was installed, along with a number of working groups, responsible for research. From 1978 onwards ocean dumping operations were accompanied by blockades and legal procedures, which led to a situation of the last dumping in 1982. The Dutch government then focused on nuclear waste storage in salt domes for which the OPLA research program was started. OPLA is the Dutch abbreviation for Storage on Land. The final report (phase 1 and 1a) of OPLA was published on 15 October 1993 as annex to the Dossier Nuclear Energy of the Dutch government. It has been decided that phase 1a is not followed by trial drillings, as planned before. Some critical remarks are made regarding the rounds of public participation and the notion of permanent retrievability of stored nuclear waste. Extensive use has been made of documentation from the Dutch government and parliament, and other literature and information sources

  17. Costs for off-site disposal of nonhazardous oil field wastes: Salt caverns versus other disposal methods

    Energy Technology Data Exchange (ETDEWEB)

    Veil, J.A.

    1997-09-01

    According to an American Petroleum Institute production waste survey reported on by P.G. Wakim in 1987 and 1988, the exploration and production segment of the US oil and gas industry generated more than 360 million barrels (bbl) of drilling wastes, more than 20 billion bbl of produced water, and nearly 12 million bbl of associated wastes in 1985. Current exploration and production activities are believed to be generating comparable quantities of these oil field wastes. Wakim estimates that 28% of drilling wastes, less than 2% of produced water, and 52% of associated wastes are disposed of in off-site commercial facilities. In recent years, interest in disposing of oil field wastes in solution-mined salt caverns has been growing. This report provides information on the availability of commercial disposal companies in oil-and gas-producing states, the treatment and disposal methods they employ, and the amounts they charge. It also compares cavern disposal costs with the costs of other forms of waste disposal.

  18. Decontaminated salt disposal as saltcrete in a landfill. Technical data summary

    International Nuclear Information System (INIS)

    1982-01-01

    This technical data summary presents a reference process for immobilizing decontaminated salt solution from the 200-Area waste storage tanks with cement, and disposing of the final waste material (called saltcrete) by burial in trenches. The saltcrete will be protected from leaching by clay and will be placed at least 3 meters above the historic high water table and beneath at least 5 meters of soil overburden. The decontaminated salt solution is a waste material which remains after the bulk of the radionuclides have been removed from waste tank supernate. This removal is effected by contacting the waste supernate with sodium tetraphenyl boron (Na-TPB) and sodium titanate (NaTi 2 O 5 H). These materials remove (by precipitation) most of the 137 Cs and 90 Sr as well as many other radioactive and non-radioactive elements. These precipitates, along with many other sludges which reside in the HLW tanks will be incorporated in borosilicate glass for eventual disposal in a geologic repository. An ion exchange process will also be used for removal of 99 Tc. The decontaminated salt solution has sufficiently low levels of radioactivity that it can be disposed of on-site. The scope of the curent effort is to describe a process for blending decontaminated salt solution with cement to form a saltcrete product which has dimensional stability and relatively low leachability. The process is to be capable of solidifying 10 gpm of supernate. About 100 million gallons of salt solution is to be solidified

  19. Temperature distributions in a salt formation used for the ultimate disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Ploumen, P.

    1980-01-01

    In the Federal Republic of Germany the works on waste disposal is focussed on the utilization of a salt formation for ultimate disposal of radioactive wastes. Heat released from the high-level waste will be dissipated in the salt and the surrounding geologic formations. The occuring temperature distributions will be calculated with computer codes. A survey of the developed computer codes will be shown; the results for a selected example, taking into account the loading sequence of the waste, the mine ventilation as well as an air gap between the waste and the salt, will be discussed. Furthermore it will be shown that by varying the disposal parameters, the maximum salt temperature can be below any described value. (Auth.)

  20. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a 238 Pu waste treatment technology that should be developed for volume reduction and recovery of 238 Pu and as an alternative to the transport and permanent disposal of 238 Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious 238 Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of 238 Pu contaminated wastes is reduced to 30 drums. Further 238 Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious 238 Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose 238 Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment

  1. Development and characterization of new high-level waste form containing LiCl KCl eutectic salts for achieving waste minimization from pyroprocessing

    International Nuclear Information System (INIS)

    Cho, Yong Zun; Kim, In Tae; Park, Hwan Seo; Ahn, Byeung Gil; Eun, Hee Chul; Son, Seock Mo; Ah, Su Na

    2011-12-01

    The purpose of this project is to develop new high level waste (HLW) forms and fabrication processes to dispose of active metal fission products that are removed from electrorefiner salts in the pyroprocessing based fuel cycle. The current technology for disposing of active metal fission products in pyroprocessing involves non selectively discarding of fission product loaded salt in a glass-bonded sodalite ceramic waste form. Selective removal of fission products from the molten salt would greatly minimize the amount of HLW generated and methods were developed to achieve selective separation of fission products during a previous I NERI research project (I NERI 2006 002 K). This I NERI project proceeds from the previous project with the development of suitable waste forms to immobilize the separated fission products. The Korea Atomic Energy Research Institute (KAERI) has focused primarily on developing these waste forms using surrogate waste materials, while the Idaho National Laboratory (INL) has demonstrated fabrication of these waste forms using radioactive electrorefiner salts in hot cell facilities available at INL. Testing and characterization of these radioactive materials was also performed to determine the physical, chemical, and durability properties of the waste forms

  2. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 16. Repository preconceptual design studies: BPNL waste forms in salt

    International Nuclear Information System (INIS)

    1978-04-01

    This volume, Volume 16, ''Repository Preconceptual Design Studies: BPNL Waste Forms in Salt,'' is one of a 23 volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provide a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The waste forms assumed to arrive at the repository were supplied by Battelle Pacific Northwest Laboratories (BPNL). The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/17, ''Drawings for Repository Preconceptual Design Studies: BPNL Waste Forms in Salt.''

  3. Modeling of waste/near field interactions for a waste repository in bedded salt: the Dynamic Network (DNET) model

    International Nuclear Information System (INIS)

    Cranwell, R.M.

    1983-01-01

    The Fuel Cycle Risk Analysis Division of Sandia National Laboratories has been funded by the US Nuclear Regulatory Commission to develop a methodology for use in assessing the long-term risk from the disposal of radioactive wastes in deep geologic formations. As part of this program, the Dynamic Network (DNET) model was developed to investigate waste/near field interactions associated with the disposal of radioactive wastes in bedded salt formations. The model is a quasi-multi-dimensional network model with capabilities for simulating processes such as fluid flow, heat transport, salt dissolution, salt creep, and the effects of thermal expansion and subsedence on the rock units surrounding the repository. The use of DNET has been demonstrated in the analysis of a hypothetical disposal site containing a bedded salt formation as the host medium for the repository. An example of this demonstration analysis is discussed. Furthermore, the outcome of sensitivity analyses performed on the DNET model are presented

  4. Interim performance specifications for conceptual waste-package designs for geologic isolation in salt repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  5. Salt type and concentration affect the viscoelasticity of polyelectrolyte solutions

    Science.gov (United States)

    Turkoz, Emre; Perazzo, Antonio; Arnold, Craig B.; Stone, Howard A.

    2018-05-01

    The addition of small amounts of xanthan gum to water yields viscoelastic solutions. In this letter, we show that the viscoelasticity of aqueous xanthan gum solutions can be tuned by different types of salts. In particular, we find that the decrease in viscoelasticity not only depends, as is known, on the salt concentration, but also is affected by the counterion ionic radius and the valence of the salt.

  6. Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mccloy, John S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lepry, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rodriguez, Carmen P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Windisch, Charles F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westman, Matthew P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rieck, Bennett T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lang, Jesse B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Olszta, Matthew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, David A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-01-17

    The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

  7. Chemical modeling of nuclear waste repositories in the salt repository project

    International Nuclear Information System (INIS)

    Jansen, G.; Raines, G.E.; Kircher, J.F.; Hubbard, N.

    1985-01-01

    Salt deposits contain small amounts of water as brine in fluid inclusions in halite and in hydrous minerals, e.g., clays, kieserite (MgSO 4 . H 2 O) and carnallite (KMgCl 3 . 6H 2 O). For the candidate salt deposits, the total amounts of water as volume % brine are: Palo Duro Basin, Texas, approximately 1.8; Paradox Basin, Utah, approximately 5.0 for the carnallite-marker zone, and less than approximately 0.5 below this zone; Gulf Coast salt domes, less than 0.15. For the Palo Duro and Paradox salt, the brines are Mg-rich (approximately 20,000 mg/L to approximately 100,000 mg/L) and sometimes Ca-rich (up to about 20,000 mg/L) NaCl brines. Brine migration calculations have been made using calculations of the time-variant thermal gradient around the waste packages and conservatively high brine volumes in the salt (5.0 volume % for the Texas and Utah sites and 0.5 volume % for the Gulf Coast) as input data. The maximum amounts of brine that eventually migrate to each waste package are about 1.0m 3 (for 5.0 volume % brine) and 0.2m 3 (for 0.5 volume % brine). With current conceptual designs for waste package overpacks (10 to 15 cm thick low-carbon steel), the waste package is not breached by uniform corrosion within 10,000 years. In brines this material thus far shows only uniform corrosion. For the expected conditions, where the brine is provided solely by brine migration, the brine is consumed by reaction with the iron of the overpack nearly as fast as it migrates to the waste package. Therefore, for the expected conditions, data about corrosion rates, radiolysis, etc., are not important. However, it is essential that accurate volumes of in-migrating brine can be calculated

  8. Solution, thermal and optical properties of bis(pyridinium salt)s as ionic liquids

    International Nuclear Information System (INIS)

    Jo, Tae Soo; Koh, Jung Jae; Han, Haesook; Bhowmik, Pradip K.

    2013-01-01

    Bis(pyridinium salt)s containing different alkyl chain lengths and various organic counterions were prepared by the ring-transmutation reaction of bis(pyrylium tosylate) with aliphatic amines in dimethyl sulfoxide at 130–135 °C for 18 h and their tosylate counterions were exchanged to other anions such as triflimide, methyl orange, and dioctyl sulfosuccinate by the metathesis reaction in a common organic solvent. Their chemical structures were established by using 1 H, 19 F, and 13 C NMR spectra. The thermal properties of bis(pyridinium salt)s were studied by DSC and TGA measurements. Some of the dicationic salts provided low melting points below 100 °C and some of them displayed amorphous properties. Polarized optical microscopy studies revealed the crystal structures prior to melting temperatures in some cases. Their optical properties were examined by using UV–Vis and photoluminescent spectrometers; and they emitted blue light both in the solution and solid states regardless of their microstructures, counterions, and the polarity of organic solvents. However, most of these salts exhibited hypsochromic shifts in their emission peaks in the solid state when compared with those of their solution spectra. Due to unique properties of methyl orange anion as a pH indicator, two of the salts showed different color change in varying concentrations of triflic acid in common organic solvents, demonstrating their potential use as an acid sensor in methanol, acetonitrile and acetone. Highlights: ► Luminescent dicationic salts were synthesized by ring-transmutation and metathesis reactions. ► Thermal and optical properties of dicationic salts are affected by the size of anion structures. ► Due to the methyl orange counterions, some dicationic salts showed pH- sensing property

  9. CO2 Capture from Flue Gas using Amino Acid Salt Solutions

    DEFF Research Database (Denmark)

    Lerche, Benedicte Mai; Stenby, Erling Halfdan; Thomsen, Kaj

    2009-01-01

    difficult. Amino acid salt solutions have emerged as an alternative to the alkanolamine solutions. A number of advantages make amino acid salt solutions attractive solvents for CO2 capture from flue gas. In the present study CO2 absorption in aqueous solutions of 0.5 M potassium glycinate and 0.5 M...

  10. Trial storage of high-level waste in the Asse II salt mine

    International Nuclear Information System (INIS)

    1984-01-01

    This report covers a second phase of the work performed by GSF and KfK in the Asse II salt mine, with a view to disposal of radioactive waste in salt formations. New items of the research were geophysical investigations of the behaviour of heated salt and preparation of a trial storage in the Asse II salt mine

  11. Mass transfer and transport in salt repositories

    International Nuclear Information System (INIS)

    Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-02-01

    Salt is a unique rock isolation of nuclear waste because it is ''dry'' and nearly impermeable. In this paper we summarize some mass-transfer and transport analyses of salt repositories. First we analyses brine migration. Heating by high-level waste can cause brine in grain boundaries to move due to pressure-gradients. We analyze brine migration treating salt as a thermoelastic solid and found that brine migration is transient and localized. We use previously developed techniques to estimate release rates from waste packages by diffusion. Interbeds exist in salt and may be conduits for radionuclide migration. We analyze steady-state migration due to brine flow in the interbed, as a function of the Peclet number. Then we analyze transient mass transfer, both into the interbed and directly to salt, due only to diffusion. Finally we compare mass transfer rates of a waste cylinder in granite facing a fracture and in salt facing an interbed. In all cases, numerical illustrations of the analytic solution are given. 10 refs., 4 figs., 3 tabs

  12. Comparison of temperature calculations for an arbitrary high-level waste disposal configuration in salt formations

    International Nuclear Information System (INIS)

    Kevenaar, J.W.A.M.; Janssen, L.G.J.; Ploumen, P.; Winske, P.

    1979-05-01

    The objective of this report is the comparison of the results of temperature analyses for an arbitrary high-level radioactive waste disposal configuration in salt formations. The analyses were carried out at the RWTH and ECN. The computer programs used are based on finite difference and finite element techniques. From the local temperature analyses that were intended to check the solution techniques, it could be concluded that both finite difference and finite elements are capable to analyse this type of problems. From the global temperature analyses it could be concluded that both analysis approaches: temperature dependent and iteratively determined temperature independent material properties, are suited to analyse the global temperature distribution in the salt formation

  13. Mercury separation from aqueous wastes

    International Nuclear Information System (INIS)

    Taylor, P.A.; Klasson, K.T.; Corder, S.L.

    1995-07-01

    This project is providing an assessment of new sorbents for removing mercury from wastes at US Department of Energy sites. Four aqueous wastes were chosen for lab-scale testing; a high-salt, acidic waste currently stored at Idaho National Engineering Laboratory (INEL); a high-salt, alkaline waste stored at the Savannah River Site (SRS); a dilute lithium hydroxide solution stored at the Oak Ridge Y-12 Plant; and a low-salt, neutral groundwater generated at the Y-12 Plant. Eight adsorbents have been identified for testing, covering a wide range of cost and capability. Screening tests have been completed, which identified the most promising adsorbents for each waste stream. Batch isotherm tests have been completed using the most promising adsorbents, and column tests are in progress. Because of the wide range of waste compositions tested, no one adsorbent is effective in all of these waste streams. Based on loading capacity and compatibility with the waste solutions. the most effective adsorbents identified to date are SuperLig 618 for the INEL tank waste stimulant; Mersorb followed by lonac SR-3 for the SRS tank waste stimulant; Durasil 70 and Ionac SR-3) for the LIOH solution; and lonac SR-3 followed by lonac SR-4 and Mersorb for the Y-12 groundwater

  14. Some geotechnical problems related to underground waste disposal in salt formations

    International Nuclear Information System (INIS)

    Berest, P.

    1993-01-01

    Nuclear waste disposal in deep salt formations is an option considered by several countries. Rock salt is a very impervious medium, but can be easily leached; selection of an appropriate disposal formation must account for natural protections of the formation as regards water movements. It must be checked that such initially favourable characteristics will not be affected by the existence of shafts and galleries, or by the important heat output generated by vitrified wastes. The discussion is uneasy, for a comprehensive rheological model for rock salt is difficult to set and to be extrapolated to large time scales; some methodological problems are raised by use of numerical computations. (author). 22 refs., 2 figs

  15. The thermo-mechanical behaviour of a salt dome with a heat-generating waste repository

    International Nuclear Information System (INIS)

    Janssen, L.G.J.; Prij, J.; Kevenaar, J.W.A.M.; Jong, C.J.T.; Klok, J.; Beemsterboer, C.

    1984-01-01

    This report reviews the analytical work on the disposal of radioactive waste in salt domes performed at ECN in the period 1 January 1980 to 31 December 1982. Chapter 4 in the main report covers the global temperature and deformation analyses of the salt dome and the surrounding rocks. The attached three topical reports cover self-contained parts of the study. The computer program TASTE developed to analyse, at acceptable cost and with, for engineering purposes, sufficient accuracies, the temperature rises in the salt dome due to the stored heat-generating waste is described in Annex 1. Annex 2 gives a description of the extended finite element program GOLIA. The program has been extended to make it suitable for the creep analysis of salt domes with repositories of heat-generating waste. The study on the closing and sealing of boreholes wit heat-generating waste is reported in Annex 3

  16. Backfill barriers for nuclear waste repositories in salt

    Energy Technology Data Exchange (ETDEWEB)

    Nowak, E J; Odoj, R; Merz, E [eds.

    1981-06-01

    Backfill materials were evaluated for containment of radionuclides, chemical modification of brine, and sensitivity to hydrothermal conditions. Experimental conditions were relevant to nuclear waste isolation in bedded salt. They were based on geologic conditions at the site of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, USA. Conclusions are: backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier; bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved ions, or the sorption properties of radionuclide species; effects of irradiation from high level waste should be investigated.

  17. Ionic solubility and solutal advection governed augmented evaporation kinetics of salt solution pendant droplets

    Science.gov (United States)

    Jaiswal, Vivek; Harikrishnan, A. R.; Khurana, Gargi; Dhar, Purbarun

    2018-01-01

    The presence of dispersed inclusions is known to modify the interfacial characteristics in liquids by adsorption-desorption of the ions at interfaces. The present article reports the influencing role of dissolved ions in a polar fluid on its evaporation dynamics. The evaporation dynamics of pendant droplets of aqueous solutions of variant simple salts and concentrations have been experimentally studied. The presence of salts is observed to enhance the evaporation rate (obeying the classical D2 law), and the enhancement has been found to hold a direct proportionality to the concentration of the dissolved salt. Furthermore, it is observed that the degree of enhancement in the evaporation rate is also directly proportional to the solubility of the salt in question. The phenomenon is explained based on the chemical kinetics and thermodynamics of hydration of the ionic species in the polar fluid. The classical evaporation rate constant formulation is found to be inadequate in modeling the enhanced species transport. Additional probing via particle image velocimetry reveals augmented internal circulation within the evaporating salt based drops compared to pure water. Mapping the dynamic surface tension reveals that a salt concentration gradient is generated between the bulk and periphery of the droplet and it could be responsible for the internal advection cells visualized. A thermo-solutal Marangoni and Rayleigh convection based mathematical formulation has been put forward, and it is shown that the enhanced solute-thermal convection could play a major role in enhanced evaporation. The internal circulation mapped from experiments is found to be in good quantitative agreement with the model predictions. Scaling analysis further reveals that the stability of the solutal Marangoni convection surpasses the thermal counterpart with higher salt concentration and solubility. The present article sheds insight into the possible domineering role of conjugate thermohydraulic and

  18. Long-term interactions of full-scale cemented waste simulates in salt solutions. Summary report; Langzeit-Wechselwirkungen von zementierten Abfallsimulaten im Originalmassstab mit Salzloesungen. Zusammenfassender Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard; Borkel, Christoph; Metz, Volker [Karlsruher Institut fuer Technologie (KIT), Karlsruhe (Germany). Inst. fuer Nukleare Entsorgung (INE)

    2015-12-04

    On March 13,.2013 the Federal Office of Radiation Protection (BfS) published a note that the responsible group of the Helmholtz Zentrums Muenchen had finished the experiments in the socalled leaching test room at the 490 m level of the Asse II mine. In this room, the previous operator the Gesellschaft fuer Strahlen- und Umweltforschung mbH (GSF) carried out leaching and corrosion experiments with cemented full-scale samples. These experiments were performed since 1979 requested by the licensing authorities. With respect to the safety case for the Asse salt mine it was a need to demonstrate the transferability of results obtained by laboratory samples to real waste forms and to investigate the effects of the industrial cementation process an the properties of the waste forms. A research program was initiated by the Nuclear Research Centre Karlsruhe (today Karlsruhe Institute of Technology, KIT) and the Institut fuer Tieflagerung of the Gesellschaft fuer Strahlenforschung m.b.H. (GSF). Since 1996 the scientific supervision of the experiments were dedicated to the Institute for Nuclear Waste Disposal (INE) of KIT. Until 2013, the corroding solutions were sampled several times. In 2006 four full-scale samples were retrieved and investigated with respect to variations of the solids. After termination of the experiments in January 2013, radioactively doped samples were transferred to KIT-INE for final evaluations. The present report summarizes the background and objectives of the experiments as well as the results of the solutions and solid state analyses.

  19. Safety evaluation of geological disposal concepts for low and medium-level wastes in rock-salt (Pacoma project)

    International Nuclear Information System (INIS)

    Prij, J.; Van Dalen, A.; Roodbergen, H.A.; Slagter, W.; Van Weers, A.W.; Zanstra, D.A.; Glasbergen, P.; Koester, H.W.; Lembrechts, J.F.; Nijhof-Pan, I.; Slot, A.F.M.

    1991-01-01

    In the framework of the Performance Assessment of Confinements for MLW and Alpha Waste (PACOMA) the disposal options dealing with rock-salt are studied by GSF and ECN (with subcontract to RIVM). The overall objectives of these studies are to develop and demonstrate procedures for the radiological safety assessment of a deep repository in salt formations. An essential objective is to show how far appropriate choices of the repository design parameters can improve the performances of the whole system. The research covers two waste inventories (the Dutch OPLA and the PACOMA reference inventory), two disposal techniques (conventional and solution mining) and three types of formations (salt dome, pillow and bedded salt). An important part of the research has been carried out in the socalled VEOS project within the framework of the Dutch OPLA study. The methodology used in the consequence analysis is a deterministic one. The models and calculation tools used to perform the consequence analysis are the codes: EMOS, METROPOL and BIOS. The results are expressed in terms of dose rates and doses to individuals as well as to groups. Detailed information with respect to the input data and the results obtained with the three codes is given in three annexes to this final report

  20. Pressure-induced brine migration in consolidated salt in a repository

    International Nuclear Information System (INIS)

    Hwang, Y.; Chambre, P.L.; Lee, W.W.L.; Pigford, T.H.

    1987-06-01

    This report describes a mathematical model for brine migration through intact salt near a radioactive waste package emplaced in salt. Solutions indicate limited movement following ten years emplacement

  1. Ion-exchange properties of cesium and strontium into zeolites from sodium salt solutions

    International Nuclear Information System (INIS)

    Kanno, Takuji; Hashimoto, Hiroyuki; Ohtani, Tozo.

    1978-01-01

    The ion-exchange properties of cesium and strontium into zeolite from sodium salt solution has been studied in zeolite A, zeolite X, zeolite Y, mordenite and clinoptilolite. The distribution of cesium into mordenite from about 1 -- 2 M sodium chloride and sodium hydroxide solutions is considerably larger than that into zeolite A. The distribution coefficient for 2 M solution of sodium salts was about 300. Therefore, the separation of cesium from sodium salt solution is possible by using mordenite. The distribution of strontium into zeolites form 1 -- 2 M solutions of sodium chloride and sodium nitrate were in the order of zeolite A>zeolite X>zeolite Y asymptoticaly equals mordenite. The distribution coefficient of 230 was obtained for 1 M solutions of sodium salts. The anion in solutions had no effect on the distribution of cesium and strontium into zeolite from sodium salt solution. (author)

  2. An improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds

    International Nuclear Information System (INIS)

    Eun, H.C.; Cho, Y.Z.; Lee, T.K.; Kim, I.T.; Park, G.I.; Lee, H.S.

    2013-01-01

    In this paper, an improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds was performed to determine optimum operating conditions. It was very important to maintain the pressure in the distillation chamber below 10 Torr for a high efficiency (salt recovery >99 %) of the salt distillation. This required increasing the salt vaporization and condensation rates in the distillation system. It was confirmed that vaporization and condensation rates could be improved controlling the given temperature of top of the condensation chamber. In the distillation tests of the salt wastes containing rare earth compounds, the operation time at a given temperature was greatly reduced changing the given temperature of top of the condensation chamber from 780 to 700 deg C. (author)

  3. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    Energy Technology Data Exchange (ETDEWEB)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a {sup 238}Pu waste treatment technology that should be developed for volume reduction and recovery of {sup 238}Pu and as an alternative to the transport and permanent disposal of {sup 238}Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious {sup 238}Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of {sup 238}Pu contaminated wastes is reduced to 30 drums. Further {sup 238}Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious {sup 238}Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose {sup 238}Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment.

  4. Rock salt as a medium for long-term isolation of radioactive wastes - a reassessment

    International Nuclear Information System (INIS)

    Chaturvedi, L.

    1985-01-01

    Rock salt has been regarded as a suitable medium for the permanent disposal of high and medium level radioactive wastes since the National Academy of Sciences recommended it in 1957. As a result of detained site-specific studies conducted for the Waste Isolation Pilot Plant (WIPP) project in New Mexico, however, several potential problems which are unique to bedded salt deposits have emerged. These include 1) the need to delineate the extent and rate of past dissolution and projections for the future, 2) the origin and significance of brines often found underlying the salt beds, 3) the rate and volume of migration of brine from the salt crystals towards the heat producing waste canisters, 4) the creep rates and implications for retrievability, and 5) the existence of potash and oil and gas resources with implications of human intrusion in the future. These questions will also be faced for sites in salt domes with added complications due to more complex structure and hydrology. The experience at WIPP shows that the site characterization process for high level waste repositories in bedded or dome salt should aim at identifying the important issues of site suitability early in the process and a clear program should be established to address these issues

  5. Reuse of hydroponic waste solution.

    Science.gov (United States)

    Kumar, Ramasamy Rajesh; Cho, Jae Young

    2014-01-01

    Attaining sustainable agriculture is a key goal in many parts of the world. The increased environmental awareness and the ongoing attempts to execute agricultural practices that are economically feasible and environmentally safe promote the use of hydroponic cultivation. Hydroponics is a technology for growing plants in nutrient solutions with or without the use of artificial medium to provide mechanical support. Major problems for hydroponic cultivation are higher operational cost and the causing of pollution due to discharge of waste nutrient solution. The nutrient effluent released into the environment can have negative impacts on the surrounding ecosystems as well as the potential to contaminate the groundwater utilized by humans for drinking purposes. The reuse of non-recycled, nutrient-rich hydroponic waste solution for growing plants in greenhouses is the possible way to control environmental pollution. Many researchers have successfully grown several plant species in hydroponic waste solution with high yield. Hence, this review addresses the problems associated with the release of hydroponic waste solution into the environment and possible reuse of hydroponic waste solution as an alternative resource for agriculture development and to control environmental pollution.

  6. Salt Repository Project waste emplacement mode decision paper: Revison 1

    International Nuclear Information System (INIS)

    1987-08-01

    This paper provides a recommendation as to the mode of waste emplacement to be used as the current basis for site characterization activity for the Deaf Smith County, Texas, high level nuclear waste repository site. It also presents a plan for implementing the recommendation so as to provide a high level of confidence in the project's success. Since evaluations of high-level waste disposal in geologic repositories began in the 1950s, most studies emplacement in salt formations employed the vertical orientation for emplacing waste packages in boreholes in the floor of the underground facility. This orientation was used in trials at Project Salt Vault in the 1960s. The Waste Isolation Pilot Plant (WIPP) has recently settled on a combination of vertical and horizontal modes for various waste types. This paper analyzes the information available and develops a project position upon which to base current site characterization activities. The position recommended is that the SRP should continue to use the vertical waste emplacement mode as the reference design and to carry the horizontal mode as a ''passive'' alternative. This position was developed based upon the conclusions of a decision analysis, risk assessment, and cost/schedule impact assessment. 52 refs., 6 figs., 1 tab

  7. Waste form dissolution in bedded salt

    International Nuclear Information System (INIS)

    Kaufman, A.M.

    1980-01-01

    A model was devised for waste dissolution in bedded salt, a hydrologically tight medium. For a typical Spent UnReprocessed Fuel (SURF) emplacement, the dissolution rate wll be diffusion limited and will rise to a steady state value after t/sub eq/ approx. = 250 (1+(1-epsilon 0 ) K/sub D//epsilon 0 ) (years) epsilon 0 is the overpack porosity and K/sub d/ is the overpack sorption coefficient. The steady state dissolution rate itself is dominated by the solubility of UO 2 . Steady state rates between 5 x 10 -5 and .5 (g/year) are achievable by SURF emplacements in bedded salt without overpack, and rates between 5 x 10 -7 and 5 x 10 -3 (g/year) with an overpack having porosity of 10 -2

  8. Performance analysis of conceptual waste package designs in salt repositories

    International Nuclear Information System (INIS)

    Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

    1984-01-01

    A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

  9. Electrodialysis-based separation process for salt recovery and recycling from waste water

    Science.gov (United States)

    Tsai, Shih-Perng

    1997-01-01

    A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid.

  10. Immobilization of LiCl-Li 2 O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, salt occlusion and hydrothermal processes were used to make chlorosodalite through reaction with a high-LiCl salt simulating a waste stream following pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and aide in densification. Hydrothermal processes included reaction of the salt simulant in an acid digestion vessel with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  11. Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Science.gov (United States)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  12. Thermophysical property characterization of aqueous amino acid salt solution containing serine

    International Nuclear Information System (INIS)

    Navarro, Shanille S.; Leron, Rhoda B.; Soriano, Allan N.; Li, Meng-Hui

    2014-01-01

    Highlights: • Thermophysical properties of aqueous potassium and sodium salt solutions of serine were studied. • Density, viscosity, refractive index and electrolytic conductivity of the solution were measured. • The concentrations of amino acid salt ranges from x 1 = 0.009 to 0.07. • The temperature range studied was (298.15 to 343.15) K. • The measured data were represented satisfactorily by using the applied correlations. - Abstract: Thermophysical property characterization of aqueous potassium and sodium salt solutions containing serine was conducted in this study; specifically the system’s density, refractive index, electrical conductivity, and viscosity. Measurements were obtained over a temperature range of (298.15 to 343.15) K and at normal atmospheric pressure. Composition range from x 1 = 0.009 to 0.07 for aqueous potassium and sodium salt solutions containing serine was used. The sensitivity of the system’s thermophysical properties on temperature and composition variation were discussed and correlated based on the equations proposed for room temperature ionic liquids. The density, viscosity, and refractive index measurements of the aqueous systems were found to decrease as the temperature increases at fixed concentration and the values increase as the salt concentration increases (water composition decreases) at fixed temperature. Whereas, a different trend was observed for the electrical conductivity data; at fixed concentration, the conductivity values increase as the temperature increases and at fixed temperature, its value generally increases as the salt concentration increases but only to a certain level (specific concentration) wherein the conductivity of the solution starts to decrease when the concentration of the salt is further increased. Calculation results show that the applied models were satisfactory in representing the measured properties in the aqueous amino acid salt solution containing serine

  13. Preconceptual design of a salt splitting process using ceramic membranes

    Energy Technology Data Exchange (ETDEWEB)

    Kurath, D.E.; Brooks, K.P.; Hollenberg, G.W.; Clemmer, R. [Pacific Northwest National Lab., Richland, WA (United States); Balagopal, S.; Landro, T.; Sutija, D.P. [Ceramatec, Inc., Salt Lake City, UT (United States)

    1997-01-01

    Inorganic ceramic membranes for salt splitting of radioactively contaminated sodium salt solutions are being developed for treating U. S. Department of Energy tank wastes. The process consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON) membranes. The primary NaSICON compositions being investigated are based on rare- earth ions (RE-NaSICON). Potential applications include: caustic recycling for sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes; reducing the volume of low-level wastes volume to be disposed of; adjusting pH and reducing competing cations to enhance cesium ion exchange processes; reducing sodium in high-level-waste sludges; and removing sodium from acidic wastes to facilitate calcining. These applications encompass wastes stored at the Hanford, Savannah River, and Idaho National Engineering Laboratory sites. The overall project objective is to supply a salt splitting process unit that impacts the waste treatment and disposal flowsheets and meets user requirements. The potential flowsheet impacts include improving the efficiency of the waste pretreatment processes, reducing volume, and increasing the quality of the final waste disposal forms. Meeting user requirements implies developing the technology to the point where it is available as standard equipment with predictable and reliable performance. This report presents two preconceptual designs for a full-scale salt splitting process based on the RE-NaSICON membranes to distinguish critical items for testing and to provide a vision that site users can evaluate.

  14. Preconceptual design of a salt splitting process using ceramic membranes

    International Nuclear Information System (INIS)

    Kurath, D.E.; Brooks, K.P.; Hollenberg, G.W.; Clemmer, R.; Balagopal, S.; Landro, T.; Sutija, D.P.

    1997-01-01

    Inorganic ceramic membranes for salt splitting of radioactively contaminated sodium salt solutions are being developed for treating U. S. Department of Energy tank wastes. The process consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NaSICON) membranes. The primary NaSICON compositions being investigated are based on rare- earth ions (RE-NaSICON). Potential applications include: caustic recycling for sludge leaching, regenerating ion exchange resins, inhibiting corrosion in carbon-steel tanks, or retrieving tank wastes; reducing the volume of low-level wastes volume to be disposed of; adjusting pH and reducing competing cations to enhance cesium ion exchange processes; reducing sodium in high-level-waste sludges; and removing sodium from acidic wastes to facilitate calcining. These applications encompass wastes stored at the Hanford, Savannah River, and Idaho National Engineering Laboratory sites. The overall project objective is to supply a salt splitting process unit that impacts the waste treatment and disposal flowsheets and meets user requirements. The potential flowsheet impacts include improving the efficiency of the waste pretreatment processes, reducing volume, and increasing the quality of the final waste disposal forms. Meeting user requirements implies developing the technology to the point where it is available as standard equipment with predictable and reliable performance. This report presents two preconceptual designs for a full-scale salt splitting process based on the RE-NaSICON membranes to distinguish critical items for testing and to provide a vision that site users can evaluate

  15. Solidification of salt solutions on a horizontal surface

    International Nuclear Information System (INIS)

    Braga, S.L.; Viskanta, R.

    1990-01-01

    The freezing of water-salt solutions on a horizontal wall is investigated experimentally and theoretically. The growth of the solid-liquid region is observed for NaCl - H sub(2)O and N H sub(4)Cl - H sub(2)O systems under different temperature and concentration conditions. A unidirectional mathematical model is used to predict the solidification process. The transport of heat is by diffusion, and convection is absent. The mass diffusion is neglected and the growth of crystal is governed by the transport of heat. In all experiments, the solution salt concentration is smaller than the eutectic composition, and the wall temperature is higher than the eutectic temperature. The predicted temperature and salt concentration profiles, as well as the interface position, are compared with experimental data. (author)

  16. Treatment Study Plan for Nitrate Salt Waste Remediation Revision 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Juarez, Catherine L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vigil-Holterman, Luciana R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Naranjo, Felicia Danielle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    The two stabilization treatment methods that are to be examined for their effectiveness in the treatment of both the unremediated and remediated nitrate salt wastes include (1) the addition of zeolite and (2) cementation. Zeolite addition is proposed based on the results of several studies and analyses that specifically examined the effectiveness of this process for deactivating nitrate salts. Cementation is also being assessed because of its prevalence as an immobilization method used for similar wastes at numerous facilities around the DOE complex, including at Los Alamos. The results of this Treatment Study Plan will be used to provide the basis for a Resource Conservation and Recovery Act (RCRA) permit modification request of the LANL Hazardous Waste Facility Permit for approval by the New Mexico Environment Department-Hazardous Waste Bureau (NMED-HWB) of the proposed treatment process and the associated facilities.

  17. Novel waste printed circuit board recycling process with molten salt

    OpenAIRE

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450?470??C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, a...

  18. Radioactive waste management solutions

    International Nuclear Information System (INIS)

    Siemann, Michael

    2015-01-01

    One of the more frequent questions that arise when discussing nuclear energy's potential contribution to mitigating climate change concerns that of how to manage radioactive waste. Radioactive waste is produced through nuclear power generation, but also - although to a significantly lesser extent - in a variety of other sectors including medicine, agriculture, research, industry and education. The amount, type and physical form of radioactive waste varies considerably. Some forms of radioactive waste, for example, need only be stored for a relatively short period while their radioactivity naturally decays to safe levels. Others remain radioactive for hundreds or even hundreds of thousands of years. Public concerns surrounding radioactive waste are largely related to long-lived high-level radioactive waste. Countries around the world with existing nuclear programmes are developing longer-term plans for final disposal of such waste, with an international consensus developing that the geological disposal of high-level waste (HLW) is the most technically feasible and safe solution. This article provides a brief overview of the different forms of radioactive waste, examines storage and disposal solutions, and briefly explores fuel recycling and stakeholder involvement in radioactive waste management decision making

  19. Assessment of tectonic hazards to waste storage in interior-basin salt domes

    International Nuclear Information System (INIS)

    Kehle, R.

    1979-01-01

    Salt domes in the northern Gulf of Mexico may make ideal sites for storage of radioactive waste because the area is tectonically quiet. The stability of such salt domes and the tectonic activity are discussed

  20. Process for the recovery of alkali metal salts from aqueous solutions thereof

    International Nuclear Information System (INIS)

    Vitner, J.

    1984-01-01

    In an integrated process for the recovery of alkakli metal phenates and carboxylates from aqueous solutions thereof, the aqueous solution is spray dried and the drying gas stream is contacted with an aqueous alkali metal salt solution which dissolves the particles of the alkali metal salt that were entrained in the drying gas stream. The salt-free inert gas stream is then dried, heated, and returned to the spray dryer

  1. Electrodialysis-ion exchange for the separation of dissolved salts

    International Nuclear Information System (INIS)

    Baroch, C.J.; Grant, P.J.

    1995-01-01

    The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. This report describes the process of electrodialysis-ion exchange (EDIX) for treating aqueous wastes streams consisting of nitrates, sodium, organics, heavy metals, and radioactive species

  2. Low disposal of radioactive wastes in salt formations of the Federal Republic of Germaany

    International Nuclear Information System (INIS)

    Albrecht, E.

    1980-01-01

    The salt formations of northern Europe are generally suitable for the storage of radioactive wastes because the region is largely free from earthquakes and the salt formations known as diapires provide effective hydrological sealing. The Federal Republic of Germany employed the Asse Salt Mine of Lower Saxony for research in waste storage. More recently, exploratory work has begun on the construction of a large recycling and disposal plant at the Gorleben salt dome. The geology, hydrology, rock mechanics, and seismicity of the two sites are briefly discussed, including a discussion of experiences gained so far from the Asse site. 11 refs

  3. Extraction, scrub, and strip test results for the salt waste processing facility caustic side solvent extraction solvent example

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-01

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.9, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  4. Analysis of corrosion data for carbon steels in simulated salt repository brines and acid chloride solutions at high temperatures

    International Nuclear Information System (INIS)

    Diercks, D.R.; Hull, A.B.; Kassner, T.F.

