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Sample records for waste simulant solutions

  1. Spray drying test of simulated borated waste solutions

    International Nuclear Information System (INIS)

    An Hongxiang; Zhou Lianquan; Fan Zhiwen; Sun Qi; Lin Xiaolong

    2007-01-01

    Performance and the effecting factors of spray drying of simulated borated waste solutions is studied for three contaeting methods between the atomized beads and the heated air, in which boron concentration is around 21000 ppm. The contacting modes are centrifugal atomizing co-current flow, pneumatic atomizing co-current flow and mixed flow. The results show that a free-flowing product in all these tests when the temperature of the solutions is between 62 degree C and 64 degree C, the inlet temperature of the spray drying chamber is between 210 degree C and 220 degree C, the temperature of the outlet of the spray drying chamber is between 110 and 120 degree C, the flow rate of the pressure air is 8.0 m 3 /h, the rotational speed of the centrifugal atomizer is 73.0 m/s. The diameters of the powder product which account for 95% of the feed range from 0.356 mm to 0.061 mm. The production capacity and water content in the powder increase in the order of pneumatic atomizing co-current flow, mixed flow and centrifugal atomizing co-current flow. The volume reduction coeffecient of spray drying is in the ranged of 0.22 and 0.27. (authors)

  2. Synergistic extraction behaviour of americium from simulated acidic waste solutions

    International Nuclear Information System (INIS)

    Pathak, P.N.; Veeraraghavan, R.; Mohapatra, P.K.; Manchanda, V.K.

    1998-01-01

    The extraction behaviour of americium has been investigated with mixtures of 3-phenyl-4-benzoyl-5-isoxazolone (PBI) and oxodonors viz. tri-n-butyl phosphate (TBP), tri-n-octyl phosphine oxide (TOPO) and di-n-butyl octanamide (DBOA) using dodecane as the diluent from 1-6 M HNO 3 media. It is observed that D Am remains unaltered with PBI concentration (in the range 0.06-0.1 M) at 1.47 M TBP in the entire range of HNO 3 concentration. PBI and TBP in combination appears more promising compared to other synergistic systems. The possibility of using this mixture for americium removal from high level liquid waste solution has been explored. Extraction studies indicated that prior removal of uranium by 20% TBP in dodecane is helpful in the quantitative recovery of americium in three contacts. Effect of lanthanides on D Am is found to be marginal. (orig.)

  3. Pore solution chemistry of simulated low-level liquid waste incorporated in cement grouts

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1995-12-01

    Expressed pore solutions from simulated low level liquid waste cement grouts cured at room temperature, 50 degree C and 90 degree C for various duration were analyzed by standard chemical methods and ion chromatography. The solid portions of the grouts were formulated with portland cement, fly ash, slag, and attapulgite clay in the ratios of 3:3:3:1. Two different solutions simulating off-gas condensates expected from vitrification of Hanford low level tank wastes were made. One is highly alkaline and contains the species Na + , P0 4 3- , N0 2 - , NO 3 - and OH - . The other is carbonated and contains the species, Na + , PO 4 3- , NO 2 - , NO 3 - , and CO 3 2- . In both cases phosphate rapidly disappeared from the pore solution, leaving behind sodium in the form of hydroxide. The carbonates were also removed from the pore solution to form calcium carbonate and possibly calcium monocarboaluminate. These reactions resulted in the increase of hydroxide ion concentration in the early period. Subsequently there was a significant reduction OH - and Na + ion concentrations. In contrast high concentration of N0 2 - and N0 3 - were retained in the pore solution indefinitely

  4. Treatment of Simulated Soil Decontamination Waste Solution by Ferrocyanide-Anion Exchange Resin Beads

    Energy Technology Data Exchange (ETDEWEB)

    Won, Hui Jun; Kim, Min Gil; Kim, Gye Nam; Jung, Chung Hun; Park, Jin Ho; Oh, Won Zin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2005-03-15

    Preparation of ferrocyanide-anion exchange resin and adsorption test of the prepared resin on the Cs{sup -} ion were performed. Adsorption capability of the prepared resin on the Cs{sup -} ion in the simulated citric acid based soil decontamination waste solution was 4 times greater than that of the commercial cation exchange resin. Adsorption equilibrium of the prepared resin on the Cs{sup -} ion reached within 360 minutes. Adsorption capability on the Cs{sup -} ion became to decrease above the necessary Co{sup 2-} ion concentration in the experimental range. Recycling test of the spent ion exchange resin by the successive application of hydrogen peroxide and hydrazine was also performed. It was found that desorption of Cs{sup -} ion from the resin occurred to satisfy the electroneutrality condition without any degradation of the resin.

  5. Preliminary results from uranium/americium affinity studies under experimental conditions for cesium removal from NPP ''Kozloduy'' simulated wastes solutions

    International Nuclear Information System (INIS)

    Nikiforova, A.; Kinova, L.; Peneva, C.; Taskaeva, I.; Petrova, P.

    2005-01-01

    We use the approach described by Westinghouse Savannah River Company using ammonium molybdophosphate (AMP) to remove elevated concentrations of radioactive cesium to facilitate handling waste samples from NPP K ozloduy . Preliminary series of tests were carried out to determine the exact conditions for sufficient cesium removal from five simulated waste solutions with concentrations of compounds, whose complexing power complicates any subsequent processing. Simulated wastes solutions contain high concentrations of nitrates, borates, H 2 C 2 O 4 , ethylenediaminetetraacetate (EDTA) and Citric acid, according to the composition of the real waste from the NPP. On this basis a laboratory treatment protocol was created. This experiment is a preparation for the analysis of real waste samples. In this sense the results are preliminary. Unwanted removal of non-cesium radioactive species from simulated waste solutions was studied with gamma spectrometry with the aim to find a compromise between on the one hand the AMP effectiveness and on the other hand unwanted affinity to AMP of Uranium and Americium. Success for the treatment protocol is defined by proving minimal uptake of U and Am, while at the same time demonstrating good removal effectiveness through the use of AMP. Uptake of U and Am were determined as influenced by oxidizing agents at nitric acid concentrations, proposed by Savannah River National laboratory. It was found that AMP does not significantly remove U and Am when concentration of oxidizing agents is more than 0.1M for simulated waste solutions and for contact times inherent in laboratory treatment protocol. Uranium and Americium affinity under experimental conditions for cesium removal were evaluated from gamma spectrometric data. Results are given for the model experiment and an approach for the real waste analysis is chosen. Under our experimental conditions simulated wastes solutions showed minimal affinity to AMP when U and Am are most probably in

  6. Simulations of water and solute movement in the buried waste repository at Vaalputs

    International Nuclear Information System (INIS)

    Hutson, J.L.

    1987-01-01

    A previous series of simulations examined the movement of water through trench cap configurations of several types. The objectives of this series are i) to extent the simulations from the surface to the bottom of the repository, accounting for the placement of drums, ii) to examine the magnitude and direction of water fluxes throughout this depth and iii) to simulate the movement of solutes, using various assumptions regarding solute adsorption. Two models were used. The first was an adaptation of a solute transport model which incorporates the transient water flow model used in previous simulations. This was used primarily to estimate the likely water fluxes in the drum placement region. Since it requires large amounts of computer time this model was used to simulate periods of one or two years only. The second model was a very simple steady state solute transport model which was used to simulate Cs distribution after a 100 year period, using flux data obtained from the transient model simulations. The most important conclusion reached from this series of simulations is that the movement of Cs in the soil under the likely water regime is extremely slow. 'Worst case' situations were simulated. Some of these situations are unlikely in reality but provide a useful indication of the rates of movement of solute under various conditions. For this reason it was assumed that plants were absent in cases when maximum percolation was simulated and present when maximum upward flow was simulated. In no case was a 'wick' (a textural barrier to unsaturated water flow) assumed to be present

  7. 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

  8. Detection of localized and general corrosion of mild steel in simulated defense nuclear waste solutions using electrochemical noise analysis

    International Nuclear Information System (INIS)

    Edgemon, G.L.; Ohl, P.C.; Bell, G.E.C.; Wilson, D.F.

    1995-12-01

    Underground waste tanks fabricated from mild steel store more than 60 million gallons of radioactive waste from 50 years of weapons production. Leaks are suspected in a significant number of tanks. The probable modes of corrosion failures are reported to be localized corrosion (e.g. nitrate stress corrosion cracking and pitting). The use of electrochemical noise (EN) for the monitoring and detection of localized corrosion processes has received considerable attention and application over the last several years. Proof of principle laboratory tests were conducted to verify the capability of EN evaluation to detect localized corrosion and to compare the predictions of general corrosion obtained from EN with those derived from other sources. Simple, pre-fabricated flat and U-bend specimens of steel alloys A516-Grade 60 (UNS K02100) and A537-CL 1 (UNS K02400) were immersed in temperature controlled simulated waste solutions. The simulated waste solution was either 5M NaNO 3 with 0.3M NaOH at 90 C or 11M NaNO 3 with 0.15M NaOH at 95 C. The electrochemical noise activity from the specimens was monitored and recorded for periods ranging between 140 and 240 hours. At the end of each test period, the specimens were metallographically examined to correlated EN data with corrosion damage

  9. Dissolution of Simulated and Radioactive Savannah River Site High-Level Waste Sludges with Oxalic Acid & Citric Acid Solutions

    International Nuclear Information System (INIS)

    STALLINGS, MARY

    2004-01-01

    This report presents findings from tests investigating the dissolution of simulated and radioactive Savannah River Site sludges with 4 per cent oxalic acid and mixtures of oxalic and citric acid previously recommended by a Russian team from the Khlopin Radium Institute and the Mining and Chemical Combine (MCC). Testing also included characterization of the simulated and radioactive waste sludges. Testing results showed the following: Dissolution of simulated HM and PUREX sludges with oxalic and citric acid mixtures at SRTC confirmed general trends reported previously by Russian testing. Unlike the previous Russian testing six sequential contacts of a mixture of oxalic acid citric acids at a 2:1 ratio (v/w) of acid to sludge did not produce complete dissolution of simulated HM and PUREX sludges. We observed that increased sludge dissolution occurred at a higher acid to sludge ratio, 50:1 (v/w), compared to the recommended ratio of 2:1 (v/w). We observed much lower dissolution of aluminum in a simulated HM sludge by sodium hydroxide leaching. We attribute the low aluminum dissolution in caustic to the high fraction of boehmite present in the simulated sludge. Dissolution of HLW sludges with 4 per cent oxalic acid and oxalic/citric acid followed general trends observed with simulated sludges. The limited testing suggests that a mixture of oxalic and citric acids is more efficient for dissolving HM and PUREX sludges and provides a more homogeneous dissolution of HM sludge than oxalic acid alone. Dissolution of HLW sludges in oxalic and oxalic/citric acid mixtures produced residual sludge solids that measured at higher neutron poison to equivalent 235U weight ratios than that in the untreated sludge solids. This finding suggests that residual solids do not present an increased nuclear criticality safety risk. Generally the neutron poison to equivalent 235U weight ratios of the acid solutions containing dissolved sludge components are lower than those in the untreated

  10. Early containment of high-alkaline solution simulating low-level radioactive waste stream in clay-bearing blended cement

    International Nuclear Information System (INIS)

    Kruger, A.A.; Olson, R.A.; Tennis, P.D.

    1995-04-01

    Portland cement blended with fly ash and attapulgite clay was mixed with high-alkaline solution simulating low-level radioactive waste stream at a one-to-one weight ratio. Mixtures were adiabatically and isothermally cured at various temperatures and analyzed for phase composition, total alkalinity, pore solution chemistry, and transport properties as measured by impedance spectroscopy. Total alkalinity is characterized by two main drops. The early one corresponds to a rapid removal of phosphorous, aluminum, sodium, and to a lesser extent potassium solution. The second drop from about 10 h to 3 days is mainly associated with the removal of aluminum, silicon, and sodium. Thereafter, the total alkalinity continues descending, but at a lower rate. All pastes display a rapid flow loss that is attributed to an early precipitation of hydrated products. Hemicarbonate appears as early as one hour after mixing and is probably followed by apatite precipitation. However, the former is unstable and decomposes at a rate that is inversely related to the curing temperature. At high temperatures, zeolite appears at about 10 h after mixing. At 30 days, the stabilized crystalline composition Includes zeolite, apatite and other minor amounts of CaCO 3 , quartz, and monosulfate Impedance spectra conform with the chemical and mineralogical data. The normalized conductivity of the pastes shows an early drop, which is followed by a main decrease from about 12 h to three days. At three days, the permeability of the cement-based waste as calculated by Katz-Thompson equation is over three orders of magnitude lower than that of ordinary portland cement paste. However, a further decrease in the calculated permeability is questionable. Chemical stabilization is favorable through incorporation of waste species into apatite and zeolite

  11. Corrosion tests of 316L and Hastelloy C-22 in simulated tank waste solutions

    International Nuclear Information System (INIS)

    Danielson, M.J.; Pitman, S.G.

    2000-01-01

    Both the 316L stainless steel and Hastelloy C-22 gave satisfactory corrosion performance in the simulated test environments. They were subjected to 100 day weight loss corrosion tests and electrochemical potentiodynamic evaluation. This activity supports confirmation of the design basis for the materials of construction of process vessels and equipment used to handle the feed to the LAW-melter evaporator. BNFL process and mechanical engineering will use the information derived from this task to select material of construction for process vessels and equipment

  12. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    Energy Technology Data Exchange (ETDEWEB)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S

    2004-07-01

    It has previously been demonstrated that immobilization of Cs{sup +} and/or Sr{sup 2+} sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO{sub 3}){sub 3} solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO{sub 3}){sub 3} losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO{sub 3}){sub 3}. In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  13. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    International Nuclear Information System (INIS)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S.

    2004-01-01

    It has previously been demonstrated that immobilization of Cs + and/or Sr 2+ sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO 3 ) 3 solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO 3 ) 3 losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO 3 ) 3 . In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  14. 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.

  15. Thermal and Physical Property Determinations for Ionsiv IE-911 Crystalline Silicotitanate and Savannah River Site Waste Simulant Solutions

    International Nuclear Information System (INIS)

    Bostick, D.T.; Steele, W.V.

    1999-01-01

    This document describes physical and thermophysical property determinations that were made in order to resolve questions associated with the decontamination of Savannah River Site (SRS) waste streams using ion exchange on crystalline silicotitanate (CST). The research will aid in the understanding of potential issues associated with cooling of feed streams within SRS waste treatment processes. Toward this end, the thermophysical properties of engineered CST, manufactured under the trade name, Ionsivereg s ign IE-911 by UOP, Mobile, AL, were determined. The heating profiles of CST samples from several manufacturers' production runs were observed using differential scanning calorimetric (DSC) measurements. DSC data were obtained over the region of 10 to 215 C to check for the possibility of a phase transition or any other enthalpic event in that temperature region. Finally, the heat capacity, thermal conductivity, density, viscosity, and salting-out point were determined for SRS waste simulants designated as Average, High NO 3 - and High OH - simulants

  16. Electrodriven selective transport of Cs+ using chlorinated cobalt dicarbollide in polymer inclusion membrane: a novel approach for cesium removal from simulated nuclear waste solution.

    Science.gov (United States)

    Chaudhury, Sanhita; Bhattacharyya, Arunasis; Goswami, Asok

    2014-11-04

    The work describes a novel and cleaner approach of electrodriven selective transport of Cs from simulated nuclear waste solutions through cellulose tri acetate (CTA)/poly vinyl chloride (PVC) based polymer inclusion membrane. The electrodriven cation transport together with the use of highly Cs+ selective hexachlorinated derivative of cobalt bis dicarbollide, allows to achieve selective separation of Cs+ from high concentration of Na+ and other fission products in nuclear waste solutions. The transport selectivity has been studied using radiotracer technique as well as atomic emission spectroscopic technique. Transport studies using CTA based membrane have been carried out from neutral solution as well as 0.4 M HNO3, while that with PVC based membrane has been carried out from 3 M HNO3. High decontamination factor for Cs+ over Na+ has been obtained in all the cases. Experiment with simulated high level waste solution shows selective transport of Cs+ from most of other fission products also. Significantly fast Cs+ transport rate along with high selectivity is an interesting feature observed in this membrane. The current efficiency for Cs+ transport has been found to be ∼100%. The promising results show the possibility of using this kind of electrodriven membrane transport methods for nuclear waste treatment.

  17. Solutions for Waste Management

    International Nuclear Information System (INIS)

    2013-01-01

    To safely and securely dispose of highlevel and long-lived radioactive waste, this material needs to be stored for a period of time that is very long compared to our everyday experience. Underground disposal facilities need to be designed and constructed in suitable geological conditions that can be confidently demonstrated to contain and isolate the hazardous waste from our environment for hundreds of thousands of years. Over this period of time, during which the safety of an underground waste repository system must be assured, the waste's radioactivity will decay to a level that cannot pose a danger to people or the environment. The archaeological record can help in visualizing such a long period of time. Climates change, oceans rise and vanish, and species evolve in the course of a one hundred millennia. Rocks bear witness to all of these changes. Geologists in their search for safe repositories for the long-term disposal of high level radioactive waste have identified rock formations that have proven stable for millions of years. These geological formations are expected to remain stable for millions of years and can serve as host formations for waste repositories.

  18. 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.

  19. Modeling of hydrologic conditions and solute movement in processed oil shale waste embankments under simulated climatic conditions

    International Nuclear Information System (INIS)

    Reeves, T.L.; Turner, J.P.; Hasfurther, V.R.; Skinner, Q.D.