    1988-03-01

    Carbon steel is currently the leading candidate material for fabrication of a container for isolation of high level nuclear waste in a salt repository. Since brine entrapped in the bedded salt can migrate to the container by several transport processes, corrosion is an important consideration in the long-term performance of the waste package. A detailed literature search was performed to compile relevant corrosion data for carbon steels in anoxic acid chloride solutions, and simulated salt repository brines at temperatures between ∼ 20 and 400 0 C. The hydrolysis of Mg 2+ ions in simulated repository brines containing high magnesium concentrations causes acidification at temperatures above 25 0 C, which, in turn, influences the protective nature of the magnetite corrosion product layer on carbon steel. The corrosion data for the steels were analyzed, and an analytical model for general corrosion was developed to calculate the amount of penetration (i.e., wall thinning) as a function of time, temperature, and the pressure of corrosion product hydrogen than can build up during exposure in a closed system (e.g., a sealed capsule). Both the temperature and pressure dependence of the corrosion rate of steels in anoxic acid chloride solutions indicate that the rate-controlling partial reaction is the cathodic reduction of water to form hydrogen. Variations in the composition and microstructure of the steels or the concentration of the ionic species in the chloride solutions (provided that they do not change the pH significantly) do not appear to strongly influence the corrosion rate

  5. Removal of salt from high-level waste tanks by density-driven circulation or mechanical agitation

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    Twenty-two high-level waste storage tanks at the Savannah River Plant are to be retired in the tank replacement/waste transfer program. The salt-removal portion of this program requires dissolution of about 19 million liters of salt cake. Steam circulation jets were originally proposed to dissolve the salt cake. However, the jets heated the waste tank to 80 to 90 0 C. This high temperature required a long cooldown period before transfer of the supernate by jet, and increased the risk of stress-corrosion cracking in these older tanks. A bench-scale investigation at the Savannah River Laboratory developed two alternatives to steam-jet circulation. One technique was density-driven circulation, which in bench tests dissolved salt at the same rate as a simulated steam circulation jet but at a lower temperature. The other technique was mechanical agitation, which dissolved the salt cake faster and required less fresh water than either density-driven circulation or the simulated steam circulation jet. Tests in an actual waste tank verified bench-scale results and demonstrated the superiority of mechanical agitation

  6. Mechanism of pitting corrosion prevention by nitrite in carbon steel exposed to dilute salt solutions. 1998 annual progress report

    International Nuclear Information System (INIS)

    Zapp, P.E.; Zee, J. van.

    1998-01-01

    'The overall goal of this project is to develop a fundamental understanding of the role of nitrite in preventing the breakdown of protective oxide(s) on carbon steel and the onset of pitting. Pitting corrosion of carbon steel exposed to dilute alkaline salt solutions can be induced by nitrate, sulfate, and chloride ions and is prevented by sufficient concentration of nitrite. A significant example of this material/electrolyte system is the storage and processing of DOE''s high-level radioactive liquid waste in carbon steel tanks. Added nitrite in the waste has a considerable downstream impact on the immobilization of the waste in a stable glass form. Waste tank integrity and glass production efficiency may benefit from the fundamental understanding of nitrite''s role in preventing pitting. This report summarizes progress after approximately six months of effort in this three-year EMSP project. Initial experimental and theoretical work has focused on the electrochemical behavior of carbon steel in simplified non-radioactive solutions that simulate complex dilute radioactive waste solutions. These solutions contain corrosion-inducing species such as nitrate and chloride and the corrosion-inhibiting nitrite at moderately alkaline pHs. The electrochemical behavior of interest here is that of the open-circuit potential of the steel specimen at equilibrium in the experimental electrolyte and the measures of the steel''s passivity and passivity breakdown.'

  7. Molt salts reactors capacity for wastes incineration and energy production

    International Nuclear Information System (INIS)

    David, S.; Nuttin, A.

    2005-01-01

    The molten salt reactors present many advantages in the framework of the IV generation systems development for the energy production and/or the wastes incineration. After a recall of the main studies realized on the molten salt reactors, this document presents the new concepts and the identified research axis: the MSRE project and experience, the incinerators concepts, the thorium cycle. (A.L.B.)

  8. Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-07-07

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

  9. Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-07-07

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

  10. Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

    International Nuclear Information System (INIS)

    Hansen, Francis D.; Kuhlman, Kristopher L.; Sobolik, Steven R.

    2016-01-01

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

  11. Test procedures for salt rock

    International Nuclear Information System (INIS)

    Dusseault, M.B.

    1985-01-01

    Potash mining, salt mining, design of solution caverns in salt rocks, disposal of waste in salt repositories, and the use of granular halite backfill in underground salt rock mines are all mining activities which are practised or contemplated for the near future. Whatever the purpose, the need for high quality design parameters is evident. The authors have been testing salt rocks in the laboratory in a number of configurations for some time. Great care has been given to the quality of sample preparation and test methodology. This paper describes the methods, presents the elements of equipment design, and shows some typical results

  12. Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

    International Nuclear Information System (INIS)

    Hull, A.B.; Williams, L.B.

    1985-07-01

    Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs

  13. Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Hull, A.B.; Williams, L.B.

    1985-07-01

    Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs.

  14. Test procedures for polyester immobilized salt-containing surrogate mixed wastes

    International Nuclear Information System (INIS)

    Biyani, R.K.; Hendrickson, D.W.

    1997-01-01

    These test procedures are written to meet the procedural needs of the Test Plan for immobilization of salt containing surrogate mixed waste using polymer resins, HNF-SD-RE-TP-026 and to ensure adequacy of conduct and collection of samples and data. This testing will demonstrate the use of four different polyester vinyl ester resins in the solidification of surrogate liquid and dry wastes, similar to some mixed wastes generated by DOE operations

  15. User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

    International Nuclear Information System (INIS)

    Lindner, E.N.; St John, C.M.; Hart, R.D.

    1984-02-01

    SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

  16. Risk assessment of nonhazardous oil-field waste disposal in salt caverns.

    Energy Technology Data Exchange (ETDEWEB)

    Elcock, D.

    1998-03-05

    In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes.

  17. Risk assessment of nonhazardous oil-field waste disposal in salt caverns

    International Nuclear Information System (INIS)

    Elcock, D.

    1998-01-01

    In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes

  18. FRACTIONAL CRYSTALLIZATION OF HANFORD SINGLE-SHELL TANK WASTES. A MODELING APPROACH

    International Nuclear Information System (INIS)

    HAMILTON, D.W.

    2006-01-01

    The Hanford site has 149 underground single-shell tanks (SST) storing mostly soluble, multi-salt, mixed wastes resulting from Cold War era weapons material production. These wastes must be retrieved and the salts immobilized before the tanks can be closed to comply with an overall site closure consent order entered into by the U.S. Department of Energy (DOE), the Environmental Protection Agency, and Washington State. Water will be used to retrieve the wastes and the resulting solution will be pumped to the proposed treatment process where a high curie (primarily 137 Cs) waste fraction will be separated from the other waste constituents. The separated waste streams will then be vitrified to allow for safe storage as an immobilized high level waste, or low level waste, borosilicate glass. Fractional crystallization, a common unit operation for production of industrial chemicals and pharmaceuticals, was proposed as the method to separate the salt wastes; it works by evaporating excess water until the solubilities of various species in the solution are exceeded (the solubility of a particular species depends on its concentration, temperature of the solution, and the presence of other ionic species in the solution). By establishing the proper conditions, selected pure salts can be crystallized and separated from the radioactive liquid phase

  19. Protein removal from waste brines generated during ham salting through acidification and centrifugation.

    Science.gov (United States)

    Gutiérrez-Martínez, Maria del Rosario; Muñoz-Guerrero, Hernán; Alcaína-Miranda, Maria Isabel; Barat, José Manuel

    2014-03-01

    The salting step in food processes implies the production of large quantities of waste brines, having high organic load, high conductivity, and other pollutants with high oxygen demand. Direct disposal of the residual brine implies salinization of soil and eutrophication of water. Since most of the organic load of the waste brines comes from proteins leaked from the salted product, precipitation of dissolved proteins by acidification and removal by centrifugation is an operation to be used in waste brine cleaning. The aim of this study is optimizing the conditions for carrying out the separation of proteins from waste brines generated in the pork ham salting operation, by studying the influence of pH, centrifugal force, and centrifugation time. Models for determining the removal of proteins depending on the pH, centrifugal force, and time were obtained. The results showed a high efficacy of the proposed treatment for removing proteins, suggesting that this method could be used for waste brine protein removal. The best pH value to be used in an industrial process seems to be 3, while the obtained results indicate that almost 90% of the proteins from the brine can be removed by acidification followed by centrifugation. A further protein removal from the brine should have to be achieved using filtrating techniques, which efficiency could be highly improved as a consequence of the previous treatment through acidification and centrifugation. Waste brines from meat salting have high organic load and electrical conductivity. Proteins can be removed from the waste brine by acidification and centrifugation. The total protein removal can be up to 90% of the initial content of the waste brine. Protein removal is highly dependent on pH, centrifugation rate, and time. © 2014 Institute of Food Technologists®

  20. Thermal decomposition of nitrate salts liquid waste for the lagoon sludge treatment

    International Nuclear Information System (INIS)

    Hwang, D. S.; Oh, J. H.; Kim, Y. K.; Lee, K. Y.; Choi, Y. D.; Hwang, S. T.; Park, J. H.

    2004-01-01

    This study investigated the thermal decomposition property of nitrate salts liquid waste which is produced in a series of the processes for the sludge treatment. Thermal decomposition property was analyzed by TG/DTA and XRD. Most ammonium nitrate in the nitrate salts liquid waste was decomposed at 250 .deg. C and calcium nitrate was decomposed and converted into calcium oxide at 550 .deg. C. Sodium nitrate was decomposed at 700 .deg. C and converted into sodium oxide which reacts with water easily. But sodium oxide was able to convert into a stable compound by adding alumina. Therefore, nitrate salts liquid waste can be treated by two steps as follows. First, ammonium nitrate is decomposed at 250 .deg. C. Second, alumina is added in residual solid sodium nitrate and calcium nitrate and these are decomposed at 900 .deg. C. Final residue consists of calcium oxide and Na 2 O.Al 2 O 3 and can be stored stably

  1. Removal of radioruthenium from alkaline intermediate level radioactive waste solution : a laboratory investigation

    International Nuclear Information System (INIS)

    Samanta, S.K.; Theyyunni, T.K.

    1994-01-01

    Various methods were investigated in the laboratory for the removal of radioruthenium from alkaline intermediate level radioactive waste solutions of reprocessing plant origin. The methods included batch equilibration with different ion exchangers and sorbents, column testing and chemical precipitation. A column method using zinc-activated carbon mixture and a chemical precipitation method using ferrous salt along with sodium sulphite were found to be promising for plant scale application. (author). 10 refs., 3 figs., 7 tabs

  2. Electrodialysis-ion exchange for the separation of dissolved salts

    Energy Technology Data Exchange (ETDEWEB)

    Baroch, C.J. [Wastren, Inc., Westminster, CO (United States); Grant, P.J. [Wastren, Inc., Hummelstown, PA (United States)

    1995-10-01

    The Department of Energy generates and stores a significant quantity of low level, high level, and mixed wastes. As some of the DOE facilities are decontaminated and decommissioned, additional and possibly different forms of wastes will be generated. A significant portion of these wastes are aqueous streams containing acids, bases, and salts, or are wet solids containing inorganic salts. Some of these wastes are quite dilute solutions, whereas others contain large quantities of nitrates either in the form of dissolved salts or acids. Many of the wastes are also contaminated with heavy metals, radioactive products, or organics. Some of these wastes are in storage because a satisfactory treatment and disposal processes have not been developed. There is considerable interest in developing processes that remove or destroy the nitrate wastes. Electrodialysis-Ion Exchange (EDIX) is a possible process that should be more cost effective in treating aqueous waste steams. This report describes the EDIX process.

  3. Review: Water recovery from brines and salt-saturated solutions: operability and thermodynamic efficiency considerations for desalination technologies.

    Science.gov (United States)

    Vane, Leland M

    2017-03-08

    When water is recovered from a saline source, a brine concentrate stream is produced. Management of the brine stream can be problematic, particularly in inland regions. An alternative to brine disposal is recovery of water and possibly salts from the concentrate. This review provides an overview of desalination technologies and discusses the thermodynamic efficiencies and operational issues associated with the various technologies particularly with regard to high salinity streams. Due to the high osmotic pressures of the brine concentrates, reverse osmosis, the most common desalination technology, is impractical. Mechanical vapor compression which, like reverse osmosis, utilizes mechanical work to operate, is reported to have the highest thermodynamic efficiency of the desalination technologies for treatment of salt-saturated brines. Thermally-driven processes, such as flash evaporation and distillation, are technically able to process saturated salt solutions, but suffer from low thermodynamic efficiencies. This inefficiency could be offset if an inexpensive source of waste or renewable heat could be used. Overarching issues posed by high salinity solutions include corrosion and the formation of scales/precipitates. These issues limit the materials, conditions, and unit operation designs that can be used.

  4. Alteration of non-metallic barriers and evolution of solution chemistry in salt formations in Germany

    International Nuclear Information System (INIS)

    Herbert, H.J.; Becker, D.; Hagemann, S.; Meyer, Th.; Noseck, U.; Rubel, A.; Mauke, R.; Wollrath, J.

    2005-01-01

    Different Engineered Barrier Systems (EBS) materials considered in Germany for the sealing of repositories in salt formations are presented. Their long term behaviour in terms of interactions with salt solutions is discussed and evaluated. The discussed EBS materials are crushed salt, self sealing salt backfill, bentonite and salt concrete. Whereas the knowledge concerning the geochemical, geomechanical, hydrological and thermal behavior of crushed salt and salt concrete is well advanced further research is needed for other EBS materials. The self healing salt backfill has also been investigated in depth recently. In order to fully qualify this material large scale in situ experiments are still needed. The present knowledge on compacted bentonites in a salt environment is not yet sufficient for reliable predictions of the long-term performance in salt formations. The sealing concept of the low- and intermediate-level Radioactive Waste Repository Morsleben (ERAM) in a former rock salt and potash mine is presented. This concept is based on cementitious materials, i.e. salt concrete. The geochemical stability of different salt concretes in contact with brines expected in ERAM is addressed. It is shown how the results from leaching experiments and geochemical modelling are used in the safety analyses and how the chemical boundary conditions prevailing in the EBS influence the development of the permeability of the sealing system and thus control the radionuclide release. As a result of modelling the behaviour of the seals in the safety assessment it is shown, that the seals are corroded within a time span of about 20 000 years. The influence of the uncertainty in the model parameters on the safety of the repository was assessed by a variation of the initial permeability of the seal. The maximum dose rate resulting from the radionuclide release from ERAM is nearly independent of the variation of the initial permeability within four orders of magnitude. (authors)

  5. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  6. Salt weathering in Egyptian limestone after laboratory simulations with continuous flow of salt solutions at different temperatures

    Science.gov (United States)

    Aly, Nevin; Gomez-Heras, Miguel; Hamed, Ayman; Alvarez de Buergo, Monica

    2013-04-01

    weathering in Egyptian limestone after laboratory simulations with continuous flow of salt solutions at different temperatures Nevin Aly Mohamed (1), Miguel Gomez - Heras(2), Ayman Hamed Ahmed (1), and Monica Alvarez de Buergo(2). (1) Faculty of Pet. & Min. Engineering- Suez Canal University, Suez, Egypt, (2) Instituto de Geociencias (CSIC-UCM) Madrid. Spain. Limestone is one of the most frequent building stones in Egypt and is used since the time of ancient Egyptians and salt weathering is one of the main threats to its conservation. Most of the limestone used in historical monuments in Cairo is a biomicrite extracted from the Mid-Eocene Mokattam Group. During this work, cylindrical samples (2.4 cm diameter and approx. 4.8 cm length) were subjected, in a purpose-made simulation chamber, to simulated laboratory weathering tests with fixed salt concentration (10% weight NaCl solution), at different temperatures, which were kept constant throughout each test (10, 20, 30, 40 oC). During each test, salt solutions flowed continuously imbibing samples by capilarity. Humidity within the simulation chamber was reduced using silica gel to keep it low and constant to increase evaporation rate. Temperature, humidity inside the simulation chamber and samples weight were digitally monitored during each test. Results show the advantages of the proposed experimental methodology using a continuous flow of salt solutions and shed light on the effect of temperature on the dynamics of salt crystallization on and within samples. Research funded by mission sector of high education ministry, Egypt and Geomateriales S2009/MAT-1629.

  7. Brine migration in salt and its implications in the geologic disposal of nuclear waste

    International Nuclear Information System (INIS)

    Jenks, G.H.; Claiborne, H.C.

    1981-12-01

    This report respresents a comprehensive review and analysis of available information relating to brine migration in salt surrounding radioactive waste in a salt repository. The topics covered relate to (1) the characteristics of salt formations and waste packages pertinent to considerations of rates, amounts, and effects of brine migration, (2) experimental and theoretical information on brine migration, and (3) means of designing to minimize any adverse effects of brine migration. Flooding, brine pockets, and other topics were not considered, since these features will presumably be eliminated by appropriate site selection and repository design. 115 references

  8. Options Assessment Report: Treatment of Nitrate Salt Waste at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-17

    This report documents the methodology used to select a method of treatment for the remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The method selected should treat the containerized waste in a manner that renders the waste safe and suitable for transport and final disposal in the Waste Isolation Pilot Plant (WIPP) repository, under specifications listed in the WIPP Waste Acceptance Criteria (DOE/CBFO, 2013). LANL recognizes that the results must be thoroughly vetted with the New Mexico Environment Department (NMED) and that a modification to the LANL Hazardous Waste Facility Permit is a necessary step before implementation of this or any treatment option. Likewise, facility readiness and safety basis approvals must be received from the Department of Energy (DOE). This report presents LANL’s preferred option, and the documentation of the process for reaching the recommended treatment option for RNS and UNS waste, and is presented for consideration by NMED and DOE.

  9. Options assessment report: Treatment of nitrate salt waste at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-16

    This report documents the methodology used to select a method of treatment for the remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The method selected should treat the containerized waste in a manner that renders the waste safe and suitable for transport and final disposal in the Waste Isolation Pilot Plant (WIPP) repository, under specifications listed in the WIPP Waste Acceptance Criteria (DOE/CBFO, 2013). LANL recognized that the results must be thoroughly vetted with the New Mexico Environment Department (NMED) and the a modification to the LANL Hazardous Waste Facility Permit is a necessary step before implementation of this or any treatment option. Likewise, facility readiness and safety basis approvals must be received from the Department of Energy (DOE). This report presents LANL's preferred option, and the documentation of the process for reaching the recommended treatment option for RNS and UNS waste, and is presented for consideration by NMED and DOE.

  10. Tetraphenylborate Solubility in High Ionic Strength Salt Solutions

    International Nuclear Information System (INIS)

    Serkiz, S.M.; Ginn, J.D.; Jurgensen, A.R.

    1998-04-01

    Solubility of sodium and potassium salts of the tetraphenylborate ion (TPB) in simulated Savannah River Site High Level Waste was investigated. Data generated from this study allow more accurate predictions of TPB solubility at the In-Tank Precipitation (ITP) facility. Because previous research showed large deviations in the observed solubility of TPB salts when compared with model predictions, additional data were generated to better understand the solubility of TPB in more complex systems of high ionic strength and those containing both potassium and sodium. These data allow evaluation of the ability of current models to accurately predict equilibrium TPB concentrations over the range of experimental conditions investigated in this study

  11. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

  12. Conceptual waste package interim product specifications and data requirements for disposal of glass commercial high-level waste forms in salt geologic repositories

    International Nuclear Information System (INIS)

    1983-10-01

    The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses and regulatory requirements become available. 13 references, 1 figure

  13. A Study of Analytical Solution for the Special Dissolution Rate Model of Rock Salt

    Directory of Open Access Journals (Sweden)

    Xin Yang

    2017-01-01

    Full Text Available By calculating the concentration distributions of rock salt solutions at the boundary layer, an ordinary differential equation for describing a special dissolution rate model of rock salt under the assumption of an instantaneous diffusion process was established to investigate the dissolution mechanism of rock salt under transient but stable conditions. The ordinary differential equation was then solved mathematically to give an analytical solution and related expressions for the dissolved radius and solution concentration. Thereafter, the analytical solution was fitted with transient dissolution test data of rock salt to provide the dissolution parameters at different flow rates, and the physical meaning of the analytical formula was also discussed. Finally, the influential factors of the analytical formula were investigated. There was approximately a linear relationship between the dissolution parameters and the flow rate. The effects of the dissolution area and initial volume of the solution on the dissolution rate equation of rock salt were computationally investigated. The results showed that the present analytical solution gives a good description of the dissolution mechanism of rock salt under some special conditions, which may provide a primary theoretical basis and an analytical way to investigate the dissolution characteristics of rock salt.

  14. Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.H.

    2001-07-11

    The HLW salt waste (salt cake and supernate) now stored at the SRS must be treated to remove insoluble sludge solids and reduce the soluble concentration of radioactive cesium radioactive strontium and transuranic contaminants (principally Pu and Np). These treatments will enable the salt solution to be processed for disposal as saltstone, a solid low-level waste.

  15. Distillation Separation of Hydrofluoric Acid and Nitric Acid from Acid Waste Using the Salt Effect on Vapor-Liquid Equilibrium

    Science.gov (United States)

    Yamamoto, Hideki; Sumoge, Iwao

    2011-03-01

    This study presents the distillation separation of hydrofluoric acid with use of the salt effect on the vapor-liquid equilibrium for acid aqueous solutions and acid mixtures. The vapor-liquid equilibrium of hydrofluoric acid + salt systems (fluorite, potassium nitrate, cesium nitrate) was measured using an apparatus made of perfluoro alkylvinylether. Cesium nitrate showed a salting-out effect on the vapor-liquid equilibrium of the hydrofluoric acid-water system. Fluorite and potassium nitrate showed a salting-in effect on the hydrofluoric acid-water system. Separation of hydrofluoric acid from an acid mixture containing nitric acid and hydrofluoric acid was tested by the simple distillation treatment using the salt effect of cesium nitrate (45 mass%). An acid mixture of nitric acid (5.0 mol · dm-3) and hydrofluoric acid (5.0 mol · dm-3) was prepared as a sample solution for distillation tests. The concentration of nitric acid in the first distillate decreased from 5.0 mol · dm-3 to 1.13 mol · dm-3, and the concentration of hydrofluoric acid increased to 5.41 mol · dm-3. This first distillate was further distilled without the addition of salt. The concentrations of hydrofluoric acid and nitric acid in the second distillate were 7.21 mol · dm-3 and 0.46 mol · dm-3, respectively. It was thus found that the salt effect on vapor-liquid equilibrium of acid mixtures was effective for the recycling of acids from acid mixture wastes.

  16. Molten salt combustion of radioactive wastes

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKenzie, D.E.; Richards, W.L.; Oldenkamp, R.D.

    1976-01-01

    The Atomics International Molten Salt Combustion Process reduces the weight and volume of combustible β-γ contaminated transuranic waste by utilizing air in a molten salt medium to combust organic materials, to trap particulates, and to react chemically with any acidic gases produced during combustion. Typically, incomplete combustion products such as hydrocarbons and carbon monoxide are below detection limits (i.e., 3 ) is directly related to the sodium chloride vapor pressure of the melt; >80% of the particulate is sodium chloride. Essentially all metal oxides (combustion ash) are retained in the melt, e.g., >99.9% of the plutonium, >99.6% of the europium, and >99.9% of the ruthenium are retained in the melt. Both bench-scale radioactive and pilot scale (50 kg/hr) nonradioactive combustion tests have been completed with essentially the same results. Design of three combustors for industrial applications are underway

  17. Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

    International Nuclear Information System (INIS)

    Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

    2011-01-01

    This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

  18. Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

    2011-12-01

    This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

  19. Mercury separation from mixed wastes. Annual report

    International Nuclear Information System (INIS)

    Taylor, P.A.; Klasson, K.T.; Corder, S.L.; Carlson, T.R.; McCandless, K.R.

    1995-11-01

    This is an assessment of new sorbents for removing Hg from wastes at US DOE sites. Four aqueous wastes were used for the laboratory tests: a simulant of a high-salt, acidic waste currently stored at INEL, a simulant of a high-salt, alkaline waste stored at Savannah River (SRS), a dilute LiOH solution stored at Y-12, and a low-salt, neutral groundwater generated at Y-12. Eight adsorbents covering a wide range of cost and capability were tested. Screening tests identified the most promising adsorbents, and column tests were performed using at least two adsorbents for each waste stream. No one adsorbent is effective in all of these waste streams. Based on loading capacity and compatibility, the most effect adsorbents to date are SuperLig 618 for the INEL tank waste simulant, Mersorb and Ionac SR-3 for the SRS tank waste simulant, Durasil 70 and Ionac SR-3 for the LiOH solution, and Ionac SR-3, followed by Ionac SR-4 and Mersorb, for the Y-12 groundwater

  20. Recovery of gold from hydrometallurgical leaching solution of electronic waste via spontaneous reduction by polyaniline

    Directory of Open Access Journals (Sweden)

    Yuanzhao Wu

    2017-08-01

    Full Text Available The present study is primarily designed to develop an environmentally-benign approach for the recovery of precious metals, especially gold, from the ever increasingly-discarded electronic wastes (e-waste. By coupling the metal reduction process with an increase in the intrinsic oxidation state of the aniline polymers, and the subsequent re-protonation and reduction of the intrinsically oxidized polymer to the protonated emeraldine (EM salt, polyaniline (PANi films and polyaniline coated cotton fibers are able to recover metallic gold from acid/halide leaching solutions of electronic wastes spontaneously and sustainably. The current technique, which does not require the use of extensive extracting reagents or external energy input, can recover as much as 90% of gold from the leaching acidic solutions. The regeneration of polyaniline after gold recovery, as confirmed by the X-ray photoelectron spectroscopy measurements, promises the continuous operation using the current approach. The as-recovered elemental gold can be further concentrated and purified by incineration in air.

  1. Possible salt mine sites for radioactive waste disposal in the northeastern states

    Energy Technology Data Exchange (ETDEWEB)

    Landes, K.K.

    1972-06-30

    The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further.

  2. Possible salt mine sites for radioactive waste disposal in the northeastern states

    International Nuclear Information System (INIS)

    Landes, K.K.

    1972-01-01

    The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further

  3. Hydrous mineral dehydration around heat-generating nuclear waste in bedded salt formations.

    Science.gov (United States)

    Jordan, Amy B; Boukhalfa, Hakim; Caporuscio, Florie A; Robinson, Bruce A; Stauffer, Philip H

    2015-06-02

    Heat-generating nuclear waste disposal in bedded salt during the first two years after waste emplacement is explored using numerical simulations tied to experiments of hydrous mineral dehydration. Heating impure salt samples to temperatures of 265 °C can release over 20% by mass of hydrous minerals as water. Three steps in a series of dehydration reactions are measured (65, 110, and 265 °C), and water loss associated with each step is averaged from experimental data into a water source model. Simulations using this dehydration model are used to predict temperature, moisture, and porosity after heating by 750-W waste canisters, assuming hydrous mineral mass fractions from 0 to 10%. The formation of a three-phase heat pipe (with counter-circulation of vapor and brine) occurs as water vapor is driven away from the heat source, condenses, and flows back toward the heat source, leading to changes in porosity, permeability, temperature, saturation, and thermal conductivity of the backfill salt surrounding the waste canisters. Heat pipe formation depends on temperature, moisture availability, and mobility. In certain cases, dehydration of hydrous minerals provides sufficient extra moisture to push the system into a sustained heat pipe, where simulations neglecting this process do not.

  4. Resistance of Coatings for Boiler Components of Waste-to-Energy Plants to Salt Melts Containing Copper Compounds

    Science.gov (United States)

    Galetz, Mathias Christian; Bauer, Johannes Thomas; Schütze, Michael; Noguchi, Manabu; Cho, Hiromitsu

    2013-06-01

    The accelerating effect of heavy metal compounds on the corrosive attack of boiler components like superheaters poses a severe problem in modern waste-to-energy plants (WTPs). Coatings are a possible solution to protect cheap, low alloyed steel substrates from heavy metal chloride and sulfate salts, which have a relatively low melting point. These salts dissolve many alloys, and therefore often are the limiting factor as far as the lifetime of superheater tubes is concerned. In this work the corrosion performance under artificial salt deposits of different coatings, manufactured by overlay welding, thermal spraying of self-fluxing as well as conventional systems was investigated. The results of our studies clearly demonstrate the importance of alloying elements such as molybdenum or silicon. Additionally, the coatings have to be dense and of a certain thickness in order to resist the corrosive attack under these severe conditions.

  5. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 8. Repository preconceptual design studies: salt

    International Nuclear Information System (INIS)

    1978-04-01

    This volume, Volume 8 ''Repository Preconceptual Design Studies: Salt,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area, and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/9, ''Drawings for Repository Preconceptual Design Studies: Salt.''

  6. Glovebox design requirements for molten salt oxidation processing of transuranic waste

    International Nuclear Information System (INIS)

    Ramsey, K.B.; Acosta, S.V.; Wernly, K.D.

    1998-01-01

    This paper presents an overview of potential technologies for stabilization of 238 Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from 238 Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented

  7. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.5

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with safety evaluation as part of the investigations regarding a repository for high-level waste in a salt dome. It is volume 5 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. Two characteristics of the waste are of special importance for the safety evaluation: the encasing of the waste in steel casks with 15 cm thick walls affording protection against corrosion, protecting the surroundings against radiation, and protecting the glass cylinders from mechanical damage resulting from the pressure at the bottom of the disposal hole, and the modest generation of heat in the waste at the time of disposal resulting in a maximum temperature increase in the salt close to the waste of approx. 40 deg. C. These characteristics proved to considerably improve the safety margin with respect to unforeseen circumstances. The character of the salt dome and of the salt in the proposed disposal area offers in itself good protection against contact with the ground water outside the dome. The relatively large depth of 1200 and 2500 m of the salt surface also means that neither dome nor disposal facility will be appreciably influenced by glaciations or earthquakes. The chalk above the proposed disposal area is very tight and to retain radioactive matter effectively even in the precence of high concentrations of NaCL. The safety investigations included a number of natural processes and probable events such as the segregation of crystal water from overlooked salt minerals, faulty sealings of disposal holes, permeable fault zones in the chalk overlying the dome, the risk in connection with human penetration into the dome. These conditions will neither lead to the destruction of the waste casks or to the release of waste from the dome. Leaching of a cavern is the only situation which proved to result in a release of radioactive material to the biosphere, but the resulting doses was found to be small

  8. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

  9. Synthetic salt cake standards for analytical laboratory quality control

    International Nuclear Information System (INIS)

    Schilling, A.E.; Miller, A.G.

    1980-01-01

    The validation of analytical results in the characterization of Hanford Nuclear Defense Waste requires the preparation of synthetic waste for standard reference materials. Two independent synthetic salt cake standards have been prepared to monitor laboratory quality control for the chemical characterization of high-level salt cake and sludge waste in support of Rockwell Hanford Operations' High-Level Waste Management Program. Each synthetic salt cake standard contains 15 characterized chemical species and was subjected to an extensive verification/characterization program in two phases. Phase I consisted of an initial verification of each analyte in salt cake form in order to determine the current analytical capability for chemical analysis. Phase II consisted of a final characterization of those chemical species in solution form where conflicting verification data were observed. The 95 percent confidence interval on the mean for the following analytes within each standard is provided: sodium, nitrate, nitrite, phosphate, carbonate, sulfate, hydroxide, chromate, chloride, fluoride, aluminum, plutonium-239/240, strontium-90, cesium-137, and water

  10. Photoionization of Sodium Salt Solutions in a Liquid Jet

    International Nuclear Information System (INIS)

    Grieves, G. A.; Petrik, Nikolay G.; Herring-Captain, J.; Olanrewaju, B.; Aleksandrov, A.; Tonkyn, Russell G.; Barlow, Stephan E.; Kimmel, Gregory A.; Orlando, Thomas M.

    2008-01-01

    A liquid microjet was employed to examine the gas/liquid interface of aqueous sodium halide (Na+X-, X=Cl, Br, I) salt solutions. Laser excitation at 193 nm produced and removed cations of the form H+(H2O)n and Na+(H2O)m from liquid jet surfaces containing either NaCl, NaBr or NaI. The protonated water cluster yield varied inversely with increasing salt concentration, while the solvated sodium ion cluster yield varied by anion type. The distribution of H+(H2O)n at low salt concentration is identical to that observed from low-energy electron irradiated amorphous ice and the production of these clusters can be accounted for using a localized ionization/Coulomb expulsion model. Production of Na+(H2O)m is not accounted for by this model but requires ionization of solvation shell waters and a contact ion/Coulomb expulsion mechanism. The reduced yields of Na+(H2O)m from high concentration (10-2 and 10-1 M) NaBr and NaI solutions indicate a propensity for Br- and I- at the solution surfaces and interfaces. This is supported by the observation of multiphoton induced production and desorption of Br+ and I+ from the 10-2 and 10-1 M solution surfaces

  11. Photoionization of Sodium Salt Solutions in a Liquid Jet

    Energy Technology Data Exchange (ETDEWEB)

    Grieves, G. A.; Petrik, Nikolay G.; Herring-Captain, J.; Olanrewaju, B.; Aleksandrov, A.; Tonkyn, Russell G.; Barlow, Stephan E.; Kimmel, Gregory A.; Orlando, Thomas M.

    2008-06-05

    A liquid microjet was employed to examine the gas/liquid interface of aqueous sodium halide (Na+X-, X=Cl, Br, I) salt solutions. Laser excitation at 193 nm produced and removed cations of the form H+(H2O)n and Na+(H2O)m from liquid jet surfaces containing either NaCl, NaBr or NaI. The protonated water cluster yield varied inversely with increasing salt concentration, while the solvated sodium ion cluster yield varied by anion type. The distribution of H+(H2O)n at low salt concentration is identical to that observed from low-energy electron irradiated amorphous ice and the production of these clusters can be accounted for using a localized ionization/Coulomb expulsion model. Production of Na+(H2O)m is not accounted for by this model but requires ionization of solvation shell waters and a contact ion/Coulomb expulsion mechanism. The reduced yields of Na+(H2O)m from high concentration (10-2 and 10-1 M) NaBr and NaI solutions indicate a propensity for Br- and I- at the solution surfaces and interfaces. This is supported by the observation of multiphoton induced production and desorption of Br+ and I+ from the 10-2 and 10-1 M solution surfaces.

  12. Chemical species of plutonium in Hanford radioactive tank waste

    International Nuclear Information System (INIS)

    Barney, G.S.

    1997-01-01

    Large quantities of radioactive wastes have been generated at the Hanford Site over its operating life. The wastes with the highest activities are stored underground in 177 large (mostly one million gallon volume) concrete tanks with steel liners. The wastes contain processing chemicals, cladding chemicals, fission products, and actinides that were neutralized to a basic pH before addition to the tanks to prevent corrosion of the steel liners. Because the mission of the Hanford Site was to provide plutonium for defense purposes, the amount of plutonium lost to the wastes was relatively small. The best estimate of the amount of plutonium lost to all the waste tanks is about 500 kg. Given uncertainties in the measurements, some estimates are as high as 1,000 kg (Roetman et al. 1994). The wastes generally consist of (1) a sludge layer generated by precipitation of dissolved metals from aqueous wastes solutions during neutralization with sodium hydroxide, (2) a salt cake layer formed by crystallization of salts after evaporation of the supernate solution, and (3) an aqueous supernate solution that exists as a separate layer or as liquid contained in cavities between sludge or salt cake particles. The identity of chemical species of plutonium in these wastes will allow a better understanding of the behavior of the plutonium during storage in tanks, retrieval of the wastes, and processing of the wastes. Plutonium chemistry in the wastes is important to criticality and environmental concerns, and in processing the wastes for final disposal. Plutonium has been found to exist mainly in the sludge layers of the tanks along with other precipitated metal hydrous oxides. This is expected due to its low solubility in basic aqueous solutions. Tank supernate solutions do not contain high concentrations of plutonium even though some tanks contain high concentrations of complexing agents. The solutions also contain significant concentrations of hydroxide which competes with other

  13. Applicability of molten salt oxidation to the destruction of actinide-contaminated wastes

    International Nuclear Information System (INIS)

    West, M.H.; Garcia, E.; Griego, W.J.; Court, D.B.; Rodriguez, L.

    1992-01-01

    A 1989 ban on incineration in the state of New Mexico caused cessation of actinide-contaminated cheesecloth, paper, and wood incineration within the Plutonium Facility (TA-55) at Los Alamos National Laboratory. Subsequently, plastic wipes were substituted for cheesecloth in the cleaning of glovebox interiors. However, waste minimization is not achieved by these measures since the wipes are discarded as Waste Isolation Pilot Plant certifiable wastes. After the ban was instituted, thermal decomposition of cheesecloth under argon at elevated temperature was examined and found satisfactory although scale of operation and speed were inferior to incineration. In 1991, the ban on incineration was lifted in New Mexico but Alamos has not chosen to pursue renewal of incineration at the Plutonium Facility. This paper reports that Los Alamos is looking from alternatives to incineration and thermal decomposition which are compatible with molten salt processing technology, historically a strength in actinide research at the Laboratory. Also, the technology must significantly reduce the volume of the waste upon treatment, i.e. waste minimization. Molten salt oxidation (MSO) has the promise of such a technology

  14. Glovebox design requirements for molten salt oxidation processing of transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    Ramsey, K.B.; Acosta, S.V. [Los Alamos National Lab., NM (United States); Wernly, K.D. [Molten Salt Oxidation Corp., Bensalem, PA (United States)

    1998-12-31

    This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented.

  15. Copper-Sulfate Pentahydrate as a Product of the Waste Sulfuric Acid Solution Treatment

    Science.gov (United States)

    Marković, Radmila; Stevanović, Jasmina; Avramović, Ljiljana; Nedeljković, Dragutin; Jugović, Branimir; Stajić-Trošić, Jasna; Gvozdenović, Milica

    2012-12-01

    The aim of this study is synthesis of copper-sulfate pentahydrate from the waste sulfuric acid solution-mother liquor generated during the regeneration process of copper bleed solution. Copper is removed from the mother liquor solution in the process of the electrolytic treatment using the insoluble lead anodes alloyed with 6 mass pct of antimony on the industrial-scale equipment. As the result of the decopperization process, copper is removed in the form of the cathode sludge and is precipitated at the bottom of the electrolytic cell. By this procedure, the content of copper could be reduced to the 20 mass pct of the initial value. Chemical characterization of the sludge has shown that it contains about 90 mass pct of copper. During the decopperization process, the very strong poison, arsine, can be formed, and the process is in that case terminated. The copper leaching degree of 82 mass pct is obtained using H2SO4 aqueous solution with the oxygen addition during the cathode sludge chemical treatment at 80 °C ± 5 °C. Obtained copper salt satisfies the requirements of the Serbian Standard for Pesticide, SRPS H.P1. 058. Therefore, the treatment of waste sulfuric acid solutions is of great economic and environmental interest.