    1992-06-01

    The scope of this program is to study interacting hydrologic, geotechnical, and chemical factors affecting the behavior and disposal of combusted processed oil shale. The research combines bench-scale testing with large scale research sufficient to describe commercial scale embankment behavior. The large scale approach was accomplished by establishing five lysimeters, each 7.3 x 3.0 x 3.0 m deep, filled with processed oil shale that has been retorted and combusted by the Lurgi-Ruhrgas (Lurgi) process. Approximately 400 tons of Lurgi processed oil shale waste was provided by RBOSC to carry out this study. Research objectives were designed to evaluate hydrologic, geotechnical, and chemical properties and conditions which would affect the design and performance of large-scale embankments. The objectives of this research are: assess the unsaturated movement and redistribution of water and the development of potential saturated zones and drainage in disposed processed oil shale under natural and simulated climatic conditions; assess the unsaturated movement of solubles and major chemical constituents in disposed processed oil shale under natural and simulated climatic conditions; assess the physical and constitutive properties of the processed oil shale and determine potential changes in these properties caused by disposal and weathering by natural and simulated climatic conditions; assess the use of previously developed computer model(s) to describe the infiltration, unsaturated movement, redistribution, and drainage of water in disposed processed oil shale; evaluate the stability of field scale processed oil shale solid waste embankments using computer models

  20. Westinghouse waste simulation and optimization software tool

    International Nuclear Information System (INIS)

    Mennicken, Kim; Aign, Jorg

    2013-01-01

    Applications for dynamic simulation can be found in virtually all areas of process engineering. The tangible benefits of using dynamic simulation can be seen in tighter design, smoother start-ups and optimized operation. Thus, proper implementation of dynamic simulation can deliver substantial benefits. These benefits are typically derived from improved process understanding. Simulation gives confidence in evidence based decisions and enables users to try out lots of 'what if' scenarios until one is sure that a decision is the right one. In radioactive waste treatment tasks different kinds of waste with different volumes and properties have to be treated, e.g. from NPP operation or D and D activities. Finding a commercially and technically optimized waste treatment concept is a time consuming and difficult task. The Westinghouse Waste Simulation and Optimization Software Tool will enable the user to quickly generate reliable simulation models of various process applications based on equipment modules. These modules can be built with ease and be integrated into the simulation model. This capability ensures that this tool is applicable to typical waste treatment tasks. The identified waste streams and the selected treatment methods are the basis of the simulation and optimization software. After implementing suitable equipment data into the model, process requirements and waste treatment data are fed into the simulation to finally generate primary simulation results. A sensitivity analysis of automated optimization features of the software generates the lowest possible lifecycle cost for the simulated waste stream. In combination with proven waste management equipments and integrated waste management solutions, this tool provides reliable qualitative results that lead to an effective planning and minimizes the total project planning risk of any waste management activity. It is thus the ideal tool for designing a waste treatment facility in an optimum manner

  1. Westinghouse waste simulation and optimization software tool

    Energy Technology Data Exchange (ETDEWEB)

    Mennicken, Kim; Aign, Jorg [Westinghouse Electric Germany GmbH, Hamburg (Germany)

    2013-07-01

    Applications for dynamic simulation can be found in virtually all areas of process engineering. The tangible benefits of using dynamic simulation can be seen in tighter design, smoother start-ups and optimized operation. Thus, proper implementation of dynamic simulation can deliver substantial benefits. These benefits are typically derived from improved process understanding. Simulation gives confidence in evidence based decisions and enables users to try out lots of 'what if' scenarios until one is sure that a decision is the right one. In radioactive waste treatment tasks different kinds of waste with different volumes and properties have to be treated, e.g. from NPP operation or D and D activities. Finding a commercially and technically optimized waste treatment concept is a time consuming and difficult task. The Westinghouse Waste Simulation and Optimization Software Tool will enable the user to quickly generate reliable simulation models of various process applications based on equipment modules. These modules can be built with ease and be integrated into the simulation model. This capability ensures that this tool is applicable to typical waste treatment tasks. The identified waste streams and the selected treatment methods are the basis of the simulation and optimization software. After implementing suitable equipment data into the model, process requirements and waste treatment data are fed into the simulation to finally generate primary simulation results. A sensitivity analysis of automated optimization features of the software generates the lowest possible lifecycle cost for the simulated waste stream. In combination with proven waste management equipments and integrated waste management solutions, this tool provides reliable qualitative results that lead to an effective planning and minimizes the total project planning risk of any waste management activity. It is thus the ideal tool for designing a waste treatment facility in an optimum manner

  2. Modeling of hydrologic conditions and solute movement in processed oil shale waste embankments under simulated climatic conditions

    International Nuclear Information System (INIS)

    Turner, J.P.; Hasfurther, V.

    1992-01-01

    The scope of the research program and the continuation is to study interacting hydrologic, geotechnical, and chemical factors affecting the behavior and disposal of combusted processed oil shale. The research combines bench-scale testing with large scale research sufficient to describe commercial scale embankment behavior. The large scale approach was accomplished by establishing five lysimeters, each 7.3 x 3.0 x 3.0 m deep, filled with processed oil shale that has been retorted and combusted by the Lurgi-Ruhrgas (Lurgi) process. Approximately 400 tons of Lurgi processed oil shale waste was provided by Rio Blanco Oil Shale Co., Inc. (RBOSC) through a separate cooperative agreement with the University of Wyoming (UW) to carry out this study. Three of the lysimeters were established at the RBOSC Tract C-a in the Piceance Basin of Colorado. Two lysimeters were established in the Environmental Simulation Laboratory (ESL) at UW. The ESL was specifically designed and constructed so that a large range of climatic conditions could be physically applied to the processed oil shale which was filled in the lysimeter cells

  3. Modeling of hydrologic conditions and solute movement in processed oil shale waste embankments under simulated climatic conditions

    International Nuclear Information System (INIS)

    Turner, J.P.; Reeves, T.L.; Skinner, Q.D.; Hasfurther, V.

    1992-11-01

    The scope of the original research program and of its continuation is to study interacting hydrologic, geotechnical, and chemical factors affecting the behavior and disposal of combusted processed oil shale. The research combines bench-scale testing with large-scale testing sufficient to describe commercial-scale embankment behavior. The large-scale testing was accomplished by constructing five lysimeters, each 7.3x3.0x3.0 m deep, filled with processed oil shale that has been retorted and combusted by the Lurgi-Ruhrgas (Lurgi) process (Schmalfield 1975). Approximately 400 tons of Lurgi processed oil shale waste was provided by Rio Blanco Oil Shale Co., Inc. to carry out this study. Three of the lysimeters were established at the RBOSC Tract C-a in the Piceance Basin near Rifle, Colorado. Two lysimeters were established in the Environmental Simulation Laboratory (ESL) at UW. The ESL was specifically designed and constructed so that a large range of climatic conditions could be physically applied to the processed oil shale which was placed in the lysimeter cells. This report discusses and summarizes results from scientific efforts conducted between October 1991 and September 1992 for Fiscal Year 1992

  4. Thermal and physical property determination for IONSIV/256 IE-911 crystalline silicotitanate and Savannah River Site waste simulant solutions

    International Nuclear Information System (INIS)

    Canada, C.C.

    1999-01-01

    This document describes physical and thermophysical property determinations that were made in order to resolve questions associated with the decontamination of Savannah River Site waste streams using ion exchange on crystalline silicotitanate

  5. 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

  6. Reaction of H atoms with chelators in highly basic solution: H2 production in high level liquid waste simulants

    International Nuclear Information System (INIS)

    Barnabas, F.; Cerny, E.; Jonah, C.D.; Meisel, D.; Sauer, M.C. Jr.

    1995-01-01

    The rate constants for hydrogen abstraction by H from ethylene-diamine tetracetic acid (EDTA), N-(2-hydroxyethyl)-ethylenediaminetriacetic acid (HEDTA), nitrilotriacetic acid (NTA), iminodiacetic acid (IDA), glycolic acid and citric acid were measured at pH 13. The predominant product of these reactions is H 2 . The rate constants obtained are more than an order of magnitude larger than the literature rates that had been measured at pH 1. These measurements are of significance for understanding the radiolytic production of H 2 in nuclear waste storage tanks. (Author)

  7. Selective separation of cesium from simulated high level liquid waste solution using 1,3-dioctyloxy calix[4]arene-benzo-crown-6

    International Nuclear Information System (INIS)

    Vikas Kumar; Sharma, J.N.; Hubli, R.C.

    2014-01-01

    The 25,27-di(octyloxy)calix[4]arenebenzocrown-6 (CBC) in 1,3-alternate conformation was synthesized indigenously starting from its intermediates in good yield and purity. The extraction studies of CBC were carried out by using two different phase modifiers namely isodecyl alcohol and ortho-nitrophenyl hexyl ether. Detailed investigations on the effect of various parameters like, concentration of phase modifiers, aqueous phase acidity, ligand concentration, nitrate ion concentration and effect of temperature on extraction of cesium have been carried out. The concentration of phase modifiers was optimized to be 30 % in n-dodecane to ensure optimum extraction of cesium. Stoichiometry of the extracted complex determined by slope analysis method reveals 1:1:1 molar ratio for CsNO 3 :CBC:HNO 3 . The extraction process was found to be exothermic as determined from the plot of log K ex versus 1/T. The solvent system with a composition 0.01 M CBC/30 % phase modifier/n-dodecane was found to be effective for selective separation of cesium from simulated high level liquid waste solution. (author)

  8. 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

  9. 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.

  10. 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

  11. Removal of radioactive materials from waste solutions via magnetic ferrites

    International Nuclear Information System (INIS)

    Boyd, T.E.; Kochen, R.L.; Price, M.Y.

    1982-01-01

    Ferrite waste treatment was found to be effective in removing actinides from simulated Rocky Flats process waste solutions. With a one-stage ferrite treatment, plutonium concentrations were consistently reduced from 10 -4 g/l to less than 10 -8 g/l, and americium concentrations were lowered from 10 -7 g/l to below 10 -10 g/l. In addition, siginficantly less solid was produced as compared with the flocculant precipitation technique now employed at Rocky Flats. Aging of ferrite solids and elevated beryllium and phosphate concentrations were identified as interferences in the ferrite treatment of process waste, but neither appeeared serious enough to prevent implementation in plant operations

  12. 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.

  13. Modeling of hydrologic conditions and solute movement in processed oil shale waste embankments under simulated climatic conditions. Final report, November 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    A study is described on the hydrological and geotechnical behavior of an oil shale solid waste. The objective was to obtain information which can be used to assess the environmental impacts of oil shale solid waste disposal in the Green River Basin. The spent shale used in this study was combusted by the Lurgi-Ruhrgas process by Rio Blanco Oil Shale Company, Inc. Laboratory bench-scale testing included index properties, such as grain size distribution and Atterberg limits, and tests for engineering properties including hydraulic conductivity and shear strength. Large-scale tests were conducted on model spent shale waste embankments to evaluate hydrological response, including infiltration, runoff, and seepage. Large-scale tests were conducted at a field site in western Colorado and in the Environmental Simulation Laboratory (ESL)at the University of Wyoming. The ESL tests allowed the investigators to control rainfall and temperature, providing information on the hydrological response of spent shale under simulated severe climatic conditions. All experimental methods, materials, facilities, and instrumentation are described in detail, and results are given and discussed. 34 refs.

  14. 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.)

  15. 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

  16. Solubilities of gases in simulated Tank 241-SY-101 wastes

    International Nuclear Information System (INIS)

    Norton, J.D.; Pederson, L.R.

    1995-09-01

    Oxygen, nitrogen, hydrogen, methane, and nitrous oxide solubilities were evaluated as a function of temperature in SYl-SIM-93B, a homogeneous simulated waste mixture containing sodium hydroxide, sodium nitrite, sodium nitrate, sodium aluminate, and sodium carbonate, the principal inorganic constituents of the wastes in Tank 241-SY-101. Ammonia solubility data for this simulated waste was obtained as a function of temperature in an earlier study. The choice of a homogeneous waste mixture in this study has the advantage of eliminating complications associated with a changing electrolyte concentration as a function of temperature that would be encountered with a slurry simulant. Dissolution is one of the means by which gases may be retained in Hanford Site wastes. While models are available to estimate gas solubilities in electrolyte solutions, few data are in existence that pertain to highly concentrated, multicomponent electrolytes such as those stored in Hanford Site waste tanks

  17. Denitrification of nitrate waste solutions

    International Nuclear Information System (INIS)

    Bertolami, R.J.; Chao, E.I.; Choi, W.M.; Johnson, B.R.; Varlet, J.L.P.

    1976-01-01

    Growth rates for the denitrifying bacteria Pseudomonas Stutzeri were studied to minimize the time necessary to start up a bacterial denitrification reactor. Batch experiments were performed in nine 250-ml Erlenmeyer flasks, a 7-liter fermentor, and a 67-liter fermentor. All reactors maintained an anaerobic environment. Initial microorganism inoculum concentration was varied over four orders of magnitude. Initial nitrate and substrate carbon concentrations were varied from 200 to 6000 ppm and from 56 to 1596 ppm, respectively, with a carbon-to-nitrogen weight ratio of 1.18. In all experiments, except those with the highest initial substrate-to-bacteria ratio, no growth was observed due to substrate depletion during the lag period. In those experiments which did exhibit an increase in bacterial population, growth also stopped due to substrate depletion. A model simulating microbe growth during the induction period was developed, but insufficient data were available to properly adjust the model constants. Because of this, the model does not accurately predict microbe growth. The metabolism of Pseudomonas Stutzeri was studied in detail. This resulted in a prediction of the denitrification stoichiometry during steady state reactor operation. Iron was found to be an important component for bacterial anabolism

  18. Simulation of ceramic materials relevant for nuclear waste management: Case of La{sub 1−x}Eu{sub x}PO{sub 4} solid solution

    Energy Technology Data Exchange (ETDEWEB)

    Kowalski, Piotr M., E-mail: p.kowalski@fz-juelich.de; Ji, Yaqi; Li, Yan; Arinicheva, Yulia; Beridze, George; Neumeier, Stefan; Bukaemskiy, Andrey; Bosbach, Dirk

    2017-02-15

    Using powerful computational resources and state-of-the-art methods of computational chemistry we contribute to the research on novel nuclear waste forms by providing atomic scale description of processes that govern the structural incorporation and the interactions of radionuclides in host materials. Here we present various results of combined computational and experimental studies on La{sub 1−x}Eu{sub x}PO{sub 4} monazite-type solid solution. We discuss the performance of DFT + U method with the Hubbard U parameter value derived ab initio, and the derivation of various structural, thermodynamic and radiation-damage related properties. We show a correlation between the cation displacement probabilities and the solubility data, indicating that the binding of cations is the driving factor behind both processes. The combined atomistic modeling and experimental studies result in a superior characterization of the investigated material.

  19. Nuclear waste repository simulation experiments

    International Nuclear Information System (INIS)

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

    1986-12-01

    This document is the third joint annual report on the Cooperative German-American 'Brine Migration Tests' that are in progress at the Asse salt mine in the Federal Republic of Germany (FRG). This Government supported mine serves as an underground test facility for research and development (R and D)-work in the field of nuclear waste repository research and simulation experiments. The tests are designed to simulate a nuclear waste repository to measure the effects of heat and gamma radiation on brine migration, salt decrepitation, disassociation of brine, and gases collected. The thermal mechanical behavior of salt, such as room closure, stresses and changes of the properties of salt are measured and compared with predicted behavior. This document covers the following sections: Issues and test objectives: This section presents issues that are investigated by the Brine Migration Test, and the test objectives derived from these issues; test site: This section describes the test site location and geology in the Asse mine; test description: A description of the test configuration, procedures, equipment, and instrumentation is given in this section; actual test chronology: The actual history of the test, in terms of the dates at which major activities occured, is presented in this section. Test results: This section presents the test results observed to data and the planned future work that is needed to complete the test; conclusions and recommendations: This section summarizes the conclusions derived to date regarding the Brine Migration Test. Additional work that would be useful to resolve the issues is discussed. (orig.)

  20. The effect of gamma radiation on reference electrodes and platinum and carbon steel bare metal electrodes in a simulated waste solution

    International Nuclear Information System (INIS)

    Danielson, M.J.

    1993-09-01

    Electrochemical potential measurements of materials in waste tanks are important in determining if the materials have a propensity for stress corrosion cracking and pitting. Potential measurement requires a reference electrode, but the effect of radiation on the potential generated by the reference electrode has been an unknown quantity. To determine the significance of the radiation effect, Pacific Northwest Laboratory conducted studies of five types of electrodes under gamma radiation at room temperature. The subjects were two types of silver/silver chloride reference electrodes (Fisher and Lazaran), a mercury/calomel reference electrode, a platinum ''flag,'' and a piece of A-537 carbon steel; the electrodes were exposed to a simulated caustic tank environment. The Fisher silver/silver chloride and mercury/calomel reference electrodes showed essentially no radiation effects up to a flux of 2.1E6 R/h and fluence of 9.4E8 R, indicating they would be useful reference electrodes for in-tank studies. The Lazaran reg-sign silver/silver chloride electrode showed serious potential deviations at fluences of 2.E8 R, but it would be the electrode of choice in many situations because it is simple to maintain. Radiation affected the open circuit potential of both the platinum and carbon steel electrodes. This effect indicates that corrosion studies without radiation may not duplicate the corrosion processes expected in a waste tank. Mixed-potential theory was used to explain the radiation effects

  1. Processing of waste solutions from electrochemical decontamination

    International Nuclear Information System (INIS)

    Charlot, L.A.; Allen, R.P.; Arrowsmith, H.W.; Hooper, J.L.