  16. Modeling internal deformation of salt structures targeted for radioactive waste disposal

    International Nuclear Information System (INIS)

    Chemia, Zurab

    2008-01-01

    This thesis uses results of systematic numerical models to argue that externally inactive salt structures, which are potential targets for radioactive waste disposal, might be internally active due to the presence of dense layers or blocks within a salt layer. The three papers that support this thesis use the Gorleben salt diapir (NW Germany), which was targeted as a future final repository for high-grade radioactive waste, as a general guideline. The first two papers present systematic studies of the parameters that control the development of a salt diapir and how it entrains a dense anhydrite layer. Results from these numerical models show that the entrainment of a dense anhydrite layer within a salt diapir depends on four parameters: sedimentation rate, viscosity of salt, perturbation width and the stratigraphic location of the dense layer. The combined effect of these four parameters, which has a direct impact on the rate of salt supply (volume/area of the salt that is supplied to the diapir with time), shape a diapir and the mode of entrainment. Salt diapirs down-built with sedimentary units of high viscosity can potentially grow with an embedded anhydrite layer and deplete their source layer (salt supply ceases). However, when salt supply decreases dramatically or ceases entirely, the entrained anhydrite layer/segments start to sink within the diapir. In inactive diapirs, sinking of the entrained anhydrite layer is inevitable and strongly depends on the rheology of the salt, which is in direct contact with the anhydrite layer. During the post-depositional stage, if the effective viscosity of salt falls below the threshold value of around 10 18 -10 19 Pa s, the mobility of anhydrite blocks might influence any repository within the diapir. However, the internal deformation of the salt diapir by the descending blocks decreases with increase in effective viscosity of salt. The results presented in this thesis suggest that it is highly likely that salt structures

  17. Definition of the waste package environment for a repository located in salt

    International Nuclear Information System (INIS)

    Clark, D.E.; Bradley, D.J.

    1983-01-01

    The expected environmental conditions for emplaced waste packages in a salt repository are simulated in the materials testing program to evaluate performance. Synthetic brines, based on the analyses of actual brines (both intrusion and inclusion), are used for corrosion and leach testing. Elevated temperatures (to 150 0 C) and radiation fields of up to 10 3 rad/h are employed as conservative conditions to bracket expected performance and provide data for worst case scenarios. Obtaining a precise definition of the waste package environment in a salt repository and its change with time is closely tied to detailed site characterization of the candidate salt repository horizon. It is expected that field testing can augment some of the materials testing currently under way and can provide increased confidence in the predicted site-specific near-field conditions. 17 references, 5 figures, 1 table

  18. Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

    International Nuclear Information System (INIS)

    Bell, J.T.; Haas, P.A.; Rudolph, J.C.

    1995-01-01

    The Oak Ridge National Laboratory (ORNL) is participating in a program to apply a molten salt oxidation (MSO) process to treatment of mixed (radioactive and RCRA) wastes. The salt residues from the MSO treatment will require further separations or other processing to prepare them for final disposal. A bench-scale MSO apparatus is being installed at ORNL and will be operated on real Oak Ridge wastes. The treatment concepts to be tested and demonstrated on the salt residues from real wastes are described

  19. Optical absorption of dilute solutions of metals in molten salts

    Energy Technology Data Exchange (ETDEWEB)

    Senatore, G.; Parrinello, M.; Tosi, M.P. (Trieste Univ. (Italy). Ist. di Fisica Teorica; Gruppo Nazionale di Struttura dell material del CNR, Trieste (Italy); International Centre for Theoretical Physics, Trieste (Italy))

    1978-12-23

    The theory of liquid structure for fluids of charged hard spheres is applied to an evaluation of the F-centre model for valence electrons in metal-molten salt solutions at high dilution. Minimization of the free energy yields the groundstate radius of the elctron bubble and hence the optical excitation energy in a Franck-Condon transition, the shift and broadening of the transition due to fluctuations in the bubble radius, the volume of mixing, and the activity of the salt in the solution.

  20. Analyses of SRS waste glass buried in granite in Sweden and salt in the United States

    International Nuclear Information System (INIS)

    Williams, J.P.; Wicks, G.G.; Clark, D.E.; Lodding, A.R.

    1991-01-01

    Simulated Savannah River Site (SRS) waste glass forms have been buried in the granite geology of the Stirpa mine in Sweden for two years. Analyses of glass surfaces provided a measure of the performance of the waste glasses as a function of time. Similar SRS waste glass compositions have also been buried in salt at the WIPP facility in Carlsbad, New Mexico for a similar time period. Analyses of the SRS waste glasses buried in-situ in granite will be presented and compared to the performance of these same compositions buried in salt at WIPP

  1. Engineering study of the potential uses of salts from selective crystallization of Hanford tank wastes

    International Nuclear Information System (INIS)

    Hendrickson, D.W.

    1996-01-01

    The Clean Salt Process (CSP) is the fractional crystallization of nitrate salts from tank waste stored on the Hanford Site. This study reviews disposition options for a CSP product made from Hanford Site tank waste. These options range from public release to onsite low-level waste disposal to no action. Process, production, safety, environment, cost, schedule, and the amount of CSP material which may be used are factors considered in each option. The preferred alternative is offsite release of clean salt. Savings all be generated by excluding the material from low-level waste stabilization. Income would be received from sales of salt products. Savings and income from this alternative amount to $1,027 million, excluding the cost of CSP operations. Unless public sale of CSP products is approved, the material should be calcined. The carbonate form of the CSP could then be used as ballast in tank closure and stabilization efforts. Not including the cost of CSP operations, savings of $632 million would be realized. These savings would result from excluding the material from low-level waste stabilization and reducing purchases of chemicals for caustic recycle and stabilization and closure. Dose considerations for either alternative are favorable. No other cost-effective alternatives that were considered had the capacity to handle significant quantities of the CSP products. If CSP occurs, full-scale tank-waste stabilization could be done without building additional treatment facilities after Phase 1 (DOE 1996). Savings in capital and operating cost from this reduction in waste stabilization would be in addition to the other gains described

  2. Forecasting the space-time stability of radioactive waste isolation in salt formations

    International Nuclear Information System (INIS)

    Anderson, E.B.; Karelin, A.I.; Krivokhatsiy, A.S.; Savonenkov, V.G.

    1992-01-01

    The possibilities to use salt formations for radioactive waste isolation are realized by creating shaft-type underground repositories in these rocks in Germany and the USA. The burial safety of low- and intermediate-level wastes for several hundred years have been substantiated for the sites chosen. Specialists of different countries presented positive properties of rock salt as a medium for isolation of radionuclides. A rich experience in building subsurface structures for different purposes in salts is accumulated in our country. Detailed investigations of salt formation have shown that far from all the saliferous areas and structures may be used for constructing burial sites. One of the reasons for this limitation is a sharp difference of individual deposits by their compositions, structures, the character of deposition and the conditions of formation. The geological criteria of safety acquire special significance in connection with the necessity to isolate radionuclides having the half-loves more than 1000 years. The time intervals required for stable isolation make up millions of years and cover great cycles of the evolution of the Earth surface and biosphere

  3. Waste processing of chemical cleaning solutions

    International Nuclear Information System (INIS)

    Peters, G.A.

    1991-01-01

    This paper reports on chemical cleaning solutions containing high concentrations of organic chelating wastes that are difficult to reduce in volume using existing technology. Current methods for evaporating low-level radiative waste solutions often use high maintenance evaporators that can be costly and inefficient. The heat transfer surfaces of these evaporators are easily fouled, and their maintenance requires a significant labor investment. To address the volume reduction of spent, low-level radioactive, chelating-based chemical cleaning solutions, ECOSAFE Liquid Volume Reduction System (LVRS) has been developed. The LVRS is based on submerged combustion evaporator technology that was modified for treatment of low-level radiative liquid wastes. This system was developed in 1988 and was used to process 180,000 gallons of waste at Oconee Nuclear Station

  4. Release consequence analysis for a hypothetical geologic radioactive waste repository in salt

    International Nuclear Information System (INIS)

    1979-08-01

    One subtask conducted under the INFCE program is to evaluate and compare the health and safety impacts of different fuel cycles in which all radioactive wastes (except those from mining and milling) are placed in a geologic repository in salt. To achieve this objective, INFCE Working Group 7 examined the radiologic dose to humans from geologic repositories containing waste arisings as defined for seven reference fuel cycles. This report examines the release consequences for a generic waste repository in bedded salt. The top of the salt formation and the top of the repository are assumed to be 250 and 600 m, respectively, below the surface. The hydrogeologic structure above the salt consists of two aquifers and two aquitards. The aquifers connect to a river 6.2 km from the repository. The regional gradient to the river is 1 m/km in all aquifers. Hydrologic, transport, and dose models were used to model two release scenarios for each fuel cycle, one without a major disturbance and one in which a major geologic perturbation breached the repository immediately after it was sealed. The purpose of the modeling was to predict the rate of transport of radioactive contaminants from the repository through the geosphere to the biosphere, and to determine the potential dose to humans. Of the many radionuclides in the waste, only 129 I and 226 Ra arrived at the river in sufficient concentrations for a measurable dose calculation. Radionuclide concentrations in the ground water pose no threat to man because the ground water is a concentrated brine and it is diluted by a factor of 10 6 to 10 7 upon entering the river

  5. Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

    International Nuclear Information System (INIS)

    St John, C.M.

    1977-04-01

    An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel

  6. Salt creep design consideration for underground nuclear waste storage

    International Nuclear Information System (INIS)

    Li, W.T.; Wu, C.L.; Antonas, N.J.

    1983-01-01

    This paper summarizes the creep consideration in the design of nuclear waste storage facilities in salt, describes the non-linear analysis method for evaluating the design adequacy, and presents computational results for the current storage design. The application of rock mechanics instrumentation to assure the appropriateness of the design is discussed. It also describes the design evolution of such a facility, starting from the conceptual design, through the preliminary design, to the detailed design stage. The empirical design method, laboratory tests and numerical analyses, and the underground in situ tests have been incorporated in the design process to assure the stability of the underground openings, retrievability of waste during the operation phase and encapsulation of waste after decommissioning

  7. Performance Assessment of a Generic Repository in Bedded Salt for DOE-Managed Nuclear Waste

    Science.gov (United States)

    Stein, E. R.; Sevougian, S. D.; Hammond, G. E.; Frederick, J. M.; Mariner, P. E.

    2016-12-01

    A mined repository in salt is one of the concepts under consideration for disposal of DOE-managed defense-related spent nuclear fuel (SNF) and high level waste (HLW). Bedded salt is a favorable medium for disposal of nuclear waste due to its low permeability, high thermal conductivity, and ability to self-heal. Sandia's Generic Disposal System Analysis framework is used to assess the ability of a generic repository in bedded salt to isolate radionuclides from the biosphere. The performance assessment considers multiple waste types of varying thermal load and radionuclide inventory, the engineered barrier system comprising the waste packages, backfill, and emplacement drifts, and the natural barrier system formed by a bedded salt deposit and the overlying sedimentary sequence (including an aquifer). The model simulates disposal of nearly the entire inventory of DOE-managed, defense-related SNF (excluding Naval SNF) and HLW in a half-symmetry domain containing approximately 6 million grid cells. Grid refinement captures the detail of 25,200 individual waste packages in 180 disposal panels, associated access halls, and 4 shafts connecting the land surface to the repository. Equations describing coupled heat and fluid flow and reactive transport are solved numerically with PFLOTRAN, a massively parallel flow and transport code. Simulated processes include heat conduction and convection, waste package failure, waste form dissolution, radioactive decay and ingrowth, sorption, solubility limits, advection, dispersion, and diffusion. Simulations are run to 1 million years, and radionuclide concentrations are observed within an aquifer at a point approximately 4 kilometers downgradient of the repository. The software package DAKOTA is used to sample likely ranges of input parameters including waste form dissolution rates and properties of engineered and natural materials in order to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia

  8. Computer simulation of an internally pressurized radioactive waste disposal room in a bedded salt formation

    International Nuclear Information System (INIS)

    Brown, W.T.; Weatherby, J.R.

    1991-01-01

    The Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico was created by the U.S. Department of Energy as an underground research and development facility to demonstrate the safe storage of transuranic waste generated from defense activities. This facility consists of storage rooms mined from a bedded salt formation at a depth of about 650 meters. Each room will accommodate about 6800 55-gallon drums filled with waste. After waste containers are emplaced, the storage rooms are to be backfilled with mined salt or other backfill materials. As time passes, reconsolidation of this backfill will reduce the hydraulic conductivity of the room. However, gases produced by decomposition and corrosion of waste and waste containers may cause a slow build-up of pressure which can retard consolidation of the waste and backfilled salt. The authors have developed a finite-element model of an idealized disposal room which is assumed to be perfectly sealed. The assumption that no gas escapes from the disposal room is a highly idealized and extreme condition which does not account for leakage paths, such as interbeds, that exist in the surrounding salt formation. This model has been used in a parametric study to determine how reconsolidation is influenced by various assumed gas generation rates and total amounts of gas generated. Results show that reductions in the gas generation, relative to the baseline case, can increase the degree of consolidation and reduce the peak gas pressure in disposal rooms. Even higher degrees of reconsolidation can be achieved by reducing both amounts and rates of gas generation. 8 refs., 4 figs., 1 tab

  9. Possible salt mine and brined cavity sites for radioactive waste disposal in the northeastern southern peninsula of Michigan

    International Nuclear Information System (INIS)

    Landes, K.K.; Bourne, H.L.

    1976-01-01

    A reconnaissance report on the possibilities for disposal of radioactive waste covers Michigan only, and is more detailed than an earlier one involving the northeastern states. Revised ''ground rules'' for pinpointing both mine and dissolved salt cavern sites for waste disposal include environmental, geologic, and economic factors. The Michigan basin is a structural bowl of Paleozoic sediments resting on downwarped Precambrian rocks. The center of the bowl is in Clare and Gladwin Counties, a short distance north of the middle of the Southern Peninsula. The strata dip toward this central area, and some stratigraphic sequences, including especially the salt-containing Silurian section, increase considerably in thickness in that direction. Lesser amounts of salt are also present in the north central part of the Lower Peninsula. Michigan has been an oil and gas producing state since 1925 and widespread exploration has had two effects on the selection of waste disposal sites: (1) large areas are leased for oil and gas; and (2) the borehole concentrations, whether producing wells, dry holes, or industrial brine wells that penetrated the salt section, should be avoided. Two types of nuclear waste, low level and high level, can be stored in man-made openings in salt beds. The storage facilities are created by (1) the development of salt mines where the depths are less than 3000 ft, and (2) cavities produced by pumping water into a salt bed, and bringing brine back out. The high level waste disposal must be confined to mines of limited depth, but the low level wastes can be accommodated in brine cavities at any depth. Seven potential prospects have been investigated and are described in detail

  10. Nuclear waste repository simulation experiments. Asse salt mine: Annual report 1984

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Feddersen, H.K.; Schwarzianeck, P.; Staupendahl, G.; Coyle, A.J.; Kalia, H.; Eckert, J.

    1985-01-01

    This is the Second Annual Report (1984) which describes experiments simulating a nuclear waste repository at the 800 meter-level of the Asse Salt Mine in the Federal Republic of Germany. The report describes the Asse Salt Mine, the test equipment, and the pretest properties of the salt in the mine and in the vicinity of the test area. Also included are test data for the first sixteen months of operation on the following: brine migration rates, thermal mechanical behavior of the salt (including room closure, stress readings and thermal profiles) and borehole gas pressures. In addition to field data laboratory analyses of results are also included in this report. The duration of the experiment will be two years, ending in December 1985. (orig.)

  11. Hydrothermal preparation of zeolite Li-A and ion exchange properties of Cs and Sr in salt waste

    International Nuclear Information System (INIS)

    Lee, S. H.; Kim, J. G.; Lee, J. H.; Kim, J. H.

    2005-01-01

    An advanced spent fuel management process that were based on Li reduction of the oxide spent fuel to a metallic form will generate a LiCl waste. Zeolite A has been reported as a promising immobilization medium for waste salt with CsCl and SrCl 2 . However, Sodium is accumulated as an ionic form (Na + -ion) in molten salt during ion exchange step between Na + -ion in zeolite A and Li + -ion in the molten salt. Therefore, zeolite Na-A need to be replaced by the Li-type zeolite for recycling the salt waste by removing the Cs and Sr ions. In this study, the hydrothermal preparation of zeolite Li-A was performed in 350ml pressure vessel by P. Norby method. The preparation characteristics of zeolite Li-A was investigated. And the ion exchange properties of Cs and Sr in molten LiCl salt were investigated under the condition of 923K using zeolite 4A and prepared zeolite Li-A

  12. Corrosion aspects of high-level waste disposal in salt domes

    International Nuclear Information System (INIS)

    Roerbo, K.

    1979-12-01

    In the ELSAM/ELKRAT waste management project it is planned that the high-level waste is glassified, encapsuled in canisters and finally deposited in a deep hole drilled in a salt dome. In the present report corrosion aspects of the canisters after deposition are discussed. The chemical environment will probably be a limited amount of brine coming from brine inclusions in the surrounding salt and moving up against the temperature gradient, the temperature at the canister surface being in the range of 100-150degC. The possible types of corrosion and the expected corrosion rates for a number of potential canister materials (mild steel, austenitic and ferritic stainless steels, Ni-base alloys, copper, titanium and a few combinations of materials) are discussed. Mild steel (possibly combined with an inner layer of copper or titanium) might possibly be an appropriate choice of material for the canister. (author)

  13. Engineering solution for the backfilling and sealing of radioactive waste repositories

    International Nuclear Information System (INIS)

    Jorda, M.; Gouvenot, D.; Bonne, A.; Lees, T.P.; Schmidt, M.

    1990-01-01

    To ensure the safety of radioactive waste deep disposal, backfilling and sealing materials (engineered barriers) have to be used to fill residual voids. For granite medium, stress is put on emplacement techniques for cement- and clay-based materials, including in-situ validation. For clay medium, mined repository and deep boreholes drilled from the surface are considered. In the case of the first solution, the thermomechanical behaviour of a clay backfill is studied. In the same way, backfill made of excavated crushed salt is considered and thermomechanical properties evaluated by means of laboratory tests and in-situ experiments. Finally, basic works on quality assurance procedures and historic concretes behaviour are reported

  14. Correlation of high-temperature stability of alpha-chymotrypsin with 'salting-in' properties of solution.

    Science.gov (United States)

    Levitsky VYu; Panova, A A; Mozhaev, V V

    1994-01-15

    A correlation between the stability of alpha-chymotrypsin against irreversible thermal inactivation at high temperatures (long-term stability) and the coefficient of Setchenov equation as a measure of salting-in/out efficiency of solutes in the Hofmeister series has been found. An increase in the concentration of salting-in solutes (KSCN, urea, guanidinium chloride, formamide) leads to a many-fold decrease of the inactivation rate of the enzyme. In contrast, addition of salting-out solutes has a small effect on the long-term stability of alpha-chymotrypsin at high temperatures. The effects of solutes are additive with respect to their salting-in/out capacities; the stabilizing action of the solutes is determined by the calculated Setchenov coefficient of solution. The correlation is explained by a solute-driven shift of the conformational equilibrium between the 'low-temperature' native and the 'high-temperature' denatured forms of the enzyme within the range of the kinetic scheme put forward in the preceding paper in this journal: irreversible inactivation of the high-temperature form proceeds much more slowly compared with the low-temperature form.

  15. High-Level Waste Salt Disposition Systems Engineering Team Final Report, Volumes I, II, and III

    International Nuclear Information System (INIS)

    Piccolo, S.F.

    1999-01-01

    This report describes the process used and results obtained by the High Level Waste Salt Disposition Systems Engineering Team to select a primary and backup alternative salt disposition method for the Savannah River Site

  16. Analysis of the corrosion of carbon steels in simulated salt repository brines and acid chloride solutions at high temperatures

    International Nuclear Information System (INIS)

    Diercks, D.R.; Kassner, T.F.

    1988-04-01

    An analysis of literature data on the corrosion of carbon steels in anoxic brines and acid chloride solutions was performed, and the results were used to assess the expected life of high-level nuclear waste package containers in a salt repository environment. The corrosion rate of carbon steels in moderately acidic aqueous chloride environments obeys an Arrhenius dependence on temperature and a (pH 2 ) -1/2 dependence on hydrogen partial pressure. The cathodic reduction of water to produce hydrogen is the rate-controlling step in the corrosion process. An expression for the corrosion rate incorporating these two dependencies was used to estimate the corrosion life of several proposed waste package configurations. 42 refs., 11 figs., 2 tabs

  17. Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-10-06

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges. This revision was created to correct an error. The previous revision used an incorrect set of temperature correction coefficients which resulted in slight deviations from the correct D(Cs) results.

  18. Analysis of the geological stability of a hypothetical radioactive waste repository in a bedded salt formation

    International Nuclear Information System (INIS)

    Tierney, M.S.; Lusso, F.; Shaw, H.R.

    1978-01-01

    This document reports on the development of mathematical models used in preliminary studies of the long-term safety of radioactive wastes deeply buried in bedded salt formations. Two analytical approaches to estimating the geological stability of a waste repository in bedded salt are described: (a) use of probabilistic models to estimate the a priori likelihoods of release of radionuclides from the repository through certain idealized natural and anthropogenic causes, and (b) a numerical simulation of certain feedback effects of emplacement of waste materials upon ground-water access to the repository's host rocks. These models are applied to an idealized waste repository for the sake of illustration

  19. SODIUM ALUMINOSILICATE FOULING AND CLEANING OF DECONTAMINATED SALT SOLUTION COALESCERS

    International Nuclear Information System (INIS)

    Poirier, M.; Thomas Peters, T.; Fernando Fondeur, F.; Samuel Fink, S.

    2008-01-01

    During initial non-radioactive operations at the Modular Caustic Side Solvent Extraction Unit (MCU), the pressure drop across the decontaminated salt solution coalescer reached ∼10 psi while processing ∼1250 gallons of salt solution, indicating possible fouling or plugging of the coalescer. An analysis of the feed solution and the 'plugged coalescer' concluded that the plugging was due to sodium aluminosilicate solids. MCU personnel requested Savannah River National Laboratory (SRNL) to investigate the formation of the sodium aluminosilicate solids (NAS) and the impact of the solids on the decontaminated salt solution coalescer. Researchers performed developmental testing of the cleaning protocols with a bench-scale coalescer container 1-inch long segments of a new coalescer element fouled using simulant solution. In addition, the authors obtained a 'plugged' Decontaminated Salt Solution coalescer from non-radioactive testing in the MCU and cleaned it according to the proposed cleaning procedure. Conclusions from this testing include the following: (1) Testing with the bench-scale coalescer showed an increase in pressure drop from solid particles, but the increase was not as large as observed at MCU. (2) Cleaning the bench-scale coalescer with nitric acid reduced the pressure drop and removed a large amount of solid particles (11 g of bayerite if all aluminum is present in that form or 23 g of sodium aluminosilicate if all silicon is present in that form). (3) Based on analysis of the cleaning solutions from bench-scale test, the 'dirt capacity' of a 40 inch coalescer for the NAS solids tested is calculated as 450-950 grams. (4) Cleaning the full-scale coalescer with nitric acid reduced the pressure drop and removed a large amount of solid particles (60 g of aluminum and 5 g of silicon). (5) Piping holdup in the full-scale coalescer system caused the pH to differ from the target value. Comparable hold-up in the facility could lead to less effective cleaning and

  20. Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

    Science.gov (United States)

    Hite, R.J.

    1977-01-01

    The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2, is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt anticlines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as 'marker beds.' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement.

  1. Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

    International Nuclear Information System (INIS)

    Hite, R.J.

    1977-01-01

    The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2 , is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt antilcines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as ''marker beds.'' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement

  2. Solution Behavior and Activity of a Halophilic Esterase under High Salt Concentration

    Science.gov (United States)

    Rao, Lang; Zhao, Xiubo; Pan, Fang; Li, Yin; Xue, Yanfen; Ma, Yanhe; Lu, Jian R.

    2009-01-01

    Background Halophiles are extremophiles that thrive in environments with very high concentrations of salt. Although the salt reliance and physiology of these extremophiles have been widely investigated, the molecular working mechanisms of their enzymes under salty conditions have been little explored. Methodology/Principal Findings A halophilic esterolytic enzyme LipC derived from archeaon Haloarcula marismortui was overexpressed from Escherichia coli BL21. The purified enzyme showed a range of hydrolytic activity towards the substrates of p-nitrophenyl esters with different alkyl chains (n = 2−16), with the highest activity being observed for p-nitrophenyl acetate, consistent with the basic character of an esterase. The optimal esterase activities were found to be at pH 9.5 and [NaCl] = 3.4 M or [KCl] = 3.0 M and at around 45°C. Interestingly, the hydrolysis activity showed a clear reversibility against changes in salt concentration. At the ambient temperature of 22°C, enzyme systems working under the optimal salt concentrations were very stable against time. Increase in temperature increased the activity but reduced its stability. Circular dichroism (CD), dynamic light scattering (DLS) and small angle neutron scattering (SANS) were deployed to determine the physical states of LipC in solution. As the salt concentration increased, DLS revealed substantial increase in aggregate sizes, but CD measurements revealed the maximal retention of the α-helical structure at the salt concentration matching the optimal activity. These observations were supported by SANS analysis that revealed the highest proportion of unimers and dimers around the optimal salt concentration, although the coexistent larger aggregates showed a trend of increasing size with salt concentration, consistent with the DLS data. Conclusions/Significance The solution α-helical structure and activity relation also matched the highest proportion of enzyme unimers and dimers. Given that

  3. Geohydrolic studies of Gulf Coast interior salt domes

    International Nuclear Information System (INIS)

    Smith, C.G. Jr.

    1977-01-01

    Disposal of high-level radioactive wastes in Gulf Coast salt domes requires that the cavities be free from groundwater dissolution for 250,000 years. Salinity variations of groundwater near selected domes were investigated. Saline groundwater anomalies (saline plumes) in aquifers pierced or uplifted by the dome may be the result of salt solution by groundwater. In the Northeast Texas salt dome basin electric logs of oil and gas wells have been used to estimate groundwater salinities in aquifers near selected domes. Thus far, the analyses have revealed saline groundwater anomalies around 4 of the 9 domes studied. Estimates of the rate of salt dissolution from domes associated with saline groundwater plumes indicate that less than 30 meters of salt will be removed from the upper surfaces of the dome in 250,000 years. Thus, these preliminary studies show that even apparently unstable domes may be sufficiently stable to serve as waste disposal sites. 6 figures

  4. Denitration of Savannah River Plant waste streams

    International Nuclear Information System (INIS)

    Orebaugh, E.G.

    1976-07-01

    Partial denitration of waste streams from Savannah River Plant separations processes was shown to significantly reduce the quantity of waste solids to be stored as an alkaline salt cake. The chemical processes involved in the denitration of nonradioactive simulated waste solutions were studied. Chemical and instrumental analytical techniques were used to define both the equilibrium concentrations and the variation of reactants and products in the denitration reaction. Mechanisms were proposed that account for the complicated chemical reactions observed in the simulated waste solutions. Metal nitrates can be denitrated by reaction with formic acid only by the release of nitric acid from hydrolysis or formate complexation of metal cations. However, eventual radiolysis of formate salts or complexes results in the formation of biocarbonate and makes complexation-denitration a nonproductive means of reducing waste solids. Nevertheless, destruction of nitrate associated with free acid and easily hydrolyzable cations such as iron, mercury, and zirconium can result in greater than 30 percent reduction in waste solids from five SRP waste streams

  5. Equipment evaluation for low density polyethylene encapsulated nitrate salt waste at the Rocky Flats Plant

    International Nuclear Information System (INIS)

    Yamada, W.I.; Faucette, A.M.; Jantzen, R.C.; Logsdon, B.W.; Oldham, J.H.; Saiki, D.M.; Yudnich, R.J.

    1993-01-01

    Mixed wastes at the Rocky Flats Plant (RFP) are subject to regulation by the Resource Conservation and Recovery Act (RCRA). Polymer solidification is being developed as a final treatment technology for several of these mixed wastes, including nitrate salts. Encapsulation nitrate salts with low density polyethylene (LDPE) has been the preliminary focus of the RFP polymer solidification effort. Literature reviews, industry surveys, and lab-scale and pilot-scale tests have been conducted to evaluate several options for encapsulating nitrate salts with LDPE. Most of the effort has focused on identifying compatible drying and extrusion technologies. Other processing options, specifically meltration and non-heated compounding machines, were also investigated. The best approach appears to be pretreatment of the nitrate salt waste brine in either a vertical or horizontal thin film evaporator followed by compounding of the dried waste with LDPE in an intermeshing, co-rotating, twin-screw extruder. Additional pilot-scale tests planned for the fall of 1993 should further support this recommendation. Preliminary evaluation work indicates that meltration is not possible at atmospheric pressure with the LDPE (Chevron PE-1409) provided by RFP. However, meltration should be possible at atmospheric pressure using another LDPE formulation with altered physical and rheological properties: Lower molecular weight and lower viscosity (Epoline C-15). Contract modifications are now in process to allow a follow-on pilot scale demonstration. Questions regarding changed safety and physical properties of the resultant LDPE waste form due to use of the Epoline C-15 will be addressed. No additional work with non-heated mixer compounder machines is planned at this time

  6. Density, viscosity, and N2O solubility of aqueous amino acid salt and amine amino acid salt solutions

    International Nuclear Information System (INIS)

    Aronu, Ugochukwu E.; Hartono, Ardi; Svendsen, Hallvard F.

    2012-01-01

    Highlights: ► Density of amino acid salt and amine amino acid salt. ► Viscosity of amino acid salt and amine amino acid salt. ► Henry’s law constant/N 2 O solubility of amino acid salt and amine amino acid salt. ► Schumpe model. Correlations for density, viscosity, and N 2 O solubility. - Abstract: Physicochemical properties of aqueous amino acid salt (AAS), potassium salt of sarcosine (KSAR) and aqueous amine amino acid salt (AAAS), 3-(methylamino)propylamine/sarcosine (SARMAPA) have been studied. Densities of KSAR were measured for sarcosine mole fraction 0.02 to 0.25 for temperature range 298.15 K to 353.15 K, the viscosities were measured for 0.02 to 0.10 mole fraction sarcosine (293.15 K to 343.15 K) while the N 2 O solubilities were measured from 0.02 to 0.10 mole fraction sarcosine solutions (298.15 K to 363.15 K). Densities of SARMAPA were measured for sarcosine mole fraction 0.02 to 0.23 for temperature range (298.15 K to 353.15 K), viscosities were measured for 0.02 to 0.16 mole fraction sarcosine (293.15 K to 343.15 K) while the N 2 O solubilities were measured from 0.02 to 0.16 mole fraction sarcosine solutions (298.15 K to 343.15 K). Experimental results were correlated well with empirical correlations and N 2 O solubility results for KSAR were predicted adequately by a Schumpe model. The solubilities of N 2 O in AAS and AAAS are significantly lower than values for amines. The solubilities vary as: amine > AAAS > AAS.

  7. Experiments on the effect of sphagnum on the pH of salt solutions

    Energy Technology Data Exchange (ETDEWEB)

    Williams, K T; Thompson, T G

    1936-01-01

    Addition of sphagnum to salt solutions produced marked increases in the concentrations of the hydrogen ions, as measured both electrometrically and colorimetrically. The greater the concentration of the salt solution, the greater the increase in hydrogen ion concentration upon the addition of sphagnum. With a given salt concentration, the hydrogen ion concentration increased with increase in quantity of sphagnum added. The divalent cations produced greater increases in the hydrogen concentration than the monovalent cations for equal weights of sphagnum. Divalent anions, while showing an increase in hydrogen ions, upon the addition of sphagnum were far less effective in increasing the hydrogen ion concentrations. Sphagnum may be a useful reagent for regulating the acidity of salt solutions for many types of scientific work. It seems probable that the adsorption of metallic and hydroxyl ions explains, at least in part, the acidity of the water of sphagnum bogs.

  8. Recovery of actinides from TBP-Na2Co3 scrub-waste solutions: the ARALEX process

    International Nuclear Information System (INIS)

    Horwitz, E.P.; Bloomquist, C.A.A.; Mason, G.W.; Leonard, R.A.; Ziegler, A.A.

    1979-08-01

    A flowsheet for the recovery of actinides from TBP-Na 2 CO 3 scrub-waste solutions has been developed, based on batch extraction data, and tested, using laboratory-scale countercurrent extraction techniques. The process, called the ARALEX process, uses 2-ethyl-1-hexanol (2-EHOH) to extract the TBP degradation products (HDBP and H 2 MBP) from acidified Na 2 CO 3 scrub waste leaving the actinides in the aqueous phase. Dibutyl and monobutyl phosphoric acids are attached to the 2-EHOH molecules through hydrogen bonds, which also diminish the ability of the HDBP and H 2 MBP to complex actinides. Thus all actinides remain in the aqueous raffinate. Dilute sodium hydroxide solutions can be used to back-extract the dibutyl and monobutyl phosphoric acid esters as their sodium salts. The 2-EHOH can then be recycled. After extraction of the acidified carbonate waste with 2-EHOH, the actinides may be readily extracted from the raffinate with DHDECMP or, in the case of tetra- and hexavalent actinides, with TBP. The ARALEX process can also be applied to other actinide waste streams which contain appreciable concentrations of polar organic compounds (e.g., detergents) that interfere with conventional actinide ion exchange and liquid-liquid extraction procedures. 20 figures, 6 tables

  9. Novel waste printed circuit board recycling process with molten salt.

    Science.gov (United States)

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going.

  10. Novel waste printed circuit board recycling process with molten salt

    Science.gov (United States)

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

  11. Molten salt oxidation as a technique for decommissioning: selection of low melting point salt mixtures

    International Nuclear Information System (INIS)

    Lainetti, Paulo E.O.; Garcia, Vitor F.; Benvegnu, Guilherme

    2013-01-01

    During the 70 and 80 years, IPEN built several facilities in pilot scale, destined to the technological domain of the Nuclear Fuel Cycle. In the nineties, radical changes in the Brazilian nuclear policy determined the interruption of the activities and the shut-down of pilot plants. Nowadays, IPEN has been facing the problem of the dismantling and decommissioning of its Nuclear Fuel Cycle old facilities. The facility CELESTE-I of the IPEN is a laboratory where reprocessing studies were accomplished during the decade of 80 and in the beginning of the 90s. The last operations occurred in 92-93. The research activities generated radioactive wastes in the form of organic and aqueous solutions of different compositions and concentrations. For the treatment of these liquid wastes it was proposed a study of waste thermal decomposition based on the molten salt oxidation process.Decomposition tests of different organic wastes have been performed in laboratory equipment developed at IPEN, in the range of temperatures of 900 to 1020 deg C, demonstrating the complete oxidation of the compounds. The reduction of the process temperatures would be of crucial importance. Besides this, the selection of lower melting point salt mixtures would have an important impact in the reduction of equipment costs. Several experiments were performed to determine the most suitable salt mixtures, optimizing costs and melting temperatures as low as possible. This paper describes the main characteristics of the molten salt oxidation process, besides the selection of salt mixtures of binary and ternary compositions, respectively Na 2 CO 3 - NaOH and Na 2 CO 3 - K 2 CO 3 -Li 2 CO 3 . (author)

  12. NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

    Energy Technology Data Exchange (ETDEWEB)

    Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D. [U.S. Nuclear Regulatory Commission (United States)

    2013-07-01

    As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in

  13. NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

    International Nuclear Information System (INIS)

    Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D.

    2013-01-01

    As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in coordination with South

  14. Isolation of transplutonium elements on ion exchangers from solutions of high salt concentration

    International Nuclear Information System (INIS)

    Guseva, L.I.; Tikhomirova, G.S.; Stepushkina, V.V.

    1985-01-01

    The behaviour of transplutonium elements (TPE) on cation and anion exchangers in aqueous alcoholic solutions of chlorides and nitrates of some alkali and alkaline earth metals depending on different factors: salt concentration, content of alcohol and of acid in the solution as well as the nature of a cation was studied. The data obtained were used to determine the optimal conditions of concentration of TPE on ion exchangers from solutions containing great quantities of salts. The advantages of the use of aqueous alcoholic solutions of nitric acid in the isolation of TPE are shown. (author)

  15. Meat batter production in an extended vane pump-grinder injecting curing salt solutions to reduce energy requirements: variation of curing salt amount injected with the solution.

    Science.gov (United States)

    Irmscher, Stefan B; Terjung, Eva-Maria; Gibis, Monika; Herrmann, Kurt; Kohlus, Reinhard; Weiss, Jochen

    2017-01-01

    The integration of a nozzle in an extended vane pump-grinder system may enable the continuous injection of curing salt solutions during meat batter production. The purpose of this work was to examine the influence of the curing salt amount injected with the solution (0-100%) on protein solubilisation, water-binding, structure, colour and texture of emulsion-type sausages. The amount of myofibrillar protein solubilised during homogenisation varied slightly from 33 to 36 g kg -1 . Reddening was not noticeably impacted by the later addition of nitrite. L * ranged from 66.9 ± 0.3 to 67.8 ± 0.3, a * from 10.9 ± 0.1 to 11.2 ± 0.1 and b * from 7.7 ± 0.1 to 8.0 ± 0.1. Although softer sausages were produced when only water was injected, firmness increased with increasing curing salt amount injected and was similar to the control when the full amount of salt was used. The substitution of two-thirds of ice with a liquid brine may enable energy savings due to reduced power consumptions of the extended vane pump-grinder system by up to 23%. The injection of curing salt solutions is feasible without affecting structure and colour negatively. This constitutes a first step towards of an 'ice-free' meat batter production allowing for substantial energy savings due to lower comminution work. © 2016 Society of Chemical Industry. © 2016 Society of Chemical Industry.

  16. Method of decomposing treatment for radioactive organic phosphate wastes

    International Nuclear Information System (INIS)

    Uki, Kazuo; Ichihashi, Toshio; Hasegawa, Akira; Sato, Tatsuaki

    1985-01-01

    Purpose: To decompose the organic phosphoric-acid ester wastes containing radioactive material, which is produced from spent fuel reprocessing facilities, into inorganic materials using a simple device, under moderate conditions and at high decomposing ratio. Method: Radioactive organic phosphate wates are oxidatively decomposed by H 2 O 2 in an aqueous phosphoric-acid solution of metal phosphate salts. Copper phosphates are used as the metal phosphate salts and the decomposed solution of the radioactive organic phosphate wastes is used as the aqueous solution of the copper phosphate. The temperature used for the oxidizing decomposition ranges from 80 to 100 0 C. (Ikeda, J.)

  17. A study on the structure of thorium salt solutions

    International Nuclear Information System (INIS)

    Magini, M.; Cabrini, A.; Di Bartolomeo, A.

    1975-01-01

    The structure of highly hydrolyzed thorium salt solutions has been investigated by large and small angle X-ray scattering techniques. The diffraction data obtained with large angle measurements show the presence in solution of microcrystalline particles with the thorium oxide structure. Particles larger than those were discovered by small angle measurements. A possible shape of these colloidal particles has been discussed

  18. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's report on Functional Design Criteria for a Repository for High-Level Radioactive Waste

    International Nuclear Information System (INIS)

    Hambley, D.F.; Russell, J.E.; Busch, J.S.; Harrison, W.; Edgar, D.E.; Tisue, M.W.

    1984-08-01

    This report summarizes Argonne's review of the Office of Nuclear Waste Isolation's (ONWI's) draft report entitled Functional Design Criteria for High-Level Nuclear Waste Repository in Salt, dated January 23, 1984. Recommendations are given for improving the ONWI draft report

  19. Reconsolidation of salt as applied to permanent seals for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Hansen, F.D.; Callahan, G.D.; Van Sembeek, L.L.

    1993-01-01

    Reconsolidated salt is a fundamental component of the permanent seals for the Waste Isolation Pilot Plant. As regulations are currently understood and seal concepts envisioned, emplaced salt is the sole long-term seal component designed to prevent the shafts from becoming preferred pathways for rating gases or liquids. Studies under way in support of the sealing function of emplaced salt include laboratory testing of crushed salt small-scale in situ tests, constitutive modeling of crushed salt, calculations of the opening responses during operation and closure, and design practicalities including emplacement techniques. This paper briefly summarizes aspects of these efforts and key areas of future work

  20. The material flow of salt

    International Nuclear Information System (INIS)

    Kostick, D.S.

    1993-01-01

    Salt (NaCl) is a universal mineral commodity used by virtually every person in the world. Although a very common mineral today, at one time it was considered as precious as gold in certain cultures. This study traces the material flow of salt from its origin through the postconsumer phase of usage. The final disposition of salt in the estimated 14,000 different uses, grouped into several macrocategories, is traced from the dispersive loss of salt into the environment to the ultimate disposal of salt-base products into the waste stream after consumption. The base year for this study is 1990, in which an estimated 196 million short tons of municipal solid waste was discarded by the US population. Approximately three-fourths of domestic salt consumed is released to the environment and unrecovered while about one-fourth is discharged to landfills and incinerators as products derived from salt. Cumulative historical domestic production, trade, and consumption data have been compiled to illustrate the long-term trends within the US salt industry and the cumulative contribution that highway deicing salt has had on the environment. Salt is an important component of drilling fluids in well drilling. It is used to flocculate and to increase the density of the drilling fluid in order to overcome high down-well gas pressures. Whenever drilling activities encounter salt formations, salt is added to the drilling fluid to saturate the solution and minimize the dissolution within the salt strata. Salt is also used to increase the set rate of concrete in cemented casings. This subsector includes companies engaged in oil, gas, and crude petroleum exploration and in refining and compounding lubricating oil. It includes SIC major groups 13 and 29. 13 refs., 14 figs., 6 tabs

  1. Implications of thermophysical properties in geoscientific investigations for the disposal of nuclear waste in a salt dome

    International Nuclear Information System (INIS)

    Kopietz, J.

    1984-01-01

    Examples from laboratory and in-situ experiments on the thermomechanical behavior of rock salt are used to discuss the implications of thermophysical properties for disposal of nuclear waste in a salt dome. The implications of thermophysical properties are also illustrated by a brief review of geothermal investigations made within the scope of geological and hydrogeological exploration of the Gorleben salt dome in northern Germany. High-resolution temperature measurements performed in shallow and deep boreholes drilled for the exploration of the Gorleben salt dome, together with thermal conductivity measurements on representative core samples from these boreholes, are contributing to a determination of groundwater flow in the covering layers of the salt dome and to the identification of zones of impurity (eg carnallitite layers) within the salt structure. Data from these experiments are used for setting up numerical models for heat propagation around a prospective waste repository in the Gorleben salt dome. Long-term creep experiments on samples of rock salt at up to 400 deg C are used to derive constitutive relations on the creep behavior of salt. In-situ heating experiments are being conducted in the Asse salt mine to determine the effect of a heat source on the integrity of the surrounding salt rock. (author)

  2. Defense Waste Processing Facility (DWPF), Modular CSSX Unit (CSSX), and Waste Transfer Line System of Salt Processing Program (U)

    International Nuclear Information System (INIS)

    CHANG, ROBERT

    2006-01-01

    All of the waste streams from ARP, MCU, and SWPF processes will be sent to DWPF for vitrification. The impact these new waste streams will have on DWPF's ability to meet its canister production goal and its ability to support the Salt Processing Program (ARP, MCU, and SWPF) throughput needed to be evaluated. DWPF Engineering and Operations requested OBU Systems Engineering to evaluate DWPF operations and determine how the process could be optimized. The ultimate goal will be to evaluate all of the Liquid Radioactive Waste (LRW) System by developing process modules to cover all facilities/projects which are relevant to the LRW Program and to link the modules together to: (1) study the interfaces issues, (2) identify bottlenecks, and (3) determine the most cost effective way to eliminate them. The results from the evaluation can be used to assist DWPF in identifying improvement opportunities, to assist CBU in LRW strategic planning/tank space management, and to determine the project completion date for the Salt Processing Program

  3. Permian salt dissolution, alkaline lake basins, and nuclear-waste storage, Southern High Plains, Texas and New Mexico

    International Nuclear Information System (INIS)

    Reeves, C.C. Jr.; Temple, J.M.

    1986-01-01

    Areas of Permian salt dissolution associated with 15 large alkaline lake basins on and adjacent to the Southern High Plains of west Texas and eastern New Mexico suggest formation of the basins by collapse of strata over the dissolution cavities. However, data from 6 other alkaline basins reveal no evidence of underlying salt dissolution. Thus, whether the basins were initiated by subsidence over the salt dissolution areas or whether the salt dissolution was caused by infiltration of overlying lake water is conjectural. However, the fact that the lacustrine fill in Mound Lake greatly exceeds the amount of salt dissolution and subsidence of overlying beds indicates that at least Mound Lake basin was antecedent to the salt dissolution. The association of topography, structure, and dissolution in areas well removed from zones of shallow burial emphasizes the susceptibility of Permian salt-bed dissolution throughout the west Texas-eastern New Mexico area. Such evidence, combined with previous studies documenting salt-bed dissolution in areas surrounding a proposed high-level nuclear-waste repository site in Deaf Smith County, Texas, leads to serious questions about the rationale of using salt beds for nuclear-waste storage

  4. Salt Repository Project: Waste Package Program (WPP) modeling activiteis: FY 1984 annual report

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Simonson, S.A.; Pulsipher, B.A.

    1987-03-01

    The Pacific Northwest Laboratory (PNL) is supporting the US Department of Energy's (DOE) Salt Repository Project (SRP) through its Waste Package Program (WPP). During FY 1984, the WPP continued its program of waste package component development and interactions testing and application of the resulting data base to develop predictive models describing waste package degradation and radionuclide release. Within the WPP, the Modeling Task (Task 04 during FY 1984) was conducted to interpret the tests in such a way that scientifically defensible models can be developed for use in qualification of the waste package

  5. Density Fluctuation in Aqueous Solutions and Molecular Origin of Salting-Out Effect for CO2

    International Nuclear Information System (INIS)

    Ho, Tuan Anh; Ilgen, Anastasia

    2017-01-01

    Using molecular dynamics simulation, we studied the density fluctuations and cavity formation probabilities in aqueous solutions and their effect on the hydration of CO 2 . With increasing salt concentration, we report an increased probability of observing a larger than the average number of species in the probe volume. Our energetic analyses indicate that the van der Waals and electrostatic interactions between CO 2 and aqueous solutions become more favorable with increasing salt concentration, favoring the solubility of CO 2 (salting in). However, due to the decreasing number of cavities forming when salt concentration is increased, the solubility of CO 2 decreases. The formation of cavities was found to be the primary control on the dissolution of gas, and is responsible for the observed CO 2 salting-out effect. Finally, our results provide the fundamental understanding of the density fluctuation in aqueous solutions and the molecular origin of the salting-out effect for real gas.

  6. Nonmetal-metal transition in metal–molten-salt solutions

    NARCIS (Netherlands)

    Silvestrelli, P.-L.; Alavi, A.; Parrinello, M.; Frenkel, D.

    1996-01-01

    The method of ab initio molecular dynamics, based on finite-temperature density-functional theory, is used to study the nonmetal-metal transition in two different metal–molten-salt solutions, Kx(KCl)1-x and Nax(NaBr)1-x. As the excess metal concentration is increased the electronic density becomes

  7. Radioactive waste disposal in the Gorleben salt deposit

    International Nuclear Information System (INIS)

    Gizycki, P. von

    1985-01-01

    In the opinion of five experts, the protective function of the overlying rock as a barrier has turned out to be questionable after borings and measurements carried through at Gorleben. Moreover, the results have also raised doubts about the geological safety of the salt deposit as a barrier in the long run. The geological multibarrier concept must be discarded. Not only critics, but also 3 advocates from the field of official research on radioactive waste disposal state their opinion. (DG) [de

  8. Mercury Phase II Study - Mercury Behavior in Salt Processing Flowsheet

    International Nuclear Information System (INIS)

    Jain, V.; Shah, H.; Wilmarth, W. R.

    2016-01-01

    Mercury (Hg) in the Savannah River Site Liquid Waste System (LWS) originated from decades of canyon processing where it was used as a catalyst for dissolving the aluminum cladding of reactor fuel. Approximately 60 metric tons of mercury is currently present throughout the LWS. Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. In February 2015, a Mercury Program Team was established at the request of the Department of Energy to develop a comprehensive action plan for long-term management and removal of mercury. Evaluation was focused in two Phases. Phase I activities assessed the Liquid Waste inventory and chemical processing behavior using a system-by-system review methodology, and determined the speciation of the different mercury forms (Hg+, Hg++, elemental Hg, organomercury, and soluble versus insoluble mercury) within the LWS. Phase II activities are building on the Phase I activities, and results of the LWS flowsheet evaluations will be summarized in three reports: Mercury Behavior in the Salt Processing Flowsheet (i.e. this report); Mercury Behavior in the Defense Waste Processing Facility (DWPF) Flowsheet; and Mercury behavior in the Tank Farm Flowsheet (Evaporator Operations). The evaluation of the mercury behavior in the salt processing flowsheet indicates, inter alia, the following: (1) In the assembled Salt Batches 7, 8 and 9 in Tank 21, the total mercury is mostly soluble with methylmercury (MHg) contributing over 50% of the total mercury. Based on the analyses of samples from 2H Evaporator feed and drop tanks (Tanks 38/43), the source of MHg in Salt Batches 7, 8 and 9 can be attributed to the 2H evaporator concentrate used in assembling the salt batches. The 2H Evaporator is used to evaporate DWPF recycle water. (2) Comparison of data between Tank 21/49, Salt Solution Feed Tank (SSFT), Decontaminated Salt Solution Hold Tank (DSSHT), and Tank 50 samples suggests that the total mercury as well as speciated

  9. Use of ferric- and ferrous-salts in liquid waste treatment processes

    International Nuclear Information System (INIS)

    Efremenkov, V.M.; Toropov, I.G.; Toropova, V.V.; Satsukevich, V.M.; Davidov, J.P.; Jabrodsky, V.N.; Prokshin, N.E.

    1995-01-01

    Treatment of spent decontamination solutions is the most complicated task in the whole problem of management of liquid radioactive waste, because quite often they have complex compositions, which makes it difficult to find for them effective and non-expensive treatment technology. New methods of treatment of such a waste is proposed based on use of specific sorption ability of ferro- and ferri-species in solution. These species are often present in solution as the by-products, and in combination with other components of decontamination solution they can be used as initial substances for synthesis of valuable sorbents directly in treating solution. Using specific compositions and conditions in solution, it is possible to make liquid waste treatment process more effective and less expensive. Particular examples of this process is presented in this work

  10. Waste removal sequencing using ProdMod

    International Nuclear Information System (INIS)

    Paul, P.K.; Gregory, M.V.; Davis, N.R.; Brooke, J.N.

    1996-01-01

    The Savannah River Site (SRS) is starting to solidify its accumulated high-level radioactive waste into borosilicate glass in stainless steel canisters for eventual permanent storage. The in-tank precipitation process (ITP) and extended sludge processing (ESP) are two key operations in the waste processing complex. The supernate and dissolved salt from the waste storage tanks are transferred to the ITP process tank where the solution is decontaminated in batch processes. Soluble radioactive cesium is precipitated with sodium tetraphenylborate and strontium, uranium, and plutonium are adsorbed on monosodium titanate. The precipitate and adsorbent solids, which now contain the radionuclides, are concentrated using crossflow filters. The concentrated solids are sent to the high-level waste vitrification process. The decontaminated salt solution is sent to the low-level waste solidification process to form cement grout. In parallel with the precipitate operations, insoluble sludges that settled originally to the bottom of the waste tanks are reslurried and sent to ESP to undergo washing to reduce soluble salt content and aluminum dissolution, if required. In the vitrification process in the Defense Waste Processing Facility (DWPF), the concentrated precipitate from the ITP is mixed with the washed sludge from ESP and glass frit in proportion to form a stable borosilicate glass. A novel and fast-running Production Planning Model (ProdMod) has been developed to simulate the waste processing operation. This paper describes the application of ProdMod in sequencing the ITP batches and scheduling the ESP batches

  11. Cerebral salt wasting following traumatic brain injury

    Directory of Open Access Journals (Sweden)

    Peter Taylor

    2017-04-01

    Full Text Available Hyponatraemia is the most commonly encountered electrolyte disturbance in neurological high dependency and intensive care units. Cerebral salt wasting (CSW is the most elusive and challenging of the causes of hyponatraemia, and it is vital to distinguish it from the more familiar syndrome of inappropriate antidiuretic hormone (SIADH. Managing CSW requires correction of the intravascular volume depletion and hyponatraemia, as well as mitigation of on-going substantial sodium losses. Herein we describe a challenging case of CSW requiring large doses of hypertonic saline and the subsequent substantial benefit with the addition of fludrocortisone.

  12. Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

    International Nuclear Information System (INIS)

    Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas; Jegou, Christophe

    2007-01-01

    Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF 3 medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

  13. Sandia studies of high-level waste canisters and overpacks applicable for a salt repository

    International Nuclear Information System (INIS)

    Molecke, M.A.; Schaefer, D.W.; Glass, R.S.; Ruppen, J.A.

    1982-01-01

    An experimental program to develop candidate materials for use as high-level waste (HLW) overpacks or canisters in a salt repository has been in progress at Sandia National Laboratories since 1976. The main objective of this program is to provide a waste package barrier having a long lifetime in the chemical and physical environment of a repository. This paper summarizes the recent corrosion and metallurgical study results for the prime overpack material, TiCode-12, in the areas of uniform corrosion (extremely low rate and extent); local attack, e.g., pits and crevices (none were found); stress corrosion cracking susceptibility (no significant changes in macroscopic tensile properties were detected); hydrogen sorption-embrittlement effects; effects of gamma irradiation in solution; and sensitization effects (testing is still in process in the last three areas). Previous candidate screening analyses on other alloys and recent work on alternate overpack alloys are reviewed. All phases of these interrelated laboratory, hot-cell, and field experimental studies are described. 16 references, 8 figures, 4 tables

  14. UK-Nuclear decommissioning authority and US Salt-stone waste management issues

    International Nuclear Information System (INIS)

    Lawless, William; Whitton, John

    2007-01-01

    Available in abstract form only. Full text of publication follows: We update two case studies of stakeholder issues in the UK and US. Earlier versions were reported at Waste Management 2006 and 2007 and at ICEM 2005. UK: The UK nuclear industry has begun to consult stakeholders more widely in recent years. Historically, methods of engagement within the industry have varied, however, recent discussions have generally been carried out with the explicit understanding that engagement with stakeholders will be 'dialogue based' and will 'inform' the final decision made by the decision maker. Engagement is currently being carried out at several levels within the industry; at the national level (via the Nuclear Decommissioning Authority's (NDA) National Stakeholder Group (NSG)); at a local site level (via Site Stakeholder Groups) and at a project level (usually via the Best Practicable Environmental Option process (BPEO)). This paper updates earlier results by the co-author with findings from a second questionnaire issued to the NSG in Phase 2 of the engagement process. An assessment is made regarding the development of stakeholder perceptions since Phase 1 towards the NDA process. US: The US case study reviews the resolution of issues on salt-stone by Department of Energy's (DOE) Savannah River Site (SRS) Citizens Advisory Board (CAB), in Aiken, SC. Recently, SRS-CAB encouraged DOE and South Carolina's regulatory Department of Health and Environmental Control (SC-DHEC) to resolve a conflict preventing SC-DHEC from releasing a draft permit to allow SRS to restart salt-stone operations. It arose with a letter sent from DOE blaming the Governor of South Carolina for delay in restarting salt processing. In reply, the Governor blamed DOE for failing to assure that Salt Waste Processing Facility (SWPF) would be built. SWPF is designed to remove most of the radioactivity from HLW prior to vitrification, the remaining fraction destined for salt-stone. (authors)

  15. SALT4: a two-dimensional displacement discontinuity code for thermomechanical analysis in bedded salt deposits

    International Nuclear Information System (INIS)

    1983-04-01

    SALT4 is a two-dimensional analytical/displacement-discontinuity code designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. This code was developed by the University of Minnesota. This documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of the computer code, SALT4. The SALT4 code takes into account: (1) viscoelastic behavior in the pillars adjacent to excavations; (2) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (2) excavation sequence. Major advantages of the SALT4 code are: (1) computational efficiency; (2) the small amount of input data required; and (3) a creep law consistent with laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of SALT4, i.e., temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT4 code can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermal and thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT4 can also be used to verify fully numerical codes. This is similar to the use of analytic solutions for code verification. Although SALT4 was designed for analysis of bedded salt, it is also applicable to crystalline rock if the creep calculation is suppressed. In Section 1.5 of this document the code custodianship and control is described along with the status of verification, validation and peer review of this report

  16. Mineral Carbonation of Phosphogypsum Waste for Production of Useful Carbonate and Sulfate Salts

    Energy Technology Data Exchange (ETDEWEB)

    Mattila, Hannu-Petteri, E-mail: hmattila@abo.fi; Zevenhoven, Ron [Thermal and Flow Engineering Laboratory, Åbo Akademi University, Turku (Finland)

    2015-11-16

    Phosphogypsum (CaSO{sub 4}·2H{sub 2}O, PG) waste is produced in large amounts during phosphoric acid (H{sub 3}PO{sub 4}) production. Minor quantities are utilized in construction or agriculture, while most of the material is stockpiled, creating an environmental challenge to prevent pollution of natural waters. In principle, the gypsum waste could be used to capture several hundred megatonnes of carbon dioxide (CO{sub 2}). For example, when gypsum is converted to ammonium sulfate [(NH{sub 4}){sub 2}SO{sub 4}] with ammonia (NH{sub 3}) and CO{sub 2}, also solid calcium carbonate (CaCO{sub 3}) is generated. The ammonium sulfate can be utilized as a fertilizer or in other mineral carbonation processes that use magnesium silicate-based rock as feedstock, while calcium carbonate has various uses as, e.g., filler material. The reaction extent of the described process was studied by thermodynamic modeling and experimentally as a function of reactant concentrations and temperature. Other essential properties such as purity and quality of the solid products are also followed. Conversion efficiencies of >95% calcium from PG to calcium carbonate are obtained. Scalenohedral, rhombohedral, and prismatic calcite particles can be produced, although the precipitates contain certain contaminants such as rare earth metals and sulfur from the gypsum. A reverse osmosis membrane cartridge is also tested as an alternative and energy-efficient method of concentrating the ammonium sulfate salt solution instead of the traditional evaporation of the process solution.

  17. Mineral carbonation of phosphogypsum waste for production of useful carbonate and sulfate salts

    Directory of Open Access Journals (Sweden)

    Hannu-Petteri eMattila

    2015-11-01

    Full Text Available Phosphogypsum (CaSO4·2H2O waste is produced in large amounts during phosphoric acid (H3PO4 production. Minor quantities are utilized in construction or agriculture, while most of the material is stockpiled, creating an environmental challenge to prevent pollution of natural waters. In principle, the gypsum waste could be used to capture several hundred Mt of carbon dioxide (CO2. For example, when gypsum is converted to ammonium sulfate ((NH42SO4 with ammonia (NH3 and CO2, also solid calcium carbonate (CaCO3 is generated. The ammonium sulfate can be utilized as a fertilizer or in other mineral carbonation processes that use magnesium silicate-based rock as feedstock, while calcium carbonate has various uses as e.g. filler material. The reaction extent of the described process was studied by thermodynamic modeling and experimentally as a function of reactant concentrations and temperature. Other essential properties such as purity and quality of the solid products are also followed. Conversion efficiencies of >95% calcium from phosphogypsum to calcium carbonate are obtained. Scalenohedral, rhombohedral and prismatic calcite particles can be produced, though the precipitates contain certain contaminants such as rare earth metals and sulfur from the gypsum. A reverse osmosis membrane cartridge is also tested as an alternative and energy-efficient method of concentrating the ammonium sulfate salt solution instead of the traditional evaporation of the process solution.

  18. Vitrification: a solution for the wastes of wastes; La vitrification: ca chauffe pour les ultimes

    Energy Technology Data Exchange (ETDEWEB)

    Guihard, B. [Europlasma, 33 - Saint Medard en Jalles (France)

    1997-07-01

    The incineration of wastes generates other wastes (fly ashes) that concentrate a large amount of polluting substances (heavy metals, salts..). French law requires a stabilization of this kind of wastes before their storage. Today vitrification can be considered as an alternative to the stabilization and storage way, the vitrified products could be seen as an interesting material in the building industry or in road works. A few years ago the municipality of Bordeaux decided to launch a demonstration program and a REFIOM (fly ashes) vitrification unit has been operating since 1997. (A.C.)

  19. Precipitation process for the removal of technetium values from nuclear waste solutions

    Science.gov (United States)

    Walker, D.D.; Ebra, M.A.

    1985-11-21

    High efficiency removal of techetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

  20. The advantages of a salt/bentonite backfill for Waste Isolation Pilot Plant disposal rooms

    International Nuclear Information System (INIS)

    Butcher, B.M.; Novak, C.F.; Jercinovic, M.

    1991-04-01

    A 70/30 wt% salt/bentonite mixture is shown to be preferable to pure crushed salt as backfill for disposal rooms in the Waste Isolation Pilot Plant (WIPP). This report discusses several selection criteria used to arrive at this conclusion: the need for low permeability and porosity after closure, chemical stability with the surroundings, adequate strength to avoid shear erosion from human intrusion, ease of emplacement, and sorption potential for brine and radionuclides. Both salt and salt/bentonite are expected to consolidate to a final state of impermeability (i.e., ≤ 10 -18 m 2 ) adequate for satisfying federal nuclear regulations. Any advantage of the salt/bentonite mixture is dependent upon bentonite's potential for sorbing brine and radionuclides. Estimates suggest that bentonite's sorption potential for water in brine is much less than for pure water. While no credit is presently taken for brine sorption in salt/bentonite backfill, the possibility that some amount of inflowing brine would be chemically bound is considered likely. Bentonite may also sorb much of the plutonium, americium, and neptunium within the disposal room inventory. Sorption would be effective only if a major portion of the backfill is in contact with radioactive brine. Brine flow from the waste out through highly localized channels in the backfill would negate sorption effectiveness. Although the sorption potentials of bentonite for both brine and radionuclides are not ideal, they are distinctly beneficial. Furthermore, no detrimental aspects of adding bentonite to the salt as a backfill have been identified. These two observations are the major reasons for selecting salt/bentonite as a backfill within the WIPP. 39 refs., 16 figs., 6 tabs

  1. FRACTIONAL CRYSTALLIZATION OF HANFORD SINGLE-SHELL TANK WASTES FROM CONCEPT TO PILOT PLANT

    International Nuclear Information System (INIS)

    GENIESSE, D.J.; NELSON, E.A.; HAMILTON, D.W.; MAJORS, J.H.; NORDAHL, T.K.

    2006-01-01

    The Hanford site has 149 underground single-shell tanks (SST) storing mostly soluble, multi-salt mixed wastes resulting from Cold War era weapons material production. These wastes must be retrieved and the salts immobilized before the tanks can be closed to comply with an overall site-closure consent order entered into by the US Department of Energy, the Environmental Protection Agency, and the State of Washington. Water will be used to retrieve the wastes and the resulting solution will be pumped to a proposed pretreatment process where a high-curie (primarily 137 Cs) waste fraction will be separated from the other waste constituents. The separated waste streams will then be vitrified to allow for safe storage as an immobilized high-level waste, or low-level waste, borosilicate glass. Fractional crystallization, a common unit operation for production of industrial chemicals and pharmaceuticals, was proposed as the method to separate the salt wastes; it works by evaporating excess water until the solubilities of various species in the solution are exceeded (the solubility of a particular species depends on its concentration, temperature of the solution, and the presence of other ionic species in the solution). By establishing the proper conditions, selected pure salts can be crystallized and separated from the radioactive liquid phase. The aforementioned parameters, along with evaporation rate, proper agitation, and residence time, determine nucleation and growth kinetics and the resulting habit and size distribution of the product crystals. These crystals properties are important considerations for designing the crystallizer and solid/liquid separation equipment. A structured program was developed to (a) demonstrate that fractional crystallization could be used to pre-treat Hanford tank wastes and (b) provide data to develop a pilot plant design

  2. Statistical methods for mechanistic model validation: Salt Repository Project

    International Nuclear Information System (INIS)

    Eggett, D.L.

    1988-07-01

    As part of the Department of Energy's Salt Repository Program, Pacific Northwest Laboratory (PNL) is studying the emplacement of nuclear waste containers in a salt repository. One objective of the SRP program is to develop an overall waste package component model which adequately describes such phenomena as container corrosion, waste form leaching, spent fuel degradation, etc., which are possible in the salt repository environment. The form of this model will be proposed, based on scientific principles and relevant salt repository conditions with supporting data. The model will be used to predict the future characteristics of the near field environment. This involves several different submodels such as the amount of time it takes a brine solution to contact a canister in the repository, how long it takes a canister to corrode and expose its contents to the brine, the leach rate of the contents of the canister, etc. These submodels are often tested in a laboratory and should be statistically validated (in this context, validate means to demonstrate that the model adequately describes the data) before they can be incorporated into the waste package component model. This report describes statistical methods for validating these models. 13 refs., 1 fig., 3 tabs

  3. Treatment of organic waste solutions containing tributyl phosphate

    International Nuclear Information System (INIS)

    Drobnik, S.

    The two processes developed in the laboratory for treating waste solutions containing TBP, namely TBP separation with phosphoric acid and saponification were tested on a semi-industrial scale. A waste solution from the first phase of the Karlsruhe reprocessing plant was used

  4. Analysis of alloys and salt solutions by 'beta'-ray back-scattering

    Energy Technology Data Exchange (ETDEWEB)

    Bahadur, A; Maji, K D; Kumar, R [National Metallurgical Lab., Jamshedpur (India)

    1975-07-01

    This investigation reports the results of a study undertaken to assess the suitability of using the GM counter for measuring the intensity of ..beta..-backscattered radiation to determine the chemical composition of binary solid alloys, and aqueous salt solutions containing a metallic radical. The results indicate that the technique is not suitable for the determination of the composition of binary alloys since the error is in the range of 1.2 to 2.3 wt-% metal. The technique can be conveniently adapted for aqueous salt solutions where the maximum error is approximately 0.2 wt-% metal for metallic elements with atomic number greater than 20.

  5. HAW project. Demonstrative disposal of high-level radioactive wastes in the Asse salt mine

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Duijves, K.; Stippler, R.

    1988-01-01

    Since 1968 the GSF has been carrying out research and development programs for the final disposal of high-level radioactive waste (HAW) in salt formations. The heat producing waste has been simulated so far by means of electrical heaters and also cobalt-60-sources. In order to improve the final concept for HAW disposal in salt formations the complete technical system of an underground repository is to be tested in an one-to-one scale test facility. To satisfy the test objectives thirty high radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. The duration of testing will be approximately five years. For the handling of the radioactive canisters and their emplacement into the boreholes a system consisting of transportation casks, transportation vehicle, disposal machine, and borehole slider will be developed and tested. The actual scientific investigation program is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This program includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. The project is funded by the BMFT and the CEC and carrier out in close co-operation with the Netherlands Energy Research Foundation (ECN)

  6. Waste Isolation Pilot Plant Salt Decontamination Testing

    Energy Technology Data Exchange (ETDEWEB)

    Rick Demmer; Stephen Reese

    2014-09-01

    On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

  7. Behavior of gellan in aqueous-salt solutions and oilfield saline water

    Directory of Open Access Journals (Sweden)

    Zhanar Nurakhmetova

    2015-09-01

    Full Text Available The influence of storage time and temperature on the behavior of low acyl gellan (LAG was studied by viscometry and 1H NMR spectroscopy without salt addition. The viscometric results revealed that the effectiveness of salts to enhance gelation of gellan changes in the following order: BaСl2>CaCl2»MgCl2>KCl>NaCl. The sol-gel and liquid-solid phase transitions of gellan solutions were observed upon addition of oilfield water containing 73 g L-1 of alkaline and alkaline earth metal ions. The effectiveness of salts to induce the separation of liquid and solid phases changes in the sequence: NaCl>KCl>MgCl2»CaCl2»BaСl2. The hydrodynamic behavior of 0.5 wt.% gellan solution injected into the sand pack model with high (20 Darcy and lower (2 Darcy permeability is useful to model the oil reservoirs in the process of enhanced oil recovery.

  8. Waste acid detoxification and reclamation

    International Nuclear Information System (INIS)

    Brouns, T.M.; Stewart, T.L.

    1988-01-01

    Economically feasible processes that reduce the volume, quantity, and toxicity of metal-bearing waste acids by reclaiming, reusing, and recycling spent acids and metal salts are being developed and demonstrated. The acids used in the demonstrations are generated during metal-finishing operations used in nuclear fuel fabrication; HF-HNO 3 , HNO 3 , and HNO 3 -H 2 SO 4 wastes result from Zr etching, Cu stripping, and chemical milling of U. At discharge, wastes contain high concentrations of acid and one major metal impurity. The waste minimization process used to reclaim acid from these three streams incorporates three processes for acid regeneration and reclamation. Normally, HNO 3 remains in the bottoms when an aqueous acid solution is distilled; however, in the presence of H 2 SO 4 , HNO 3 will distill to the overhead stream. In this process, nitrates and fluorides present as free acid and metal salts can be reclaimed as acid for recycle to the metal-finishing processes. Uranium present in the chemical milling solution can be economically recovered from distillation bottoms and refined. Using acid distillation, the volume of chemical milling solution discharged as waste can be reduced by as much as 60% depending on the H 2 SO 4 concentration. A payback period of 2.2 years has been estimated for this process. The development and demonstration of precipitation and distillation processes for detoxification and reclamation of waste acid is supported by the US Department of Energy's Hazardous Waste Remedial Actions Program (HAZWRAP)

  9. Polyethylene encapsulatin of nitrate salt wastes: Waste form stability, process scale-up, and economics

    International Nuclear Information System (INIS)

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1991-07-01

    A polyethylene encapsulation system for treatment of low-level radioactive, hazardous, and mixed wastes has been developed at Brookhaven National Laboratory. Polyethylene has several advantages compared with conventional solidification/stabilization materials such as hydraulic cements. Waste can be encapsulated with greater efficiency and with better waste form performance than is possible with hydraulic cement. The properties of polyethylene relevant to its long-term durability in storage and disposal environments are reviewed. Response to specific potential failure mechanisms including biodegradation, radiation, chemical attack, flammability, environmental stress cracking, and photodegradation are examined. These data are supported by results from extensive waste form performance testing including compressive yield strength, water immersion, thermal cycling, leachability of radioactive and hazardous species, irradiation, biodegradation, and flammability. The bench-scale process has been successfully tested for application with a number of specific ''problem'' waste streams. Quality assurance and performance testing of the resulting waste form confirmed scale-up feasibility. Use of this system at Rocky Flats Plant can result in over 70% fewer drums processed and shipped for disposal, compared with optimal cement formulations. Based on the current Rocky Flats production of nitrate salt per year, polyethylene encapsulation can yield an estimated annual savings between $1.5 million and $2.7 million, compared with conventional hydraulic cement systems. 72 refs., 23 figs., 16 tabs

  10. Brine flow in heated geologic salt.

    Energy Technology Data Exchange (ETDEWEB)

    Kuhlman, Kristopher L.; Malama, Bwalya

    2013-03-01

    This report is a summary of the physical processes, primary governing equations, solution approaches, and historic testing related to brine migration in geologic salt. Although most information presented in this report is not new, we synthesize a large amount of material scattered across dozens of laboratory reports, journal papers, conference proceedings, and textbooks. We present a mathematical description of the governing brine flow mechanisms in geologic salt. We outline the general coupled thermal, multi-phase hydrologic, and mechanical processes. We derive these processes governing equations, which can be used to predict brine flow. These equations are valid under a wide variety of conditions applicable to radioactive waste disposal in rooms and boreholes excavated into geologic salt.

  11. CO2 Capture from Flue gas using Amino acid salt solutions

    DEFF Research Database (Denmark)

    Lerche, Benedicte Mai

    to storage. Typical solvents for the process are based on aqueous solutions of alkanolamines, such as mono-ethanolamine (MEA), but their use implies economic disadvantages and environmental complications. Amino acid salt solutions have emerged as an alternative to the alkanolamines, partlybecause...... for measuring of CO2 solubility based on the semi-flow method. A validation study of CO2 solubility in aqueous solutions of MEA is presented. Chapter 5 focuses on the determination of the chemical compositions of the precipitations, which arise in the five amino acid salt solutions upon CO2 absorption...

  12. Amino acid salt solutions as solvents in CO2 capture from flue gas

    DEFF Research Database (Denmark)

    Lerche, Benedicte Mai; Thomsen, Kaj; Stenby, Erling Halfdan

    New solvents based on the salts of amino acids have emerged as an alternative to the alkanolamine solutions, for the chemical absorption of CO2 from flue gas. But only few studies on amino acids as CO2 capturing agents have been performed so far. One of the interesting features of amino acid salt...... solutions is their ability to form solid precipitates upon the absorption of CO2. The occurrence of crystallization offers the possibility of increasing the CO2 loading capacity of the solvent. However, precipitation can also have negative effect on the CO2 capture process. The chemical nature of the solid...... of glycine, taurine, and lysine, while in the case of proline, and glutamic acid, the precipitate was found to be bicarbonate. These results give an important contribution to further understanding the potential of amino acid salt solutions in CO2 capture from flue gas....

  13. The HAW project. Demonstrative disposal of high-level radioactive wastes in the Asse salt mine

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Duijves, K.

    1988-04-01

    Since 1968 the GSF has been carrying out research and development programs for the final disposal of high-level radioactive waste (HAW) in salt formations. The heat producing waste has been simulated so far by means of electrical heaters and also cobalt-60-sources. In order to improve the final concept for HAW disposal in salt formations the complete technical system of an underground repository is to be tested in a one-to-one scale test facility. To satisfy the test objectives thirty high radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. The duration of testing will be approximately five years. For the handling of the radioactive canisters and their emplacement into the boreholes a system consisting of transportation casks, transportation vehicle, disposal machine, and borehole slider will be developed and tested. The actual scientific investigation program is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This program includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./HP)

  14. Crystallization of sodium nitrate from radioactive waste

    International Nuclear Information System (INIS)

    Krapukhin, V.B.; Krasavina, E.P.; Pikaev, A.K.

    1997-07-01

    From the 1940s to the 1980s, the Institute of Physical Chemistry of the Russian Academy of Sciences (IPC/RAS) conducted research and development on processes to separate acetate and nitrate salts and acetic acid from radioactive wastes by crystallization. The research objective was to decrease waste volumes and produce the separated decontaminated materials for recycle. This report presents an account of the IPC/RAS experience in this field. Details on operating conditions, waste and product compositions, decontamination factors, and process equipment are described. The research and development was generally related to the management of intermediate-level radioactive wastes. The waste solutions resulted from recovery and processing of uranium, plutonium, and other products from irradiated nuclear fuel, neutralization of nuclear process solutions after extractant recovery, regeneration of process nitric acid, equipment decontamination, and other radiochemical processes. Waste components include nitric acid, metal nitrate and acetate salts, organic impurities, and surfactants. Waste management operations generally consist of two stages: volume reduction and processing of the concentrates for storage, solidification, and disposal. Filtration, coprecipitation, coagulation, evaporation, and sorption were used to reduce waste volume. 28 figs., 40 tabs

  15. The salting-out of molibdoferrats(II from aqueous solutions by the organic solvents

    Directory of Open Access Journals (Sweden)

    Mykola V. Nikolenko

    2016-12-01

    Full Text Available The aim of this work was to develop a method for producing of molybdoferrate(II precipitates by salting-out them from aqueous solutions by means of organic solvents. Dependence of the composition of molybdoferrate(II precipitates on the pH of the reaction solutions was studied. Experiments on salting-out of molybdoferrate(II with various organic solvents were carried out. As a result it was found that the best reagent for the molybdoferrate(II salting-out is acetone. By its use, lowest quantity of the ammonium sulfate impurities was obtained. It is also of importance that by using of acetone the process of regeneration by distillation of the reaction solutions is characterized by the lowest energy consumption. A functional relationship between the solubility of molybdoferrates(II and dielectric constant of the medium was established. By increasing the dielectric constant of the solvent solubility of molybdoferrates(II rapidly increases. The linearized dependence ln(lnS–ln(1/e was proposed to predict the solubility of molybdoferrates(II in various aqueous-organic solutions.

  16. Laboratory scale vitrification of low-level radioactive nitrate salts and soils from the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Shaw, P.; Anderson, B.

    1993-07-01

    INEL has radiologically contaminated nitrate salt and soil waste stored above and below ground in Pad A and the Acid Pit at the Radioactive Waste Management Complex. Pad A contain uranium and transuranic contaminated potassium and sodium nitrate salts generated from dewatered waste solutions at the Rocky Flats Plant. The Acid Pit was used to dispose of liquids containing waste mineral acids, uranium, nitrate, chlorinated solvents, and some mercury. Ex situ vitrification is a high temperature destruction of nitrates and organics and immobilizes hazardous and radioactive metals. Laboratory scale melting of actual radionuclides containing INEL Pad A nitrate salts and Acid Pit soils was performed. The salt/soil/additive ratios were varied to determine the range of glass compositions (resulted from melting different wastes); maximize mass and volume reduction, durability, and immobilization of hazardous and radioactive metals; and minimize viscosity and offgas generation for wastes prevalent at INEL and other DOE sites. Some mixtures were spiked with additional hazardous and radioactive metals. Representative glasses were leach tested and showed none. Samples spiked with transuranic showed low nuclide leaching. Wasteforms were two to three times bulk densities of the salt and soil. Thermally co-processing soils and salts is an effective remediation method for destroying nitrate salts while stabilizing the radiological and hazardous metals they contain. The measured durability of these low-level waste glasses approached those of high-level waste glasses. Lab scale vitrification of actual INEL contaminated salts and soils was performed at General Atomics Laboratory as part of the INEL Waste Technology Development and Environmental Restoration within the Buried Waste Integrated Demonstration Program

  17. Evaporation of a sessile water drop and a drop of aqueous salt solution.

    Science.gov (United States)

    Misyura, S Y

    2017-11-07

    The influence of various factors on the evaporation of drops of water and aqueous salt solution has been experimentally studied. Typically, in the studies of drop evaporation, only the diffusive vapor transfer, radiation and the molecular heat conduction are taken into account. However, vapor-gas convection plays an important role at droplet evaporation. In the absence of droplet boiling, the influence of gas convection turns out to be the prevailing factor. At nucleate boiling, a prevailing role is played by bubbles generation and vapor jet discharge at a bubble collapse. The gas convection behavior for water and aqueous salt solution is substantially different. With a growth of salt concentration over time, the influence of the convective component first increases, reaches an extremum and then significantly decreases. At nucleate boiling in a salt solution it is incorrect to simulate the droplet evaporation and the heat transfer in quasi-stationary approximation. The evaporation at nucleate boiling in a liquid drop is divided into several characteristic time intervals. Each of these intervals is characterized by a noticeable change in both the evaporation rate and the convection role.

  18. Waste immobilization process development at the Savannah River Plant

    International Nuclear Information System (INIS)

    Charlesworth, D.L.