    1979-09-01

    The use of electropolishing as a decontamination technique will be effective only if we can minimize the amount of secondary waste requiring disposal and economically recycle part of the decontamination electrolyte. Consequently, a solution purification method is needed to remove the dissolved contamination and metal in the electrolyte. This report describes the selection of a purification method for a phosphoric acid electrolyte from the following possible acid reclamation processes: ion exchange, solvent extraction, precipitation, distillation, electrolysis, and membrane separation

  2. 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

  3. 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.

  4. Comparison of Waste Feed Delivery Small Scale Mixing Demonstration Simulant to Hanford Waste

    Energy Technology Data Exchange (ETDEWEB)

    Wells, Beric E.; Gauglitz, Phillip A.; Rector, David R.

    2012-07-10

    The Hanford double-shell tank (DST) system provides the staging location for waste that will be transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Specific WTP acceptance criteria for waste feed delivery describe the physical and chemical characteristics of the waste that must be met before the waste is transferred from the DSTs to the WTP. One of the more challenging requirements relates to the sampling and characterization of the undissolved solids (UDS) in a waste feed DST because the waste contains solid particles that settle and their concentration and relative proportion can change during the transfer of the waste in individual batches. A key uncertainty in the waste feed delivery system is the potential variation in UDS transferred in individual batches in comparison to an initial sample used for evaluating the acceptance criteria. To address this uncertainty, a number of small-scale mixing tests have been conducted as part of Washington River Protection Solutions' Small Scale Mixing Demonstration (SSMD) project to determine the performance of the DST mixing and sampling systems. A series of these tests have used a five-part simulant composed of particles of different size and density and designed to be equal or more challenging than AY-102 waste. This five-part simulant, however, has not been compared with the broad range of Hanford waste, and thus there is an additional uncertainty that this simulant may not be as challenging as the most difficult Hanford waste. The purpose of this study is to quantify how the current five-part simulant compares to all of the Hanford sludge waste, and to suggest alternate simulants that could be tested to reduce the uncertainty in applying the current testing results to potentially more challenging wastes.

  5. Scientific Solutions to Nuclear Waste Environmental Challenges

    International Nuclear Information System (INIS)

    Johnson, Bradley R.

    2014-01-01

    The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount

  6. Cold Dissolved Saltcake Waste Simulant Development, Preparation, and Analysis

    International Nuclear Information System (INIS)

    Rassat, Scot D.; Mahoney, Lenna A.; Russell, Renee L.; Bryan, Samuel A.; Sell, Rachel L.

    2003-01-01

    CH2M HILL Hanford Group, Inc. is identifying and developing supplemental process technologies to accelerate the Hanford tank waste cleanup mission. Bulk vitrification, containerized grout, and steam reforming are three technologies under consideration for treatment of the radioactive saltcake wastes in 68 single-shell tanks. To support development and testing of these technologies, Pacific Northwest National Laboratory (PNNL) was tasked with developing a cold dissolved saltcake simulant formulation to be representative of an actual saltcake waste stream, preparing 25- and 100-L batches of the simulant, and analyzing the composition of the batches to ensure conformance to formulation targets. Lacking a defined composition for dissolved actual saltcake waste, PNNL used available tank waste composition information and an equilibrium chemistry model (Environmental Simulation Program [ESP(trademark)]) to predict the concentrations of analytes in solution. Observations of insoluble solids in initial laboratory preparations for the model-predicted formulation prompted reductions in the concentration of phosphate and silicon in the final simulant formulation. The analytical results for the 25- and 100-L simulant batches, prepared by an outside vendor to PNNL specifications, agree within the expected measurement accuracy (∼10%) of the target concentrations and are highly consistent for replicate measurements, with a few minor exceptions. In parallel with the production of the 2nd simulant batch (100-L), a 1-L laboratory control sample of the same formulation was carefully prepared at PNNL to serve as an analytical standard. The instrumental analyses indicate that the vendor prepared batches of solution adequately reflect the as-formulated simulant composition. In parallel with the simulant development effort, a nominal 5-M (molar) sodium actual waste solution was prepared at the Hanford Site from a limited number of tank waste samples. Because this actual waste solution w

  7. Electrochemical studies of the corrosion behavior of a low-carbon steel in aqueous chloride solutions simulating accident conditions of radioactive waste disposal

    International Nuclear Information System (INIS)

    Farvaque-Bera, A.M.; Leistikow, S.

    1991-01-01

    The fine-grained structural steel DIN W.Nr. 1.0566 was exposed to various sulfate and chloride-containing aqueous solutions, the latter ones simulating the potential accidental environment of water intrusion into a salt mine. By electrochemical measurements in salt brines, the following results were achieved: (1) The corrosion rate is highly dependent on salt brine composition, pH and temperature. (2) Active metal dissolution led to formation of shallow pits as surface corrosion phenomenon. Thus, the application of electrochemical techniques - under non-polarized as well as under potentiodynamic conditions - proved to be suitable for fast qualitative testing of the influence of various environmental parameters on steel corrosion. (orig.)

  8. Secondary Waste Simulant Development for Cast Stone Formulation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rinehart, Donald E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Swanberg, David J. [Washington River Protection Solutions, Richland, WA (United States); Mahoney, J. [Washington River Protection Solutions, Richland, WA (United States)

    2015-04-01

    Washington River Protection Solutions, LLC (WRPS) funded Pacific Northwest National Laboratory (PNNL) to conduct a waste form testing program to implement aspects of the Secondary Liquid Waste Treatment Cast Stone Technology Development Plan (Ashley 2012) and the Hanford Site Secondary Waste Roadmap (PNNL 2009) related to the development and qualification of Cast Stone as a potential waste form for the solidification of aqueous wastes from the Hanford Site after the aqueous wastes are treated at the Effluent Treatment Facility (ETF). The current baseline is that the resultant Cast Stone (or grout) solid waste forms would be disposed at the Integrated Disposal Facility (IDF). Data and results of this testing program will be used in the upcoming performance assessment of the IDF and in the design and operation of a solidification treatment unit planned to be added to the ETF. The purpose of the work described in this report is to 1) develop simulants for the waste streams that are currently being fed and future WTP secondary waste streams also to be fed into the ETF and 2) prepare simulants to use for preparation of grout or Cast Stone solid waste forms for testing.

  9. 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.

  10. Global solutions through simulation for better decommissioning

    International Nuclear Information System (INIS)

    Scoto Di Suoccio, Ines; Testard, Vincent

    2016-01-01

    Decommissioning is a new activity in sense that it only exists a limited experience. Moreover, each facility is different due to their own history and there is no rule about choosing a decommissioning strategy. There are three major decommissioning strategies. First, 'immediate dismantling', which means the action of decommissioning begins immediately after the transfer of waste and nuclear material. Second, 'deferred dismantling strategy', which means that the facility is maintained into a containment zone from thirty to one hundred years before being decommissioned. Finally, 'entombment', means the facility is placed into a reinforced containment until the radionuclides decay and reach a level allowing the site release. When a strategy is decided many factors have to be taken into account. Into a major project such as a reactor decommissioning, there are many smaller projects. The decommissioning strategy can be different among these smaller projects. For some reasons, some entry data are not perfectly known. For example, dosimetric activity has not been updated through time or after specific events. Indeed, because of uncertainties and/or hypothesis existing around projects and their high level of interdependency, global solutions are a good way to choose the best decommissioning strategy. Actually, each entry data has consequences on output results whether it is on costs, cumulated dose, waste or delays. These output data are interdependent and cannot be taken apart from each other. Whether the dose, delays or waste management, all have impact on costs. To obtain an optimal scenario into a special environment, it is necessary to deal with all these items together. This global solution can be implemented thanks to simulation in dedicated software which helps to define the global strategy, to optimize the scenario, and to prevent contingencies. As a complete scenario simulation can be done quickly and efficiently, many strategies can

  11. Nuclear waste management. Pioneering solutions from Finland

    International Nuclear Information System (INIS)

    Rasilainen, Kari

    2016-01-01

    Presentation outline: Background: Nuclear energy in Finland; Nuclear Waste Management (NWM) Experiences; Low and Intermediate Level Waste (LILW); High Level Waste - Deep Geological Repository (DGR); NWM cost estimate in Finland; Conclusions: World-leading expert services

  12. Analysis of Discharged Gas from Incinerator using Simulated Organic Solution

    International Nuclear Information System (INIS)

    Kim, Seungil; Kim, Hyunki; Heo, Jun; Kang, Dukwon; Kim, Yunbok; Kwon, Youngbock

    2014-01-01

    Korea has no experience of treatment of RI organic waste and appropriate measures for treatment of organic waste did not suggested. RI organic wastes which are occurring in KOREA are stored at the RI waste storage building of KORAD. But they can't no more receive the RI organic waste because the storage facility for RI organic waste was saturated with these organic wastes. In case of Japan, they recognized the dangerousness of long-term storage for RI organic wastes. In case of Korea, the released concentration of gaseous pollutant from the incinerator is regulated by attached table No.1 of the Notification No. 2012-60 of Nuclear Safety Commission and attached table No.8 of Clean Air Conservation Act. And the dioxin from the incinerator is regulated by attached table No.3 of Persistent Organic Pollutants Control Act. This experiment was performed to examine whether the incinerator introduced from Japan is manufactured suitably for municipal law regulation and to confirm the compliance about the gaseous pollutant released from incinerator with the above-mentioned laws especially attached table No.1 of NSC using simulated organic waste solution. In this experiment, we examined whether the incinerator was manufactured suitably for municipal law regulation and confirmed the compliance about the gaseous pollutant released from incinerator with the above-mentioned laws using simulated organic waste solution. The design requirement of incinerator for RI organic waste in the municipal law regulation is proposed briefly but the requirements for more detail about the incinerator are proposed in regulation of Japan. The incinerator used in this experiment is satisfied with all clauses of the domestic as well as Japan. Multiple safety functions were installed in the incinerator such as air purge system to remove unburned inflammable gases in the furnace and earthquake detector. Also, perfect combustion of RI organic waste is achieved because the temperature in the furnace

  13. Analysis of Discharged Gas from Incinerator using Simulated Organic Solution

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seungil; Kim, Hyunki; Heo, Jun; Kang, Dukwon [HaJI Co., Ltd., Radiation Eng. Center, Siheung (Korea, Republic of); Kim, Yunbok; Kwon, Youngbock [KORAD, Daejeon (Korea, Republic of)

    2014-05-15

    Korea has no experience of treatment of RI organic waste and appropriate measures for treatment of organic waste did not suggested. RI organic wastes which are occurring in KOREA are stored at the RI waste storage building of KORAD. But they can't no more receive the RI organic waste because the storage facility for RI organic waste was saturated with these organic wastes. In case of Japan, they recognized the dangerousness of long-term storage for RI organic wastes. In case of Korea, the released concentration of gaseous pollutant from the incinerator is regulated by attached table No.1 of the Notification No. 2012-60 of Nuclear Safety Commission and attached table No.8 of Clean Air Conservation Act. And the dioxin from the incinerator is regulated by attached table No.3 of Persistent Organic Pollutants Control Act. This experiment was performed to examine whether the incinerator introduced from Japan is manufactured suitably for municipal law regulation and to confirm the compliance about the gaseous pollutant released from incinerator with the above-mentioned laws especially attached table No.1 of NSC using simulated organic waste solution. In this experiment, we examined whether the incinerator was manufactured suitably for municipal law regulation and confirmed the compliance about the gaseous pollutant released from incinerator with the above-mentioned laws using simulated organic waste solution. The design requirement of incinerator for RI organic waste in the municipal law regulation is proposed briefly but the requirements for more detail about the incinerator are proposed in regulation of Japan. The incinerator used in this experiment is satisfied with all clauses of the domestic as well as Japan. Multiple safety functions were installed in the incinerator such as air purge system to remove unburned inflammable gases in the furnace and earthquake detector. Also, perfect combustion of RI organic waste is achieved because the temperature in the furnace

  14. Mechanical compaction of Waste Isolation Pilot Plant simulated waste

    International Nuclear Information System (INIS)

    Butcher, B.M.; Thompson, T.W.; VanBuskirk, R.G.; Patti, N.C.

    1991-06-01

    The investigation described in this report acquired experimental information about how materials simulating transuranic (TRU) waste compact under axial compressive stress, and used these data to define a model for use in the Waste Isolation Pilot Plant (WIPP) disposal room analyses. The first step was to determine compaction curves for various simultant materials characteristic of TRU waste. Stress-volume compaction curves for various combinations of these materials were than derived to represent the combustible, metallic, and sludge waste categories. Prediction of compaction response in this manner is considered essential for the WIPP program because of the difficulties inherent in working with real (radioactive) waste. Next, full-sized 55-gallon drums of simulated combustible, metallic, and sludge waste were axially compacted. These results provided data that can be directly applied to room consolidation and data for comparison with the predictions obtained in Part 1 of the investigation. Compaction curves, which represent the combustible, metallic, and sludge waste categories, were determined, and a curve for the averaged waste inventory of the entire repository was derived. 9 refs., 31 figs., 12 tabs

  15. An MCNP simulation for API applications to waste management issues

    International Nuclear Information System (INIS)

    Tunnell, L.N.

    1994-01-01

    Issues associated with waste management have increasingly become a focal point of attention for both the government and private sector since the end of the cold war. The problem are difficult to solve; the solutions are expensive to implement. Consequently, the development of a data simulation system capable of predicting the performance of a real system can save many thousands of dollars in travel expenses, optimization of experimental parameters, etc.. In this effort, computer codes were developed to simulate the production of associated particle imaging data so that its performance in a typical waste management application can be assessed

  16. Quantitative measurement of cyanide species in simulated ferrocyanide Hanford waste

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pool, K.H.; Matheson, J.D.

    1993-02-01

    Analytical methods for the quantification of cyanide species in Hanford simulated high-level radioactive waste were pursued in this work. Methods studied include infrared spectroscopy (solid state and solution), Raman spectroscopy, Moessbauer spectroscopy, X-ray diffraction, scanning electron microscopy-electron dispersive spectroscopy (SEM-EDS), and ion chromatography. Of these, infrared, Raman, X-ray diffraction, and ion chromatography techniques show promise in the concentration range of interest. Quantitation limits for these latter four techniques were demonstrated to be approximately 0.1 wt% (as cyanide) using simulated Hanford wastes

  17. Study on cementation of simulated radioactive borated liquid wastes

    International Nuclear Information System (INIS)

    Sun Qina; Li Junfeng; Wang Jianlong

    2010-01-01

    To compare sulfoaluminate cement with ordinary Portland cement on their cementation of radioactive borated liquid waste and to provide more data for formula optimization, simulated radioactive borated liquid waste were solidified by the two cements. 28 d compressive strength and strength losses after water/freezing/irradiation resistance tests were investigated. Leaching test and X-ray diffraction analysis were also conducted. The results show that it is feasible to solidify borated liquid wastes with sulfoaluminate cement and ordinary Portland cement with formulas used in the study. The 28 d compressive strengths, strength losses after tests and simulated nuclides leaching rates of the solidified waste forms meet the demand of GB 14569.1-93. The sulfoaluminate cement formula show better retention of Cs + than ordinary Portland cement formula. Boron, in form of B (OH) 4 - , incorporate in ettringite as solid solutions. (authors)

  18. Extraction of technetium from simulated Hanford tank wastes

    International Nuclear Information System (INIS)

    Chaiko, D.J.; Vojta, Y.; Takeuchi, M.

    1993-01-01

    Aqueous biphasic separation systems are being developed for the treatment of liquid radioactive wastes. These extraction systems are based on the use of polyethylene glycols (PEGs) for the selective extraction and recovery of long-lived radionuclides, such as 129 I, 75 Se, and 99 Tc, from caustic solutions containing high concentrations of nitrate, nitrite, and carbonate. Because of the high ionic strengths of supernatant liquids in Hanford underground storage tanks, aqueous biphasic systems can be generated by simply adding aqueous PEG solutions directly to the waste solution. In the process, anionic species like I - and TcO 4 - are selectively transferred to the less dense PEG phase. The partition coefficient for a wide range of inorganic cations and anions, such as sodium, potassium, aluminum, nitrate, nitrate, and carbonate, are all less than one. The authors present experimental data on extraction of technetium from several simulated Hanford tank wastes at 25 degree and 50 degree C

  19. 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

  20. Solidification of radioactive waste solutions by pelletization technique

    International Nuclear Information System (INIS)

    Akbar, A.H.; Koester, R.; Rudolph, G.

    1980-04-01

    A possible way of performing the cement fixation of radioactive wastes is the incorporation into cement pellets on a pan pelletizer, followed by embedding the pellets into an inactive cement matrix. This procedure is suitable for various types of waste, particularly for medium level liquid wastes, and can be used both at drum disposal and at in-situ solidification. This report describes some initial studies on the pelletization technique using a laboratory pelletizer. Formation and size of the pellets have been found to be determined by speed, angle, and load of the pan, ratio and mode of addition of the liquid and solid components, ect. Pellets in various compositions have been produced from cement and water or simulated waste solution, in some cases with the addition of bentonite for improving cesium retention. Some mechanical properties of the pellets such as fall height of fresh pellets, development of hardness (crush test), impact and abrasion resistance, have been determined. Some preliminary experiments were done on backfilling the void space between the pellets - about 40 per cent of the bulk volume - with cement grouts of appropriate compositions. (orig.) [de

  1. Waste and Simulant Precipitation Issues

    International Nuclear Information System (INIS)

    Steele, W.V.