    1986-01-01

    Processes to immobilize various wasteforms, including waste salt solution, transuranic waste, and low-level incinerator ash, are being developed. Wasteform characteristics, process and equipment details, and results from field/pilot tests and mathematical modeling studies are discussed

  19. Cast iron deterioration with time in various aqueous salt solutions

    Indian Academy of Sciences (India)

    Unknown

    various aqueous salt solutions have been carried out using total immersion test ... circuit potential, Icorr, Tafel slopes, corrosion rate, have been calculated by standard methods. ..... Rao B V S 1980 in Maintenance for reliability (Bombay: Media.

  20. Disposal of decontaminated salts at the Savannah River Plant by solidification and burial

    International Nuclear Information System (INIS)

    Dukes, M.D.; Wolf, H.C.; Langton, C.A.

    1983-01-01

    The current plan for disposal of waste salt at the Savannah River Plant (SRP) is to immobilize the decontaminated salt solution by mixing with cement and SRP soil, and bury the resulting grout (saltstone) in a landfill. The grout which contains 37.8 wt % salt solution, 22.8 wt % Portland I-P cement, and 39.2 wt % SRP soil, was specially formulated to have a low permeability ( -10 cm/sec). This material will be mixed and placed in trenches. After setting, the saltstone will be covered with a clay cap, and an overburden of compacted native soil will be replaced. 6 references

  1. Permeability of natural rock salt from the Waste Isolation Pilot Plant (WIPP) during damage evolution and healing

    International Nuclear Information System (INIS)

    Pfeifle, T.W.; Hurtado, L.D.

    1998-06-01

    The US Department of Energy has developed the Waste Isolation Pilot Plant (WIPP) in the bedded salt of southeastern New Mexico to demonstrate the safe disposal of radioactive transuranic wastes. Four vertical shafts provide access to the underground workings located at a depth of about 660 meters. These shafts connect the underground facility to the surface and potentially provide communication between lithologic units, so they will be sealed to limit both the release of hazardous waste from and fluid flow into the repository. The seal design must consider the potential for fluid flow through a disturbed rock zone (DRZ) that develops in the salt near the shafts. The DRZ, which forms initially during excavation and then evolves with time, is expected to have higher permeability than the native salt. The closure of the shaft openings (i.e., through salt creep) will compress the seals, thereby inducing a compressive back-stress on the DRZ. This back-stress is expected to arrest the evolution of the DRZ, and with time will promote healing of damage. This paper presents laboratory data from tertiary creep and hydrostatic compression tests designed to characterize damage evolution and healing in WIPP salt. Healing is quantified in terms of permanent reduction in permeability, and the data are used to estimate healing times based on considerations of first-order kinetics

  2. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.4

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with construction, operation and sealing of disposal facilities for high-level waste in a salt dome. It is volume 4 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. The safety investigations were carried out for a deep-hole disposal facility located in the salt dome on Mors. In principle the results of the investigations also apply to a shaft/mine disposal facility. The facility is designed for the disposal of vitrified high-level waste in the shape of glass canisters. There is a low concentration of waste in each canister, approx. 10%. Furthermore, it was selected to place the waste in an intermediate storage for about 40 years prior to its final disposal. Consequently, heat generation in the waste at the time of final disposal will be modest, resulting in low temperature increase in the salt. As an example, the highest temperature increase will be approx. 40 deg. C. and it will occur at the edge of the hole five years after disposal has taken place. Prior to disposal, the glass canisters are encased in steel casks with 15 cm thick walls. Three canisters are placed in each cask, and 215 casks are stacked on top on one another in each deep-hole from a depth of 1200 m to 2500 m underground. The additional encasing is designed to protect the glass from dissolution should any brine reach the disposal facility. Furthermore, the steel cask protects the glass canisters against pressure from the wall of the hole. The technical design of the disposal facility gives it a considerable safety margin against unexpected events. The investigations proved Cretaceous strata to constitute an effective secondary barrier that would prevent radioactive matter from travelling from the underlying disposal facility to the biosphere. (BP)

  3. Results For The Third Quarter Calendar Year 2016 Tank 50H Salt Solution Sample

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-10-13

    In this memorandum, the chemical and radionuclide contaminant results from the Third Quarter Calendar Year 2016 (CY16) sample of Tank 50H salt solution are presented in tabulated form. The Third Quarter CY16 Tank 50H samples (a 200 mL sample obtained 6” below the surface (HTF-5-16-63) and a 1 L sample obtained 66” from the tank bottom (HTF-50-16-64)) were obtained on July 14, 2016 and received at Savannah River National Laboratory (SRNL) on the same day. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours, and the samples were pulled immediately after pump shut down. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Third Quarter CY16 sampling of Tank 50H were requested by Savannah River Remediation (SRR) personnel and details of the testing are presented in the SRNL TTQAP.

  4. Studies of the suitability of salt domes in east Texas basin for geologic isolation of nuclear wastes

    International Nuclear Information System (INIS)

    Kreitler, C.W.

    1979-01-01

    The suitability of salt domes in the east Texas basin (Tyler basin), Texas, for long-term isolation of nulear wastes is being evaluated. The major issues concern hydrogeologic and tectonic stability of the domes and potential natural resources in the basin. These issues are being approached by integration of dome-specific and regional hydrogeolgic, geologic, geomorphic, and remote-sensing investigations. Hydrogeologic studies are evaluating basinal hydrogeology and ground-water flow around the domes in order to determine the degree to which salt domes may be dissolving, their rates of solution, and the orientation of saline plumes in the fresh-water aquifers. Subsurface geologic studies are being conducted: (1) to determine the size and shape of specific salt domes, the geology of the strata immediately surrounding the domes, and the regional geology of the east Texas basin; (2) to understand the geologic history of dome growth and basin infilling; and (3) to evaluate potential natural resources. Geomorphic and surficial geology studies are determining whether there has been any dome growth or tectonic movement in the basin during the Quaternary. Remote-sensing studies are being conducted to determine: (1) if dome uplift has altered regional lineation patterns in Quaternary sediments; and (2) whether drainage density indicates Quaternary structural movement. On the basis of the screening criteria of Brunton et al (1978), Oakwood and Keechi domes have been chosen as possible candidate domes. Twenty-three domes have been eliminated because of insufficient size, too great a depth to salt, major hydrocarbon production, or previous use (such as liquid propane storage or salt mining or brining). Detailed geologic, hydrogeologic, and geomorphic investigations are now being conducted around Oakwood and Keechi salt domes

  5. 21 CFR 864.2875 - Balanced salt solutions or formulations.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Balanced salt solutions or formulations. 864.2875 Section 864.2875 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES HEMATOLOGY AND PATHOLOGY DEVICES Cell And Tissue Culture Products § 864.2875...

  6. Temperature calculations on different configurations for disposal of high-level reprocessing waste in a salt dome model

    International Nuclear Information System (INIS)

    Hamstra, J.; Kevenaar, J.W.A.M.

    1978-06-01

    A medium size salt dome is considered as a structure in which a repository can be located for all radioactive wastes to be produced within the scope of a postulated nuclear power program. A dominating design factor for the lay-out of such a waste repository is the temperature distribution in the salt dome resulting from decay heat released from the buried solidified high-level reprocessing waste. Two numerical models are presented for the calculation of both global and local rock salt temperatures. The results of calculations performed with these models are demonstrated to be compatible. Rock salt temperatures related to several types of burial configurations, ranging from two layer configurations with various vertical distances between the layers via a three and a four layer repository to deep bore hole concepts varying from 100 to 600 m bore hole depth, can therefore be calculated with one rather simple unit cell model. The results of these calculations indicate that rock salt temperatures can be kept within acceptable limits to realize a repository using standard mining techniques. The temperatures at mine galery level prove to be a dominating factor in the selection of a repository configuration. More detailed calculations of these temperatures taking into account the loading sequence and the cooling capacity of the mine ventilation are recommended. Finally the apparent advantages of a deep bore hole concept emphasize the need for R and D work with respect to advanced drilling techniques in order to achieve deep dry disposal bore holes that can be realized from a burial mine in the salt dome. (Auth.)

  7. Nuclear waste repository simulation experiments, Asse Salt Mine, Federal Republic of Germany. Annual report, 1983

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Luebker, D.; Coyle, A.; Kalia, H.

    1984-10-01

    This is the First Annual report (1983) which describes experiments simulating a nuclear waste respository at the 800-meter level of the Asse Salt Mine in the Federal Republic of Germany. The report describes the test equipment, the Asse Salt Mine, the pretest properties of the salt in the test gallery, and the mine proper. Also included are test data for the first six months of operations on brine migration rates, room closure rates, extensometer readings, stress measurements, and thermal mechanical behavior of the salt. The duration of the experiments will be two years, ending in December 1985. 3 references, 34 figures, 13 tables

  8. The function of packing materials in a high-level nuclear waste repository and some candidate materials: Salt Repository Project

    International Nuclear Information System (INIS)

    Bunnell, L.R.; Shade, J.W.

    1987-03-01

    Packing materials should be included in waste package design for a high-level nuclear waste repository in salt. A packing material barrier would increase confidence in the waste package by alleviating possible shortcomings in the present design and prolonging confinement capabilities. Packing materials have been studied for uses in other geologic repositories; appropriately chosen, they would enhance the confinement capabilities of salt repository waste packages in several ways. Benefits of packing materials include retarding or chemically modifying brines to reduce corrosion of the waste package, providing good thermal conductivity between the waste package and host rock, retarding or absorbing radionuclides, and reducing the massiveness of the waste package. These benefits are available at low percentage of total repository cost, if the packing material is properly chosen and used. Several candidate materials are being considered, including oxides, hydroxides, silicates, cement-based mixtures, and clay mixtures. 18 refs

  9. Ageing behaviour of unary hydroxides in trivalent metal salt solutions

    Indian Academy of Sciences (India)

    LDH)-like phases on ageing in solutions of Al or Cr salts. This reaction is similar to acid leaching and proceeds by a dissolution–reprecipitation mechanism offering a simple method of LDH synthesis, with implications for the accepted theories of ...

  10. Cerebral salt wasting: a report of three cases

    International Nuclear Information System (INIS)

    Younas, H.; Sabir, O.; Tarif, N.

    2015-01-01

    Hyponatremia secondary to the Syndrome of Inappropriate Anti-Diuretic Hormone (SIADH) secretion is commonly observed in patients with various neurological disorders. Cerebral Salt Wasting (CSW) resulting in hyponatremia is also an infrequent occurrence in some patients with neurological disorders. Confusion in differentiating CSW from SIADH may arise since both results in similar electrolyte disturbances. Herein, we report three cases of CSW with intracranial afflictions. CSW was diagnosed on the basis of fractional excretion of urinary sodium and uric acid along with extremely low serum uric acid. Improvements in serum sodium levels after saline hydration and fludrocortisone administration further supported the diagnosis. (author)

  11. Radioactive waste management

    International Nuclear Information System (INIS)

    Blomek, D.

    1980-01-01

    The prospects of nuclear power development in the USA up to 2000 and the problems of the fuel cycle high-level radioactive waste processing and storage are considered. The problems of liquid and solidified radioactive waste transportation and their disposal in salt deposits and other geologic formations are discussed. It is pointed out that the main part of the high-level radioactive wastes are produced at spent fuel reprocessing plants in the form of complex aqueous mixtures. These mixtures contain the decay products of about 35 isotopes which are the nuclear fuel fission products, about 18 actinides and their daughter products as well as corrosion products of fuel cans and structural materials and chemical reagents added in the process of fuel reprocessing. The high-level radioactive waste management includes the liquid waste cooling which is necessary for the short and middle living isotope decay, separation of some most dangerous components from the waste mixture, waste solidification, their storage and disposal. The conclusion is drawn that the seccessful solution of the high-level radioactive waste management problem will permit to solve the problem of the fuel cycle radioactive waste management as a whole. The salt deposits, shales and clays are the most suitable for radioactive waste disposal [ru

  12. Anion bridges drive salting out of a simple amphiphile from aqueous solution

    International Nuclear Information System (INIS)

    Bowron, D.T.; Finney, J.L.

    2002-01-01

    Neutron diffraction with isotope substitution has been used to determine the structural changes that occur on the addition of a simple salting-out agent to a dilute aqueous alcohol solution. The striking results obtained demonstrate a relatively simple process occurs in which interamphiphile anionic salt bridges are formed between the polar groups of the alcohol molecules. These ion bridges drive an increase in the exposure of the alcohol molecule nonpolar surface to the solvent water and hence point the way to their eventual salting out by the hydrophobic effect

  13. Waste Minimization Study on Pyrochemical Reprocessing Processes

    International Nuclear Information System (INIS)

    Boussier, H.; Conocar, O.; Lacquement, J.

    2006-01-01

    Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal

  14. The nonmetal-metal transition in solutions of metals in molten salts

    International Nuclear Information System (INIS)

    Tosi, M.P.

    1997-04-01

    Solutions of metals in molten salts present a rich phenomenology: localization of electrons in disordered ionic media, activated electron transport increasing with metal concentration towards a nonmetal-metal (NM-M) transition, and liquid-liquid phase separation. A brief review of progress in the study of these systems is given in this article, with main focus on the NM-M transition. After recalling the known NM-M behaviour of the component elements in the case of expanded fluid alkali metals and mercury and of solid halogens under pressure, the article focuses on liquid metal - molten salt solutions and traces the different NM-M behaviours of the alkalis in their halides and of metals added to polyvalent metal halides. (author). 51 refs, 2 figs

  15. Boron Removal in Radioactive Liquid Waste by Forward Osmosis Membrane

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Choi, Hei Min; Lee, Kune Woo; Moon Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    These wastes contain about 0.3-0.8 wt% boric acid and have been concentrated through an evaporation treatment. Boric acid tends to crystallize owing to its solubility, and to plug the evaporator. The volume reduction obtained through evaporation is limited by the amount of boric acid in the waste. As an emerging technology, forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, very few studies on the removal of boron by FO have been performed. The objective of this study is to evaluate the possibility of boron separation in radioactive liquid waste by FO. In this study, the performance of FO was investigated to separate boron in the simulated liquid waste under the factors such as pH, osmotic pressure, ionic strength of the solution, and membrane characteristic. The boron separation in radioactive borate liquid waste was investigated with an FO membrane. When the feed solution containing boron is treated by the FO membrane, the boron permeation depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7, and increases with an increase in the osmotic driving force. The boron flux of the CTA-ES and ALFD membrane orientation is higher than those of the CTA-NW and ALFF orientation, respectively. The boron permeation rate is constant regardless of the osmotic pressure and membrane orientation. The boron flux decreases slightly with the salt concentration, but it is not heavily influenced at a low salt concentration.

  16. Boron Removal in Radioactive Liquid Waste by Forward Osmosis Membrane

    International Nuclear Information System (INIS)

    Hwang, Dooseong; Choi, Hei Min; Lee, Kune Woo; Moon Jeikwon

    2014-01-01

    These wastes contain about 0.3-0.8 wt% boric acid and have been concentrated through an evaporation treatment. Boric acid tends to crystallize owing to its solubility, and to plug the evaporator. The volume reduction obtained through evaporation is limited by the amount of boric acid in the waste. As an emerging technology, forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination. FO is a membrane process in which water flows across a semi-permeable membrane from a feed solution of lower osmotic pressure to a draw solution of higher osmotic pressure. However, very few studies on the removal of boron by FO have been performed. The objective of this study is to evaluate the possibility of boron separation in radioactive liquid waste by FO. In this study, the performance of FO was investigated to separate boron in the simulated liquid waste under the factors such as pH, osmotic pressure, ionic strength of the solution, and membrane characteristic. The boron separation in radioactive borate liquid waste was investigated with an FO membrane. When the feed solution containing boron is treated by the FO membrane, the boron permeation depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7, and increases with an increase in the osmotic driving force. The boron flux of the CTA-ES and ALFD membrane orientation is higher than those of the CTA-NW and ALFF orientation, respectively. The boron permeation rate is constant regardless of the osmotic pressure and membrane orientation. The boron flux decreases slightly with the salt concentration, but it is not heavily influenced at a low salt concentration

  17. An experimental study on Sodalite and SAP matrices for immobilization of spent chloride salt waste

    Science.gov (United States)

    Giacobbo, Francesca; Da Ros, Mirko; Macerata, Elena; Mariani, Mario; Giola, Marco; De Angelis, Giorgio; Capone, Mauro; Fedeli, Carlo

    2018-02-01

    In the frame of Generation IV reactors a renewed interest in pyro-processing of spent nuclear fuel is underway. Molten chloride salt waste arising from the recovering of uranium and plutonium through pyro-processing is one of the problematic wastes for direct application of vitrification or ceramization. In this work, Sodalite and SAP have been evaluated and compared as potential matrices for confinement of spent chloride salt waste coming from pyro-processing. To this aim Sodalite and SAP were synthesized both in pure form and mixed with different glass matrices, i.e. commercially available glass frit and borosilicate glass. The confining matrices were loaded with mixed chloride salts to study their retention capacities with respect to the elements of interest. The matrices were characterized and leached for contact times up to 150 days at room temperature and at 90 °C. SEM analyses were also performed in order to compare the matrix surface before and after leaching. Leaching results are discussed and compared in terms of normalized releases with similar results reported in literature. According to this comparative study the SAP matrix with glass frit binder resulted in the best matrix among the ones studied, with respect to retention capacities for both matrix and spent fuel elements.

  18. Homemade sugar-salt solution for oral rehydration: knowledge of ...

    African Journals Online (AJOL)

    Up to 95% of these cases can be treated successfully with oral rehydration therapy. The aim of the study was to evaluate caregivers' knowledge of, attitudes to and use of homemade sugar and salt solution (SSS) in order to address the shortfalls. Differences between the knowledge, attitudes and practices in urban, rural and ...

  19. Flotation of copper-bearing shale in solutions of inorganic salts and organic reagents

    Directory of Open Access Journals (Sweden)

    Ratajczak Tomasz

    2017-01-01

    Full Text Available Flotation data on copper-bearing shale in aqueous solutions of inorganic electrolytes (NaCl, Na2SO4, KPF6, NH4Cl and organic reagents (ethylamine, propylamine as frothers were presented and discussed. The relationships between shale flotation, surface tension of aqueous solution and foam height during bubbling with air in the flotation system were presented. It has been found that flotation of shale in the presence of inorganic salts the yield was directly proportional to the surface tension of the aqueous solution of salt and inversely proportional to the height of the foam. On the other hand, for organic reagents solutions (short chain amines, a reverse effect has been observed in relation to the inorganic compounds studied, that is the yield of copper-bearing shale flotation and the foam height were inversely proportional to the surface tension of the amine solution.

  20. Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

    Energy Technology Data Exchange (ETDEWEB)

    Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-08

    An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

  1. Performance assessment of confinements for medium-level and α-contaminated waste. PACOMA project. Rock salt option

    International Nuclear Information System (INIS)

    Hirsekorn, R.P; Nies, A.; Rausch, H.; Storck, R.

    1991-03-01

    The objective of the contribution to the PACOMA project is to develop and demonstrate procedures for radiological safety of repositories in salt domes. An analogue study is performed by the Netherlands Energy Research Foundation ECN, where alternative disposal concepts in different salt formations were investigated. It is discussed, how far appropriate choice of the repository design parameters can improve the whole systems. The research covers deterministic calculations for three scenarios, the normal evolution scenario with subrosion of the salt dome, the combined brine intrusion scenario with brine intrusion from brine pockets and via an anhydrite vein, and the human intrusion scenario of solution mining of a storage cavern. For the combined brine intrusion scenario alternative waste inventories, different disposal concepts, variants of the layout of dams and sealings are investigated, and results obtained from variations of parameter values are discussed. Additionally, comprehensive probabilistic calculations have been carried out with the help of a Monte-Carlo simulation. Results are discussed in form of an uncertainty analysis of the maximum dose and global sensitivity studies of system parameters. The assessments main result is, that the reference case, where the reference repository design and the reference disposal concept are applied, deterministic calculations with best estimate values as well as probabilistic calculations do not manifest unacceptable risk. Investigation of alternative concepts and design variants indicate a high potential for system optimization. (orig./HP)

  2. Bentonite-Clay Waste Form for the Immobilization of Cesium and Strontium from Fuel Processing Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States); Mertz, Carol J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-01-01

    The physical properties of a surrogate waste form containing cesium, strontium, rubidium, and barium sintered into bentonite clay were evaluated for several simulant feed streams: chlorinated cobalt dicarbollide/polyethylene glycol (CCD-PEG) strip solution, nitrate salt, and chloride salt feeds. We sintered bentonite clay samples with a loading of 30 mass% of cesium, strontium, rubidium, and barium to a density of approximately 3 g/cm3. Sintering temperatures of up to 1000°C did not result in volatility of cesium. Instead, there was an increase in crystallinity of the waste form upon sintering to 1000ºC for chloride- and nitrate-salt loaded clays. The nitrate salt feed produced various cesium pollucite phases, while the chloride salt feed did not produce these familiar phases. In fact, many of the x-ray diffraction peaks could not be matched to known phases. Assemblages of silicates were formed that incorporated the Sr, Rb, and Ba ions. Gas evolution during sintering to 1000°C was significant (35% weight loss for the CCD-PEG waste-loaded clay), with significant water being evolved at approximately 600°C.

  3. Conductometric investigation of salt-free solutions of polyriboguanylic acid

    International Nuclear Information System (INIS)

    Kozlov, A.G.; Davydova, O.V.; Kargov, S.I.

    1993-01-01

    Salt-free solutions of various ionic forms of polyriboguanylic acid (poly(G)) were studied by the methods of conductometry and spectroscopy of annular dichroism. The Manning approach was employed to calculate transport characteristics and structural parameters of poly(G) on the basis of spectra permit putting poly(G) salts in two groups: the first one comprising NH 4 + -, Rb + -, K + -, Na + -, the second one - Cs + -, and Li + -poly(G). The assumption is made that Li + and Cs + ions, bound with concrete groups of polyanion in a specific way, can promote formation of a stable structure different from the one observed in the presence of the first group counterions. 25 refs., 3 figs

  4. Radioactive waste isolation in salt: Peer review of the Golder Associates draft test plan for in situ testing in an exploratory shaft in salt

    International Nuclear Information System (INIS)

    Hambley, D.F.; Mraz, D.Z.; Unterberter, R.R.

    1987-01-01

    This report documents the peer review conducted by Argonne National Laboratory of a document entitled ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared for Battelle Memorial Institute's Office of Nuclear Waste Isolation by Golder Associates, Inc. In general, the peer review panelists found the test plan to be technically sound, although some deficiencies were identified. Recommendations for improving the test plan are presented in this review report. A microfiche copy of the following unpublished report is attached to the inside back cover of this report: ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared by Golder Associates, Inc., for Office of Nuclear Waste Isolation, Battelle Memorial Institute, Columbus, Ohio (March 1985)

  5. Importance of waste composition for Life Cycle Assessment of waste management solutions

    DEFF Research Database (Denmark)

    Bisinella, Valentina; Götze, Ramona; Conradsen, Knut

    2017-01-01

    The composition of waste materials has fundamental influence on environmental emissions associated with waste treatment, recycling and disposal, and may play an important role also for the Life Cycle Assessment (LCA) of waste management solutions. However, very few assessments include effects...... of the waste composition and waste LCAs often rely on poorly justified data from secondary sources. This study systematically quantifiesy the influence and uncertainty on LCA results associated with selection of waste composition data. Three archetypal waste management scenarios were modelled with the waste...... LCA model EASETECH based on detailed waste composition data from the literature. The influence from waste composition data on the LCA results was quantified with a step-wise Global Sensitivity Analysis (GSA) approach involving contribution, sensitivity, uncertainty and discernibility analyses...

  6. Chemical durability and characterization of nuclear waste forms in a hydrothermal environment

    International Nuclear Information System (INIS)

    Braithwaite, J.W.; Johnstone, J.K.

    1979-01-01

    The chemical durability of a simulated copper borosilicate waste glass and titanate waste ceramic has been studied in hydrothermal environments which could possibly be encountered in a bedded salt or sub-sealed waste isolation repository. The major parameters investigated which affect matrix corrosion and cesium solubilization include solution saturation and equilibrium phenomena, solution composition (especially the Mg +2 ion concentration), pH, particle size, temperature, and time

  7. Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Robinson, Bruce Alan; Funk, David John; Stevens, Patrice Ann

    2016-01-01

    This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop®1, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options, ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a prescreening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented.

  8. Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-17

    This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop®1, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options, ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a prescreening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented.

  9. Comparison of slagging pyrolysis and molten salt incinerators for treating TRU waste at the INEL

    International Nuclear Information System (INIS)

    1977-11-01

    For the comparison, it is assumed that the waste product is required to meet the acceptance criteria of the Waste Isolation Pilot Plant, i.e., low leachability. Slagging pyrolysis incinerates combustible waste and melts noncombustible waste; the resulting slag forms a glass of low leachability. In the molten salt incinerator, combustion occurs at low temperatures with no accumulation of explosive gases, but the waste must have been previously sorted into combustibles and noncombustibles and then shredded. The economics, safety, and technical features are compared. Advantages, disadvantages, and areas of technical uncertainty of the two systems are listed. Development costs and schedules for the two types of incinerators are discussed

  10. Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

    International Nuclear Information System (INIS)

    Fallis, S.M.

    1973-12-01

    With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr 90 , Cs 137 , and Pu 239 . Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150 0 C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated

  11. Defense waste solidification studies, 200-S area. Savannah River Plant work request 860504, Project S-1780

    International Nuclear Information System (INIS)

    1977-05-01

    A scope of work and a venture guidance appraisal were prepared for a conceptual process and plant facilities for the solidification and long-term storage of radioactive wastes removed from underground storage tanks in the 241 F and H Areas at the Savannah River Plant. Conceptual design was based on incorporating the highly radioactive waste components in a borosilicate type glass. The scope of work describes facilities for: reclaiming liquid and sludge wastes from F and H area tank farms; separating the sludge from the liquid salt solution by physical processes; removing radioactive cesium from the salt solution by ion exchange techniques; incorporating the dried sludge and cesium in a borosilicate glass in stainless steel containers; evaporating the liquid salt solution and encapsulating the resulting salt cake in a stainless steel container; and storing two years' worth of glass and salt containing cyclinders in separate retrievable surface storage facilities. Operations are to be located in a new area, designated the 200-S area. A full complement of power, general, and service facilities are provided. The venture guidance appraisal based on FY 82 authorization and FY 87 turnover is $2,900,000,000. The figure is suitable for planning purposes only. The Glass-form Waste Case is a variation of the concrete-form waste case (or the Reference Plant Case) reported in DPE--3410. The new venture guidance appraisal for the concrete-form case (updated to a consistent time basis with the glass-form case) is $2,900,000,000, indicating no apparent cost advantage between the two waste product forms

  12. Radioactive waste management turning options into solution

    International Nuclear Information System (INIS)

    Neubauer, J.

    2000-10-01

    Most of the statements from representatives of different countries and institutions focused on the status of high level radioactive waste management, including spent fuel repositories. Speakers dealing with such topics were representatives from countries applying nuclear power for electricity production. They all reported about there national programs on technical and safety aspects of radioactive waste management. The panel discussion extended to questions on political sensitivities and public acceptance; in this respect, interesting developments are taking place in Finland and Sweden. It is expected that Finland will operate a final repository for spent fuel in 10 - 15 years from now, followed close by Sweden. Other countries, however, face decisions by policy makers and elected officials to postpone dealing with waste disposal concerns. In this connection there is relevant experience in our country, too - even in the absence of spent fuel or other high level waste to be dealt with. During personal discussions with representatives of other countries not using nuclear power it was confirmed that there are similar or shared experiences. Development of publicly -accepted solutions to radioactive waste management remains an important issue. Independent of the amount or the activity of radioactive waste, the public at large remains skeptical despite the agreement among experts that disposal can be safe, technically feasible and environmentally sound. In countries not using nuclear power there are only small quantities of low and intermediate level radioactive waste. Therefore, international co-operation among such countries should be an option. There was common understanding by representatives from Norway, Italy and Austria that international co-operation should be developed for treatment and disposal of such waste. For the moment however it has to be accepted that, for political reasons, it is not possible. Forced to deal with the lack of near-term solutions, the

  13. Selection of Technical Solutions for the Management of Radioactive Waste

    International Nuclear Information System (INIS)

    2017-07-01

    The objectives of this publication are to identify and critically review the criteria to be considered while selecting waste management technologies; summarize, evaluate, rank and compare the different technical solutions; and offer a systematic approach for selecting the best matching solution. This publication covers the management of radioactive waste from all nuclear operations, including waste generated from research reactors, power reactors, and nuclear fuel cycle activities including high level waste (HLW) arising from reprocessing and spent nuclear fuel declared as waste (SFW), as well as low level waste (LLW) and intermediate level waste (ILW) arising from the production and use of radionuclides in industry, agriculture, medicine, education and research.

  14. Adaptive resolution simulation of an atomistic DNA molecule in MARTINI salt solution

    NARCIS (Netherlands)

    Zavadlav, J.; Podgornik, R.; Melo, M.n.; Marrink, S.j.; Praprotnik, M.

    2016-01-01

    We present a dual-resolution model of a deoxyribonucleic acid (DNA) molecule in a bathing solution, where we concurrently couple atomistic bundled water and ions with the coarse-grained MAR- TINI model of the solvent. We use our fine-grained salt solution model as a solvent in the inner shell

  15. Slag-based materials for toxic metal and radioactive waste stabilization

    International Nuclear Information System (INIS)

    Langton, C.A.

    1989-01-01

    This paper discusses a salt solution that is a hazardous waste and has both corrosive and metal toxicity characteristics. Objectives of a wasteform designed to stabilize this solution are presented. Disposal site characterization studies are examined

  16. Mass transport in bedded salt and salt interbeds

    International Nuclear Information System (INIS)

    Hwang, Y.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-08-01

    Salt is the proposed host rock for geologic repositories of nuclear waste in several nations because it is nearly dry and probably impermeable. Although experiments and experience at potential salt sites indicate that salt may contain brine, the low porosity, creep, and permeability of salt make it still a good choice for geologic isolation. In this paper we summarize several mass-transfer and transport analyses of salt repositories. The mathematical details are given in our technical reports

  17. Hydrogen production under salt stress conditions by a freshwater Rhodopseudomonas palustris strain.

    Science.gov (United States)

    Adessi, Alessandra; Concato, Margherita; Sanchini, Andrea; Rossi, Federico; De Philippis, Roberto

    2016-03-01

    Hydrogen represents a possible alternative energy carrier to face the growing request for energy and the shortage of fossil fuels. Photofermentation for the production of H2 constitutes a promising way for integrating the production of energy with waste treatments. Many wastes are characterized by high salinity, and polluted seawater can as well be considered as a substrate. Moreover, the application of seawater for bacterial culturing is considered cost-effective. The aims of this study were to assess the capability of the metabolically versatile freshwater Rhodopseudomonas palustris 42OL of producing hydrogen on salt-containing substrates and to investigate its salt stress response strategy, never described before. R. palustris 42OL was able to produce hydrogen in media containing up to 3 % added salt concentration and to grow in media containing up to 4.5 % salinity without the addition of exogenous osmoprotectants. While the hydrogen production performances in absence of sea salts were higher than in their presence, there was no significant difference in performances between 1 and 2 % of added sea salts. Nitrogenase expression levels indicated that the enzyme was not directly inhibited during salt stress, but a regulation of its expression may have occurred in response to salt concentration increase. During cell growth and hydrogen production in the presence of salts, trehalose was accumulated as a compatible solute; it protected the enzymatic functionality against salt stress, thus allowing hydrogen production. The possibility of producing hydrogen on salt-containing substrates widens the range of wastes that can be efficiently used in production processes.

  18. Direct Analysis of Proteins from Solutions with High Salt Concentration Using Laser Electrospray Mass Spectrometry

    Science.gov (United States)

    Karki, Santosh; Shi, Fengjian; Archer, Jieutonne J.; Sistani, Habiballah; Levis, Robert J.

    2018-05-01

    The detection of lysozyme, or a mixture of lysozyme, cytochrome c, and myoglobin, from solutions with varying salt concentrations (0.1 to 250 mM NaCl) is compared using laser electrospray mass spectrometry (LEMS) and electrospray ionization-mass spectrometry (ESI-MS). Protonated protein peaks were observed up to a concentration of 250 mM NaCl in the case of LEMS. In the case of ESI-MS, a protein solution with salt concentration > 0.5 mM resulted in predominantly salt-adducted features, with suppression of the protonated protein ions. The constituents in the mixture of proteins were assignable up to 250 mM NaCl for LEMS and were not assignable above a NaCl concentration of 0.5 mM for ESI. The average sodium adducts () bound to the 7+ charge state of lysozyme for LEMS measurements from salt concentrations of 2.5, 25, 50, and 100 mM NaCl are 1.71, 5.23, 5.26, and 5.11, respectively. The conventional electrospray measurements for lysozyme solution containing salt concentrations of 0.1, 1, 2, and 5 mM NaCl resulted in of 2.65, 6.44, 7.57, and 8.48, respectively. LEMS displays an approximately two orders of magnitude higher salt tolerance in comparison with conventional ESI-MS. The non-equilibrium partitioning of proteins on the surface of the charged droplets is proposed as the mechanism for the high salt tolerance phenomena observed in the LEMS measurements. [Figure not available: see fulltext.

  19. Conceptual design of retrieval systems for emplaced transuranic waste containers in a salt bed depository. Final report

    International Nuclear Information System (INIS)

    Fogleman, S.F.

    1980-04-01

    The US Department of Energy and the Nuclear Regulatory Commission have jurisdiction over the nuclear waste management program. Design studies were previously made of proposed repository site configurations for the receiving, processing, and storage of nuclear wastes. However, these studies did not provide operational designs that were suitable for highly reliable TRU retrieval in the deep geologic salt environment for the required 60-year period. The purpose of this report is to develop a conceptual design of a baseline retrieval system for emplaced transuranic waste containers in a salt bed depository. The conceptual design is to serve as a working model for the analysis of the performance available from the current state-of-the-art equipment and systems. Suggested regulations would be based upon the results of the performance analyses

  20. Reconsolidated Salt as a Geotechnical Barrier

    International Nuclear Information System (INIS)

    Hansen, Francis D.; Gadbury, Casey

    2015-01-01

    Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt

  1. Reconsolidated Salt as a Geotechnical Barrier

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gadbury, Casey [USDOE Carlsbad Field Office, NM (United States)

    2015-11-01

    Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt

  2. Waste package reference conceptual designs for a repository in salt

    International Nuclear Information System (INIS)

    1986-02-01

    This report provides the reference conceptual waste package designs for the Office of Nuclear Waste Isolation to baseline these designs, thereby establishing the configuration and interface controls necessary, within the Civilian Radioactive Waste Management Program, formerly the National Waste Terminal Storage Program, to proceed in an orderly manner with preliminary design. Included are designs for the current reference defense high-level waste form from the Savannah River Plant, an optimized commercial high-level waste form, and spent fuel which has been disassembled and compacted into a circular bundle containing either 12 pressurized-water reactor or 30 boiling-water reactor assemblies. For compacted spent fuel, it appears economically attractive to standardize the waste package diameter for all fuel types. The reference waste packages consist of the containerized waste form, a low carbon steel overpack, and, after emplacement, a cover of salt. The overpack is a hollow cylinder with a flat head welded to each end. Its design thickness is the sum of the structural thickness required to resist the 15.4-MPa lithostatic pressure plus the corrosion allowance necessary to assure the required structural thickness will exist through the 1000-year containment period. Based on available data and completed analyses, the reference concepts described in this report satisfy all requirements of the US Department of Energy and the US Nuclear Regulatory Commission with reasonable assurance. In addition, sufficient design maturity exists to form a basis for preliminary design; these concepts can be brought under configuration control to serve as reference package designs. Development programs are identified that will be required to support these designs during the licensing process. 19 refs., 37 figs., 31 tabs

  3. Methods for separating actinides from reprocessing and refabrication plant wastes

    International Nuclear Information System (INIS)

    Tedder, D.W.; Finney, B.C.; Blomeke, J.O.

    1979-01-01

    Chemical processing flowsheets have been developed to partition actinides from all actinide-bearing LWR fuel reprocessing and refabrication plant wastes. These wastes include high-activity-level liquids, scrap recovery liquors, HEPA filters and incinerator ashes, and chemical salt wastes such as sodium carbonate scrub solutions, detergent cleanup streams, and alkaline off-gas scrubber liquors. The separations processes that were adopted for this study are based on solvent extraction, cation exchange chromatography, and leaching with Ce 4+ -HNO 3 solution

  4. Identifying suitable piercement salt domes for nuclear waste storage sites

    International Nuclear Information System (INIS)

    Kehle, R.; e.

    1980-08-01

    Piercement salt domes of the northern interior salt basins of the Gulf of Mexico are being considered as permanent storage sites for both nuclear and chemically toxic wastes. The suitable domes are stable and inactive, having reached their final evolutionary configuration at least 30 million years ago. They are buried to depths far below the level to which erosion will penetrate during the prescribed storage period and are not subject to possible future reactivation. The salt cores of these domes are themselves impermeable, permitting neither the entry nor exit of ground water or other unwanted materials. In part, a stable dome may be recognized by its present geometric configuration, but conclusive proof depends on establishing its evolutionary state. The evolutionary state of a dome is obtained by reconstructing the growth history of the dome as revealed by the configuration of sedimentary strata in a large area (commonly 3,000 square miles or more) surrounding the dome. A high quality, multifold CDP reflection seismic profile across a candidate dome will provide much of the necessary information when integrated with available subsurface control. Additional seismic profiles may be required to confirm an apparent configuration of the surrounding strata and an interpreted evolutionary history. High frequency seismic data collected in the near vicinity of a dome are also needed as a supplement to the CDP data to permit accurate depiction of the configuration of shallow strata. Such data must be tied to shallow drill hole control to confirm the geologic age at which dome growth ceased. If it is determined that a dome reached a terminal configuration many millions of years ago, such a dome is incapable of reactivation and thus constitutes a stable storage site for nuclear wastes

  5. Adaptive resolution simulation of salt solutions

    International Nuclear Information System (INIS)

    Bevc, Staš; Praprotnik, Matej; Junghans, Christoph; Kremer, Kurt

    2013-01-01

    We present an adaptive resolution simulation of aqueous salt (NaCl) solutions at ambient conditions using the adaptive resolution scheme. Our multiscale approach concurrently couples the atomistic and coarse-grained models of the aqueous NaCl, where water molecules and ions change their resolution while moving from one resolution domain to the other. We employ standard extended simple point charge (SPC/E) and simple point charge (SPC) water models in combination with AMBER and GROMOS force fields for ion interactions in the atomistic domain. Electrostatics in our model are described by the generalized reaction field method. The effective interactions for water–water and water–ion interactions in the coarse-grained model are derived using structure-based coarse-graining approach while the Coulomb interactions between ions are appropriately screened. To ensure an even distribution of water molecules and ions across the simulation box we employ thermodynamic forces. We demonstrate that the equilibrium structural, e.g. radial distribution functions and density distributions of all the species, and dynamical properties are correctly reproduced by our adaptive resolution method. Our multiscale approach, which is general and can be used for any classical non-polarizable force-field and/or types of ions, will significantly speed up biomolecular simulation involving aqueous salt. (paper)

  6. Nuclear waste repository simulation experiments, Asse salt mine, Federal Republic of Germany. Annual report 1984

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Feddersen, H.K.; Schwarzianeck, P.; Staupendahl, G.; Coyle, A.J.; Eckert, J.; Kalia, H.