    2000-01-01

    As Savannah River Site (SRS) personnel have studied methods of preparing high-level waste for vitrification in the Defense Waste Processing Facility (DWPF), questions have arisen with regard to the formation of insoluble waste precipitates at inopportune times. One option for decontamination of the SRS waste streams employs the use of an engineered form of crystalline silicotitanate (CST). Testing of the process during FY 1999 identified problems associated with the formation of precipitates during cesium sorption tests using CST. These precipitates may, under some circumstances, obstruct the pores of the CST particles and, hence, interfere with the sorption process. In addition, earlier results from the DWPF recycle stream compatibility testing have shown that leaching occurs from the CST when it is stored at 80 C in a high-pH environment. Evidence was established that some level of components of the CST, such as silica, was leached from the CST. This report describes the results of equilibrium modeling and precipitation studies associated with the overall stability of the waste streams, CST component leaching, and the presence of minor components in the waste streams

  2. The different solutions for the waste storage

    International Nuclear Information System (INIS)

    Fillion, E.

    2001-01-01

    Created in 1979, the National agency for the management of radioactive waste (A.N.D.R.A.) is a public establishment in charge of the management of radioactive waste produced in France. It is independent from waste producers and watches over the long term protection of man and his environment, at any step of radioactive waste management. It has for mission to check the waste quality and to conceive, to establish, to build and to manage storage centers where waste are stored according their characteristics. (N.C.)

  3. 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.

  4. 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

  5. Energy from waste: a wholly acceptable waste-management solution

    International Nuclear Information System (INIS)

    Porteous, A.

    1997-01-01

    This paper briefly reviews the 'waste management hierarchy' and why it should be treated as a checklist and not a piece of unquestioning dogma. The role of energy from waste (EfW) is examined in depth to show that it is a rigorous and environmentally sound waste-management option which complements other components of the waste-management hierarchy and assists resource conservation. (Copyright (c) 1997 Elsevier Science B.V., Amsterdam. All rights reserved.)

  6. Radioactive waste management - a safe solution

    International Nuclear Information System (INIS)

    1993-01-01

    This booklet sets out current United Kingdom government policy regarding radioactive waste management and is aimed at reassuring members of the public concerned about the safety of radioactive wastes. The various disposal or, processing or storage options for low, intermediate and high-level radioactive wastes are explained and sites described, and the work of the Nuclear Industry Radioactive Waste Executive (NIREX) is outlined. (UK)

  7. Low-level waste management - suggested solutions for problem wastes

    International Nuclear Information System (INIS)

    Pechin, W.H.; Armstrong, K.M.; Colombo, P.

    1984-01-01

    Problem wastes are those wastes which are difficult or require unusual expense to place into a waste form acceptable under the requirements of 10 CFR 61 or the disposal site operators. Brookhaven National Laboratory has been investigating the use of various solidification agents as part of the DOE Low-Level Waste Management Program for several years. Two of the leading problem wastes are ion exchange resins and organic liquids. Ion exchange resins can be solidified in Portland cement up to about 25 wt % resin, but waste forms loaded to this degree exhibit significantly reduced compressive strength and may disintegrate when immersed in water. Ion exchange resins can also be incorporated into organic agents. Mound Laboratory has been investigating the use of a joule-heated glass melter as a means of disposing of ion exchange resins and organic liquids in addition to other combustible wastes

  8. WASTES: Waste System Transportation and Economic Simulation--Version 2:

    International Nuclear Information System (INIS)

    Sovers, R.A.; Shay, M.R.; Ouderkirk, S.J.; McNair, G.W.; Eagle, B.G.

    1988-02-01

    The Waste System Transportation and Economic Simulation (WASTES) Technical Reference Manual was written to describe and document the algorithms used within the WASTES model as implemented in Version 2.23. The manual will serve as a reference for users of the WASTES system. The intended audience for this manual are knowledgeable users of WASTES who have an interest in the underlying principles and algorithms used within the WASTES model. Each algorithm is described in nonprogrammers terminology, and the source and uncertainties of the constants in use by these algorithms are described. The manual also describes the general philosophy and rules used to: 1) determine the allocation and priority of spent fuel generation sources to facility destinations, 2) calculate transportation costs, and 3) estimate the cost of at-reactor ex-pool storage. A detailed description of the implementation of many of the algorithms is also included in the WASTES Programmers Reference Manual (Shay and Buxbaum 1986a). This manual is separated into sections based on the general usage of the algorithms being discussed. 8 refs., 14 figs., 2 tabs

  9. Corrosion of simulated nuclear waste glass

    International Nuclear Information System (INIS)

    Music, S.; Ristic, M.; Gotic, M.; Foric, J.

    1988-01-01

    In this study the preparation and characterization of borosilicate glasses of different chemical composition were investigated. Borosilicate glasses were doped with simulated nuclear waste oxides. The chemical corrosion in water of these glasses was followed by measuring the leach rates as a function of time. It was found that a simulated nuclear waste glass with the chemical composition (weight %), 15.61% Na 2 O, 10.39% B 2 O 3 , 45.31% SiO 2 , 13.42% ZnO, 6.61% TiO 2 and 8.66% waste oxides, is characterized by low melting temperature and with good corrosion resistance in water. Influence of passive layers on the leaching behaviour of nuclear waste glasses is discussed. (author) 20 refs.; 7 figs.; 4 tabs

  10. 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.

  11. Electrochemical processing of nitrate waste solutions

    International Nuclear Information System (INIS)

    Genders, D.; Weinberg, N.; Hartsough, D.

    1992-01-01

    The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F - ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions

  12. 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

  13. Waste Evaporator Accident Simulation Using RELAP5 Computer Code

    International Nuclear Information System (INIS)

    POLIZZI, L.M.

    2004-01-01

    An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs

  14. 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...

  15. Technical solutions for waste treatment in the Belene project

    International Nuclear Information System (INIS)

    Büttner, K.; Eichhorn, H.

    2011-01-01

    Outline: In June 2010 NUKEM Technologies GmbH was awarded a contract from ATOMSTROYEXPORT JSC to perform the complete work package related to designing and completion of the equipment for treatment of radioactive waste on the turn-key basis for Belene NPP. Technical Solutions: Waste Streams and Technologies at UKC and UKS; Concentration Plant; Thermal Treatment of Resins Sorting Facility; Biological Waste Water Treatment; Conditioning – Cementation • Sorting of Radwaste; Plasma Facility; Grouting; Filter Press; Monitoring and Tracking

  16. 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

  17. Industrial Water Waste, Problems and the Solution

    Directory of Open Access Journals (Sweden)

    Alif Noor Anna

    2004-01-01

    Full Text Available Recently, the long term development in Indonesia has changed agricultural sector to the industrial sector. This development can apparently harm our own people. This is due to the waste that is produced from factories. The waste from various factories seems to have different characteristics. This defference encourages us to be able to find out different of methods of managing waste so that cost can be reduced, especially in water treatment. In order that industrial development and environmental preservation can run together in balance, many institutions involved should be consider, especially in the industrial chain, the environment, and human resource, these three elements can be examined in terms of their tolerance to waste.

  18. Environment friendly solutions of plastics waste management

    International Nuclear Information System (INIS)

    Pirzada, F.N.; Riffat, T.; Pirzada, M.D.S.

    1997-01-01

    The use of plastics is growing worldwide. Consequently, the volume of plastic waste is also increasing. Presently, more than 100 million tons per year of plastic is being produced globally. In U.S. alone more than 10 million tons of plastic is being dumped in landfills as waste, where it can persist for decades. This has resulted in exhausting old landfills. Public awareness on environment is also making it difficult to find new sites for landfills. This has led to increased emphasis on treatment and recycling of plastic wastes. Volume reduction of plastic waste has some unique problems. They arise from the intrinsic chemical inertness of polymeric materials and toxic nature of their degradation byproducts. The paper reviews the present state of plastic waste management including land filling, incineration and recycling technologies. The technical problems associated with each of these processes have been discussed. There is also brief description of ongoing R and D for finding improved methods of plastic waste handling with their promises and problems. The role of tougher legislation in developing better recycling methods and degradable plastics has also been evaluated. The claims made by the proponents of degradable polymers have also been critically reviewed. (authors)

  19. Hanford solid waste management system simulation

    International Nuclear Information System (INIS)

    Shaver, S.R.; Armacost, L.L.; Konynenbelt, H.S.; Wehrman, R.R.

    1994-12-01

    This paper describes systems analysis and simulation model development for a proposed solid waste management system at a U.S. Department of Energy Site. The proposed system will include a central storage facility, four treatment facilities, and three disposal sites. The material managed by this system will include radioactive, hazardous, and mixed radioactive and hazardous wastes. The objective of the modeling effort is to provide a means of evaluating throughput and capacity requirements for the proposed treatment, storage, and disposal facilities. The model is used to evaluate alternative system configurations and the effect on the alternatives of changing waste stream characteristics and receipt schedules. An iterative modeling and analysis approach is used that provides macro-level models early in the project and establishes credibility with the customer. The results from the analyses based on the macro models influence system design decisions and provide information that helps focus subsequent model development. Modeling and simulation of alternative system configurations and operating strategies yield a better understanding of the solid waste system requirements. The model effectively integrates information obtained through systems analysis and waste characterization to provide a consistent basis for system and facility planning

  20. Characterisation of Plasma Vitrified Simulant Plutonium Contaminated Material Waste

    International Nuclear Information System (INIS)

    Hyatt, Neil C.; Morgan, Suzy; Stennett, Martin C.; Scales, Charlie R.; Deegan, David

    2007-01-01

    The potential of plasma vitrification for the treatment of a simulant Plutonium Contaminated Material (PCM) was investigated. It was demonstrated that the PuO 2 simulant, CeO 2 , could be vitrified in the amorphous calcium iron aluminosilicate component of the product slag with simultaneous destruction of the organic and polymer waste fractions. Product Consistency Tests conducted at 90 deg. C in de-ionised water and buffered pH 11 solution show the PCM slag product to be durable with respect to release of Ce. (authors)

  1. Simulation of construction and demolition waste leachate

    Energy Technology Data Exchange (ETDEWEB)

    Townsend, T.G.; Jang, Y.; Thurn, L.G.

    1999-11-01

    Solid waste produced from construction and demolition (C and D) activities is typically disposed of in unlined landfills. Knowledge of C{ampersand}D debris landfill leachate is limited in comparison to other types of wastes. A laboratory study was performed to examine leachate resulting from simulated rainfall infiltrating a mixed C and D waste stream consisting of common construction materials (e.g., concrete, wood, drywall). Lysimeters (leaching columns) filled with the mixed C and D waste were operated under flooded and unsaturated conditions. Leachate constituent concentrations in the leachate from specific waste components were also examined. Leachate samples were collected and analyzed for a number of conventional water quality parameters including pH, alkalinity, total organic carbon, total dissolved solids, and sulfate. In experiments with the mixed C and D waste, high concentrations of total dissolved solids (TDS) and sulfate were detected in the leachate. C and D leachates produced as a result of unsaturated conditions exhibited TDS concentrations in the range of 570--2,200 mg/L. The major contributor to the TDS was sulfate, which ranged in concentration between 280 and 930 mg/L. The concentrations of sulfate in the leachate exceeded the sulfate secondary drinking water standard of 250 mg/L.

  2. New sorbents and ion exchangers for nuclear waste solution remediation

    International Nuclear Information System (INIS)

    Clearfield, A.; Peng, G.Z.; Cahill, R.A.; Bellinghausen, P.; Aly, H.I.; Scott, K.; Wang, J.D.

    1993-01-01

    There is now a concerted effort underway to clean up the accumulated nuclear wastes as the major sites around the country. Because of the complexity of the mixtures in the holding tanks highly specific exchangers are required to fulfill a multitude of desired tasks. These include removal of Cs + , Sr 2+ , Tc, Actinides and possible recovery of rare and precious metals. No one exchanger or sequestrant can accomplish these tasks and a variety of exchangers in a multistep process will be required. The behavior of a number of inorganic ion exchangers in a multistep process will be required. The behavior of a number of inorganic ion exchangers and new organo-inorganic exchangers towards Cs + , Sr 2+ and rare-earth ions in acid and basic media will be described. Preliminary data on the effect of high levels of sodium nitrate on the uptake of these ions will also be presented, as well as the changes observed in selectivity in simulated waste solutions. A possible separation scheme based on these data will be described

  3. Fluctuation Solution Theory Properties from Molecular Simulation

    DEFF Research Database (Denmark)

    Abildskov, Jens; Wedberg, R.; O’Connell, John P.

    2013-01-01

    The thermodynamic properties obtained in the Fluctuation Solution Theory are based on spatial integrals of molecular TCFs between component pairs in the mixture. Molecular simulation, via either MD or MC calculations, can yield these correlation functions for model inter- and intramolecular...

  4. Mathematical simulation of a waste rock heap

    International Nuclear Information System (INIS)

    Scharer, J.M.; Pettit, C.M.; Chambers, D.B.; Kwong, E.C.

    1994-01-01

    A computer model has been developed to simulate the generation of acidic drainage in waste rock piles. The model considers the kinetic rates of biological and chemical oxidation of sulfide minerals (pyrite, pyrrhotite) present as fines and rock particles, as well as chemical processes such as dissolution (kinetic or equilibrium controlled), complexation (from equilibrium and stoichiometry of several complexes), and precipitation (formation of complexes and secondary minerals). Through mass balance equations and solubility constraints (e.g., pH, phase equilibria) the model keeps track of the movement of chemical species through the waste pile and provides estimates of the quality of seepage (pH, sulfate, iron, acidity, etc.) leaving the heap. The model has been expanded to include the dissolution (thermodynamic and sorption equilibrium), adsorption and coprecipitation of uranium and radium. The model was applied to simulate waste rock heaps in British Columbia, Canada and in Thueringia, Germany. To improve the accuracy and confidence of long-term predictions of seepage quality, the entire history of the heaps was simulated. Cumulative acidity loads and water treatment considerations were used as a basis for evaluation of various decommissioning alternatives. Simulation of the technical leaching history of a heap in Germany showed it will generate contaminated leachate requiring treatment for acidity and radioactivity for several hundred years; cover installation was shown to provide a significant reduction of potential burdens, although chemical treatment would still be required beyond 100 years

  5. Engineering solutions to the management of solid radioactive waste

    International Nuclear Information System (INIS)

    1991-01-01

    The management of radioactive waste, its safe handling and ultimate disposal, is of vital concern to engineers in the nuclear industry. The international conference 'Engineering Solutions to the Management of Solid Radioactive Waste', organized by the Institution of Mechanical Engineers and held in Manchester in November 1991, provided a forum for the discussion and comparison of the different methods of waste management used in Europe and America. Papers presented and discussed included: the interaction between the design of containers for low level radioactive waste and the design of a deep repository, commercial low level waste disposal sites in the United States, and the development of radioactive waste monitoring systems at the Sellafield reprocessing complex. This volume is a collection of 22 papers presented at the conference. All are indexed separately. (author)

  6. Sustainable solutions for solid waste management in Southeast Asian countries

    International Nuclear Information System (INIS)

    Uyen Nguyen Ngoc; Schnitzer, Hans

    2009-01-01

    Human activities generate waste and the amounts tend to increase as the demand for quality of life increases. Today's rate in the Southeast Asian Nations (ASEANs) is alarming, posing a challenge to governments regarding environmental pollution in the recent years. The expectation is that eventually waste treatment and waste prevention approaches will develop towards sustainable waste management solutions. This expectation is for instance reflected in the term 'zero emission systems'. The concept of zero emissions can be applied successfully with today's technical possibilities in the agro-based processing industry. First, the state-of-the-art of waste management in Southeast Asian countries will be outlined in this paper, followed by waste generation rates, sources, and composition, as well as future trends of waste. Further on, solutions for solid waste management will be reviewed in the discussions of sustainable waste management. The paper emphasizes the concept of waste prevention through utilization of all wastes as process inputs, leading to the possibility of creating an ecosystem in a loop of materials. Also, a case study, focusing on the citrus processing industry, is displayed to illustrate the application of the aggregated material input-output model in a widespread processing industry in ASEAN. The model can be shown as a closed cluster, which permits an identification of opportunities for reducing environmental impacts at the process level in the food processing industry. Throughout the discussion in this paper, the utilization of renewable energy and economic aspects are considered to adapt to environmental and economic issues and the aim of eco-efficiency. Additionally, the opportunities and constraints of waste management will be discussed.

  7. 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.)

  8. ALLIANCES: simulation platform for radioactive waste disposal

    International Nuclear Information System (INIS)

    Deville, E.; Montarnal, Ph.; Loth, L.; Chavant, C.

    2009-01-01

    CEA, ANDRA and EDF are jointly developing the software platform ALLIANCES whose aim is to produce a tool for the simulation of nuclear waste storage and disposal. This type of simulations deals with highly coupled thermo-hydro-mechanical-chemical and radioactive (T-H-M-C-R) processes. ALLIANCES' aim is to accumulate within the same simulation environment the already acquired knowledge and to gradually integrate new knowledge. The current version of ALLIANCES contains the following modules: - Hydraulics and reactive transport in unsaturated and saturated media; - Multi-phase flow; - Mechanical thermal-hydraulics; - Thermo-Aeraulics; - Chemistry/Transport coupling in saturated media; - Alteration of waste package coupled with the environment; - Sensitivity analysis tools. The next releases will include more physical phenomena like: reactive transport in unsaturated flow and multicomponent multiphase flow; incorporation of responses surfaces in sensitivity analysis tools; integration of parallel numerical codes for flow and transport. Since the distribution of the first release of ALLIANCES (December 2003), the platform was used by ANDRA for his safety simulation program and by CEA for reactive transport simulations (migration of uranium in a soil, diffusion of different reactive species on laboratory samples, glass/iron/clay interaction). (authors)

  9. Overview on the Multinational Collaborative Waste Storage and Disposal Solutions

    International Nuclear Information System (INIS)

    MARGEANU, C.A.

    2013-01-01

    The main drivers for a Safe, Secure and Global Energy future become clear and unequivocal: Security of supply for energy sources, Low-carbon electricity generation and Extended nuclear power assuring economic nuclear energy production, safe nuclear facilities and materials, safe and secure radioactive waste management and public acceptance. Responsible use of nuclear power requires that – in addition to safety, security and environmental protection associated with NPPs operation – credible solutions to be developed for dealing with the radioactive waste produced and especially for a responsible long term radioactive waste management. The paper deals with the existing multinational initiative in nuclear fuel cycle and the technical documents sustaining the multinational/regional disposal approach. Meantime, the paper far-reaching goal is to highlight on: What is offering the multinational waste storage and disposal solutions in terms of improved nuclear security ‽

  10. Photochemical oxidation: A solution for the mixed waste dilemma

    Energy Technology Data Exchange (ETDEWEB)

    Prellberg, J.W.; Thornton, L.M.; Cheuvront, D.A. [Vulcan Peroxidation Systems, Inc., Tucson, AZ (United States)] [and others

    1995-12-31

    Numerous technologies are available to remove organic contamination from water or wastewater. A variety of techniques also exist that are used to neutralize radioactive waste. However, few technologies can satisfactorily address the treatment of mixed organic/radioactive waste without creating unacceptable secondary waste products or resulting in extremely high treatment costs. An innovative solution to the mixed waste problem is on-site photochemical oxidation. Liquid-phase photochemical oxidation has a long- standing history of successful application to the destruction of organic compounds. By using photochemical oxidation, the organic contaminants are destroyed on-site leaving the water, with radionuclides, that can be reused or disposed of as appropriate. This technology offers advantages that include zero air emissions, no solid or liquid waste formation, and relatively low treatment cost. Discussion of the photochemical process will be described, and several case histories from recent design testing, including cost analyses for the resulting full-scale installations, will be presented as examples.

  11. 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

  12. Modeling VOC transport in simulated waste drums

    International Nuclear Information System (INIS)

    Liekhus, K.J.; Gresham, G.L.; Peterson, E.S.; Rae, C.; Hotz, N.J.; Connolly, M.J.

    1993-06-01

    A volatile organic compound (VOC) transport model has been developed to describe unsteady-state VOC permeation and diffusion within a waste drum. Model equations account for three primary mechanisms for VOC transport from a void volume within the drum. These mechanisms are VOC permeation across a polymer boundary, VOC diffusion across an opening in a volume boundary, and VOC solubilization in a polymer boundary. A series of lab-scale experiments was performed in which the VOC concentration was measured in simulated waste drums under different conditions. A lab-scale simulated waste drum consisted of a sized-down 55-gal metal drum containing a modified rigid polyethylene drum liner. Four polyethylene bags were sealed inside a large polyethylene bag, supported by a wire cage, and placed inside the drum liner. The small bags were filled with VOC-air gas mixture and the VOC concentration was measured throughout the drum over a period of time. Test variables included the type of VOC-air gas mixtures introduced into the small bags, the small bag closure type, and the presence or absence of a variable external heat source. Model results were calculated for those trials where the VOC permeability had been measured. Permeabilities for five VOCs [methylene chloride, 1,1,2-trichloro-1,2,2-trifluoroethane (Freon-113), 1,1,1-trichloroethane, carbon tetrachloride, and trichloroethylene] were measured across a polyethylene bag. Comparison of model and experimental results of VOC concentration as a function of time indicate that model accurately accounts for significant VOC transport mechanisms in a lab-scale waste drum

  13. 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

  14. Effects of soluble organic complexants and their degradation products on the removal of selected radionuclides from high-level waste. Part II: Distributions of Sr, Cs, Tc, and Am onto 32 absorbers from four variations of Hanford tank 101-SY simulant solution

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1995-04-01

    Many of the radioactive waste storage tanks at U.S. Department of Energy facilities contain organic compounds that have been degraded by radiolysis and chemical reactions during decades of storage. In this second part of our three-part investigation of the effects of soluble organic complexants and their degradation products, we measured the sorption of strontium, cesium, technetium, and americium onto 32 absorbers that offer high sorption of these elements in the absence of organic complexants. The four solutions tested were (1) a simulant for a 3:1 dilution of Hanford Tank 101-SY contents that initially contained ethylenediaminetetraacetic acid (EDTA), (2) this simulant after gamma-irradiation to 34 Mrads, (3) the unirradiated simulant after treatment with a hydrothermal organic-destruction process, and (4) the irradiated simulant after hydrothermal processing. For each of 512 element/absorber/solution combinations, we measured distribution coefficients (Kds) twice for each period for dynamic contact periods of 30 min, 2 h, and 6 h to obtain information about sorption kinetics. On the basis of our 3,072 measured Kd values, the sorption of strontium and americium is significantly decreased by the organic components of the simulant solutions, whereas the sorption of cesium and technetium appears unaffected by the organic components of the simulant solutions

  15. Solutions for energy recovery of animal waste from leather industry

    International Nuclear Information System (INIS)

    Lazaroiu, Gheorghe; Pană, Constantin; Mihaescu, Lucian; Cernat, Alexandru; Negurescu, Niculae; Mocanu, Raluca; Negreanu, Gabriel

    2017-01-01

    Highlights: • Animal fats in blend with diesel fuel for energy valorification through combustion. • Animal waste from tanneries as fuel and for biogas production. • Experimental tests using animal fats as fuel for diesel engines. • Experimental tests modifying the characteristic parameters. - Abstract: Secondary products from food and leather industries are regarded as animal wastes. Conversion of these animal wastes into fuels represents an energy recovery solution not only because of their good combustion properties, but also from the viewpoint of supply stability. A tannery factory usually processes 60–70 t/month of crude leathers, resulting in 12–15 t/month of waste. Fats, which can be used as the input fuel for diesel engines (in crude state or as biodiesel), represent 10% of this animal waste, while the rest are proteins that can be used to generate biogas through anaerobic digestion. Herein, we analyse two approaches to the use of animal waste from tanneries: as fuel for diesel engines and for biogas generation for heat production. Diesel fuelling and fuelling by animal wastes are compared in terms of the engine performance and pollutant emissions. The effects of animal waste usage on the pollutant emissions level, exhaust gas temperature, indicated mean effective pressure, maximum pressure, and engine efficiency are analysed. The energy recovery technologies for animal waste, which are analysed in this work, can be easily implemented and can simultaneously solve the problem posed by animal wastes by using them as an alternative to fossil fuels. Animal fats can be considered an excellent alternative fuel for diesel engines without major constructive modifications.

  16. Incineration of Non-radioactive Simulated Waste

    International Nuclear Information System (INIS)

    Ahmed, A.Z.; Abdelrazek, I.D.

    1999-01-01

    An advanced controlled air incinerator has been investigated, developed and put into successful operation for both non radioactive simulated and other combustible solid wastes. Engineering efforts concentrated on providing an incinerator which emitted a clean, easily treatable off-gas and which produced minimum amounts of secondary waste. Feed material is fed by gravity into the gas reactor without shredding or other pretreatment. The temperature of the waste is gradually increased in a reduced oxygen atmosphere as the resulting products are introduced into the combustion chamber. Steady burning is thus accomplished under easily controlled excess air conditions with the off-gas then passing through a simple dry cleaning-up system. Experimental studies showed that, at lower temperature, CO 2 , and CH 4 contents in gas reactor effluent increase by the increase of glowing bed temperature, while H 2 O, H 2 and CO decrease . It was proved that, a burn-out efficiency (for ash residues) and a volume reduction factor appeared to be better than 95.5% and 98% respectively. Moreover, high temperature permits increased volumes of incinerated material and results in increased gasification products. It was also found that 8% by weight of ashes are separated by flue gas cleaning system as it has chemical and size uniformity. This high incineration efficiency has been obtained through automated control and optimization of process variables like temperature of the glowing bed and the oxygen feed rate to the gas reactor

  17. Radioactive Waste...The Problem and Some Possible Solutions

    Science.gov (United States)

    Olivier, Jean-Pierre

    1977-01-01

    Nuclear safety is a highly technical and controversial subject that has caused much heated debate and political concern. This article examines the problems involved in managing radioactive wastes and the techniques now used. Potential solutions are suggested and the need for international cooperation is stressed. (Author/MA)

  18. Food waste in Central Europe – challenges and solutions

    Directory of Open Access Journals (Sweden)

    den Boer Jan

    2017-01-01

    Full Text Available Food waste is an important issue in the global economy. In the EU many activities aimed at this topic are carried out, however in Central Europe is still quite pristine. There is lack of reliable data on food waste quantities in this region, and not many preventive actions are taken. To improve this situation the STREFOWA (Strategies to Reduce and Manage Food Waste in Central Europe was initiated. It is an international project (Austria, Czech Republic, Hungary, Italy, Poland, founded by the Interreg Central Europe programme, running from July 2016 to June 2019. Its main purpose is to provide solutions to prevent and manage food waste throughout the entire food supply chain. The results of STREFOWA will have positive economical, social and environmental impacts.

  19. Food waste in Central Europe - challenges and solutions

    Science.gov (United States)

    den Boer, Jan; Kobel, Przemysław; Dyjakon, Arkadiusz; Urbańska, Klaudia; Obersteiner, Gudrun; Hrad, Marlies; Schmied, Elisabeth; den Boer, Emilia

    2017-11-01

    Food waste is an important issue in the global economy. In the EU many activities aimed at this topic are carried out, however in Central Europe is still quite pristine. There is lack of reliable data on food waste quantities in this region, and not many preventive actions are taken. To improve this situation the STREFOWA (Strategies to Reduce and Manage Food Waste in Central Europe) was initiated. It is an international project (Austria, Czech Republic, Hungary, Italy, Poland), founded by the Interreg Central Europe programme, running from July 2016 to June 2019. Its main purpose is to provide solutions to prevent and manage food waste throughout the entire food supply chain. The results of STREFOWA will have positive economical, social and environmental impacts.

  20. Corrosion of a carbon steel in simulated liquid nuclear wastes

    International Nuclear Information System (INIS)

    Saenz Gonzalez, Eduardo

    2005-01-01

    This work is part of a collaboration agreement between CNEA (National Atomic Energy Commission of Argentina) and USDOE (Department of Energy of the United States of America), entitled 'Tank Corrosion Chemistry Cooperation', to study the corrosion behavior of carbon steel A537 class 1 in different simulated non-radioactive wastes in order to establish the safety concentration limits of the tank waste chemistry at Hanford site (Richland-US). Liquid high level nuclear wastes are stored in tanks made of carbon steel A537 (ASTM nomenclature) that were designed for a service life of 20 to 50 years. A thickness reduction of some tank walls, due to corrosion processes, was detected at Hanford site, beyond the existing predicted values. Two year long-term immersion tests were started using non radioactive simulated liquid nuclear waste solutions at 40 C degrees. This work extends throughout the first year of immersion. The simulated solutions consist basically in combinations of the 10 most corrosion significant chemical components: 5 main components (NaNO 3 , NaCl, NaF, NaNO 2 and NaOH) at three concentration levels and 5 secondary components at two concentration levels. Measurements of the general corrosion rate with time were performed for carbon steel coupons, both immersed in the solutions and in the vapor phases, using weight loss and electrochemistry impedance spectroscopy techniques. Optic and scanning electron microscopy examination, analysis of U-bend samples and corrosion potential measurements, were also done. Localized corrosion susceptibility (pitting and crevice corrosion) was assessed in isolated short-term tests by means of cyclic potentiodynamic polarization curves. The effect of the simulated waste composition on the corrosion behavior of A537 steel was studied based on statistical analyses. The Surface Response Model could be successfully applied to the statistical analysis of the A537 steel corrosion in the studied solutions. General corrosion was not

  1. 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)

  2. Removal of fluoride ions from aqueous solution by waste mud

    International Nuclear Information System (INIS)

    Kemer, Baris; Ozdes, Duygu; Gundogdu, Ali; Bulut, Volkan N.; Duran, Celal; Soylak, Mustafa

    2009-01-01

    The present study was carried out to assess the ability of original waste mud (o-WM) and different types of activated waste mud which are acid-activated (a-WM) and precipitated waste mud (p-WM), in order to remove excess of fluoride from aqueous solution by using batch technique. The p-WM exhibited greater performance than the others. Adsorption studies were conducted as a function of pH, contact time, initial fluoride concentration, adsorbent concentration, temperature, etc. Studies were also performed to understand the effect of some co-existing ions present in aqueous solutions. Adsorption process was found to be almost independent of pH for all types of waste mud. Among the kinetic models tested for p-WM, pseudo-second-order model fitted the kinetic data well with a perfect correlation coefficient value of 1.00. It was found that the adequate time for the adsorption equilibrium of fluoride was only 1 h. Thermodynamic parameters including the Gibbs free energy (ΔG o ), enthalpy (ΔH o ), and entropy (ΔS o ) revealed that adsorption of fluoride ions on the p-WM was feasible, spontaneous and endothermic in the temperature range of 0-40 deg. C. Experimental data showed a good fit with the Langmuir and Freundlich adsorption isotherm models. Results of this study demonstrated the effectiveness and feasibility of WM for removal of fluoride ions from aqueous solution.

  3. Removal of fluoride ions from aqueous solution by waste mud

    Energy Technology Data Exchange (ETDEWEB)

    Kemer, Baris; Ozdes, Duygu; Gundogdu, Ali; Bulut, Volkan N.; Duran, Celal [Karadeniz Technical University, Faculty of Arts and Sciences, Department of Chemistry, 61080 Trabzon (Turkey); Soylak, Mustafa, E-mail: soylak@erciyes.edu.tr [Erciyes University, Faculty of Arts and Sciences, Department of Chemistry, 38039 Kayseri (Turkey)

    2009-09-15

    The present study was carried out to assess the ability of original waste mud (o-WM) and different types of activated waste mud which are acid-activated (a-WM) and precipitated waste mud (p-WM), in order to remove excess of fluoride from aqueous solution by using batch technique. The p-WM exhibited greater performance than the others. Adsorption studies were conducted as a function of pH, contact time, initial fluoride concentration, adsorbent concentration, temperature, etc. Studies were also performed to understand the effect of some co-existing ions present in aqueous solutions. Adsorption process was found to be almost independent of pH for all types of waste mud. Among the kinetic models tested for p-WM, pseudo-second-order model fitted the kinetic data well with a perfect correlation coefficient value of 1.00. It was found that the adequate time for the adsorption equilibrium of fluoride was only 1 h. Thermodynamic parameters including the Gibbs free energy ({Delta}G{sup o}), enthalpy ({Delta}H{sup o}), and entropy ({Delta}S{sup o}) revealed that adsorption of fluoride ions on the p-WM was feasible, spontaneous and endothermic in the temperature range of 0-40 deg. C. Experimental data showed a good fit with the Langmuir and Freundlich adsorption isotherm models. Results of this study demonstrated the effectiveness and feasibility of WM for removal of fluoride ions from aqueous solution.

  4. The waste management at research laboratories - problems and solutions

    International Nuclear Information System (INIS)

    Dellamano, Jose Claudio; Vicente, Roberto

    2011-01-01

    The radioactive management in radioactive installations must be planned and controlled. However, in the case of research laboratories, that management is compromised due to the common use of materials and installations, the lack of trained personnel and the nonexistence of clear and objective orientations by the regulator organism. Such failures cause an increasing of generated radioactive wastes and the imprecision or nonexistence of record of radioactive substances, occasioning a financial wastage, and the cancelling of licences for use of radioactive substances. This paper discusses and proposes solutions for the problems found at radioactive waste management in research laboratories

  5. Mixed waste: An alternative solution. The utility perspective

    International Nuclear Information System (INIS)

    Seizert, R.D.

    1988-01-01

    The issue of mixed waste is one of significant interest to the utility industry. The interest is focused on the current regulatory scheme of dual regulation. A fundamental concern of the commercial nuclear utilities resulting from dual regulation is that there are currently no facilities in the US to dispose of mixed low-level radioactive and hazardous waste. The lack of available sites renders mixed waste an orphan, requiring generators of such material to store the waste on-site. This in turn causes commercial nuclear power plants to be subjected to the full gamut of Environmental Protection Agency (EPA) Resource Conservation and Recovery Act (RCRA) regulation in addition to the existing Nuclear Regulatory Commission (NRC) regulations. Superimposing dual regulatory schemes will have impacts which extend far beyond the mere management of mixed waste. Certainly the burdens, complexities and costs of complying with the overlapping regulatory schemes will not have a commensurate increase in protection from the real risks being addressed. For these reasons, the commercial nuclear utility industry is working toward an alternative solution which will protect the public health and the environment from all hazards of mixed waste and will minimize the impacts on both the regulators and the regulated community

  6. Ferrocyanide Safety Project: Comparison of actual and simulated ferrocyanide waste properties

    International Nuclear Information System (INIS)

    Scheele, R.D.; Burger, L.L.; Sell, R.L.; Bredt, P.R.; Barrington, R.J.