    1986-07-01

    This is the second joint annual report (1984) on experiments simulating a nuclear waste repository at the 800-m (2624-ft) level of the Asse salt mine in the Federal Republic of Germany. This report describes the Asse salt mine, the test equipment, and the pretest properties of the salt in the mine and in the vicinity of the test area. Also included are test data for the first 19 months of operation on the following: brine migration rates, thermal mechanical behavior of the salt (including room closure, stress reading, and thermal profiles), and borehole gas pressures. In addition to field data, laboratory analyses of results are included in this report. The duration of the experiment will be 2 years, ending in December 1985

  7. Hanford Waste Physical and Rheological Properties: Data and Gaps

    Energy Technology Data Exchange (ETDEWEB)

    Wells, Beric E.; Kurath, Dean E.; Mahoney, Lenna A.; Onishi, Yasuo; Huckaby, James L.; Cooley, Scott K.; Burns, Carolyn A.; Buck, Edgar C.; Tingey, Joel M.; Daniel, Richard C.; Anderson, K. K.

    2011-08-01

    The Hanford Site in Washington State manages 177 underground storage tanks containing approximately 250,000 m3 of waste generated during past defense reprocessing and waste management operations. These tanks contain a mixture of sludge, saltcake and supernatant liquids. The insoluble sludge fraction of the waste consists of metal oxides and hydroxides and contains the bulk of many radionuclides such as the transuranic components and 90Sr. The saltcake, generated by extensive evaporation of aqueous solutions, consists primarily of dried sodium salts. The supernates consist of concentrated (5-15 M) aqueous solutions of sodium and potassium salts. The 177 storage tanks include 149 single-shell tanks (SSTs) and 28 double -hell tanks (DSTs). Ultimately the wastes need to be retrieved from the tanks for treatment and disposal. The SSTs contain minimal amounts of liquid wastes, and the Tank Operations Contractor is continuing a program of moving solid wastes from SSTs to interim storage in the DSTs. The Hanford DST system provides the staging location for waste feed delivery to the Department of Energy (DOE) Office of River Protection’s (ORP) Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP is being designed and constructed to pretreat and then vitrify a large portion of the wastes in Hanford’s 177 underground waste storage tanks.

  8. From cerebral salt wasting to diabetes insipidus with adipsia: case report of a child with craniopharyngioma.

    Science.gov (United States)

    Raghunathan, Veena; Dhaliwal, Maninder Singh; Gupta, Aditya; Jevalikar, Ganesh

    2015-03-01

    Craniopharyngioma is associated with a wide and interesting variety of sodium states both by itself and following surgical resection. These are often challenging to diagnose, especially given their dynamic nature during the perioperative course. We present the case of a boy with craniopharyngioma who had hyponatremia due to cerebral salt wasting preoperatively, developed diabetes insipidus (DI) intraoperatively and proceeded to develop hypernatremia with adipsic DI. Cerebral salt wasting is a rare presenting feature of craniopharyngioma. Postoperative DI can be associated with thirst abnormalities including adipsia due to hypothalamic damage; careful monitoring and a high index of suspicion are required for its detection. Adipsic DI is a difficult condition to manage; hence a conservative surgical approach is suggested.

  9. Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-May 1986

    International Nuclear Information System (INIS)

    1986-10-01

    DOE/CH/10140-05 is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

  10. Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-December 1986

    International Nuclear Information System (INIS)

    1987-06-01

    This document is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

  11. Worth its salt?

    Science.gov (United States)

    The idea that all underground salt deposits can serve as storage sites for toxic and nuclear waste does not always hold water—literally. According to Daniel Ronen and Brian Berkowitz of Israel's Weizmann Institute of Science and Yoseph Yechieli of the Geological Survey of Israel, some buried salt layers are in fact highly conductive of liquids, suggesting that wastes buried in their confines could easily leech into groundwater and nearby soil.When drilling three wells into a 10,000-year-old salt layer near the Dead Sea, the researchers found that groundwater had seeped into the layer and had absorbed some of its salt.

  12. Prostaglandin-E2 Mediated Increase in Calcium and Phosphate Excretion in a Mouse Model of Distal Nephron Salt Wasting.

    Directory of Open Access Journals (Sweden)

    Manoocher Soleimani

    Full Text Available Contribution of salt wasting and volume depletion to the pathogenesis of hypercalciuria and hyperphosphaturia is poorly understood. Pendrin/NCC double KO (pendrin/NCC-dKO mice display severe salt wasting under basal conditions and develop profound volume depletion, prerenal renal failure, and metabolic alkalosis and are growth retarded. Microscopic examination of the kidneys of pendrin/NCC-dKO mice revealed the presence of calcium phosphate deposits in the medullary collecting ducts, along with increased urinary calcium and phosphate excretion. Confirmatory studies revealed decreases in the expression levels of sodium phosphate transporter-2 isoforms a and c, increases in the expression of cytochrome p450 family 4a isotypes 12 a and b, as well as prostaglandin E synthase 1, and cyclooxygenases 1 and 2. Pendrin/NCC-dKO animals also had a significant increase in urinary prostaglandin E2 (PGE-2 and renal content of 20-hydroxyeicosatetraenoic acid (20-HETE levels. Pendrin/NCC-dKO animals exhibit reduced expression levels of the sodium/potassium/2chloride co-transporter 2 (NKCC2 in their medullary thick ascending limb. Further assessment of the renal expression of NKCC2 isoforms by quantitative real time PCR (qRT-PCR reveled that compared to WT mice, the expression of NKCC2 isotype F was significantly reduced in pendrin/NCC-dKO mice. Provision of a high salt diet to rectify volume depletion or inhibition of PGE-2 synthesis by indomethacin, but not inhibition of 20-HETE generation by HET0016, significantly improved hypercalciuria and salt wasting in pendrin/NCC dKO mice. Both high salt diet and indomethacin treatment also corrected the alterations in NKCC2 isotype expression in pendrin/NCC-dKO mice. We propose that severe salt wasting and volume depletion, irrespective of the primary originating nephron segment, can secondarily impair the reabsorption of salt and calcium in the thick ascending limb of Henle and/or proximal tubule, and reabsorption of

  13. Preliminary area selection considerations for radioactive waste repositories in bedded salt

    International Nuclear Information System (INIS)

    Wagoner, J.L.; Steinborn, T.L.

    1979-01-01

    This guide describes an approach to selection of areas of bedded salt which contain potentially suitable sites for the storage of radioactive waste. To evaluate a site selected by a license applicant, it is necessary to understand the technical site characteristics which should be considered in the preliminary phase of site selection. These site characteristics are presented here in checklist form, and each item is accompanied by a discussion which explains its significance. These qualitative considerations are used first to select an area of interest within a broad geologic or geomorphic region. Once an area has been selected, more quantitative information must be acquired to determine whether the proposed site meets the resultations for storage of nuclear waste

  14. Crushed Salt Constitutive Model

    International Nuclear Information System (INIS)

    Callahan, G.D.

    1999-01-01

    The constitutive model used to describe the deformation of crushed salt is presented in this report. Two mechanisms -- dislocation creep and grain boundary diffusional pressure solution -- are combined to form the basis for the constitutive model governing the deformation of crushed salt. The constitutive model is generalized to represent three-dimensional states of stress. Upon complete consolidation, the crushed-salt model reproduces the Multimechanism Deformation (M-D) model typically used for the Waste Isolation Pilot Plant (WIPP) host geological formation salt. New shear consolidation tests are combined with an existing database that includes hydrostatic consolidation and shear consolidation tests conducted on WIPP and southeastern New Mexico salt. Nonlinear least-squares model fitting to the database produced two sets of material parameter values for the model -- one for the shear consolidation tests and one for a combination of the shear and hydrostatic consolidation tests. Using the parameter values determined from the fitted database, the constitutive model is validated against constant strain-rate tests. Shaft seal problems are analyzed to demonstrate model-predicted consolidation of the shaft seal crushed-salt component. Based on the fitting statistics, the ability of the model to predict the test data, and the ability of the model to predict load paths and test data outside of the fitted database, the model appears to capture the creep consolidation behavior of crushed salt reasonably well

  15. Sealing of Anodised Aluminium Alloys with Rare Earth Metal Salt Solutions

    OpenAIRE

    Mansfield, C.; Chen, F.; Breslin, Carmel B.; Dull, D.

    1998-01-01

    Boric‐sulfuric acid anodized (BSAA) aluminum alloys have been sealed in hot solutions of cerium or yttrium salts. For comparison, sealing has also been performed in the presently used dilute chromate solution, boiling water, and a cold nickel fluoride solution. The corrosion resistance of the sealed BSAA Al alloys Al 2024, Al 6061, and Al 7075 has been evaluated by recording impedance spectra during exposure in 0.5 N NaCl for 7 days. Shorter or longer exposure times have also been used depend...

  16. Mössbauer spectroscopic studies of frozen aqueous solutions of Fe3+ salts

    DEFF Research Database (Denmark)

    Mørup, Steen; Knudsen, J. E.; Nielsen, M. K.

    1976-01-01

    Frozen aqueous solutions (FAS) of Fe3+ salts have been investigated by use of Mössbauer spectroscopy in order to study the conditions for formation of glasses. A general discussion of spin–spin relaxation in glasses is given, and we discuss how changes in the spin–spin relaxation time can...... be attributed to changes in the average separation between the iron ions. In the FeCl3–H2O system, it was found that homogeneous glasses are easily formed when the salt concentration is larger than 3.5 moles FeCl3 per 100 moles H2O. In more dilute samples, ice crystallizes during cooling, while the salt...

  17. Recovery of uranium from analytical waste solution

    International Nuclear Information System (INIS)

    Kumar, Pradeep; Anitha, M.; Singh, D.K.

    2016-01-01

    Dispersion fuels are considered as advance fuel for the nuclear reactor. Liquid waste containing significant quantity of uranium gets generated during chemical characterization of dispersion fuel. The present paper highlights the effort in devising a counter current solvent extraction process based on the synergistic mixture of D2EHPA and Cyanex 923 to recover uranium from such waste solutions. A typical analytical waste solution was found to have the following composition: U 3 O 8 (∼3 g/L), Al: 0.3 g/L, V: 15 ppm, Phosphoric acid: 3M, sulphuric acid : 1M and nitric acid : 1M. The aqueous solution is composed of mixture of either 3M phosphoric acid and 1M sulphuric acid or 1M sulphuric acid and 1M nitric acid, keeping metallic concentrations in the above mentioned range. Different organic solvents were tested. Based on the higher extraction of uranium with synergistic mixture of 0.5M D2EHPA + 0.125M Cyanex 923, it was selected for further investigation in the present work

  18. Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

    Energy Technology Data Exchange (ETDEWEB)

    Fallis, S.M.

    1973-12-01

    With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr/sup 90/, Cs/sup 137/, and Pu/sup 239/. Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150/sup 0/C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated.

  19. Removal of radioactive ions from nuclear waste solutions by electrodialysis

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, S [Radia Industries Co. Ltd., Takasaki, Gunma (Japan)

    1978-10-01

    Removal of radioactive ions was studied from low and medium level radioactive waste solutions by electrodialysis using ion exchange membranes. The test solutions contained /sup 137/Cs/sup +/, /sup 106/Ru/sup 3 +/ or fission products (F.P.) as active ions and NaCl, Na/sub 2/SO/sub 4/ or Ca(NO/sub 3/)/sub 2/ as inactive coexisting salts. The decontamination factor of the active ions was in the order: /sup 137/Cs/sup +/ (greater than 99%) > /sup 90/Sr/sup 2 +/ > F.P. > /sup 106/Ru/sup 3 +/. The dialysis time required to attain the saturation was the shortest for monovalent cations K/sup +/, Cs/sup +/ and Na/sup +/, intermediate for divalent cation Sr/sup 2 +/, and the longest for trivalent cation Ru/sup 3 +/. The ratio of the decontamination factor of an active ion eta sub( a) to the desalination factor of an inactive ion eta sub( b) was nearly equal to unity for /sup 24/Na, /sup 42/K, /sup 137/Cs and /sup 90/Sr. On the other hand, the apparent selective permeability of an active ion (A/sup +/) against Na/sup +/ ion, T sub(Na/sup +/) sup( a) was higher than unity for all the active ions tested, and was in the order of /sup 137/Cs > /sup 90/Sr > /sup 42/K > /sup 24/Na, where T sub(Na/sup +/) sup( a) is defined by the ratio of ..gamma..sub( a) to ..gamma..sub(Na/sup +/) with ..gamma..sub( a) being the ratio of dilution of A in the diluate the ..gamma..sub(Na/sup +/) being that of Na/sup +/ in the same diluate. The decontamination factor of the active ions did not depend significantly on the species and concentration of the coexistent salts or on the concentration of the active ions.

  20. The solubilities of significant organic compounds in HLW tank supernate solutions -- FY 1995 progress report

    International Nuclear Information System (INIS)

    Barney, G.S.

    1996-01-01

    At the Hanford Site organic compounds were measured in tank supernate simulant solutions during FY 1995. This solubility information will be used to determine if these organic salts could exist in solid phases (saltcake or sludges) in the waste where they might react violently with the nitrate or nitrite salts present in the tanks. Solubilities of sodium glycolate, succinate, and caproate salts; iron and aluminum and butylphosphate salts; and aluminum oxalate were measured in simulated waste supernate solutions at 25 degree C, 30 degree C, 40 degree C, and 50 degree C. The organic compounds were selected because they are expected to exist in relatively high concentrations in the tanks. The solubilities of sodium glycolate, succinate, caproate, and butylphosphate in HLW tank supernate solutions were high over the temperature and sodium hydroxide concentration ranges expected in the tanks. High solubilities will prevent solid sodium salts of these organic acids from precipitating from tank supernate solutions. The total organic carbon concentrations (YOC) of actual tank supernates are generally much lower than the TOC ranges for simulated supernate solutions saturated (at the solubility limit) with the organic salts. This is so even if all the dissolved carbon in a given tank and supernate is due to only one of these eight soluble compounds (an unlikely situation). Metal ion complexes of and butylphosphate and oxalate in supernate solutions were not stable in the presence of the hydroxide concentrations expected in most tanks. Iron and aluminum dibutylphosphate compounds reacted with hydroxide to form soluble sodium dibutylphosphate and precipitated iron and aluminum hydroxides. Aluminum oxalate complexes were also not stable in the basic simulated supernate solutions. Solubilities of all the organic salts decrease with increasing sodium hydroxide concentration because of the common ion effect of Na+. Increasing temperatures raised the solubilities of the organic

  1. Extraction of lithium from neutral salt solutions with fluorinated. beta. -diketones

    Energy Technology Data Exchange (ETDEWEB)

    Seeley, F.G.; Baldwin, W.H.

    1976-01-01

    Lithium was selectively extracted from near-neutral aqueous solutions of alkali metal salts. The mechanism by which this was achieved involves the formation of the trioctylphosphine oxide adduct of a lithium chelate of a fluorinated ..beta..-diketone, which is then readily extractable into an organic diluent. High separation factors were obtained from sodium, potassium, rubidium, and cesium. The selectivity of the fluorinated ..beta..-diketones for lithium over the alkaline earths was found to be poor. A suggested general flowsheet for the recovery of lithium from a salt brine concentrate is included.

  2. Extraction of lithium from neutral salt solutions with fluorinated β-diketones

    International Nuclear Information System (INIS)

    Seeley, F.G.; Baldwin, W.H.

    1976-01-01

    Lithium was selectively extracted from near-neutral aqueous solutions of alkali metal salts. The mechanism by which this was achieved involves the formation of the trioctylphosphine oxide adduct of a lithium chelate of a fluorinated β-diketone, which is then readily extractable into an organic diluent. High separation factors were obtained from sodium, potassium, rubidium, and cesium. The selectivity of the fluorinated β-diketones for lithium over the alkaline earths was found to be poor. A suggested general flowsheet for the recovery of lithium from a salt brine concentrate is included. (author)

  3. Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment

    International Nuclear Information System (INIS)

    Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

    1983-10-01

    Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables

  4. Development of advanced treatment technologies of radio-aqueous waste by an environmental friendly decomposition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwang Wook; Lee, E. H.; Moon, J. K. and others

    2006-01-15

    This project was aimed at the technology developments of electrode fabrication, electrolytic reactor design and fabrication, electrolytic processes and the analyses of electroytic reaction mechanisms, which were essential elements for the development of electrolytic systems to decompose or teat environmentally- friendly the several salts contained in waste solutions which are to be generated in the fields of nuclear/non-nuclear industries. Major research items carried our in this project were as follows; - Development of technologies to choose and fabricate the anodes and cathodes for the treatments of waste solutions containing nitrogen compounds and organics. - Development of a membrane electrolyzer stacked by mono-polar unit cells with independent series flow path of electrolytes - Development of an electrolyzer with a self-pH adjustment and an electrolytic process for ammonia decomposition by using the electrolyzer - Analysis of electrolytic reaction mechanism of ammonia - Development of an ion exchange membrane electrolyzer with only one discharge of pH-controlled electrolyte solution - Development of electrolytic dechlorination technology for the treatment of chloride molten salt waste salt from pyroprocess. - Development of technologies for treatment of high concentration nitric acid and recovery of waste organic solvent.

  5. Computer-aided evaluation of waste disposal cavern construction methods. ISBN 3-9801713-0-2.

    International Nuclear Information System (INIS)

    Knissel, W.; Fahlbusch, M.

    1991-01-01

    The disposal of hazardous radioactive and toxic wastes in deep geological formations is considered the safest solution in many countries. The Federal Republic of Germany prefers salt formations for underground disposal on account of the special advantages of the rock salt. Calculation methods are presented for the mathematical description of mining techniques for the construction of waste disposal salt caverns. The developed calculation model allows one to evaluate different construction methods with regard to expenses and time. (orig./DG) [de

  6. Influence of complexing on physicochemical properties of polymer-salt solutions. Vliyanie kompleksoobrazovaniya na fiziko-khimicheskie svojstva polimerno-solevykh rastvorov

    Energy Technology Data Exchange (ETDEWEB)

    Ostroushko, A A; Yushkova, S M; Koridze, N V; Skobkoreva, N V; Zhuravleva, L I; Palitskaya, T A; Antropova, S V; Ostroushko, I P [Ural' skij Gosudarstvennyj Univ., Sverdlovsk (Russian Federation) AN SSSR, Moscow (Russian Federation). Inst. Obshchej i Neorganicheskoj Khimii

    1993-06-01

    Using the methods of spectrophotometry, viscosimetry, conductometry the influence of salt-polymer complexing processes on physicochemical prperties of aqueous solutions of yttrium, barium, copper nitrates and formates with polyvinyl alcohol was studied. Change of dynamic viscosity, specific electric conductivity of solutions in the process of complexing was shown. Thermal effects of salt-polymer interaction were measured. It is shown that decrease of transition temperature of polymer to plastic state in films, temperature and effective activation energy of salt decomposition is also connected with complexing. Effective values of surface tension on the boundary with air are measured. Coefficients of cation diffusion in polymer-salt solutions are estimated.

  7. Carbon Market and Integrated Waste Solutions : a Case Study of ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Carbon Market and Integrated Waste Solutions : a Case Study of Indonesia ... dual purpose of helping developing countries achieve sustainable development ... with a view to devising integrated waste management solutions in urban centres ... and disseminate them through national, regional and international networks.

  8. Spectroscopic Characterization of HAN-Based Liquid Gun Propellants and Nitrate Salt Solutions

    Science.gov (United States)

    1989-01-15

    spectra were recorded of bubbles of a concentrated aqueous nitrate solution, mineral oil, and an aqueous surfactant solution. Polymethacrylic acid ...FTIR spectra of droplets of a concentrated aqueous nitrate salt based solution (LGP1845), of solid particles cf polymethacrylic acid packing IO, 3... polymethacrylic acid low density packing foam cut to a 3x4 mnn rectangle was levitated with a low acoustic power. The sample was easily I positioned in the

  9. Radiant energy dissipation during final storage of high-level radioactive waste in rock salt

    International Nuclear Information System (INIS)

    Ramthun, H.

    1981-08-01

    A final disposal concept is assumed where the high-active waste from 1400 t of uranium, remaining after conditioning, is solidified in borosilicate glass and distributed in 1.760 waste casks. These containers 1.2 m in height and 0.3 m in diameter are to be buried 10 years after the fuel is removed from the reactor in the 300 m deep boreholes of a salt dome. For this design the mean absorbed dose rates are calculated in the glass die (3.9 Gy/s), the steel mantle (0.26 Gy/s) and in the salt rock (0.12 Gy/s at a distance of 1 cm and 0.034 Gy/s at a distance of 9 cm from the container surface) valid at the beginning of disposal. The risk involved with these amounts of stored lattice energy is shortly discussed. (orig.) [de

  10. Electrochemical processing of low-level waste solutions

    International Nuclear Information System (INIS)

    Hobbs, D.T.; Ebra, M.A.

    1987-01-01

    The feasibility of treating low-level Savannah River Plant (SRP) waste solutions by an electrolytic process has been demonstrated. Although the economics of the process are marginal at the current densities investigated at the laboratory scale, there are a number of positive environmental benefits. These benefits include: (1) reduction in the levels of nitrate and nitrite in the waste, (2) further decontamination of 99 Tc and 106 Ru, and (3) reduction in the volume of waste

  11. Investigation of complexing in solutions of salt mixture In(NO3)3-NaVO3

    International Nuclear Information System (INIS)

    Nakhodnova, A.N.; Listratenko, I.V.

    1987-01-01

    Spectrophotometry, conductometry and pH-metry are used to investigate properties and composition of the solid phases of isomolar series of In(NO 3 ) 3 -NaVO 3 salt mixture solutions and series of solutions having constant concentration of one of the components and varied of the other. Results of investigation are presented. It is stated that in the investigated solution series in weakly acid media HPA with the ratios [In 3+ ]:[V 5+ ] being equal to 11:1, 6:1, and 1:9, are formed. Composition of the complexes is mainly defined by the ratio of the components in In(NO 3 ) 3 and NaVO 3 salt mixture solutions and the medium acidity. Compounds of Na 2 OxIn 2 O 3 x2.5V 2 O 5 x8.5H 2 O and Cs 2 OxIn 2 O 3 x6V 2 O 5 x6.5H 2 O empirical formulae are separated. Results of IR spectroscopy, derivatography and X-ray phase analysis of the corresponding salts are presented

  12. Economics of defense high level waste management in the United States

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1987-01-01

    Life-cycle costs of defense waste disposal, as presented in the foregoing sections, are summarized. Expressed as incremental costs per canister of waste deposited in a Federal geologic repository and per gallon of decontaminated salt solution immobilized in onsite concrete vaults, the tabulated values provide a measure of waste management costs relatively independent of the inventories of waste processed. Total values are about $350,000 per glass waste canister processed and $4.68 per gallon of decontaminated salt immobilized. These costs do not generally include contributions of fixed charges, such as capital costs, except in the case of transport and repository charges for which the quantities of waste handled determine allocation of fixed costs included in the fee assessments. 14 refs., 2 figs., 3 tabs

  13. History of waste tank 9 , 1955--1974

    International Nuclear Information System (INIS)

    Tharin, D.W.; Lohr, D.R.

    1979-01-01

    Tank 9 was placed in service as a receiver for Purex HLW on July 19, 1955. Filling was essentially completed in December 1955, and this original complement of waste remained in the tank until December 1965, when most of the liquid was decanted to allow refilling. In July 1966, the remaining liquid and approximately 15 inches of sludge were removed using 3000 to 3500 psi water introduced through nozzles to mobilize the sludge. The tank was then used as a receiver and cooler for aged HLW solution concentrated by the tank farm evaporator; the resulting crystallized salt, covered with saturated solution, is now stored in this tank. Inspections have been made of the tank interior and annulus by direct observation and with a 40-ft optical periscope. Analytical samples have been taken of the sludge, supernate, vapor, and leaked material in the annulus. Top-to-bottom profiles of radiation and temperature have been obtained in the annulus and tank, respectively, and measurements have been made of roof deflection caused by salt adhering to roof-supported cooling coils. Leaked waste was discovered in the annulus pan in October 1957. During 1958-59, the annulus pan was flushed nine times with water in 2000-gallon batches, jetting the waste and flush water into the primary tank. However, waste leakage into the annulus continued. The maximum liquid depth reached in the annulus was about 12 inches. This was jetted out in 1961., but some leakage continued theeeafter as indicated by roddings. The roddings showed no standing liquid by August 1964, but some liquid may have been present undera salt crust. In March 1972, salt depth in the annulus was measured to be 8 to 10 in., and the bottom 3 in. was quite wet. The salt remains although most of the liquid has been removed

  14. Chemical decontamination method for radioactive metal waste

    International Nuclear Information System (INIS)

    Onuma, Tsutomu; Tanaka, Akio; Shibuya, Sadao.

    1991-01-01

    When contaminants mainly composed of copper remained on the surface of stainless steel wastes sent from an electrolytic reduction as a first step are chemically decontaminated, metal wastes are discriminated to carbon steel wastes and stainless steel wastes. Then, the carbon steel wastes are applied only with the first step of immersing in a sulfuric acid solution, and stainless steel wastes are applied with a first step of immersing into a sulfuric acid solution for electrolytic reduction for a predetermined period of time and a second step of immersing into a liquid in which an oxidative metal salt is added to sulfuric acid. The decontamination liquid which is used for immersing the stainless steel wastes in the second step and the oxidation force of which is lowered is used as the sulfuric acid solution in the first step for the carbon steel wastes. In view of the above, the decontamination liquid of the second step can be utilized most effectively, enabling to greatly decrease the secondary wastes and to improve decontamination efficiency. (T.M.)

  15. Thickening agent for flood water in secondary recovery of oil and for other aqueous salt solutions

    Energy Technology Data Exchange (ETDEWEB)

    Roth, H H

    1966-04-14

    Alkenyl-aromatic polymer sulfonates are good thickeners for some aqueous solutions, but addition of salts to such solutions reduces the desirable viscosity. High-molecular, water-soluble alkenyl-aromatic polymers which carry sulfonic acid or sulfonate groups substituted at the aromatic nuclei yield thickened solutions (e.g., for waterflooding) which are not influenced by the presence of water-soluble salts. Such polymers are derivatives of polyvinyltoluene, alone or in combination with about 5% acrylonitrile. It was also found that such thickening agents are less adsorbed on the rock matrix in a waterflood formation. (1 claim)

  16. A reactive distillation process for the treatment of LiCl-KCl eutectic waste salt containing rare earth chlorides

    Energy Technology Data Exchange (ETDEWEB)

    Eun, H.C., E-mail: ehc2004@kaeri.re.kr; Choi, J.H.; Kim, N.Y.; Lee, T.K.; Han, S.Y.; Lee, K.R.; Park, H.S.; Ahn, D.H.

    2016-11-15

    The pyrochemical process, which recovers useful resources (U/TRU metals) from used nuclear fuel using an electrochemical method, generates LiCl-KCl eutectic waste salt containing radioactive rare earth chlorides (RECl{sub 3}). It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste salt in a hot-cell facility. For this reason, a reactive distillation process using a chemical agent was achieved as a method to separate rare earths from the LiCl-KCl waste salt. Before conducting the reactive distillation, thermodynamic equilibrium behaviors of the reactions between rare earth (Nd, La, Ce, Pr) chlorides and the chemical agent (K{sub 2}CO{sub 3}) were predicted using software. The addition of the chemical agent was determined to separate the rare earth chlorides into an oxide form using these equilibrium results. In the reactive distillation test, the rare earth chlorides in LiCl-KCl eutectic salt were decontaminated at a decontamination factor (DF) of more than 5000, and were mainly converted into oxide (Nd{sub 2}O{sub 3}, CeO{sub 2}, La{sub 2}O{sub 3}, Pr{sub 2}O{sub 3}) or oxychloride (LaOCl, PrOCl) forms. The LiCl-KCl was purified into a form with a very low concentration (<1 ppm) for the rare earth chlorides.

  17. Colloidal agglomerates in tank sludge: Impact on waste processing

    International Nuclear Information System (INIS)

    Bunker, B.C.; Martin, J.E.

    1998-01-01

    'Insoluble colloidal sludges in hazardous waste streams such as tank wastes can pose serious problems for waste processing, interfering with retrieval, transport, separation, and solidification procedures. Properties of sediment layers and sludge suspensions such as slurry viscosities, sedimentation rates, and final sediment densities can vary by orders of magnitude depending on the particle types present, the degree to which the particles agglomerate or stick to each other, and on a wide range of processing parameters such as solution shear rates, pH, salt content, and temperature. The objectives of this work are to: (1) understand the factors controlling the nature and extent of colloidal agglomeration under expected waste processing conditions; (2) determine how agglomeration phenomena influence physical properties relevant to waste processing including rheology, sedimentation, and filtration; and (3) develop strategies for optimizing processing conditions via control of agglomeration phenomena. Insoluble colloidal sludges in hazardous waste streams such as tank wastes can pose serious problems for waste processing, interfering with retrieval, transport, separation, and solidification procedures. Properties of sediment layers and sludge suspensions such as slurry viscosities, sedimentation rates, and final sediment densities can vary by orders of magnitude depending on the particle types present, the degree to which the particles agglomerate or stick to each other, and on a wide range of processing parameters such as solution shear rates, pH, salt content, and temperature. The objectives of this work are to: (1) understand the factors controlling the nature and extent of colloidal agglomeration under expected waste processing conditions; (2) determine how agglomeration phenomena influence physical properties relevant to waste processing including rheology, sedimentation, and filtration; and (3) develop strategies for optimizing processing conditions via control

  18. Management and deposition of nuclear, toxic and hazardous wastes

    International Nuclear Information System (INIS)

    Dalston, Regina Celia Reboucas; Montalvao, Renata; Nascimento, Igor; Oliveira, Maristela Aparecida de; Motta, Rondineli; Morais, Magda de; Dantas, Alberto Pinheiro

    2005-01-01

    the main guidelines of the management program of toxic, radioactive and hazardous wastes which are applicable to the graduation laboratories at the Catholic University of Brasilia (UCB) are presented. The main advantages and possibilities of applications of processes for inertization of salts of heavy metals by precipitation with sodium metasilicate solution and methods of adsorption of toxic wastes in minerals such as bentonites and silicates are discussed. In the treatment of waste, the use of effective technologies enables solid wastes to be processed and prepared in accordance with the existing rules and resolutions. The applicability of supports of polymeric resins catalyzed for moulding and final disposal of toxic wastes, previously treated and converted in the form of insoluble salts is presented. It is also suggested the use of polymeric supports for the containment of radioactive wastes

  19. Waste management outlook for mountain regions: Sources and solutions.

    Science.gov (United States)

    Semernya, Larisa; Ramola, Aditi; Alfthan, Björn; Giacovelli, Claudia

    2017-09-01

    Following the release of the global waste management outlook in 2015, the United Nations Environment Programme (UN Environment), through its International Environmental Technology Centre, is elaborating a series of region-specific and thematic waste management outlooks that provide policy recommendations and solutions based on current practices in developing and developed countries. The Waste Management Outlook for Mountain Regions is the first report in this series. Mountain regions present unique challenges to waste management; while remoteness is often associated with costly and difficult transport of waste, the potential impact of waste pollutants is higher owing to the steep terrain and rivers transporting waste downstream. The Outlook shows that waste management in mountain regions is a cross-sectoral issue of global concern that deserves immediate attention. Noting that there is no 'one solution fits all', there is a need for a more landscape-type specific and regional research on waste management, the enhancement of policy and regulatory frameworks, and increased stakeholder engagement and awareness to achieve sustainable waste management in mountain areas. This short communication provides an overview of the key findings of the Outlook and highlights aspects that need further research. These are grouped per source of waste: Mountain communities, tourism, and mining. Issues such as waste crime, plastic pollution, and the linkages between exposure to natural disasters and waste are also presented.

  20. Solid solutions of platinum(II) and palladium(II) oxalato-complex salt as precursors of nanoalloys

    Science.gov (United States)

    Zadesenets, A. V.; Asanova, T. I.; Vikulova, E. S.; Filatov, E. Yu.; Plyusnin, P. E.; Baidina, I. A.; Asanov, I. P.; Korenev, S. V.

    2013-03-01

    A solid solution of platinum (II) and palladium (II) oxalato-complex salt, (NH4)2[Pt0.5Pd0.5(C2O4)2]·2H2O, has been synthesized and studied as a precursor for preparing bimetallic PtPd nanoparticles through its thermal decomposition. The smallest homogenous bimetallic PtPd nanoparticles were found to form in hydrogen and helium atmospheres. The annealing temperature and time have low effect on the bimetallic particles size. Comparative analysis of structural and thermal properties of the solid solution and individual Pt, Pd oxalato-complex salts was performed to investigate a mechanism of thermal decomposition of (NH4)2[Pt0.5Pd0.5(C2O4)2]·2H2O. Based on in situ X-ray photoemission spectroscopy investigation it was proposed a mechanism of formation of bimetallic PtPd nanoparticles from the solid-solution oxalato-complex salt during thermal decomposition.

  1. Evaluation of iron-base materials for waste package containers in a salt repository

    International Nuclear Information System (INIS)

    Westerman, R.E.; Nelson, J.L.; Kuhn, W.L.; Basham, S.G.; Moak, D.A.; Pitman, S.G.

    1983-11-01

    Design studies for high-level nuclear waste packages for salt repositories have identified low-carbon steel as a candidate material for containers. Among the requirements are strength, corrosion resistance, and fabricability. The studies of the corrosion resistance and structural stability of iron-base materials (particularly low-carbon steel) are treated in this paper. The materials have been exposed in brines that are characteristic of the potential sites for salt repositories. The effects of temperature, radiation level, oxygen level and other parameters are under investigation. The initial development of corrosion models for these environments is presented with discussion of the key mechanisms under consideration. 6 references, 5 figures

  2. Sedimentary modelling and nuclear-waste disposal

    International Nuclear Information System (INIS)

    Van Loon, A.J.

    1982-01-01

    Nuclear energy is an important source of energy. Recently a slow down is experienced in its growth rate, due to the following factors: a) the supposed shortage of uranium; b) the fear for the consequences of nuclear accident, and c) the problem of nuclear wastes. Two types of waste are distinguished: a) fission products and actinides, and b) operational waste. The United States have started a program that must lead in 1989 to the first final storage of such waste in salt. Open-pit mines and oil-well drilling are discussed as possible solutions for operational waste storage

  3. Heat transfer analysis of the waste-container sleeve/salt configuration

    International Nuclear Information System (INIS)

    Callahan, G.D.; Ratigan, J.L.; Russell, J.E.; Fossum, A.F.

    1975-01-01

    Prior to this investigation, the heat transport considered was only that of straight conduction. The waste container, air gap, and sleeve arrangement was considered to be a single, consistent, time-dependent, heat-generating unit in intimate contact with the salt. The conduction model does not accurately model the heat transfer mechanisms available. Thus radiation and combined radiation and convection must also be considered in the determination of the temperature field. As would be expected, the canister temperatures are higher for the case of radiation across the airgap than those that result from conduction when the canister is in intimate contact with the salt. For the radiation case, the canister temperatures rise rapidly to a temperature of approximately 1,140 0 F and maintain an almost steady state condition for one year whereafter the temperatures slowly decrease. The far field temperatures, near the pillar centerline, are essentially equivalent for all cases. As time proceeds, the far field temperatures of the conduction models are about 15% different

  4. Ion clustering in aqueous salt solutions near the liquid/vapor interface

    Directory of Open Access Journals (Sweden)

    J.D. Smith

    2016-03-01

    Full Text Available Molecular dynamics simulations of aqueous NaCl, KCl, NaI, and KI solutions are used to study the effects of salts on the properties of the liquid/vapor interface. The simulations use the models which include both charge transfer and polarization effects. Pairing and the formation of larger ion clusters occurs both in the bulk and surface region, with a decreased tendency to form larger clusters near the interface. An analysis of the roughness of the surface reveals that the chloride salts, which have less tendency to be near the surface, have a roughness that is less than pure water, while the iodide salts, which have a greater surface affinity, have a larger roughness. This suggests that ions away from the surface and ions near the surface affect the interface in opposite ways.

  5. NUCLEAR WASTE GLASSES: CONTINUOUS MELTING AND BULK VITRIFICAITON

    International Nuclear Information System (INIS)

    KRUGER, A.A.

    2008-01-01

    This contribution addresses various aspects of nuclear waste vitrification. Nuclear wastes have a variety of components and composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee the glass can be easily made and resist environmental degradation. Glass formulation is facilitated by developing property-composition models, and the strategy of model development and application is reviewed. However, the large variability of waste compositions presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slow down the process; molten salts may accumulate in container refractory walls; the glass on cooling may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed

  6. Conditioning of radioactive waste solutions by cementation

    International Nuclear Information System (INIS)

    Vejmelka, P.; Rudolph, G.; Kluger, W.; Koester, R.

    1992-02-01

    For the cementation of the low and intermediate level evaporator concentrates resulting from the reprocessing of spent fuel numerous experiments were performed to optimize the waste form composition and to characterize the final waste form. Concerning the cementation process, properties of the waste/cement suspension were investigated. These investigations include the dependence of viscosity, bleeding, setting time and hydration heat from the waste cement slurry composition. For the characterization of the waste forms, the mechanical, thermal and chemical stability were determined. For special cases detailed investigations were performed to determine the activity release from waste packages under defined mechanical and thermal stresses. The investigations of the interaction of the waste forms with aqueous solutions include the determination of the Cs/Sr release, the corrosion resistance and the release of actinides. The Cs/Sr release was determined in dependence of the cement type, additives, setting time and sample size. (orig./DG) [de

  7. The effectiveness of sodium hydroxide (NaOH) and sodium carbonate (Na2CO3) on the impurities removal of saturated salt solution

    Science.gov (United States)

    Pujiastuti, C.; Ngatilah, Y.; Sumada, K.; Muljani, S.

    2018-01-01

    Increasing the quality of salt can be done through various methods such as washing (hydro-extraction), re-crystallization, ion exchange methods and others. In the process of salt quality improvement by re-crystallization method where salt product diluted with water to form saturated solution and re-crystallized through heating process. The quality of the salt produced is influenced by the quality of the dissolved salt and the crystallization mechanism applied. In this research is proposed a concept that before the saturated salt solution is recrystallized added a chemical for removal of the impurities such as magnesium ion (Mg), calcium (Ca), potassium (K) and sulfate (SO4) is contained in a saturated salt solution. The chemical reagents that used are sodium hydroxide (NaOH) 2 N and sodium carbonate (Na2CO3) 2 N. This research aims to study effectiveness of sodium hydroxide and sodium carbonate on the impurities removal of magnesium (Mg), calcium (Ca), potassium (K) and sulfate (SO4). The results showed that the addition of sodium hydroxide solution can be decreased the impurity ions of magnesium (Mg) 95.2%, calcium ion (Ca) 45%, while the addition of sodium carbonate solution can decreased magnesium ion (Mg) 66.67% and calcium ion (Ca) 77.5%, but both types of materials are not degradable sulfate ions (SO4). The sodium hydroxide solution more effective to decrease magnesium ion than sodium carbonate solution, and the sodium carbonate solution more effective to decrease calcium ion than sodium hydroxide solution.

  8. ''FIXBOX'' - a new technique for the reliable conditioning of plutonium waste solutions

    International Nuclear Information System (INIS)

    Bruchertseifer, H.; Sommer, E.; Steinemann, M.; Bart, G.

    1994-01-01

    ''FIXBOX'' - A new technique and facility for the conditioning of plutonium waste solutions has been developed and brought into operation in the Hot-laboratory at PSI, for the solidification of the waste from the research programmes. The facility is situated in glove-boxes for handling alpha activity and gamma-shielded for conditioning of fission product-containing waste. This report gives a brief description of the FIXBOX facility, the procedure and the first results of the cementation of plutonium waste solutions. As a result of this solidification, the actinide waste is homogeneous and strongly bound in the cement. The presence of gluconic acid and other complexing agents in the waste solution will not disturb this process. (author) figs., tabs., refs

  9. The dispersal and impact of salt from surface storage piles the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Reith, C.C.; Louderbough, E.T.

    1986-01-01

    A comprehensive program of ecological studies occurs at the Waste Isolation Pilot Plant (WIPP) in an effort to detect and quantify impacts of excavated salt which is stored on the surface in two piles: one having originated in 1980, the other in 1984. Both piles are surrounded by berms which channel runoff to holding ponds, so nearly all dispersal is due to the resuspension, transport, and deposition of salt particles by wind. Ecological parameters which have been monitored since 1984 include: visual evidence (via photography), soil properties, microbial activity, leaf-litter decomposition, seedling emergence, plant foliar cover, and plant species diversity. These are periodically assessed at experimental plots near the salt piles, and at control plots several kilometers away

  10. Areal thermal loading recommendations for nuclear waste repositories in salt

    International Nuclear Information System (INIS)

    Russell, J.E.

    1979-06-01

    This document gives a wider understanding of the history of the recommended thermal loadings in salt for both high-level waste (HLW) from fresh UO 2 -fueled, light-water reactors (LWR) with no recycle and spent unreprocessed fuel (SURF) from LWRs. Aspects of the current recommendations that need further study are identified. Finally, an interim set of design thermal-loading recommendations are given that have a common rationale of satisfying performance limits within our current state of knowledge. These recommendations are made on a generic rather than a site-specific basis. 11 figures, 5 tables

  11. Salt-assisted and salt-suppressed sol-gel transitions of methylcellulose in water.

    Science.gov (United States)

    Xu, Y; Wang, C; Tam, K C; Li, L

    2004-02-03

    The effects of various salts on the sol-gel transition of aqueous methylcellulose (MC) solutions have been studied systematically by means of a micro differential scanning calorimeter. It was found that the heating process was endothermic while the cooling process was exothermic for both MC solutions with and without salts. The addition of salts did not change the patterns of gelation and degelation of MC. However, the salts could shift the sol-gel transition and the gel-sol transition to lower or higher temperatures from a pure MC solution, depending on the salt type. These opposite effects were termed the salt-assisted and salt-suppressed sol-gel transitions. Either the salt-assisted transition or the salt-suppressed sol-gel transition was a function of salt concentration. In addition, each salt was found to have its own concentration limit for producing a stable aqueous solution of MC at a given concentration of MC, which was related to the anionic charge density of the salt. Cations were proved to have weaker effects than anions. The "salt-out strength", defined as the salt effect per mole of anion, was obtained for each anion studied. The thermodynamic mechanisms involved in the salt-assisted and salt-suppressed sol-gel transitions are discussed.

  12. Calculations on the development in space and time of the temperature field around a repository of medium and high active wastes in a salt formation

    International Nuclear Information System (INIS)

    Delisle, G.

    1980-01-01

    The concept of nuclear waste disposal of th of the Federal Republic of Germany calls for the burial of the wastes within a salt formation. A small portion of the wastes will generate heat after the disposal procedure. A temperature rise within the salt formation, in space and time limited, will be the consequence. The temperature change at any point in the near or far field of the disporal area can be calculated with the aid of numerical models. The thermal parameters representative for the bulk material of the Zechstein formation in NW-Germany, on which the calculations are based, will be discussed in detail. The interrelation between the concentration of heat producing wastes in the disposal field and the maximum average temperature in the salt formation will be treated. By defining numerical models, which are based on assumed shapes of a salt dome and a disposal area, the temperature development in the near and far field of a nuclear repository are shown. (orig.) [de

  13. Characterization of Laboratory Prepared Concrete Pastes Exposed to High Alkaline and High Sodium Salt Solutions

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-06-30

    The objective of this study was to identify potential chemical degradation mechanisms for the Saltstone Disposal Unit (SDU) concretes, which over the performance life of the structures may be exposed to highly alkaline sodium salt solutions containing sulfate, hydroxide, and other potentially corrosive chemicals in salt solution and saltstone flush water, drain water, leachate and / or pore solution. The samples analyzed in this study were cement pastes prepared in the SIMCO Technologies, Inc. concrete laboratory. They were based on the paste fractions of the concretes used to construct the Saltstone Disposal Units (SDUs). SDU 1 and 4 concrete pastes were represented by the PV1 test specimens. The paste in the SDU 2, 3, 5, and 6 concrete was represented by the PV2 test specimens. SIMCO Technologies, Inc. selected the chemicals and proportions in the aggressive solutions to approximate proportions in the saltstone pore solution [2, 3, 5, and 6]. These test specimens were cured for 56 days in curing chamber before being immersed in aggressive solutions. After exposure, the samples were frozen to prevent additional chemical transport and reaction. Selected archived (retrieved from the freezer) samples were sent to the Savannah River National Laboratory (SRNL) for additional characterization using x-ray diffraction (XRD), scanning electron microscopy (SEM), and energy dispersive x-ray (EDX) spectroscopy. Characterization results are summarized in this report. In addition, a correlation between the oxide composition of the pastes and their chemical durability in the alkaline salt solutions is provided.

  14. Quenching characteristics of bathocuproinedisulfonic acid, disodium salt in aqueous solution and copper sulfate plating solution

    Science.gov (United States)

    Koga, Toshiaki; Hirakawa, Chieko; Takeshita, Michinori; Terasaki, Nao

    2018-04-01

    Bathocuproinedisulfonic acid, disodium salt (BCS) is generally used to detect Cu(I) through a color reaction. We newly found BCS fluorescence in the visible blue region in an aqueous solution. However, the fluorescence mechanism of BCS is not well known, so we should investigate its fundamental information. We confirmed that the characteristics of fluorescence are highly dependent on the molecular concentration and solvent properties. In particular, owing to the presence of the copper compound, the fluorescence intensity extremely decreases. By fluorescence quenching, we observed that a copper compound concentration of 10-6 mol/L or less could easily be measured in an aqueous solution. We also observed BCS fluorescence in copper sulfate plating solution and the possibility of detecting monovalent copper by fluorescence reabsorption.

  15. Partitioning high-level waste from alkaline solution: A literature survey

    International Nuclear Information System (INIS)

    Marsh, S.F.

    1993-05-01

    Most chemical partitioning procedures are designed for acidic feed solutions. However, the high-level waste solutions in the underground storage tanks at US Department of Energy defense production sites are alkaline. Effective partitioning procedures for alkaline solutions could decrease the need to acidify these solutions and to dissolve the solids in acid, which would simplify subsequent processing and decrease the generation of secondary waste. The author compiles candidate technologies from his review of the chemical literature, experience, and personal contacts. Several of these are recommended for evaluation

  16. Cementation of liquid radioactive waste with high content of borate salts

    International Nuclear Information System (INIS)

    Gorbunova, O.

    2015-01-01

    The report reviews the ways of optimization of cementation of boron-containing liquid radioactive waste. The most common way to hardening the low-level liquid radioactive waste (LRW) is the cementation. However, boron-containing liquid radioactive waste with low pH values cannot be cemented without alkaline additives, to neutralize acid forms of borate compounds. Cement setting without additives happens only on 14-56 days, the compounds have low strength, and hence an insufficient reliability of radionuclides fixation in the cement matrix. The alkaline additives increase the volume of the final cement compound which enhances financial and operational costs. In order to control the speed of hardening of cement solution with a boron-containing liquid radioactive waste and to remove the components that prevent hardening of cement solution, it is proposed an electromagnetic treatment of LRW in the vortex layer of ferromagnetic particles. The results of infrared spectroscopy show, that electromagnetic treatment of liquid radioactive waste changes the ionic forms of the borates and raises the pH due to the dissociation of the oxygen and hydrogen bonds in the aqueous solutions of the boron compounds. The various types of ferromagnetic activators of the vortex layer have been investigated, including the highly dispersed nano-powders and the magnetic phases of the iron oxides. It has been determined the technological parameters of the electromagnetic treatment of liquid radioactive waste and the subsequent cementation of this type of LRW. By using the method of scanning electron microscopy it has been shown, that the nano-particles of magnetic phases of the ferric oxides are involved in phase formation of hydro-aluminum-calcium ferrites in the early stages of hardening and improving strength of the cement compounds with liquid radioactive waste. (authors)

  17. Recovery of fission products from waste solutions utilizing controlled cathodic potential electrolysis

    International Nuclear Information System (INIS)

    Carlin, W.W.; Darlington, W.B.

    1975-01-01

    Fission products, e.g., palladium, rhodium and technetium, are recovered from aqueous waste solutions thereof, e.g., aged Purex alkaline waste solutions. The metal values from the waste solutions are extracted by ion exchange techniques. The metals adsorbed by the ion exchange resin are eluted and selectively recovered by controlled cathodic potential electrolysis. The metal values deposited on the cathode are recovered and, if desired, further purified

  18. Length scale dependence of the dynamic properties of hyaluronic acid solutions in the presence of salt.

    Science.gov (United States)

    Horkay, Ferenc; Falus, Peter; Hecht, Anne-Marie; Geissler, Erik

    2010-12-02

    In solutions of the charged semirigid biopolymer hyaluronic acid in salt-free conditions, the diffusion coefficient D(NSE) measured at high transfer momentum q by neutron spin echo is more than an order of magnitude smaller than that determined by dynamic light scattering, D(DLS). This behavior contrasts with neutral polymer solutions. With increasing salt content, D(DLS) approaches D(NSE), which is independent of ionic strength. Contrary to theoretical expectation, the ion-polymer coupling, which dominates the low q dynamics of polyelectrolyte solutions, already breaks down at distance scales greater than the Debye-Hückel length.

  19. Vitrification in the presence of salts

    International Nuclear Information System (INIS)

    Marra, J.C.; Andrews, M.K.; Schumacher, R.F.

    1994-01-01

    Glass is an advantageous material for the immobilization of nuclear wastes because of the simplicity of processing and its unique ability to accept a wide variety of waste elements into its network structure. Unfortunately, some anionic species which are present in the nuclear waste streams have only limited solubility in oxide glasses. This can result in either vitrification concerns or it can affect the integrity, of the final vitrified waste form. The presence of immiscible salts can also corrode metals and refractories in the vitrification unit as well as degrade components in the off-gas system. The presence of a molten salt layer on the melt may alter the batch melting rate and increase operational safety concerns. These safety concerns relate to the interaction of the molten salt and the melter cooling fluids. Some preliminary data from ongoing experimental efforts examining the solubility of molten salts in glasses and the interaction of salts with melter component materials is included

  20. Use of zinc and copper (I) salts to reduce sulfur and nitrogen impurities during the pyrolysis of plastic and rubber waste to hydrocarbons

    Science.gov (United States)

    Wingfield, Jr., Robert C.; Braslaw, Jacob; Gealer, Roy L.

    1984-01-01

    An improvement in a process for the pyrolytic conversion of rubber and plastic waste to hydrocarbon products which results in reduced levels of nitrogen and sulfur impurities in these products. The improvement comprises pyrolyzing the waste in the presence of at least about 1 weight percent of salts, based on the weight of the waste, preferably chloride or carbonate salts, of zinc or copper (I). This invention was made under contract with or subcontract thereunder of the Department of Energy Contract #DE-AC02-78-ER10049.

  1. Corrosion of candidate iron-base waste package structural barrier materials in moist salt environments

    International Nuclear Information System (INIS)

    Westerman, R.E.; Pitman, S.G.

    1984-11-01

    Mild steels are considered to be strong candidates for waste package structural barrier (e.g., overpack) applications in salt repositories. Corrosion rates of these materials determined in autoclave tests utilizing a simulated intrusion brine based on Permian Basin core samples are low, generally <25 μm (1 mil) per year. When the steels are exposed to moist salts containing simulated inclusion brines, the corrosion rates are found to increase significantly. The magnesium in the inclusion brine component of the environment is believed to be responsible for the increased corrosion rates. 1 reference, 4 figures, 2 tables

  2. Safety related aspects of ultimate disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Goemmel, R.

    1992-01-01

    Solutions and questions related to nuclear waste management are presented. In particular, long-term safety of repositories in Germany and Sweden is considered, with special attention being paid to methods of detection, geotechnical barriers and post-operational phase of salt dome repositories, and conditioning of wastes to make them fit for ultimate disposal. (DG) [de

  3. Salt formations offer disposal alternative

    International Nuclear Information System (INIS)

    Funderburk, R.

    1990-01-01

    This paper discusses how three U.S. firms are spending millions to permit and build underground disposal sites in salt formations. These companies claim salt is the ideal geological medium for holding hazardous wastes. Two Texas locations and one in Michigan have been targeted as future sites for hazardous waste disposal. The Michigan site, outside Detroit, is a former salt mine 2,000 feet beneath the Ford Motor Co. (Detroit) assembly works in Dearborn. Both Texas sites are atop salt domes---one east and one west of Houston

  4. Method of decomposing radioactive organic solvent wastes

    International Nuclear Information System (INIS)

    Uki, Kazuo; Ichihashi, Toshio; Hasegawa, Akira; Sato, Tatsuaki

    1986-01-01

    Purpose: To decompose radioactive organic solvent wastes or radioactive hydrocarbon solvents separated therefrom into organic materials under moderate conditions, as well as greatly decrease the amount of secondary wastes generated. Method: Radioactive organic solvent wastes comprising an organic phosphoric acid ester ingredient and a hydrocarbon ingredient as a diluent therefor, or radioactive hydrocarbon solvents separated therefrom are oxidatively decomposed by hydrogen peroxide in an aqueous phosphoric acid solution of phosphoric acid metal salts finally into organic materials to perform decomposing treatment for the radioactive organic solvent wastes. The decomposing reaction is carried out under relatively moderate conditions and cause less burden to facilities or the likes. Further, since the decomposed liquid after the treatment can be reused for the decomposing reaction as a catalyst solution secondary wastes can significantly be decreased. (Yoshihara, H.)

  5. Hydraulic conductivity in response to exchangeable sodium percentage and solution salt concentration

    Directory of Open Access Journals (Sweden)

    Jefferson Luiz de Aguiar Paes

    2014-10-01

    Full Text Available Hydraulic conductivity is determined in laboratory assays to estimate the flow of water in saturated soils. However, the results of this analysis, when using distilled or deionized water, may not correspond to field conditions in soils with high concentrations of soluble salts. This study therefore set out to determine the hydraulic conductivity in laboratory conditions using solutions of different electrical conductivities in six soils representative of the State of Pernambuco, with the exchangeable sodium percentage adjusted in the range of 5-30%. The results showed an increase in hydraulic conductivity with both decreasing exchangeable sodium percentage and increasing electrical conductivity in the solution. The response to the treatments was more pronounced in soils with higher proportion of more active clays. Determination of hydraulic conductivity in laboratory is routinely performed with deionized or distilled water. However, in salt affected soils, these determinations should be carried out using solutions of electrical conductivity different from 0 dS m-1, with values close to those determined in the saturation extracts.

  6. OPERATIONS REVIEW OF THE SAVANNAH RIVER SITE INTEGRATED SALT DISPOSITION PROCESS - 11327

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T.; Poirier, M.; Fondeur, F.; Fink, S.; Brown, S.; Geeting, M.

    2011-02-07

    The Savannah River Site (SRS) is removing liquid radioactive waste from its Tank Farm. To treat waste streams that are low in Cs-137, Sr-90, and actinides, SRS developed the Actinide Removal Process and implemented the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The Actinide Removal Process contacts salt solution with monosodium titanate to sorb strontium and select actinides. After monosodium titanate contact, the resulting slurry is filtered to remove the monosodium titanate (and sorbed strontium and actinides) and entrained sludge. The filtrate is transferred to the MCU for further treatment to remove cesium. The solid particulates removed by the filter are concentrated to {approx} 5 wt %, washed to reduce the sodium concentration, and transferred to the Defense Waste Processing Facility for vitrification. The CSSX process extracts the cesium from the radioactive waste using a customized solvent to produce a Decontaminated Salt Solution (DSS), and strips and concentrates the cesium from the solvent with dilute nitric acid. The DSS is incorporated in grout while the strip acid solution is transferred to the Defense Waste Processing Facility for vitrification. The facilities began radiological processing in April 2008 and started processing of the third campaign ('MarcoBatch 3') of waste in June 2010. Campaigns to date have processed {approx}1.2 million gallons of dissolved saltcake. Savannah River National Laboratory (SRNL) personnel performed tests using actual radioactive samples for each waste batch prior to processing. Testing included monosodium titanate sorption of strontium and actinides followed by CSSX batch contact tests to verify expected cesium mass transfer. This paper describes the tests conducted and compares results from facility operations. The results include strontium, plutonium, and cesium removal, cesium concentration, and organic entrainment and recovery data. Additionally, the poster describes lessons learned during

  7. Field experiments in salt formations

    International Nuclear Information System (INIS)

    Kuehn, K.

    1986-01-01

    Field experiments in salt formations started as early as 1965 with Project Salt Vault in the Lyons Mine, Kansas, U.S.A., and with the purchase of the Asse salt mine by the German Federal Government. Underground tests concentrated on the heat dissipation around buried high-level radioactive wastes and the geomechanical consequences of their disposal. Near-field investigations cover the properties of water and gas release, radiolysis and corrosion. Further objectives of field experiments are the development and underground testing of a handling system for high-level wastes. The performance of an underground test disposal for such wastes is not only considered to be necessary for technical and scientific reasons but also for improving public acceptance of the concept of radioactive waste disposal. (author)

  8. Tank Waste Transport Stability: Summary of Slurry and Salt-Solution Studies for FY 2001

    Energy Technology Data Exchange (ETDEWEB)

    Welch, T.D.

    2002-06-07

    Despite over 50 years of experience in transporting radioactive tank wastes to and from equipment and tanks at the Department of Energy's Hanford, Savannah River, and Oak Ridge sites, waste slurry transfer pipelines and process piping become plugged on occasion. At Hanford, several tank farm pipelines are no longer in service because of plugs. At Savannah River, solid deposits in the outlet line of the 2H evaporator have resulted in an unplanned extended downtime. Although waste transfer criteria and guidelines intended to prevent pipeline plugging are in place, they are not always adequate. To avoid pipeline plugging in the future, other factors that are not currently embodied in the transfer criteria may need to be considered. The work summarized here is being conducted to develop a better understanding of the chemical and waste flow dynamics during waste transfer. The goal is to eliminate pipeline plugs by improving analysis and engineering tools in the field that incorporate this understanding.

  9. Congenital primary adrenal insufficiency and selective aldosterone defects presenting as salt-wasting in infancy: a single center 10-year experience.

    Science.gov (United States)

    Bizzarri, Carla; Olivini, Nicole; Pedicelli, Stefania; Marini, Romana; Giannone, Germana; Cambiaso, Paola; Cappa, Marco

    2016-08-02

    Salt-wasting represents a relatively common cause of emergency admission in infants and may result in life-threatening complications. Neonatal kidneys show low glomerular filtration rate and immaturity of the distal nephron leading to reduced ability to concentrate urine. A retrospective chart review was conducted for infants hospitalized in a single Institution from 1(st) January 2006 to 31(st) December 2015. The selection criterion was represented by the referral to the Endocrinology Unit for hyponatremia (serum sodium <130 mEq/L) of suspected endocrine origin at admission. Fifty-one infants were identified. In nine infants (17.6 %) hyponatremia was related to unrecognized chronic gastrointestinal or renal salt losses or reduced sodium intake. In 10 infants (19.6 %) hyponatremia was related to central nervous system diseases. In 19 patients (37.3 %) the final diagnosis was congenital adrenal hyperplasia (CAH). CAH was related to 21-hydroxylase deficiency in 18 patients, and to 3β-Hydroxysteroid dehydrogenase (3βHSD) deficiency in one patient. Thirteen patients (25.5 %) were affected by different non-CAH salt-wasting forms of adrenal origin. Four familial cases of X-linked adrenal hypoplasia congenita due to NROB1 gene mutation were identified. Two unrelated girls showed aldosterone synthase deficiency due to mutation of the CYP11B2 gene. Two unrelated infants were affected by familial glucocorticoid deficiency due to MC2R gene mutations. One girl showed pseudohypoaldosteronism related to mutations of the SCNN1G gene encoding for the epithelial sodium channel. Transient pseudohypoaldosteronism was identified in two patients with renal malformations. In two infants the genetic aetiology was not identified. Emergency management of infants presenting with salt wasting requires correction of water losses and treatment of electrolyte imbalances. Nevertheless, the differential diagnosis may be difficult in emergency settings, and sometimes hospitalized infants

  10. Effect of salt solutions on the radiosensitivity of mammalian cells as a function of the state of adhesion and the water structure

    Energy Technology Data Exchange (ETDEWEB)

    Moggach, P G; Lepock, J R; Kruuv, J [Waterloo Univ., Ontario (Canada). Dept. of Physics

    1979-11-01

    The radiation isodose survival curve of attached Chinese hamster (V79) cells, subjected to a wide concentration range of salt or sucrose solutions, was characterized by two maxima separated by a minimum. Cells were radioprotected at the maxima (high and low hypertonic salt concentrations) while they were radiosensitized at the minimum (intermediate hypertonic salt concentrations). Both cations and anions could alter the cellular radiosensitivity above and beyond the (osmotic) effect observed for cells treated with sucrose solutions. However, the basic curve shape, except in the case of sulphate salts, remained the same. When these experiments were repeated with single cells in suspension, the isodose survival curve was quite different in that high salt concentrations did not protect cells in suspension unlike the case with attached cells. The curve shape was also altered in that the second maximum was absent with many salt solutions. When multicellular spheroids were used for these experiments, the data resembled those for single cell suspensions rather than for attached cells. The radiation survival data for cells in suspension in salt solutions correlated with water proton spin lattice relaxation time (T/sub 1/) and, in hypo- and iso-tonic solutions, with cell volume.

  11. Geohydrology of the Keechi, Mount Sylvan, Oakwood, and Palestine salt domes in the northeast Texas salt-dome basin

    International Nuclear Information System (INIS)

    Carr, J.E.; Halasz, S.J.; Peters, H.B.

    1980-01-01

    The salt within these domes has penetrated as much as 20,000 feet of Mesozoic and Cenozoic strata, and presently extends to within 120 to 800 feet of the land surface. The salt penetrates or closely underlies major freshwater and salinewater aquifers within the basin. To provide a safe repository for radioactive wastes within one or more of these domes, a thorough understanding of the geohydrology needs to be obtained, and the hydrologic stability of the domes needs to be established for the expected life of the storage facility. Dissolution may exist at all four candidate salt domes, possibly through contact with Cretaceous or Tertiary aquifers, or through fault systems in the vicinity of the domes. Strata overlying and surrounding Palestine and Keechi Salt Domes have been arched into steeply-dipping folds that are complexly faulted. Similar conditions exist at Oakwood and Mount Sylvan Domes, except that the Tertiary strata have been only moderately disturbed. Additional problems concerning the hydrologic stability of Oakwood and Palestine Salt Domes have resulted from the disposal of oil-field salinewater in the cap rock at the Oakwood Dome and previous solution mining of salt at the Palestine Dome

  12. Results Of The Extraction-Scrub-Strip Testing Using An Improved Solvent Formulation And Salt Waste Processing Facility Simulated Waste

    International Nuclear Information System (INIS)

    Peters, T.; Washington, A.; Fink, S.

    2012-01-01

    The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D Cs in an ESS test, using the baseline solvent formulation and the typical waste feed, is ∼15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under

  13. Sodium concentration in home made salt – sugar – solution (sss ...

    African Journals Online (AJOL)

    In a cohort of 210 young mothers, selected through cluster sampling technique from Ogida health district of Egor Local Government Area of Edo State, the electrolyte concentration of prepared salt-sugar-solutions (SSS) were evaluated. This was predicated on the need to determine the effects of introduction of various ...

  14. China's Scientific Investigation for Liquid Waste Treatment Solutions

    International Nuclear Information System (INIS)

    Liangjin, B.; Meiqiong, L.; Kelley, D.

    2006-01-01

    Post World War II created the nuclear age with several countries developing nuclear technology for power, defense, space and medical applications. China began its nuclear research and development programs in 1950 with the establishment of the China Institute of Atomic Energy (CIAE) located near Beijing. CIAE has been China's leader in nuclear science and technical development with its efforts to create advanced reactor technology and upgrade reprocessing technology. In addition, with China's new emphasis on environmental safety, CIAE is focusing on waste treatment options and new technologies that may provide solutions to legacy waste and newly generated waste from the full nuclear cycle. Radioactive liquid waste can pose significant challenges for clean up with various treatment options including encapsulation (cement), vitrification, solidification and incineration. Most, if not all, nuclear nations have found the treatment of liquids to be difficult, due in large part to the high economic costs associated with treatment and disposal and the failure of some methods to safely contain or eliminate the liquid. With new environmental regulations in place, Chinese nuclear institutes and waste generators are beginning to seek new technologies that can be used to treat the more complex liquid waste streams in a form that is safe for transport and for long-term storage or final disposal. [1] In 2004, CIAE and Pacific Nuclear Solutions, a division of Pacific World Trade, USA, began discussions about absorbent technology and applications for its use. Preliminary tests were conducted at CIAE's Department of Radiochemistry using generic solutions, such as lubricating oil, with absorbent polymers for solidification. Based on further discussions between both parties, it was decided to proceed with a more formal test program in April, 2005, and additional tests in October, 2005. The overall objective of the test program was to apply absorbent polymers to various waste streams

  15. The best solution to our Nation's waste management problem: Education

    International Nuclear Information System (INIS)

    Mikel, C.J.

    1992-01-01

    In addition to the Waste Isolation Pilot Plant (WIPP) being the best solution today to the Nation's problem of permanent storage of transuranic radioactive waste produced by the defense industry, WIPP is also involved in finding the solution for another national problem: the education of our youth. The youth of America have grown up thinking that science and math are too hard, or not interesting. We, the parents of our Nation's leaders of tomorrow, must find a solution to this dilemma. It is the mission of the Waste Isolation Division Educational Programs to create programs to promote quality education in the classroom and to enhance each student's interest in mathematics and the sciences

  16. Radioactive wastes. The groundwork of current solutions

    International Nuclear Information System (INIS)

    Grevoz, A.; Boullis, B.; Devezeaux de Lavergne, J.G.; Butez, M.; Bordier, G.; Vitart, X.; Hablot, I.; Chastagnet, F.

    2005-01-01

    Today the groundwork laid down by research has made processes available for the durable treatment and conditioning of all types of radioactive waste. This document illustrates the today situations in five presentations. Now standing as a national reference, the french inventory of radioactive waste, drawn up by ANDRA, has not only expanded to cover recoverable material but also features predictions of waste arisings for 2010 and 2020, including waste from the decommissioning of current installations. The current process used for spent fuel reprocessing allows extraction for recycling purpose, of uranium and plutonium, with very high recovery and purification rates. Advances in characterization and decontamination allow improvements in sorting and retrieval and conditioning to be considered for older wastes. The french National radioactive waste management agency (ANDRA) is already providing optimum industrial solutions for all short-lived, low and very low level waste on its Soulaines and Morvillers sites. For several decades, Areva has been reprocessing spent fuel and conditioning ultimate waste in its La Hague plants. (A.L.B.)

  17. Development of integrated radioactive waste packaging and conditioning solutions in the UK

    Energy Technology Data Exchange (ETDEWEB)

    Sibley, Peter; Butter, Kevin; Zimmerman, Ian [EnergySolutions EU Ltd., Swindon, Wiltshire (United Kingdom); Viermann, Joerg [GNS Gesellschaft fur Nuklear-Service mbH, Essen (Germany); Messer, Matthias [GNS Gesellschaft fur Nuklear-Service mbH, Bristol (United Kingdom)

    2013-07-01

    In order to offer a more cost effective, safer and efficient Intermediate Level Waste (ILW) management service, EnergySolutions EU Ltd. and Gesellschaft fur Nuklear-Service mbH (GNS) have been engaged in the development of integrated radioactive waste retrieval, packaging and conditioning solutions in the UK. Recognising the challenges surrounding regulatory endorsement and on-site implementation in particular, this has resulted in an alternative approach to meeting customer, safety regulator and disposability requirements. By working closely with waste producers and the organisation(s) responsible for endorsing radioactive waste management operations in the UK, our proposed solutions are now being implemented. By combining GNS' off-the-shelf, proven Ductile Cast Iron Containers (DCICs) and water removal technologies, with EnergySolutions EU Ltd.'s experience and expertise in waste retrieval, safety case development and disposability submissions, a fully integrated service offering has been developed. This has involved significant effort to overcome technical challenges such as onsite equipment deployment, active commissioning, conditioning success criteria and disposability acceptance. Our experience in developing such integrated solutions has highlighted the importance of working in collaboration with all parties to achieve a successful and viable outcome. Ultimately, the goal is to ensure reliable, safe and effective delivery of waste management solutions. (authors)

  18. Effects of heating on salt-occluded zeolite

    International Nuclear Information System (INIS)

    Lewis, M.A.; Hash, M.C.; Pereira, C.; Ackerman, J.P.

    1996-01-01

    The electrometallurgical treatment of spent nuclear fuel generates a waste stream of fission products in the electrolyte, LiCl-KCl eutectic salt. Argonne National Laboratory is developing a mineral waste form for this waste stream. The waste form consists of a composite formed by hot pressing salt-occluded zeolite and a glass binder. Pressing conditions must be judiciously chosen. For a given pressure, increasing temperatures and hold times give denser products but the zeolite is frequently converted to sodalite. Reducing the temperature or hold time leads to a porous zeolite composite. Therefore, conditions that affect the thermal stability of salt-occluded zeolite both with and without glass are being investigated in an ongoing study. The parameters varied in this stage of the work were heating time, temperature, salt loading, and glass content. The heat-treated samples were examined primarily by X-ray diffraction. Large variations were found in the rate at which salt-occluded zeolite converted to other phases such as nepheline, salt, and sodalite. The products depended on the initial salt loading. Heating times required for these transitions depended on the procedure and temperature used to prepare the salt-occluded zeolite. Mixtures of glass and zeolite reacted much faster than the pure salt-occluded zeolite and were almost always converted to sodalite

  19. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    Science.gov (United States)

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

  20. Final status of the salt repository project waste package program experimental database

    International Nuclear Information System (INIS)

    Thornton, B.M.; Reimus, P.W.

    1988-03-01

    This report describes the final status of the Salt Repository Project Waste Package Program Experimental Database. The data base serves as a clearinghouse for all data collected within the Waste Package Program (WPP) and its predecessor programs at Pacific Northwest Laboratory (PNL). The database was maintained using RS/1 database management software. Documented assurance that the entries in the database were consistent with experimental records was provided by having each experimentalist inspect the entries and signify that they were in agreement with the records. The inspection and signoff were done per PNL technical procedures. Data for which it was impossible to obtain the experimentalist's inspection and signature were segregated from the rest of the database, although they could still be accessed by WPP staff. The WPPED contains two groups of subdirectories. One group contains data taken prior to the installation of quality assurance procedures at PNL. The other group of subdirectories contains data taken under the NQA-1 procedures since their installation in April 1985. As part of closeout activities in the Salt Repository Project, the WPP database has been archived onto magnetic media. The data in the database are available by request on magnetic media or in hardcopy form. 2 refs

  1. Fluid inclusions in salt: an annotated bibliography

    International Nuclear Information System (INIS)

    Isherwood, D.J.

    1979-01-01

    An annotated bibliography is presented which was compiled while searching the literature for information on fluid inclusions in salt for the Nuclear Regulatory Commission's study on the deep-geologic disposal of nuclear waste. The migration of fluid inclusions in a thermal gradient is a potential hazard to the safe disposal of nuclear waste in a salt repository. At the present time, a prediction as to whether this hazard precludes the use of salt for waste disposal can not be made. Limited data from the Salt-Vault in situ heater experiments in the early 1960's (Bradshaw and McClain, 1971) leave little doubt that fluid inclusions can migrate towards a heat source. In addition to the bibliography, there is a brief summary of the physical and chemical characteristics that together with the temperature of the waste will determine the chemical composition of the brine in contact with the waste canister, the rate of fluid migration, and the brine-canister-waste interactions

  2. Innovative tank emptying system for the retrieval of salt, sludge and IX resins from storage tanks of NPPs

    International Nuclear Information System (INIS)

    Karl Froschauer; Holger Witing; Bernhard Christ

    2006-01-01

    RWE NUKEM recently developed a new Tank Emptying System (TESY) for the extraction of stored radioactive boric acid/borate salt blocks, sludge and IX resin from NPP stainless steel tanks of several hundred cubic meters content in Russia. RWE NUKEM has chosen the emptying concept consisting of a tracked submersible vehicle ('Crawler'), with jet nozzles for solution, agitation and fluidization, and a suction head to pick up the generated solution or suspension respectively. With the employment of RWE NUKEM's TESY system, spent radioactive salt deposits, ion-exchange resins and sludge, can be emptied and transferred out of the tank. The sediment, crystallized and settled during storage, will be agitated with increased temperature and suitable pH value and then picked up in form of a suspension or solution directly at the point of mobilization. This new Tank Emptying System concept enables efficiently to retrieve stored salt and other sediment waste, reduces operating time, safes cost for spare parts, increases the safety of operation and minimizes radiation exposure to personnel. All emptying tasks are performed remotely from a panel board and TV monitor located in a central control room. The TESY system consists of the following main components: glove box, crawler, submersible pump, heater, TV camera and spot light, control panel and monitor, water separation and feed unit, sodium hydroxide dosing unit. The system is specially requested for the removal of more than 2,500 cubic meter salt solution generated from the dissolution of some 300 cubic meter crystallized salt deposit per tank and per year. The TESY system is able to dissolve efficiently the salts and retrieve solutions and other liquefied suspensions. TESY is adaptable to all liquid waste storage facilities and especially deployable for tanks with limited access openings (<550 mm)

  3. THE IMPACT OF DISSOLVED SALTS ON PASTES CONTAINING FLY ASH, CEMENT AND SLAG

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, J.; Edwards, T.; Williams, V.

    2009-09-21

    The degree of hydration of a mixture of cementitious materials (Class F fly ash, blast furnace slag and portland cement) in highly concentrated alkaline salt solutions is enhanced by the addition of aluminate to the salt solution. This increase in the degree of hydration, as monitored with isothermal calorimetry, leads to higher values of dynamic Young's modulus and compressive strength and lower values of total porosity. This enhancement in performance properties of these cementitious waste forms by increased hydration is beneficial to the retention of the radionuclides that are also present in the salt solution. The aluminate ions in the solution act first to retard the set time of the mix but then enhance the hydration reactions following the induction period. In fact, the aluminate ions increase the degree of hydration by {approx}35% over the degree of hydration for the same mix with a lower aluminate concentration. An increase in the blast furnace slag concentration and a decrease in the water to cementitious materials ratio produced mixes with higher values of Young's modulus and lower values of total porosity. Therefore, these operational factors can be fine tuned to enhance performance properties of cementitious waste form. Empirical models for Young modulus, heat of hydration and total porosity were developed to predict the values of these properties. These linear models used only statistically significant compositional and operational factors and provided insight into those factors that control these properties.

  4. Air Emissions Sampling from Vacuum Thermal Desorption for Mixed Wastes Designated with a Combustion Treatment Code for the Energy Solutions LLC Mixed Waste Facility

    International Nuclear Information System (INIS)

    Christensen, M.E.; Willoughby, O.H.

    2009-01-01

    EnergySolutions LLC is permitted by the State of Utah to treat organically-contaminated Mixed Waste by a vacuum thermal desorption (VTD) treatment process at its Clive, Utah treatment, storage, and disposal facility. The VTD process separates organics from organically-contaminated waste by heating the material in an inert atmosphere, and captures them as concentrated liquid by condensation. The majority of the radioactive materials present in the feed to the VTD are retained with the treated solids; the recovered aqueous and organic condensates are not radioactive. This is generally true when the radioactivity is present in solid form such as inorganic salts, metals or metallic oxides. The exception is when volatile radioactive materials are present such as radon gas, tritium, or carbon-14 organic chemicals. Volatile radioactive materials are a small fraction of the feed material. On August 28, 2006, EnergySolutions submitted a request to the USEPA for a variance to the Land Disposal Restrictions (LDR) standards for wastes designated with the combustion treatment code (CMBST). The final rule granting a site specific treatment variance was effective June 13, 2008. This variance is an alternative treatment standard to treatment by CMBST required for these wastes under USEPA's rules. The State of Utah provides oversight of the VTD processing operations. A demonstration test for treating CMBST-coded wastes was performed on April 29, 2008 through May 1, 2008. Three separate process cycles were conducted during this test. Both solid/liquid samples and emission samples were collected each day during the demonstration test. To adequately challenge the unit, feed material was spiked with trichloroethylene, o-cresol, dibenzofuran, and coal tar. Emission testing was conducted by EnergySolutions' emissions test contractor and sampling for radioactivity within the off-gas was completed by EnergySolutions' Health Physics department. This report discusses the emission testing

  5. Analytical Chemistry and Materials Characterization Results for Debris Recovered from Nitrate Salt Waste Drum S855793

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, Patrick Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chamberlin, Rebecca M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schwartz, Daniel S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worley, Christopher Gordon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Garduno, Katherine [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lujan, Elmer J. W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Borrego, Andres Patricio [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Castro, Alonso [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Colletti, Lisa Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fulwyler, James Brent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Holland, Charlotte S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Keller, Russell C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Klundt, Dylan James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Alexander [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martin, Frances Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montoya, Dennis Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Porterfield, Donivan R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schake, Ann Rene [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schappert, Michael Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Soderberg, Constance B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Spencer, Khalil J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanley, Floyd E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Thomas, Mariam R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Townsend, Lisa Ellen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Xu, Ning [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-16

    Solid debris was recovered from the previously-emptied nitrate salt waste drum S855793. The bulk sample was nondestructively assayed for radionuclides in its as-received condition. Three monoliths were selected for further characterization. Two of the monoliths, designated Specimen 1 and 3, consisted primarily of sodium nitrate and lead nitrate, with smaller amounts of lead nitrate oxalate and lead oxide by powder x-ray diffraction. The third monolith, Specimen 2, had a complex composition; lead carbonate was identified as the predominant component, and smaller amounts of nitrate, nitrite and carbonate salts of lead, magnesium and sodium were also identified. Microfocused x-ray fluorescence (MXRF) mapping showed that lead was ubiquitous throughout the cross-sections of Specimens 1 and 2, while heteroelements such as potassium, calcium, chromium, iron, and nickel were found in localized deposits. MXRF examination and destructive analysis of fragments of Specimen 3 showed elevated concentrations of iron, which were broadly distributed through the sample. With the exception of its high iron content and low carbon content, the chemical composition of Specimen 3 was within the ranges of values previously observed in four other nitrate salt samples recovered from emptied waste drums.