    1994-09-01

    In the 1950s, additional high-level radioactive waste storage capacity was needed to accommodate the wastes that would result from the production of recovery of additional nuclear defense materials. To provide this additional waste storage capacity, the Hanford Site operating contractor developed a process to decontaminate aqueous wastes by precipitating radiocesium as an alkali nickel ferrocyanide; this process allowed disposal of the aqueous waste. The radiocesium scavenging process as developed was used to decontaminate (1) first-cycle bismuth phosphate (BiPO 4 ) wastes, (2) acidic wastes resulting from uranium recovery operations, and (3) the supernate from neutralized uranium recovery wastes. The radiocesium scavenging process was often coupled with other scavenging processes to remove radiostrontium and radiocobalt. Because all defense materials recovery processes used nitric acid solutions, all of the wastes contained nitrate, which is a strong oxidizer. The variety of wastes treated, and the occasional coupling of radiostrontium and radiocobalt scavenging processes with the radiocesium scavenging process, resulted in ferrocyanide-bearing wastes having many different compositions. In this report, we compare selected physical, chemical, and radiochemical properties measured for Tanks C-109 and C-112 wastes and selected physical and chemical properties of simulated ferrocyanide wastes to assess the representativeness of stimulants prepared by WHC

  7. 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

  8. Analysis of mercury in simulated nuclear waste

    International Nuclear Information System (INIS)

    Policke, T.A.; Johnson, L.C.; Best, D.R.

    1991-01-01

    Mercury, Hg, is a non-radioactive component in the High Level Waste at the Savannah River Site (SRS). Thus, it is a component of the Defense Waste Processing Facility's (DWPF) process streams. It is present because mercuric nitrate (Hg(NO 3 ) 2 ) is used to dissolve spent fuel rods. Since mercury halides are extremely corrosive, especially at elevated temperatures such as those seen in a melter (1150 degrees C), its concentration throughout the process needs to be monitored so that it is at an acceptable level prior to reaching the melter off-gas system. The Hg can be found in condensates and sludge feeds and throughout the process and process lines, i.e., at any sampling point. The different samples types that require Hg determinations in the process streams are: (1) sludges, which may be basic or acidic and may or may not include aromatic organics, (2) slurries, which are sludges with frit and will always contain organics (formate and aromatics), and (3) condensates, from feed prep and melter off-gas locations. The condensates are aqueous and the mercury may exist as a complex mixture of halides, oxides, and metal, with levels between 10 and 100 ppm. The mercury in the sludges and slurries can be Hg 0 , Hg +1 , or Hg +2 , with levels between 200 and 3000 ppm, depending upon the location, both time and position, of sampling. For DWPF, both total and soluble Hg concentrations need to be determined. The text below describes how these determinations are being made by the Defense Waste Processing Technology (DWPT) Analytical Laboratory at the Savannah River Site. Both flame atomic absorption (FAA) and cold vapor atomic (CVAA) measurements are discussed. Also, the problems encountered in the steps toward measuring HG in these samples types of condensates and sludges are discussed along with their solutions

  9. Hanford tank waste operation simulator operational waste volume projection verification and validation procedure

    International Nuclear Information System (INIS)

    HARMSEN, R.W.

    1999-01-01

    The Hanford Tank Waste Operation Simulator is tested to determine if it can replace the FORTRAN-based Operational Waste Volume Projection computer simulation that has traditionally served to project double-shell tank utilization. Three Test Cases are used to compare the results of the two simulators; one incorporates the cleanup schedule of the Tri Party Agreement

  10. Solution exchange corrosion testing with the glass-zeolite ceramic waste form in demineralized water at 900C

    International Nuclear Information System (INIS)

    Simpson, L. J.

    1998-01-01

    A ceramic waste form of glass-bonded zeolite is being developed for the long-term disposition of fission products and transuranic elements in wastes from the U.S. Department of Energy's spent nuclear fuel conditioning activities. Solution exchange corrosion tests were performed on the ceramic waste form and its potential base constituents of glass, zeolite 5A, and sodalite as part of an effort to qualify the ceramic waste form for acceptance into the Civilian Radioactive Waste Management System. Solution exchange tests were performed at 90 C by replacing 80 to 90% of the leachate with fresh demineralized water after set time intervals. The results from these tests provide information about corrosion mechanisms and the ability of the ceramic waste form and its constituent materials to retain waste components. The results from solution exchange tests indicate that radionuclides will be preferentially retained in the zeolites without the glass matrix and in the ceramic waste form, with respect to cations like Li, K, and Na. Release results have been compared for simulated waste from candidate ceramic waste forms with zeolite 5A and its constituent materials to determine the corrosion behavior of each component

  11. Women, e-waste, and technological solutions to climate change.

    Science.gov (United States)

    McAllister, Lucy; Magee, Amanda; Hale, Benjamin

    2014-06-14

    In this paper, we argue that a crossover class of climate change solutions (which we term "technological solutions") may disproportionately and adversely impact some populations over others. We begin by situating our discussion in the wider climate discourse, particularly with regard to the Millennium Development Goals (MDGs) and the Basel Convention. We then suggest that many of the most attractive technological solutions to climate change, such as solar energy and electric car batteries, will likely add to the rapidly growing stream of electronic waste ("e-waste"). This e-waste may have negative downstream effects on otherwise disenfranchised populations. We argue that e-waste burdens women unfairly and disproportionately, affecting their mortality/morbidity and fertility, as well as the development of their children. Building on this, we claim that these injustices are more accurately captured as problems of recognition rather than distribution, since women are often institutionally under-acknowledged both in the workplace and in the home. Without institutional support and representation, women and children are deprived of adequate safety equipment, health precautions, and health insurance. Finally, we return to the question of climate justice in the context of the human right to health and argue for greater inclusion and recognition of women waste workers and other disenfranchised groups in forging future climate agreements. Copyright © 2014 McAllister, Magee. This is an open access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/by-nc/3.0/), which permits unrestricted non-commercial use, distribution, and reproduction in any medium, provided the original author and source are credited.

  12. An eco friendly solution to the food waste disposal

    Science.gov (United States)

    Babu, G. Reddy; Kumar, G. Madhav

    2017-07-01

    In recent years, waste disposal at workmen camp is one of the major problems being faced by many nations across the world. In the workmen colony at Chittapur, a series of kitchens were built for cooking purpose and a number of small canteens are also functioning. Considerable quantity of food waste is collected daily from these eateries and disposed at a faraway place. Food waste is highly degradable in nature, if not disposed properly it causes problems related to environmental pollution. Hence, it is very important to identify an environment friendly process rather than opt for land filling or any disposal method. We worked together to find a suitable eco-friendly solution for the food waste disposal at Chittapur site and suggested that biogas production through anaerobic digestion is a solution for the disposal and utilization of food waste for better purpose. This resulted in setting up a 500 kg per day food waste treatment biogas plant at Chittapur. This establishment is the first time in the construction industry at workmen camp in India. Anaerobic Digestion has been recognized as one of the best options that is available for treating food waste, as it generates two valuable end products, biogas and compost. Biogas is a mixture of CH4 and CO2 about (55:45). Biogas generated can be used for thermal applications such as cooking or for generating electricity. The digested slurry is a well stabilized organic manure and can be used as soil fertilizer. Plant design is to handle 500 kg of food waste /day. 27 kg LPG is obtained from 500kg of kitchen waste. The Value of 27 kg of LPG is Rs.2700/day. Daily 1000 litres of digested effluent was obtained. It is good organic manure with plant micro nutrients and macro nutrients. This can be used for growing plants and in agriculture. The value of manure per day is Rs.250/-. The annual revenue is Rs.10.62 lakhs and the annual expenditure is 1.8 lakhs. The net benefit is 8.82 lakhs. Payback period is 2.1 years. This process

  13. 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.

  14. 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.

  15. Distillation as a pretreatment process of waste scintillation solutions

    International Nuclear Information System (INIS)

    Dellamano, J.C.

    1988-05-01

    A process to pretreat scintillation solutions composed basically of PPO, POPOP, TOLUENE and ANTAROX, utilized by radioimmunoassay laboratories, is described. The technique employed is distillation which permits a waste reduction to about 40% of the initial volume with the recovery of the solvent (toluene). The recovered toluene can be resued for the same purpose, since it is free of radioactive material as assured by quality control procedures. (author) [pt

  16. 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

  17. Mixing of Process Heels, Process Solutions and Recycle Streams: Small-Scale Simulant

    International Nuclear Information System (INIS)

    Kaplan, D.I.

    2001-01-01

    The overall objective of this small-scale simulant mixing study was to identify the processes within the Hanford Site River Protection Project - Waste Treatment Plant (RPP-WTP) that may generate precipitates and to identify the types of precipitates formed. This information can be used to identify where mixtures of various solutions will cause precipitation of solids, potentially causing operational problems such as fouling equipment or increasing the amount of High Level Waste glass produced. Having this information will help guide protocols for flushing or draining tanks, mixing internal recycle streams, and mixing waste tank supernates. This report contains the discussion and thermodynamic chemical speciation modeling of the raw data

  18. Fuzzy Simulation-Optimization Model for Waste Load Allocation

    Directory of Open Access Journals (Sweden)

    Motahhare Saadatpour

    2006-01-01

    Full Text Available This paper present simulation-optimization models for waste load allocation from multiple point sources which include uncertainty due to vagueness of the parameters and goals. This model employs fuzzy sets with appropriate membership functions to deal with uncertainties due to vagueness. The fuzzy waste load allocation model (FWLAM incorporate QUAL2E as a water quality simulation model and Genetic Algorithm (GA as an optimization tool to find the optimal combination of the fraction removal level to the dischargers and pollution control agency (PCA. Penalty functions are employed to control the violations in the system.  The results demonstrate that the goal of PCA to achieve the best water quality and the goal of the dischargers to use the full assimilative capacity of the river have not been satisfied completely and a compromise solution between these goals is provided. This fuzzy optimization model with genetic algorithm has been used for a hypothetical problem. Results demonstrate a very suitable convergence of proposed optimization algorithm to the global optima.

  19. Simulation of unsaturated flow and nonreactive solute transport in a heterogeneous soil at the field scale

    International Nuclear Information System (INIS)

    Rockhold, M.L.

    1993-02-01

    A field-scale, unsaturated flow and solute transport experiment at the Las Cruces trench site in New Mexico was simulated as part of a ''blind'' modeling exercise to demonstrate the ability or inability of uncalibrated models to predict unsaturated flow and solute transport in spatially variable porous media. Simulations were conducted using a recently developed multiphase flow and transport simulator. Uniform and heterogeneous soil models were tested, and data from a previous experiment at the site were used with an inverse procedure to estimate water retention parameters. A spatial moment analysis was used to provide a quantitative basis for comparing the mean observed and simulated flow and transport behavior. The results of this study suggest that defensible predictions of waste migration and fate at low-level waste sites will ultimately require site-specific data for model calibration

  20. Decontamination of waste radioactive polluted solutions in radiation treatment

    International Nuclear Information System (INIS)

    Simova, G.; Boyadzhiev, A.; Mikhajlov, M.G.; Shopov, N.

    1979-01-01

    The decontamination capacity of solutions of the trivial cleaning Bulgarian preparations ''Mipro'', ''Sana'', ''Synthek'' and ''Univer'' for different surfaces (steel, glass, PVC and linoleum) contaminated with cesium-134, strontium-85 or cerium-144 chlorides, was studied. Concentrations from 5 to 15 g/l of the solutions used in this study displayed a degree of cleaning over 90%. Higher concentration of the solution does not improve its cleaning capacity. For evaluation of foam formation by the solutions, the so called ''foam column stability coefficient'' has been adopted. This coefficient represents the ratio between the height of the foam column and the time of its half life, referred to the time for the foam column formation when blown through with a constant air current. On the basis of this index, solutions of the preparation ''Mipro'' proved to be the best ones for decontamination - in the whole investigated concentration span, the foam column stability coefficient for the solutions of this preparation is with two orders lower than the respective coefficient of the other preparations. It was experimentally established that radiation treatment of radio-contaminated solutions reduces the foam column stability coefficient. Radiation treatment should be carried out in a gamma field, realizing at least one megarad within an acceptable for the liquid wastes time period. (A.B.)

  1. Leaching behavior of a simulated bituminized radioactive waste form under deep geological conditions

    International Nuclear Information System (INIS)

    Nakayama, Shinichi; Iida, Yoshihisa; Nagano, Tetsushi; Akimoto, Toshiyuki

    2003-01-01

    The leaching behavior of a simulated bituminized waste form was studied to acquire data for the performance assessment of the geologic disposal of bituminized radioactive waste. Laboratory-scale leaching tests were performed for radioactive and non-radioactive waste specimens simulating bituminized waste of a French reprocessing company, COGEMA. The simulated waste was contacted with deionized water, an alkaline solution (0.03-mol/l KOH), and a saline solution (0.5-mol/l KCl) under atmospheric and anoxic conditions. The concentrations of Na, Ba, Cs, Sr, Np, Pu, NO 3 , SO 4 and I in the leachates were determined. Swelling of the bituminized waste progressed in deionized water and KOH. The release of the soluble components, Na and Cs, was enhanced by the swelling, and considered to be diffusion-controlled in the swelled layers of the specimens. The release of sparingly soluble components such as Ba and Np was solubility-limited in addition to the progression of leaching. Neptunium, a redox-sensitive element, showed a distinct difference in release between anoxic and atmospheric conditions. The elemental release from the bituminized waste specimens leached in the KCl was very low, which is likely due to the suppression of swelling of the specimens at high ionic strength. (author)

  2. Geopolymerisation of fly ashes with waste aluminium anodising etching solutions.

    Science.gov (United States)

    Ogundiran, M B; Nugteren, H W; Witkamp, G J

    2016-10-01

    Combined management of coal combustion fly ash and waste aluminium anodising etching solutions using geopolymerisation presents economic and environmental benefits. The possibility of using waste aluminium anodising etching solution (AES) as activator to produce fly ash geopolymers in place of the commonly used silicate solutions was explored in this study. Geopolymerisation capacities of five European fly ashes with AES and the leaching of elements from their corresponding geopolymers were studied. Conventional commercial potassium silicate activator-based geopolymers were used as a reference. The geopolymers produced were subjected to physical, mechanical and leaching tests. The leaching of elements was tested on 28 days cured and crushed geopolymers using NEN 12457-4, NEN 7375, SPLP and TCLP leaching tests. After 28 days ambient curing, the geopolymers based on the etching solution activator showed compressive strength values between 51 and 84 MPa, whereas the commercial potassium silicate based geopolymers gave compressive strength values between 89 and 115 MPa. Based on the regulatory limits currently associated with the used leaching tests, all except one of the produced geopolymers (with above threshold leaching of As and Se) passed the recommended limits. The AES-geopolymer geopolymers demonstrated excellent compressive strength, although less than geopolymers made from commercial activator. Additionally, they demonstrated low element leaching potentials and therefore can be suitable for use in construction works. Copyright © 2016. Published by Elsevier Ltd.

  3. Leaching characteristics of actinides from simulated reactor waste glass

    International Nuclear Information System (INIS)

    Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W.; Schweiger, J.S.

    1979-01-01

    Two methods for measuring the leach rates of simulated high level waste glass are compared. One is a modification of the standard IAEA method and the other is a one-pass method in which fresh leachant solution is pumped over the sample at a controlled flow rate and temperature. For times up to 3 days, there is close agreement between results from the two methods at 25.0 0 C. Leach rates from the one-pass method show a correlation with flow rate only on day 1 at 25.0 0 C, whereas they show a correlation with flow rate for all three days at 75.0 0 C. 237 Np rates at 75.0 0 C are greater than those at 25.0 0 C, but 239 Pu rates at 75.0 0 C are less than or equal to those at 25.0 0 C

  4. The waste management program VUB-AZ: An integrated solution for nuclear biomedical waste management

    International Nuclear Information System (INIS)

    Covens, P.; Sonck, M.; Eggermont, G.; Meert, D.

    2001-01-01

    unit will be compared with the MDA obtained by different handheld monitors. All results will be finally correlated to the different proposed clearance levels. These clearance levels can easily be met through on-site storage for radionuclides with half-life less than 1 year. For a waste stream of 1000 packages or more a year, a management software is indispensable. The software 'WasteMan' was developed on-site. This user-friendly software takes care of the entire storage procedure and allows a complete bookkeeping of the daily nuclear waste streams. Based on the sophisticated waste collection procedure, the WasteMan software allows both a complete inventory of the storage facility and a full traceability of all waste packages from production to either clearance or disposal. At the same time all necessary documents for either clearance or disposal are generated automatically. The data-exchange between several interfaces enables timesaving administration. In addition to these technical aspects a general analysis of the economic impact of such an on- site decay program will be made for a medium sized university with hospital, yielding a serious reduction of waste handling costs. This waste storage program, including the complete measurement set-up and the necessary management software, was recently installed in a second university, proving the general applicability of the whole concept for biomedical nuclear waste. Many hospitals and other biomedical centres however produce small quantities of nuclear waste for which investments, like measurement equipment and decay rooms, are not cost-effective. The installation of a regional centre for nuclear biomedical waste will be presented here as an alternative solution for this problem

  5. Retrieval process development and enhancements waste simulant compositions and defensibility

    International Nuclear Information System (INIS)

    Powell, M.R.; Golcar, G.R.; Geeting, J.G.H.