  6. Contribution to the study of uranyl salts in butyl phosphate solutions

    International Nuclear Information System (INIS)

    Coulon, A.

    1965-06-01

    A spectroscopic study in the normal infrared region and involving the following associations: tri-alkyl phosphates (tri-butyl, tri-ethyl, tri-methyl), uranyl salts (nitrate, chloride, acetate) has confirmed the existence of an interaction between the phosphoryl group and the uranium atom, as shown by a movement of absorption band for the valency P = 0 from ∼ 1270 cm -1 to ∼ 1180 cm -1 . A study of the preparation, analysis and spectroscopy of the solids obtained by the precipitation of uranyl salts by acid butyl phosphates has been carried out. By infrared spectrophotometry it has been shown that the tri-butyl and di-butyl phosphates are associated in non-polar diluents even before the uranium is introduced. The extraction of uranyl salts from acid aqueous solutions by a diluted mixture of tri-butyl and di-butyl phosphates proceeds by different mechanisms according to the nature of the ion (nitrate or chloride). (author) [fr

  7. Dynamic flow method to study the CO2 loading capacity of amino acid salt solutions

    DEFF Research Database (Denmark)

    Lerche, Benedicte Mai; Stenby, Erling Halfdan; Thomsen, Kaj

    Due to a number of advantages amino acid salt solutions have emerged as alternatives to the alkanolamine solvents for the chemical absorption of CO2 from flue gas. The use of amino acids in CO2 capture is a bio-mimetic process, as it is similar to CO2 binding by proteins in the blood......, such as hemoglobin. Amino acid salt solutions have the same amine functionality as alkanolamines, and are thus expected to behave similar towards CO2 in flue gas. Despite rising interest, few studies have been performed so far on amino acids as CO2 absorbents....

  8. Buckling design criteria for waste package disposal containers in mined salt repositories: Technical report

    International Nuclear Information System (INIS)

    Mallett, R.H.

    1986-12-01

    This report documents analytical and experimental results from a survey of the technical literature on buckling of thick-walled cylinders under external pressure. Based upon these results, a load factor is suggested for the design of waste package containers for disposal of high-level radioactive waste in repositories mined in salt formations. The load factor is defined as a ratio of buckling pressure to allowable pressure. Specifically, a load factor which ranges from 1.5 for plastic buckling to 3.0 for elastic buckling is included in a set of proposed buckling design criteria for waste disposal containers. Formulas are given for buckling design under axisymmetric conditions. Guidelines are given for detailed inelastic buckling analyses which are generally required for design of disposal containers

  9. Whispering Gallery Mode Based Optical Fiber Sensor for Measuring Concentration of Salt Solution

    Directory of Open Access Journals (Sweden)

    Chia-Chin Chiang

    2013-01-01

    Full Text Available An optical fiber solution-concentration sensor based on whispering gallery mode (WGM is proposed in this paper. The WGM solution-concentration sensors were used to measure salt solutions, in which the concentrations ranged from 1% to 25% and the wavelength drifted from the left to the right. The experimental results showed an average sensitivity of approximately 0.372 nm/% and an R2 linearity of 0.8835. The proposed WGM sensors are of low cost, feasible for mass production, and durable for solution-concentration sensing.

  10. Problems and risks involved in the projected storage of radioactive waste in a salt dome in the northwest of the FRG

    International Nuclear Information System (INIS)

    Mauthe, F.

    1979-01-01

    Current planning envisages long-term intermediate storage of radioactive waste and the exploration of the Gorleben salt dome by deep drilling in order to start appropriate mining work in case of favourable drilling results. The statements presented here on the problem of the 'Feasibility of ultimate storage of radioactive waste in salt deposits' (subject selected by the Government of the land Lower-Saxony) are aimed at informing the general public about the difficulties and problems involved in this waste disposal project and critically assess the arguments put forward by industry and licensing authorities in order to gain acceptance for this politically delicate project; the argumentation discussed here mainly refers to the field of geological science. (orig.) [de

  11. Geochemical and geomechanical solid-solutions interactions in unsaturated media. Prospects for the storage of nuclear waste

    International Nuclear Information System (INIS)

    Bouzid, M.

    2010-01-01

    Porous materials, especially the unsaturated ones, are complex systems in which several physicochemical parameters interact (eg relative humidity, T C, pore solution composition, geometry of the pore network). The precipitation of secondary phases inside and the associated changes (e.g. topology of the porous spaces) are important to understand for several applied topics: civil engineering, soil science or geology of deep wastes disposal. This experimental work was undertaken to better understand the mechanisms linking geochemical phase transitions and physicochemical properties of multiphasic porous media. The precipitation of salts in porous synthetic materials allowed us to identify two types of geochemistry-geomechanics coupling: the crystallization pressure (compression phenomenon, already known in the literature), and the capillary traction. These secondary precipitates are also responsible for a porous networks heterogenization which modifies the transfer functions. But we also show that the portions of liquid may be isolated by salts 'corks' and thus develop new thermochemical properties. In particular, we have observed cavitation events in some of these occluded solutions which indicate that they underwent a metastable superheated state. Finally, differential extraction experiments showed that the solubility changes with the pore size, and an interpretation based on pore geometry (solid curvature) has been proposed. Some evidence that these phenomena may actually be active in natural processes were collected, and this extension to the natural environment must now be treated extensively. (authors)

  12. Water purification using organic salts

    Science.gov (United States)

    Currier, Robert P.

    2004-11-23

    Water purification using organic salts. Feed water is mixed with at least one organic salt at a temperature sufficiently low to form organic salt hydrate crystals and brine. The crystals are separated from the brine, rinsed, and melted to form an aqueous solution of organic salt. Some of the water is removed from the aqueous organic salt solution. The purified water is collected, and the remaining more concentrated aqueous organic salt solution is reused.

  13. Composite Properties of Polyimide Resins Made From "Salt-Like" Solution Precursors

    Science.gov (United States)

    Cano, Roberto J.; Weiser, Erik S.; SaintClair, Terry L.; Echigo, Yoshiaki; Kaneshiro, Hisayasu

    1997-01-01

    Recent work in high temperature materials at NASA Langley Research Center (LaRC (trademark)) have led to the development of new polyimide resin systems with very attractive properties. The majority of the work done with these resin systems has concentrated on determining engineering mechanical properties of composites prepared from a poly(amide acid) precursor. Three NASA Langley-developed polyimide matrix resins, LaRC (trademark) -IA, LaRC (trademark) -IAX, and LaRC (trademark) -8515, were produced via a salt-like process developed by Unitika Ltd. The 'salt-like' solutions (sixty-five percent solids in NMP) were prepregged onto Hexcel IM7 carbon fiber using the NASA LaRC Multipurpose Tape Machine. Process parameters were determined and composite panels fabricated. Mechanical properties are presented for these three intermediate modulus carbon fiber/polyimide matrix composites and compared to existing data on the same polyimide resin systems and IM7 carbon fiber manufactured via poly(amide acid) solutions (thirty-five percent solids in NMP). This work studies the effects of varying the synthetic route on the processing and mechanical properties of polyimide composites.

  14. Assessment of drug salt release from solutions, suspensions and in situ suspensions using a rotating dialysis cell

    DEFF Research Database (Denmark)

    Parshad, Henrik; Frydenvang, Karla; Liljefors, Tommy

    2003-01-01

    buffer is used as release media. Generally, the initial release of the drug salt from in situ suspensions occurred faster as compared to conventional suspensions, probably due to incomplete precipitation of the drug salt, and hence formation of supersaturated solutions where the rate of release......A rotating dialysis cell consisting of a small (10 ml) and a large compartment (1000 ml) was used to study the release of drug salt (bupivacaine 9-anthracene carboxylate) from (i). solutions, (ii). suspensions and (iii). in situ formed suspensions. Initial release experiments from suspensions...... indicated that the release of drug salt in deionized water was predominantly limited by the diffusion across the membrane whereas it is essentially dissolution rate controlled in 0.05 M phosphate buffer (pH 7.40). Thus, the in vitro model appears to have a potential in formulation screening when phosphate...

  15. Precipitation-filtering technology for uranium waste solution generated on washing-electrokinetic decontamination

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gye-Nam, E-mail: kimsum@kaeri.re.kr; Park, Uk-Ryang; Kim, Seung-Soo; Moon, Jei-Kwon

    2015-05-15

    Graphical abstract: A recycling process diagram for the volume reduction of waste solution generated from washing-electrokinetic decontamination. - Highlights: • A process for recycling a waste solution generated was developed. • The total metal precipitation rate by NaOH in a supernatant after precipitation was the highest at pH 9. • The uranium radioactivity in the treated solution upon injection of 0.2 g of alum was lower. • After drying, the volume of sludge was reduced to 35% of the initial sludge volume. - Abstract: Large volumes of uranium waste solution are generated during the operation of washing-electrokinetic decontamination equipment used to remove uranium from radioactive soil. A treatment technology for uranium waste solution generated upon washing-electrokinetic decontamination for soil contaminated with uranium has been developed. The results of laboratory-size precipitation experiments were as follows. The total amount of metal precipitation by NaOH for waste solution was highest at pH 11. Ca(II), K(I), and Al(III) ions in the supernatant partially remained after precipitation, whereas the concentration of uranium in the supernatant was below 0.2 ppm. Also, when NaOH was used as a precipitant, the majority of the K(I) ions in the treated solution remained. The problem of CaO is to need a long dissolution time in the precipitation tank, while Ca(OH){sub 2} can save a dissolution time. However, the volume of the waste solution generated when using Ca(OH){sub 2} increased by 8 mL/100 mL (waste solution) compared to that generated when using CaO. NaOH precipitant required lower an injection volume lower than that required for Ca(OH){sub 2} or CaO. When CaO was used as a precipitant, the uranium radioactivity in the treated solution at pH 11 reached its lowest value, compared to values of uranium radioactivity at pH 9 and pH 5. Also, the uranium radioactivity in the treated solution upon injection of 0.2 g of alum with CaO or Ca(OH){sub 2} was

  16. Inefficacy of osmotic backwash induced by sodium chloride salt solution in controlling SWRO membrane fouling

    Science.gov (United States)

    Farooque, A. Mohammed; Al-Jeshi, Subhi; Saeed, Mohamed O.; Alreweli, Ali

    2014-12-01

    A study was conducted to evaluate the efficacy of osmotic backwash induced by high salt (NaCl) concentration solution on feed side of seawater reverse osmosis (SWRO) membranes, online and offline, in controlling membrane fouling and therefore minimizing/eliminating the need for chemical cleaning. SWRO membranes were deliberately fouled by feeding seawater from an open intake located on the Arabian Gulf Coast without dosing chemicals. The fouled membranes were subjected to offline cleaning with the salt solution of up to 25 % concentration. Despite the partial removal of foulants from the membrane surface, SWRO membrane performance could not be restored, indicating the ineffectiveness of osmotic backwash in aiding offline salt cleaning. Similarly, online osmotic backwash was found to be not only ineffective in removing foulants from membrane surfaces but actually increased the fouling rate, as indicated by faster fouling rates compared to other cases. Although the driving force required for the osmotic backwash existed, the generated back flow proved to be insufficient to detach foulants from membrane surfaces. During the study period, the average SWRO membrane flux was maintained between 19 and 23 LMH, whereas the average generated back flow flux by high salt concentration solution was only 11 LMH, which was not adequate to remove foulants from membrane surfaces. Moreover, it seems that the membrane configuration as well as inherent microstructure of SWRO membrane places certain constraints on the osmotic backwash process and renders osmotic backwash ineffective in tackling SWRO membrane fouling. Hence, chemical cleaning is essential to restore SWRO membrane performance whenever fouling occurs, and the use of highly concentrated salt solution does not have any significant benefit. Membrane autopsy revealed only an insignificant accumulation of biofouling layer despite the absence of disinfection. However, it was shown that culturable biofilm bacteria species

  17. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    International Nuclear Information System (INIS)

    Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

    1985-10-01

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document

  18. Process innovations to minimize waste volumes at Savannah River

    International Nuclear Information System (INIS)

    Doherty, J.P.

    1986-01-01

    In 1983 approximately 1.6 x 10 3 m 3 (427,000 gallons) of radioactive salt solution were decontaminated in a full-scale demonstration. The cesium decontamination factor (DF) was in excess of 4 x 10 4 vs. a goal of 1 x 10 4 . Data from this test were combined with pilot data and used to design the permanent facilities currently under construction. Startup of the Salt Decontamination Process is scheduled for 1987 and will decontaminate 2 x 10 4 m 3 (5.2 million gallons) of radioactive salt solution and generate 2 x 10 3 m 3 (520,000 gallons) of concentrated and washed precipitate per year. The Defense Waste Processing Facility (DWPF) will begin processing this concentrate in the Precipitate Hydrolysis Process starting in 1989. Laboratory data using simulated salt solution and nonradioactive cesium are being used to design this process. A 1/5-scale pilot plant is under construction and will be used to gain large-scale operating experience using nonradioactive simulants. This pilot plant is scheduled to startup in early 1987. The incentives to reduce the volume of waste that must be treated are self-evident. At Savannah River process development innovations to minimize the DWPF feed volumes have directly improved the economics of the process. The integrity of the final borosilicate glass water form has not been compromised by these developments. Many of the unit operations are familiar to chemical engineers and were put to use in a unique environment. As a result, tax dollars have been saved, and the objective of safely disposing of the nation's high-level defense waste has moved forward

  19. Direct examination of cadmium bonding in rat tissues dosed with mine wastes and cadmium-containing solutions

    International Nuclear Information System (INIS)

    Diacomanolis, V.; Ng, J. C.; Sadler, R.; Harris, H. H.; Nomura, M.; Noller, B. N.

    2010-01-01

    Direct examination by XANES and EXAFS of metal bonding in tissue can be demonstrated by examining cadmium uptake and bonding in animal tissue maintained at cryogenic temperatures. XANES at the K-edge of cadmium were collected at the Photon Factory Advanced Ring (PF-AR), NW10A beam line at KEK-Tsukuba-Japan. Rats fed with 1g mine waste containing 8-400 mg/kg cadmium per 200g body weight (b.w.) or dosed by oral gavage with either cadmium chloride solution alone (at 6 mg/kg b.w.) or in combination with other salts (As, Cu or Zn), 5 days/week for 6 weeks, had 0.1-7.5 and 8-86 mg/kg cadmium in the liver or kidney, respectively. Rats given intraperitoneally (ip) or intravenously (iv) 1-4 times with 1 mg/kg b.w. cadmium solution had 30-120 mg/kg cadmium in the liver or kidney. Tissues from rats were kept and transferred at cryogenic temperature and XANES were recorded at 20 K. The spectra for rat liver samples suggested conjugation of cadmium with glutathione or association with the sulfide bond (Cd-S) of proteins and peptides. EXAFS of rat liver fed by Cd and Zn solutions showed that Cd was clearly bound to S ligands with an inter-atomic distance of 2.54 A ring for Cd-S that was similar to cadmium sulfide with an inter-atomic distance of 2.52 A ring for Cd-S. Liver or kidney of rats fed with mine wastes did not give an edge in the XANES spectra indicating little uptake of cadmium by the animals. Longer and higher dosing regimen may be required in order to observe the same Cd-S bond in the rat tissue from mine wastes, including confirmation by EXAFS.

  20. Technical bases for establishing a salt test facility

    International Nuclear Information System (INIS)

    1985-05-01

    The need for a testing facility in which radioactive materials may be used in an underground salt environment is explored. No such facility is currently available in salt deposits in the United States. A salt test facility (STF) would demonstrate the feasibility of safely storing radioactive waste in salt and would provide data needed to support the design, construction, licensing, and operation of a radioactive waste repository in salt. Nineteen issues that could affect long-term isolation of waste materials in a salt repository are identified from the most pertinent recent literature. The issues are assigned an overall priority and a priority relative to the activities of the STF. Individual tests recommended for performance in the STF to resolve the 19 issues are described and organized under three groups: waste package performance, repository design and operation, and site characterization and evaluation. The requirements for a salt test facility are given in the form of functional criteria, and the approach that will be used in the design, execution, interpretation, and reporting of tests is discussed

  1. Stabilization/Solidification of radioactive molten salt waste by using xSiO2-yAl2O3-zP2O5 material

    International Nuclear Information System (INIS)

    Hwan-Seo Park; In-Tae Kim; Yong-Zun Cho; Seong-Won Park; Eung-Ho Kim

    2008-01-01

    Molten salt waste generated from the electro metallurgical process to recover uranium and transuranic elements is considered as one of problematic wastes to be difficult to immobilize into a durable for final disposal. As an alternative, this study suggested a new method performed at molten state, where dechlorination was achieved with a new inorganic material containing SiO 2 , Al 2 O 3 and P 2 O 5 (SAP). The SAP as a reactive material to molten salt was prepared by a conventional sol-gel process. The prepared SAPs were reacted with each metal chloride, LiCl, CsCl, SrCl 2 and CeCl 3 at 650 deg. C for 6 hours and also were reacted with simulated salt waste consisting of 90 wt% LiCl, 6.8 wt% CsCl and 3.2 wt% SrCl 2 at different waste loading. All the reactions were carried out in oxidative atmosphere and metal chlorides were effectively converted into stable products under a reasonable reaction ratio

  2. LITERATURE REVIEWS TO SUPPORT ION EXCHANGE TECHNOLOGY SELECTION FOR MODULAR SALT PROCESSING

    Energy Technology Data Exchange (ETDEWEB)

    King, W

    2007-11-30

    This report summarizes the results of literature reviews conducted to support the selection of a cesium removal technology for application in a small column ion exchange (SCIX) unit supported within a high level waste tank. SCIX is being considered as a technology for the treatment of radioactive salt solutions in order to accelerate closure of waste tanks at the Savannah River Site (SRS) as part of the Modular Salt Processing (MSP) technology development program. Two ion exchange materials, spherical Resorcinol-Formaldehyde (RF) and engineered Crystalline Silicotitanate (CST), are being considered for use within the SCIX unit. Both ion exchange materials have been studied extensively and are known to have high affinities for cesium ions in caustic tank waste supernates. RF is an elutable organic resin and CST is a non-elutable inorganic material. Waste treatment processes developed for the two technologies will differ with regard to solutions processed, secondary waste streams generated, optimum column size, and waste throughput. Pertinent references, anticipated processing sequences for utilization in waste treatment, gaps in the available data, and technical comparisons will be provided for the two ion exchange materials to assist in technology selection for SCIX. The engineered, granular form of CST (UOP IE-911) was the baseline ion exchange material used for the initial development and design of the SRS SCIX process (McCabe, 2005). To date, in-tank SCIX has not been implemented for treatment of radioactive waste solutions at SRS. Since initial development and consideration of SCIX for SRS waste treatment an alternative technology has been developed as part of the River Protection Project Waste Treatment Plant (RPP-WTP) Research and Technology program (Thorson, 2006). Spherical RF resin is the baseline media for cesium removal in the RPP-WTP, which was designed for the treatment of radioactive waste supernates and is currently under construction in Hanford, WA

  3. LITERATURE REVIEWS TO SUPPORT ION EXCHANGE TECHNOLOGY SELECTION FOR MODULAR SALT PROCESSING

    International Nuclear Information System (INIS)

    King, W.

    2007-01-01

    This report summarizes the results of literature reviews conducted to support the selection of a cesium removal technology for application in a small column ion exchange (SCIX) unit supported within a high level waste tank. SCIX is being considered as a technology for the treatment of radioactive salt solutions in order to accelerate closure of waste tanks at the Savannah River Site (SRS) as part of the Modular Salt Processing (MSP) technology development program. Two ion exchange materials, spherical Resorcinol-Formaldehyde (RF) and engineered Crystalline Silicotitanate (CST), are being considered for use within the SCIX unit. Both ion exchange materials have been studied extensively and are known to have high affinities for cesium ions in caustic tank waste supernates. RF is an elutable organic resin and CST is a non-elutable inorganic material. Waste treatment processes developed for the two technologies will differ with regard to solutions processed, secondary waste streams generated, optimum column size, and waste throughput. Pertinent references, anticipated processing sequences for utilization in waste treatment, gaps in the available data, and technical comparisons will be provided for the two ion exchange materials to assist in technology selection for SCIX. The engineered, granular form of CST (UOP IE-911) was the baseline ion exchange material used for the initial development and design of the SRS SCIX process (McCabe, 2005). To date, in-tank SCIX has not been implemented for treatment of radioactive waste solutions at SRS. Since initial development and consideration of SCIX for SRS waste treatment an alternative technology has been developed as part of the River Protection Project Waste Treatment Plant (RPP-WTP) Research and Technology program (Thorson, 2006). Spherical RF resin is the baseline media for cesium removal in the RPP-WTP, which was designed for the treatment of radioactive waste supernates and is currently under construction in Hanford, WA

  4. Effects of brine migration on waste storage systems. Final report

    International Nuclear Information System (INIS)

    Gaffney, E.S.; Nickell, R.E.

    1979-01-01

    Processes which can lead to mobilization of brine adjacent to spent fuel or nuclear waste canisters and some of the thermomechanical consequences have been investigated. Velocities as high as 4 x 10 -7 m s -1 (13 m y -1 ) are calculated at the salt/canister boundary. As much as 40 liters of pure NaCl brine could accumulate around each canister during a 10-year storage period. Accumulations of bittern brines would probably be less, in the range of 2 to 5 liters. With 0.5% water, NaCl brine accumulation over a 10-year storage cycle around a spent fuel canister producing 0.6 kW of heat is expected to be less than 1 liter for centimeter-size inclusions and less than 0.5 liter for millimeter-size inclusions. For bittern brines, about 25 years would be required to accumulate 0.4 liter. The most serious mechanical consequence of brine migration would be the increased mobility of the waste canister due to pressure solution. In pressure solution enhanced deformation, the existence of a thin film of fluid either between grains or between media (such as between a canister and the salt) provides a pathway by which the salt can be redistributed leading to a marked increase in strain rates in wet rock relative to dry rock. In salt, intergranular water will probably form discontinuous layers rather than films so that they would dominate pressure solution. A mathematical model of pressure solution indicates that pressure solution will not lead to appreciable canister motions except possibly in fine grained rocks (less than 10 -4 m). In fine grained salts, details of the contact surface between the canister and the salt bed may lead to large pressure solution motions. A numerical model indicates that heat transfer in the brine layer surrounding a spent fuel canister is not conduction dominated but has a significant convective component

  5. Recent studies on radiation damage formation in synthetic NaCl and natural rock salt for radioactive waste disposal applications

    International Nuclear Information System (INIS)

    Swyler, K.J.; Klaffky, R.W.; Levy, P.W.

    1980-01-01

    Radiation damage formation in natural rock salt is described as a function of irradiation temperature and plastic deformation. F-center formation decreases with increasing temperature while significant colloidal sodium formation occurs over a restricted temperature range around 150 0 C. Plastic deformation increases colloid formation; it is estimated that colloid concentrations may be increased by a factor of 3 if the rock salt near radioactive waste disposal canisters is heavily deformed. Optical bandshape analysis indicates systematic differences between the colloids formed in synthetic and natural rock salts

  6. Results for the first quarter calendar year 2017 tank 50H salt solution sample

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-04-12

    In this memorandum, the chemical and radionuclide contaminant results from the First Quarter Calendar Year 2017 (CY17) sample of Tank 50H salt solution are presented in tabulated form. The First Quarter CY17 Tank 50H samples [a 200 mL sample obtained 6” below the surface (HTF-50-17-7) and a 1 L sample obtained 66” from the tank bottom (HTF-50-17-8)] were obtained on January 15, 2017 and received at Savannah River National Laboratory (SRNL) on January 16, 2017. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours and the samples were pulled immediately after pump shut down. All volatile organic analysis (VOA) and semi-volatile organic analysis (SVOA) were performed on the surface sample and all other analyses were performed on the variable depth sample. The information from this characterization will be used by Savannah River Remediation (SRR) for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. The chemical and radionuclide contaminant results from the characterization of the First Quarter CY17 sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan (TTQAP). This memorandum is part of Deliverable 2 from SRR request. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the TTQAP for the Tank 50H saltstone task.

  7. Methods for removing transuranic elements from waste solutions

    International Nuclear Information System (INIS)

    Slater, S.A.; Chamberlain, D.B.; Connor, C.; Sedlet, J.; Srinivasan, B.; Vandegrift, G.F.

    1994-11-01

    This report outlines a treatment scheme for separating and concentrating the transuranic (TRU) elements present in aqueous waste solutions stored at Argonne National Laboratory (ANL). The treatment method selected is carrier precipitation. Potential carriers will be evaluated in future laboratory work, beginning with ferric hydroxide and magnetite. The process will result in a supernatant with alpha activity low enough that it can be treated in the existing evaporator/concentrator at ANL. The separated TRU waste will be packaged for shipment to the Waste Isolation Pilot Plant

  8. Test Results and Comparison of Triaxial Strength Testing of Waste Isolation Pilot Plant Clean Salt

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Stuart A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    This memorandum documents laboratory thermomechanical triaxial strength testing of Waste Isolation Pilot Plant (WIPP) clean salt. The limited study completed independent, adjunct laboratory tests in the United States to assist in validating similar testing results being provided by the German facilities. The testing protocol consisted of completing confined triaxial, constant strain rate strength tests of intact WIPP clean salt at temperatures of 25°C and 100°C and at multiple confining pressures. The stratigraphy at WIPP also includes salt that has been labeled “argillaceous.” The much larger test matrix conducted in Germany included both the so-called clean and argillaceous salts. When combined, the total database of laboratory results will be used to develop input parameters for models, assess adequacy of existing models, and predict material behavior. These laboratory studies are also consistent with the goals of the international salt repository research program. The goal of this study was to complete a subset of a test matrix on clean salt from the WIPP undertaken by German research groups. The work was performed at RESPEC in Rapid City, South Dakota. A rigorous Quality Assurance protocol was applied, such that corroboration provides the potential of qualifying all of the test data gathered by German research groups.

  9. Preparation of porous monolayer film by immersing the stearic acid Langmuir-Blodgett monolayer on mica in salt solution

    Energy Technology Data Exchange (ETDEWEB)

    Wang, S. [Institute of Near-Field Optics and Nano Technology, School of Physics and Optoelectronic Technology, Dalian University of Technology, Street No. 2 Linggong Road, Dalian 116024 (China); Li, Y.L.; Zhao, H.L.; Liang, H. [Institute of Photo-Biophysics, School of Physics and Electronic, Henan University, Jinming, Kaifeng 475004, Henan (China); Liu, B., E-mail: boliu@henu.edu.cn [Institute of Photo-Biophysics, School of Physics and Electronic, Henan University, Jinming, Kaifeng 475004, Henan (China); Pan, S., E-mail: span@dlut.edu.cn [Institute of Near-Field Optics and Nano Technology, School of Physics and Optoelectronic Technology, Dalian University of Technology, Street No. 2 Linggong Road, Dalian 116024 (China)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer Porous film has been prepared by immersing the stearic acid Langmuir-Blodgett monolayer on mica in salt solution. Black-Right-Pointing-Pointer The mechanism relies on the electrostatic screening effect of the cations in salt solution. Black-Right-Pointing-Pointer The factors influencing the size and area of the pores were investigated. - Abstract: Porous materials have drawn attention from scientists in many fields such as life sciences, catalysis and photonics since they can be used to induce some materials growth as expected. Especially, porous Langmuir-Blodgett (LB) film is an ideal material with controlled thickness and flat surface. In this paper, stearic acid (SA), which has been extensively explored in LB film technique, is chosen as the template material with known parameters to prepare the LB film, and then the porous SA monolayer film is obtained by means of etching in salt solution. The main etching mechanism is suggested that the cations in the solution block the electrostatic interaction between the polar carboxyl group of SA and the electronegative mica surface. The influencing factors (such as concentration of salt solution, valence of cation and surface pressure) of the porous SA film are systematically studied in this work. The novel method proposed in this paper makes it convenient to prepare porous monolayer film for designed material growth or cell culture.

  10. The potential/pH diagram of silver in aqueous ammonium salt solution

    NARCIS (Netherlands)

    Sluyters, J.H.; Wijnen, M.D.; Hul, H.J. van den

    1961-01-01

    The potential/pH diagram of silver in aqueous ammonium salt solution at 25°C has been calculated and verified experimentally. Calculations were carried out on the basis of the standard potential of the silver/silver-ion couple, the dissociation constants of the silver mono- and di-ammonia

  11. Recent studies of uranium and plutonium chemistry in alkaline radioactive waste solutions

    International Nuclear Information System (INIS)

    King, William D.; Wilmarth, William R.; Hobbs, David T.; Edwards, Thomas B.

    2008-01-01

    Solubility studies of uranium and plutonium in a caustic, radioactive Savannah River Site tank waste solution revealed the existence of uranium supersaturation in the as-received sample. Comparison of the results to predictions generated from previously published models for solubility in these waste types revealed that the U model poorly predicts solubility while Pu model predictions are quite consistent with experimental observations. Separate studies using simulated Savannah River Site evaporator feed solution revealed that the known formation of sodium aluminosilicate solids in waste evaporators can promote rapid precipitation of uranium from supersaturated solutions

  12. Coagulation mechanism of salt solution-extracted active component in Moringa oleifera seeds

    OpenAIRE

    Okuda, Tetsuji; Baes, Aloysius U.; Nishijima, Wataru; Okada, Mitsumasa

    2001-01-01

    This study focuses on the coagulation mechanism by the purified coagulant solution (MOC-SC-PC) with the coagulation active component extracted from M. oleifera seeds using salt solution. The addition of MOC-SC-PC into tap water formed insoluble matters. The formation was responsible for kaolin coagulation. On the other hand, insoluble matters were not formed when the MOC-SC-PC was added into distilled water. The formation was affected by Ca2+ or other bivalent cations which may connect each m...

  13. Salt disposition alternatives filtration at SRTC

    International Nuclear Information System (INIS)

    Walker, B. W.; Hobbs, D.

    2000-01-01

    Several of the prospective salt disposition alternative technologies require a monosodium titanate (MST) contact to remove strontium and actinides from inorganic salt solution feedstock. This feedstock also contains sludge solids from waste removal operations and may contain defoamers added in the evaporator systems. Filtration is required to remove the sludge and MST solids before sending the salt solution for further processing. This report describes testing performed using the Parallel Theological Experimental Filter (PREF). The PREF contains two single tube Mott sintered metal crossflow filters. For this test one filter was isolated so that the maximum velocities could be achieved. Previous studies showed slurries of MST and sludge in the presence of sodium tetraphenylborate (NaTPB) were filterable since the NaTPB slurry formed a filter cake which aided in removing the smaller MST and sludge particles. Some of the salt disposition alternative technologies do not use NaTPB raising the question of how effective crossflow filtration is with a feed stream containing only sludge and MST. Variables investigated included axial velocity, transmembrane pressure, defoamer effects, and solids concentration (MST and sludge). Details of the tests are outlined in the technical report WSRC-RP-98-O0691. Key conclusions from this study are: (1) Severe fouling of the Mott sintered metal filter did not occur with any of the solutions filtered. (2) The highest fluxes, in the range of .46 to 1.02 gpm/f 2 , were obtained when salt solution decanted from settled solids was fed to the filter. These fluxes would achieve 92 to 204 gpm filtrate production for the current ITP filters. The filtrate fluxes were close to the flux of 0.42 gpm/f 2 reported for In Tank Precipitation Salt Solution by Morrisey. (3) For the range of solids loading studied, the filter flux ranged from .04 to .17 gpm/f 2 which would result in a filtrate production rate of 9 to 31 gpm for the current HP filter. (4

  14. The use of marine aquaculture solid waste for nursery production of the salt marsh plants Spartina alterniflora and Juncus roemerianus

    Directory of Open Access Journals (Sweden)

    H.M. Joesting

    2016-05-01

    Full Text Available Recent technological advances in marine shrimp and finfish aquaculture alleviate many of the environmental risks associated with traditional aquaculture, but challenges remain in cost-effective waste management. Liquid effluent from freshwater aquaculture systems has been shown to be effective in agricultural crop production (i.e., aquaponics, but few studies have explored the potential for reuse of marine aquaculture effluent, particularly the solid fraction. The purpose of this study was to investigate the use of marine aquaculture solid waste as a nutrient source for the nursery production of two salt tolerant plants commonly used in coastal salt marsh restoration, Spartina alterniflora (smooth cordgrass and Juncus roemerianus (black needlerush. Specifically, measurements of plant biomass and tissue nitrogen and phosphorus allocation were compared between plants fertilized with dried shrimp biofloc solids and unfertilized controls, as well as between plants fertilized with dried fish solids and unfertilized controls. In both experiments, S. alterniflora plants fertilized with marine aquaculture solids showed few significant differences from unfertilized controls, whereas fertilized J. roemerianus plants had significantly greater biomass and absorbed and incorporated more nutrients in plant tissue compared to unfertilized controls. These results suggest that J. roemerianus may be a suitable plant species for the remediation of marine aquaculture solid waste. Keywords: Marine aquaculture, Salt marsh plants, Solid waste, Phytoremediation

  15. Solution-derived sodalite made with Si- and Ge-ethoxide precursors for immobilizing electrorefiner salt

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J., E-mail: brian.riley@pnnl.gov; Lepry, William C.; Crum, Jarrod V.

    2016-01-15

    Chlorosodalite has the general form of Na{sub 8}(AlSiO{sub 4}){sub 6}Cl{sub 2} and this paper describes experiments conducted to synthesize sodalite with a solution-based approach to immobilize a simulated spent electrorefiner salt solution containing a mixture of alkali, alkaline earth, and lanthanide chlorides. The reactants used were the salt solution, NaAlO{sub 2}, and either Si(OC{sub 2}H{sub 5}){sub 4} or Ge(OC{sub 2}H{sub 5}){sub 4}. Additionally, seven different glass sintering aids (at loadings of 5 mass%) were evaluated as sintering aids for consolidating the as-made powders using a cold-press-and-sinter technique. This process of using alkoxide additives for the Group IV component can be used to produce large quantities of sodalite at near-room temperature as compared to a method where colloidal silica was used as the silica source. However, the small particle sizes inhibited densification during heat treatments.

  16. Molecular dynamics study of charged dendrimers in salt-free solution : effect of counterions

    NARCIS (Netherlands)

    Gurtovenko, A.A.; Lyulin, S.V.; Karttunen, M.E.J.; Vattulainen, I.

    2006-01-01

    Polyamidoamine dendrimers, being protonated under physiological conditions, represent a promising class of nonviral, nanosized vectors for drug and gene delivery. We performed extensive molecular dynamics simulations of a generic model dendrimer in a salt-free solution with dendrimer’s terminal

  17. Ice crystallization in ultrafine water-salt aerosols: nucleation, ice-solution equilibrium, and internal structure.

    Science.gov (United States)

    Hudait, Arpa; Molinero, Valeria

    2014-06-04

    Atmospheric aerosols have a strong influence on Earth's climate. Elucidating the physical state and internal structure of atmospheric aqueous aerosols is essential to predict their gas and water uptake, and the locus and rate of atmospherically important heterogeneous reactions. Ultrafine aerosols with sizes between 3 and 15 nm have been detected in large numbers in the troposphere and tropopause. Nanoscopic aerosols arising from bubble bursting of natural and artificial seawater have been identified in laboratory and field experiments. The internal structure and phase state of these aerosols, however, cannot yet be determined in experiments. Here we use molecular simulations to investigate the phase behavior and internal structure of liquid, vitrified, and crystallized water-salt ultrafine aerosols with radii from 2.5 to 9.5 nm and with up to 10% moles of ions. We find that both ice crystallization and vitrification of the nanodroplets lead to demixing of pure water from the solutions. Vitrification of aqueous nanodroplets yields nanodomains of pure low-density amorphous ice in coexistence with vitrified solute rich aqueous glass. The melting temperature of ice in the aerosols decreases monotonically with an increase of solute fraction and decrease of radius. The simulations reveal that nucleation of ice occurs homogeneously at the subsurface of the water-salt nanoparticles. Subsequent ice growth yields phase-segregated, internally mixed, aerosols with two phases in equilibrium: a concentrated water-salt amorphous mixture and a spherical cap-like ice nanophase. The surface of the crystallized aerosols is heterogeneous, with ice and solution exposed to the vapor. Free energy calculations indicate that as the concentration of salt in the particles, the advance of the crystallization, or the size of the particles increase, the stability of the spherical cap structure increases with respect to the alternative structure in which a core of ice is fully surrounded by

  18. Preliminary investigation results as applied to utilization of Ukrainian salt formations for disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Shekhunova, S.B.; Khrushchov, D.P.; Petrichenko, O.I.

    1994-01-01

    The salt-bearing formations have been investigated in five regions of Ukraine. Upper Devonian and Lower Permian evaporite formations in Dnieper-Donets Depression and in the NW part of Donets basin are considered to be promising for disposal of high-level radioactive waste (HLRW). Rock salt occurs there either as bedded salts or as salt pillows and salt diapirs. Preliminary studies have resulted in selection of several candidate sites that show promise for construction of a subsurface pilot lab. Ten salt domes and two sites in bedded salts have been proposed for further exploration. Based on microstructural studies it is possible to separate the body of a salt structure and to locate within its limits the rock salt structure and to locate within its limits the rock salt blocks of different genesis, i.e.: (a) blocks characteristic of initial undisturbed sedimentary structure; (b) flow zones; (c) sliding planes; (d) bodies of loose or uncompacted rock salt. Ultramicrochemical examination of inclusions in halite have shown that they are composed of more than 40 minerals. It is emphasized that to assess suitability of a structure for construction of subsurface lab, and also the potential construction depth intervals, account should be taken of the results of ultra microchemical and microstructural data

  19. Concept development for saltstone and low level waste disposal

    International Nuclear Information System (INIS)

    Wilhite, E.L.

    1987-03-01

    A low-level alkaline salt solution will be a byproduct in the processing of high-level waste at the Savannah River Plant (SRP). This solution will be incorporated into a cement wasteform, saltstone, and placed in surface vaults. Laboratory and field testing and mathematical modeling have demonstrated the predictability of contaminant release from cement wasteforms. Saltstone disposal in surface vaults will meet drinking water standards in shallow groundwater at the disposal area boundary. Planning for new Low-Level Waste (LLW) disposal could incorporate concepts developed for saltstone disposal

  20. Boron removal in radioactive liquid waste by forward osmosis membrane

    Energy Technology Data Exchange (ETDEWEB)

    Doo Seong Hwang; Hei Min Choi; Kune Woo Lee; Jei Kwon Moon [KAERI, Daejeon (Korea, Republic of)

    2013-07-01

    This study investigated the treatment of boric acid contained in liquid radioactive waste using a forward osmosis membrane. The boron permeation through the membrane depends on the type of membrane, membrane orientation, pH of the feed solution, salt and boron concentration in the feed solution, and osmotic pressure of the draw solution. The boron flux begins to decline from pH 7 and increases with an increase of the osmotic driving force. The boron flux decreases slightly with the salt concentration, but is not heavily influenced by a low salt concentration. The boron flux increases linearly with the concentration of boron. No element except for boron was permeated through the FO membrane in the multi-component system. The maximum boron flux is obtained in an active layer facing a draw solution orientation of the CTA-ES membrane under conditions of less than pH 7 and high osmotic pressure. (authors)