    1997-09-01

    The purpose of this report is to document the physical waste simulant development efforts of the EM-50 Tanks Focus Area at the Hanford Site. Waste simulants are used in the testing and development of waste treatment and handling processes because performing such tests using actual tank waste is hazardous and prohibitively expensive. This document addresses the simulant development work that supports the testing of waste retrieval processes using simulants that mimic certain key physical properties of the tank waste. Development and testing of chemical simulants are described elsewhere. This work was funded through the EM-50 Tanks Focus Area as part of the Retrieval Process Development and Enhancements (RPD ampersand E) Project at the Pacific Northwest National Laboratory (PNNL). The mission of RPD ampersand E is to understand retrieval processes, including emerging and existing processes, gather performance data on those processes, and relate the data to specific tank problems to provide end users with the requisite technical bases to make retrieval and closure decisions. Physical simulants are prepared using relatively nonhazardous and inexpensive materials rather than the chemicals known to be in tank waste. Consequently, only some of the waste properties are matched by the simulant. Deciding which properties need to be matched and which do not requires a detailed knowledge of the physics of the process to be tested using the simulant. Developing this knowledge requires reviews of available literature, consultation with experts, and parametric tests. Once the relevant properties are identified, waste characterization data are reviewed to establish the target ranges for each property. Simulants are then developed that possess the desired ranges of properties

  6. Low-level tank waste simulant data base

    International Nuclear Information System (INIS)

    Lokken, R.O.

    1996-04-01

    The majority of defense wastes generated from reprocessing spent N- Reactor fuel at Hanford are stored in underground Double-shell Tanks (DST) and in older Single-Shell Tanks (SST) in the form of liquids, slurries, sludges, and salt cakes. The tank waste remediation System (TWRS) Program has the responsibility of safely managing and immobilizing these tank wastes for disposal. This report discusses three principle topics: the need for and basis for selecting target or reference LLW simulants, tanks waste analyses and simulants that have been defined, developed, and used for the GDP and activities in support of preparing and characterizing simulants for the current LLW vitrification project. The procedures and the data that were generated to characterized the LLW vitrification simulants were reported and are presented in this report. The final section of this report addresses the applicability of the data to the current program and presents recommendations for additional data needs including characterization and simulant compositional variability studies

  7. Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-01

    stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfates in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to demonstrate evaporation of a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations, in order to predict the composition of the effluents from the EMF evaporator to aid in planning for their disposition. This document describes the results of that test using the core simulant. This simulant formulation is designated as the “core simulant”; other additives will be included for specific testing, such as volatiles for evaporation or hazardous metals for measuring leaching properties of waste forms. The results indicate that the simulant can easily be concentrated via evaporation. During that the pH adjustment step in simulant preparation, ammonium is quickly converted to ammonia, and most of the ammonia was stripped from the simulated waste and partitioned to the condensate. Additionally, it was found that after concentrating (>12x) and cooling that a small amount of LiF and Na3(SO4)F precipitate out of solution. With the exception of ammonia, analysis of the condensate indicated very low to below detectable levels of many of the constituents in the simulant, yielding very high decontamination factors (DF).

  8. Measurement of Solute Diffusion Behavior in Fractured Waste Glass Media

    International Nuclear Information System (INIS)

    Saripalli, Kanaka P.; Lindberg, Michael J.; Meyer, Philip D.

    2008-01-01

    Determination of aqueous phase diffusion coefficients of solutes through fractured media is essential for understanding and modeling contaminants transport at many hazardous waste disposal sites. No methods for earlier measurements are available for the characterization of diffusion in fractured glass blocks. We report here the use of time-lag diffusion experimental method to assess the diffusion behavior of three different solutes (Cs, Sr and Pentafluoro Benzoic Acid or PFBA) in fractured, immobilized low activity waste (ILAW) glass forms. A fractured media time-lag diffusion experimental apparatus that allows the measurement of diffusion coefficients has been designed and built for this purpose. Use of time-lag diffusion method, a considerably easier experimental method than the other available methods, was not previously demonstrated for measuring diffusion in any fractured media. Hydraulic conductivity, porosity and diffusion coefficients of a solute were experimentally measured in fractured glass blocks using this method for the first time. Results agree with the range of properties reported for similar rock media earlier, indicating that the time-lag experimental method can effectively characterize the diffusion coefficients of fractured ILAW glass media

  9. Inhibition of nuclear waste solutions containing multiple aggressive anions

    International Nuclear Information System (INIS)

    Congdon, J.W.

    1987-01-01

    The inhibition of localized corrosion of carbon steel in caustic, high-level radioactive waste solutions was studied using cyclic potentiodynamic polarization scans, supplemented by partially immersed coupon tests. The electrochemical tests provided a rapid and accurate means of determining the relationship between the minimum inhibitor requirements and the concentration of the aggressive anions in this system. Nitrate, sulfate, chloride, and fluoride were identified as aggressive anions, however, no synergistic effects were observed between these anions. This observation may have important theoretical implications because it tends to contradict the behavior of aggressive anions as predicted by existing theories for localized corrosion. 10 refs., 5 figs., 2 tabs

  10. Cs separation from nitric acid solutions of radioactive waste

    International Nuclear Information System (INIS)

    Heckmann, K.; Pieronczyk, W.; Strnad, J.; Feldmaier, F.

    1989-01-01

    It was the objective of this study to selectively separate active caesium (Cs-134 and Cs-137) from acid radioactive waste solutions (especially MAW and HAWC). The following 'strategy' was designed for a separation process: synthesis of reagents which are acid-resistant and selective for caesium; precipitation of Cs + and separation of the precipitates by filtration or centrifugation or precipitation of Cs + and separation of the precipitates by flotation; caesium separation by liquid-liquid extraction. As precipitating agents, sodium tetraphenylborate (kalignost) and several of its fluorine derivatives were examined. (orig./RB) [de

  11. 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

  12. Photometric estimation of plutonium in product solutions and acid waste solutions using flow injection analysis technique

    International Nuclear Information System (INIS)

    Dhas, A.J.A.; Dharmapurikar, G.R.; Kumaraguru, K.; Vijayan, K.; Kapoor, S.C.; Ramanujam, A.

    1995-01-01

    Flow injection analysis technique is employed for the measurement of plutonium concentrations in product nitrate solutions by measuring the absorbance of Pu(III) at 565 nm and of Pu(IV) at 470 nm, using a Metrohm 662 photometer, with a pyrex glass tube of 2 nm (ID) inserted in the light path of the detector serving as a flow cell. The photometer detector never comes in contact with radioactive solution. In the case of acid waste solutions Pu is first purified by extraction chromatography with 2-ethyl hexyl hydrogen 2 ethyl hexyl phosphonate (KSM 17)- chromosorb and the Pu in the eluate in complexed with Arsenazo III followed by the measured of absorbance at 665 nm. Absorbance of reference solutions in the desired concentration ranges are measured to calibrate the system. The results obtained agree with the reference values within ±2.0%. (author). 3 refs., 1 tab

  13. Community Solutions to Solid Waste Pollution. Operation Waste Watch: The New Three Rs for Elementary School. Grade 6. [Second Edition.

    Science.gov (United States)

    Virginia State Dept. of Waste Management, Richmond. Div. of Litter & Recycling.

    This publication, the last in a series of seven for elementary schools, is an environmental education curriculum guide with a focus on waste management issues. It contains a unit of exercises selected for sixth grade students focusing on community solutions to solid waste pollution. Waste management activities included in this unit seek to…

  14. Induction melting of simulated transuranic waste

    International Nuclear Information System (INIS)

    Tenaglia, R.D.; McCall, J.L.

    1983-06-01

    Coreless induction melting was investigated as a method to melt and consolidate waste material representative of the transuranic waste (TRU) stored at the Idaho National Engineering Laboratory (INEL). Waste material was introduced onto the surface of a molten cast iron bath in a coreless induction furnace. Waste metallics were incorporated into the bath. Noncombustibles formed a slag which was poured or skimmed from the bath surface. Stack sampling was performed to characterize the off-gas and particulate matter evolved. Experimental melting tests were performed for a variety of types of wastes including metallics, chemical sludge, soil, concrete, and glass. Each test also included a representative level of combustible materials consisting of paper, wood, cloth, polyvinyl chloride and polyethylene. Metallic wastes were readily processed by induction melting with a minimum of slag production. Test waste consisting primarily of chemical sludge provided fluid slags which could be poured from the bath surface. Processing of wastes consisting of soil, concrete, or glass was limited by the inability to achieve fluid slags. It appears from test results that coreless induction melting is a feasible method to process INEL-type waste materials if two problems can be resolved. First, slag fluidity must be improved to facilitate the collection of slags formed from soil, concrete, or glass containing wastes. Secondly, refractory life must be further optimized to permit prolonged processing of the waste materials. The use of a chrome-bearing high-alumina refractory was found to resist slag line attach much better than a magnesia refractory, although some attack was still noted

  15. 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.

  16. Using simulation to assess the opportunities of dynamic waste collection

    NARCIS (Netherlands)

    Mes, Martijn R.K.; Bangsow, S.

    2012-01-01

    In this chapter, we illustrate the use of discrete event simulation to evaluate how dynamic planning methodologies can be best applied for the collection of waste from underground containers. We present a case study that took place at the waste collection company Twente Milieu, located in The

  17. Using Simulation to Assess the Opportunities of Dynamic Waste Collection

    NARCIS (Netherlands)

    Mes, Martijn R.K.

    In this paper, we illustrate the use of discrete event simulation to evaluate how dynamic planning methodologies can be best applied for the collection of waste from underground containers. We present a case study that took place at the waste collection company Twente Milieu, located in The

  18. Low-Level Radioactive Waste siting simulation information package

    International Nuclear Information System (INIS)

    1985-12-01

    The Department of Energy's National Low-Level Radioactive Waste Management Program has developed a simulation exercise designed to facilitate the process of siting and licensing disposal facilities for low-level radioactive waste. The siting simulation can be conducted at a workshop or conference, can involve 14-70 participants (or more), and requires approximately eight hours to complete. The exercise is available for use by states, regional compacts, or other organizations for use as part of the planning process for low-level waste disposal facilities. This information package describes the development, content, and use of the Low-Level Radioactive Waste Siting Simulation. Information is provided on how to organize a workshop for conducting the simulation. 1 ref., 1 fig

  19. Radiolytic gas formation in high-level liquid waste solutions

    International Nuclear Information System (INIS)

    Brodda, B.-G.; Dix, Siegfried; Merz, E.R.

    1989-01-01

    High-level fission product waste solutions originating from the first-cycle raffinate stream of spent fast breeder reactor fuel reprocessing have been investigated gas chromatographically for their radiolytic and chemical gas production. The solutions showed considerable formation of hydrogen, carbon dioxide and dinitrogen oxide, whereas atmospheric oxygen was consumed completely within a short time. In particular, carbon dioxide resulted from the radiolytic degradation of entrained organic solvent. After nearly complete degradation of the organic solvent, the influence of hydrazine and nitrogen dioxide on hydrogen formation was investigated. Hydrazinium hydroxide led to the formation of dinitrogen oxide and nitrogen. After 60 d, the concentration of dinitrogen oxide had reduced to zero, whereas the amount of nitrogen formed had reached a maximum. This may be explained by simultaneous chemical and radiolytic reactions leading to the formation of dinitrogen oxide and nitrogen and photolytic fission of dinitrogen oxide. Addition of sodium nitrite resulted in the rapid formation of dinitrogen oxide. The rate of hydrogen production was not changed significantly after the addition of hydrazine or nitrite. The results indicate that under normal operating conditions no dangerous hydrogen radiolysis yields should develop in the course of reprocessing and high-level liquid waste tank storage. Organic entrainment may lead to enhanced radiolytic decomposition and thus to considerable hydrogen production rates and pressure build-up in closed systems. (author)

  20. Corrosion of dissimilar metal crevices in simulated concentrated ground water solutions at elevated temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, B.M.; Quinn, M.J

    2003-01-01

    The disposal of high-level nuclear waste in the Yucca Mountain, Nevada is under consideration by the US Department of Energy. The proposed facility will be located in the unsaturated zone approximately 300 m below the surface and 300 m above the water table. The proposed waste container consists of an outer corrosion-resistant Alloy 22 shell surrounding a 316 NG stainless steel structural inner container that encapsulates the used nuclear fuel waste. A titanium drip shield is proposed to protect the waste container from ground water seepage arid rock-fail. A cycle of dripping/evaporation could result in the generation of concentrated aggressive solutions, which could contact the waste container. The waste container material could be susceptible to crevice corrosion from such solutions. The experiments described in this report support the modeling of waste package degradation processes. The intent was to provide parameter values that are required to model crevice corrosion chemistry, as it relates to hydrogen pick-up, and stress corrosion cracking for selected candidate waste package materials. The purpose of the experiments was to study the crevice corrosion behavior of various candidate materials under near freely corroding conditions and to determine the pH developed in crevice solutions. Experimental results of crevice corrosion of dissimilar metal pairs (Alloy 22, Grade-7 and -16 titanium and 316 stainless steel) immersed in a simulated concentrated ground water at {approx}90{sup o}C are reported. The corrosion potential was measured during exposure periods of between 330 and 630 h. Following the experiments, the pH of the crevice solution was measured. The results indicate that a limited degree of crevice acidification occurred during the experiment. The values for corrosion potential suggest that crevice corrosion may have initiated. The total corrosion was limited, with little visible evidence for crevice corrosion being observed on the sample coupon faces

  1. Corrosion of dissimilar metal crevices in simulated concentrated ground water solutions at elevated temperature

    International Nuclear Information System (INIS)

    Ikeda, B.M.; Quinn, M.J.

    2003-01-01

    The disposal of high-level nuclear waste in the Yucca Mountain, Nevada is under consideration by the US Department of Energy. The proposed facility will be located in the unsaturated zone approximately 300 m below the surface and 300 m above the water table. The proposed waste container consists of an outer corrosion-resistant Alloy 22 shell surrounding a 316 NG stainless steel structural inner container that encapsulates the used nuclear fuel waste. A titanium drip shield is proposed to protect the waste container from ground water seepage arid rock-fail. A cycle of dripping/evaporation could result in the generation of concentrated aggressive solutions, which could contact the waste container. The waste container material could be susceptible to crevice corrosion from such solutions. The experiments described in this report support the modeling of waste package degradation processes. The intent was to provide parameter values that are required to model crevice corrosion chemistry, as it relates to hydrogen pick-up, and stress corrosion cracking for selected candidate waste package materials. The purpose of the experiments was to study the crevice corrosion behavior of various candidate materials under near freely corroding conditions and to determine the pH developed in crevice solutions. Experimental results of crevice corrosion of dissimilar metal pairs (Alloy 22, Grade-7 and -16 titanium and 316 stainless steel) immersed in a simulated concentrated ground water at ∼90 o C are reported. The corrosion potential was measured during exposure periods of between 330 and 630 h. Following the experiments, the pH of the crevice solution was measured. The results indicate that a limited degree of crevice acidification occurred during the experiment. The values for corrosion potential suggest that crevice corrosion may have initiated. The total corrosion was limited, with little visible evidence for crevice corrosion being observed on the sample coupon faces. The

  2. Determination of microamounts of uranium in waste solutions

    International Nuclear Information System (INIS)

    Birringer, K.J.; Netzer, S.; Kuhn, E.; Groll, P.

    1975-07-01

    A method for the determination of microamounts of uranium in presence of high amounts of fission and corrosion products is described. Uranium is separated by reversed-phase chromatography on a small column, packed with Voltalef micro and impregnated with TOPO. For the direct photometric determination uranium is eluted by TAM dissolved in ethanol/pyridine. The efficiency of the separation, using a suitable scrub-solution, was tested with solutions of simulated inactive fission and corrosion products. The reproducibility of the method, with 24 μg of uranium, is +- 2,5%. (orig.) [de

  3. 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

  4. Mixed Waste Treatment Project: Computer simulations of integrated flowsheets

    International Nuclear Information System (INIS)

    Dietsche, L.J.

    1993-12-01

    The disposal of mixed waste, that is waste containing both hazardous and radioactive components, is a challenging waste management problem of particular concern to DOE sites throughout the United States. Traditional technologies used for the destruction of hazardous wastes need to be re-evaluated for their ability to handle mixed wastes, and in some cases new technologies need to be developed. The Mixed Waste Treatment Project (MWTP) was set up by DOE's Waste Operations Program (EM30) to provide guidance on mixed waste treatment options. One of MWTP's charters is to develop flowsheets for prototype integrated mixed waste treatment facilities which can serve as models for sites developing their own treatment strategies. Evaluation of these flowsheets is being facilitated through the use of computer modelling. The objective of the flowsheet simulations is to provide mass and energy balances, product compositions, and equipment sizing (leading to cost) information. The modelled flowsheets need to be easily modified to examine how alternative technologies and varying feed streams effect the overall integrated process. One such commercially available simulation program is ASPEN PLUS. This report contains details of the Aspen Plus program

  5. Leaching of actinides from simulated nuclear waste glass

    International Nuclear Information System (INIS)

    Pickering, S.; Walker, C.T.; Offermann, P.

    1982-01-01

    Two types of simulated nuclear waste glass doped with actinides were leached at 200 0 C in distilled water and salt solutions. Am, Np, Pu and U were all preferentially retained in the surface layer on the glass. Leaching ratios of 0.1 to 0.2 for Np and approx. 0.02 for Am were measured. The losses of Am and Np to the leachant were proportional to the total weight loss of the glass and were larger at 10 ml leachant/cm 2 glass than at 5 ml/cm 2 . Weight loss from the glass occurred only at the start of the experiments for periods ranging from 10 h to 10 days according to leachant composition and volume. Wt losses from the C31-3-EC glass were much greater in saturated NaCl solution than in distilled water. Enrichment in the outer surface layer of Al or Ca according to glass type could be correlated with leachant pH, glass composition and weight loss measurements

  6. Effects of tuff waste package components on release from 76-68 simulated waste glass: Final report

    International Nuclear Information System (INIS)

    McVay, G.L.; Robinson, G.R.

    1984-04-01

    An experimental matrix has been conducted that will allow evaluation of the effects of waste package constituents on the waste form release behavior in a tuff repository environment. Tuff rock and groundwater were used along with 304L, 316, and 1020M ferrous metals to evaluate release from uranium-doped MCC 76-68 simulated waste glass. One of the major findings was that in the absence of 1020M mild steel, tuff rock powder dominates the system. However, when 1020M mild steel is present, it appears to dominate the system. The rock-dominated system results in suppressed glass-water reaction and leaching while the 1020M-dominated system results in enhanced leaching - but the metal effectively scavenges uranium from solution. The 300-series stainless steels play no significant role in affecting glass leaching characteristics. 6 refs., 28 figs., 5 tabs

  7. Chemodynamics of EDTA in a simulated mixed waste: the Hanford Site's complex concentrate waste

    International Nuclear Information System (INIS)

    Toste, A.P.; Ohnuki, Toshihiko

    1999-01-01

    Enormous stockpiles of mixed wastes at the USDOE's Hanford Site, the original US plutonium production facility, await permanent disposal. One mixed waste derived from reprocessing spent fuel was found to contain numerous nuclear related organics including chelating agents like EDTA and complexing agents, which have been used as decontamination agents, etc. Their presence in actual mixed wastes indicates that the organic content of nuclear wastes is dynamic and complicate waste management efforts. The subjects of this report is the chemo-degradation of EDTA degradation in a simulant Hanford's complex concentrate waste. The simulant was prepared by adding EDTA to an inorganic matrix, which was formulated based on past analyses of the actual waste. Aliquots of the EDTA simulant were withdrawn at different time points, derivatized via methylation and analyzed by gas chromatography and Gc/MS to monitor the disappearance of EDTA and the appearance of its' degradation products. This report also compares the results of EDTA's chemo-degradation to the g-radiolysis of EDTA in the simulant, the subject of a recently published article. Finally based on the results of these two studies, an assesment of the potential impact of EDTA degradation on the management of mixed wastes is offered. (J.P.N.)

  8. Determination of uranium distribution in the evaporation of simulated Savannah River Site waste

    International Nuclear Information System (INIS)

    Barnes, M.J.; Chandler, G.T.

    1995-01-01

    The results of an experimental program addressing the distribution of uranium in saltcake and supernate for two Savannah River Site waste compositions are presented. Successive batch evaporations were performed on simulated H-Area Modified Purex low-heat and post-aluminum dissolution wastes spiked with depleted uranium. Waste compositions and physical data were obtained for supernate and saltcake samples. For the H-Area Modified Purex low-heat waste, the product saltcake contained 42% of the total uranium from the original evaporator feed solution. However, precipitated solids only accounted for 10% of the original uranium mass; the interstitial liquid within the saltcake matrix contained the remainder of the uranium. In the case of the simulated post-aluminum dissolution waste; the product saltcake contained 68% of the total uranium from the original evaporator feed solution. Precipitated solids accounted for 52% of the original uranium mass; again, the interstitial liquid within the saltcake matrix contained the remainder of the uranium. An understanding of the distribution of uranium between supernatant liquid, saltcake, and sludge is required to develop a material balance for waste processing operations. This information is necessary to address nuclear criticality safety concerns

  9. Development of Simulants to Support Mixing Tests for High Level Waste and Low Activity Waste

    International Nuclear Information System (INIS)

    EIBLING, RUSSELLE.

    2004-01-01

    The objectives of this study were to develop two different types of simulants to support vendor agitator design studies and mixing studies. The initial simulant development task was to develop rheologically-bounding physical simulants and the final portion was to develop a nominal chemical simulant which is designed to match, as closely as possible, the actual sludge from a tank. The physical simulants to be developed included a lower and upper rheologically bounded: pretreated low activity waste (LAW) physical simulant; LAW melter feed physical simulant; pretreated high level waste (HLW) physical simulant; HLW melter feed physical simulant. The nominal chemical simulant, hereafter referred to as the HLW Precipitated Hydroxide simulant, is designed to represent the chemical/physical composition of the actual washed and leached sludge sample. The objective was to produce a simulant which matches not only the chemical composition but also the physical properties of the actual waste sample. The HLW Precipitated Hydroxide simulant could then be used for mixing tests to validate mixing, homogeneity and representative sampling and transferring issues. The HLW Precipitated Hydroxide simulant may also be used for integrated nonradioactive testing of the WTP prior to radioactive operation

  10. Innovative waste management solutions: An outlook for the future

    Energy Technology Data Exchange (ETDEWEB)

    1993-01-01

    At a conference on various aspects of waste management, papers were presented on the effects of landfills, plastic debris in the marine environment, skills development and training in the waste industry, composting, remediation of contaminated soils, disposal methods, sludge derived byproducts, recycling, waste management economics, municipal solid waste management, ship-generated waste, managing waste in national parks, and septic tank sludge treatment. Separate abstracts have been prepared for five papers from this conference.

  11. 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

  12. Simulating the structure of gypsum composites using pulverized basalt waste

    Directory of Open Access Journals (Sweden)

    Buryanov Аleksandr

    2017-01-01

    Full Text Available This paper examines the possibility of simulating the structure of gypsum composite modified with basalt dust waste to make materials and products based on it. Structural simulating of the topological space in gypsum modified composite by optimizing its grain-size composition highly improves its physical and mechanical properties. Strength and density tests have confirmed the results of the simulation. The properties of modified gypsum materials are improved by obtaining of denser particle packing in the presence of hemihydrate of finely dispersed basalt and plasticizer particles in the system, and by engaging basalt waste in the structuring process of modified gypsum stone.

  13. WIPP waste package testing on simulated DHLW: emplacement

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1984-01-01

    Several series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests. These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs. These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplace under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced under accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass. 9 refs., 1 fig

  14. Ammonia nitrogen removal from aqueous solution by local agricultural wastes

    Science.gov (United States)

    Azreen, I.; Lija, Y.; Zahrim, A. Y.

    2017-06-01

    Excess ammonia nitrogen in the waterways causes serious distortion to environment such as eutrophication and toxicity to aquatic organisms. Ammonia nitrogen removal from synthetic solution was investigated by using 40 local agricultural wastes as potential low cost adsorbent. Some of the adsorbent were able to remove ammonia nitrogen with adsorption capacity ranging from 0.58 mg/g to 3.58 mg/g. The highest adsorption capacity was recorded by Langsat peels with 3.58 mg/g followed by Jackfruit seeds and Moringa peels with 3.37 mg/g and 2.64 mg/g respectively. This experimental results show that the agricultural wastes can be utilized as biosorbent for ammonia nitrogen removal. The effect of initial ammonia nitrogen concentration, pH and stirring rate on the adsorption process were studied in batch experiment. The adsorption capacity reached maximum value at pH 7 with initial concentration of 500 mg/L and the removal rate decreased as stirring rate was applied.

  15. Challenges and solutions for realistic room simulation

    Science.gov (United States)

    Begault, Durand R.

    2002-05-01

    Virtual room acoustic simulation (auralization) techniques have traditionally focused on answering questions related to speech intelligibility or musical quality, typically in large volumetric spaces. More recently, auralization techniques have been found to be important for the externalization of headphone-reproduced virtual acoustic images. Although externalization can be accomplished using a minimal simulation, data indicate that realistic auralizations need to be responsive to head motion cues for accurate localization. Computational demands increase when providing for the simulation of coupled spaces, small rooms lacking meaningful reverberant decays, or reflective surfaces in outdoor environments. Auditory threshold data for both early reflections and late reverberant energy levels indicate that much of the information captured in acoustical measurements is inaudible, minimizing the intensive computational requirements of real-time auralization systems. Results are presented for early reflection thresholds as a function of azimuth angle, arrival time, and sound-source type, and reverberation thresholds as a function of reverberation time and level within 250-Hz-2-kHz octave bands. Good agreement is found between data obtained in virtual room simulations and those obtained in real rooms, allowing a strategy for minimizing computational requirements of real-time auralization systems.

  16. An innovative simulation tool for waste to energy generation opportunities

    Directory of Open Access Journals (Sweden)

    Bilal Abderezzak

    2017-03-01

    Full Text Available The new world energy policies encourage the use of renewable energy sources with clean technologies, and abandon progressively the fossil fuel dependence. Another energy generation trend called commonly the “Waste-to-Energy” solution, uses organic waste as a response for two major problems: energy generation and waste management. Thanks to the anaerobic digestion, the organic waste can provide a biogas composed essentially from Carbone dioxide (CO2 and Methane (CH4. This work aims essentially to help students, researchers and even decision makers to consider the importance of biogas generation. The proposed tool is the last version of our previous tool which is enhanced and completed. It presents the potential to produce biogas of any shortlisted kind of waste, including also some energy valorization ways. A technical economical data are introduced for eventual feasibility studies.

  17. Effects of simulant mixed waste on EPDM and butyl rubber

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Dickens, T.G.

    1997-11-01

    The authors have developed a Chemical Compatibility Testing Program for the evaluation of plastic packaging components which may be used in transporting mixed waste forms. In this program, they have screened 10 plastic materials in four liquid mixed waste simulants. These plastics were butadiene-acrylonitrile copolymer (Nitrile) rubber, cross-linked polyethylene, epichlorohydrin rubber, ethylene-propylene (EPDM) rubber, fluorocarbons (Viton and Kel-F trademark), polytetrafluoro-ethylene (Teflon), high-density polyethylene, isobutylene-isoprene copolymer (Butyl) rubber, polypropylene, and styrene-butadiene (SBR) rubber. The selected simulant mixed wastes were (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. The screening testing protocol involved exposing the respective materials to approximately 3 kGy of gamma radiation followed by 14-day exposures to the waste simulants at 60 C. The rubber materials or elastomers were tested using Vapor Transport Rate measurements while the liner materials were tested using specific gravity as a metric. The authors have developed a chemical compatibility program for the evaluation of plastic packaging components which may be incorporated in packaging for transporting mixed waste forms. From the data analyses performed to date, they have identified the thermoplastic, polychlorotrifluoroethylene, as having the greatest chemical compatibility after having been exposed to gamma radiation followed by exposure to the Hanford Tank simulant mixed waste. The most striking observation from this study was the poor performance of polytetrafluoroethylene under these conditions. In the evaluation of the two elastomeric materials they have concluded that while both materials exhibit remarkable resistance to these environmental conditions, EPDM has a greater resistance to this corrosive simulant mixed waste

  18. Uncertainty analysis of NDA waste measurements using computer simulations

    International Nuclear Information System (INIS)

    Blackwood, L.G.; Harker, Y.D.; Yoon, W.Y.; Meachum, T.R.

    2000-01-01

    Uncertainty assessments for nondestructive radioassay (NDA) systems for nuclear waste are complicated by factors extraneous to the measurement systems themselves. Most notably, characteristics of the waste matrix (e.g., homogeneity) and radioactive source material (e.g., particle size distribution) can have great effects on measured mass values. Under these circumstances, characterizing the waste population is as important as understanding the measurement system in obtaining realistic uncertainty values. When extraneous waste characteristics affect measurement results, the uncertainty results are waste-type specific. The goal becomes to assess the expected bias and precision for the measurement of a randomly selected item from the waste population of interest. Standard propagation-of-errors methods for uncertainty analysis can be very difficult to implement in the presence of significant extraneous effects on the measurement system. An alternative approach that naturally includes the extraneous effects is as follows: (1) Draw a random sample of items from the population of interest; (2) Measure the items using the NDA system of interest; (3) Establish the true quantity being measured using a gold standard technique; and (4) Estimate bias by deriving a statistical regression model comparing the measurements on the system of interest to the gold standard values; similar regression techniques for modeling the standard deviation of the difference values gives the estimated precision. Actual implementation of this method is often impractical. For example, a true gold standard confirmation measurement may not exist. A more tractable implementation is obtained by developing numerical models for both the waste material and the measurement system. A random sample of simulated waste containers generated by the waste population model serves as input to the measurement system model. This approach has been developed and successfully applied to assessing the quantity of

  19. VS2DRTI: Simulating Heat and Reactive Solute Transport in Variably Saturated Porous Media.

    Science.gov (United States)

    Healy, Richard W; Haile, Sosina S; Parkhurst, David L; Charlton, Scott R

    2018-01-29

    Variably saturated groundwater flow, heat transport, and solute transport are important processes in environmental phenomena, such as the natural evolution of water chemistry of aquifers and streams, the storage of radioactive waste in a geologic repository, the contamination of water resources from acid-rock drainage, and the geologic sequestration of carbon dioxide. Up to now, our ability to simulate these processes simultaneously with fully coupled reactive transport models has been limited to complex and often difficult-to-use models. To address the need for a simple and easy-to-use model, the VS2DRTI software package has been developed for simulating water flow, heat transport, and reactive solute transport through variably saturated porous media. The underlying numerical model, VS2DRT, was created by coupling the flow and transport capabilities of the VS2DT and VS2DH models with the equilibrium and kinetic reaction capabilities of PhreeqcRM. Flow capabilities include two-dimensional, constant-density, variably saturated flow; transport capabilities include both heat and multicomponent solute transport; and the reaction capabilities are a complete implementation of geochemical reactions of PHREEQC. The graphical user interface includes a preprocessor for building simulations and a postprocessor for visual display of simulation results. To demonstrate the simulation of multiple processes, the model is applied to a hypothetical example of injection of heated waste water to an aquifer with temperature-dependent cation exchange. VS2DRTI is freely available public domain software. © 2018, National Ground Water Association.

  20. Electrical resistivities of glass melts containing simulated SRP waste sludges

    International Nuclear Information System (INIS)

    Wiley, J.R.

    1978-08-01

    One option for the long-term management of radioactive waste at the Savannah River Plant is to solidify the waste in borosilicate glass by using a continuous, joule-heated, ceramic melter. Electrical resistivities that are needed for melter design were measured for melts of two borosilicate, glass-forming mixtures, each of which was combined with various amounts of several simulated-waste sludges. The simulated sludge spanned the composition range of actual sludges sampled from SRP waste tanks. Resistivities ranged from 6 to 10 ohm-cm at 500 0 C. Melt composition and temperature were correlated with resistivity. Resistivity was not a simple function of viscosity. 15 figures, 4 tables

  1. Polymer solution phase separation: Microgravity simulation

    Science.gov (United States)

    Cerny, Lawrence C.; Sutter, James K.

    1989-01-01

    In many multicomponent systems, a transition from a single phase of uniform composition to a multiphase state with separated regions of different composition can be induced by changes in temperature and shear. The density difference between the phase and thermal and/or shear gradients within the system results in buoyancy driven convection. These differences affect kinetics of the phase separation if the system has a sufficiently low viscosity. This investigation presents more preliminary developments of a theoretical model in order to describe effects of the buoyancy driven convection in phase separation kinetics. Polymer solutions were employed as model systems because of the ease with which density differences can be systematically varied and because of the importance of phase separation in the processing and properties of polymeric materials. The results indicate that the kinetics of the phase separation can be performed viscometrically using laser light scattering as a principle means of following the process quantitatively. Isopycnic polymer solutions were used to determine the viscosity and density difference limits for polymer phase separation.

  2. Full-scale retrieval of simulated buried transuranic waste

    International Nuclear Information System (INIS)

    Valentich, D.J.

    1993-09-01

    This report describes the results of a field test conducted to determine the effectiveness of using conventional type construction equipment for the retrieval of buried transuranic (TRU) waste. A cold (nonhazardous and nonradioactive) test pit (1,100 yd 3 volume) was constructed with boxes and drums filled with simulated waste materials, such as metal, plastic, wood, concrete, and sludge. Large objects, including truck beds, tanks, vaults, pipes, and beams, were also placed in the pit. These materials were intended to simulate the type of wastes found in TRU buried waste pits and trenches. A series of commercially available equipment items, such as excavators and tracked loaders outfitted with different end effectors, were used to remove the simulated waste. Work was performed from both the abovegrade and belowgrade positions. During the demonstration, a number of observations, measurements, and analyses were performed to determine which equipment was the most effective in removing the waste. The retrieval rates for the various excavation techniques were recorded. The inherent dust control capabilities of the excavation methods used were observed. The feasibility of teleoperating reading equipment was also addressed

  3. Combining metadynamics simulation and experiments to characterize dendrimers in solution

    NARCIS (Netherlands)

    Pavan, G.M.; Barducci, A.; Albertazzi, L.; Parrinello, M.

    2013-01-01

    We report a combined theoretical-experimental approach to characterize dendrimers and Polyethylene Glycol (PEG)-dendrimer hybrids in solution. Well-tempered metadynamics simulation allows for an exhaustive sampling of the conformational fluctuations in solution. In contrast to classical molecular

  4. Investigation of thermolytic hydrogen generation rate of tank farm simulated and actual waste

    Energy Technology Data Exchange (ETDEWEB)

    Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Howe, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-15

    To support resolution of Potential Inadequacies in the Safety Analysis for the Savannah River Site (SRS) Tank Farm, Savannah River National Laboratory conducted research to determine the thermolytic hydrogen generation rate (HGR) with simulated and actual waste. Gas chromatography methods were developed and used with air-purged flow systems to quantify hydrogen generation from heated simulated and actual waste at rates applicable to the Tank Farm Documented Safety Analysis (DSA). Initial simulant tests with a simple salt solution plus sodium glycolate demonstrated the behavior of the test apparatus by replicating known HGR kinetics. Additional simulant tests with the simple salt solution excluding organics apart from contaminants provided measurement of the detection and quantification limits for the apparatus with respect to hydrogen generation. Testing included a measurement of HGR on actual SRS tank waste from Tank 38. A final series of measurements examined HGR for a simulant with the most common SRS Tank Farm organics at temperatures up to 140 °C. The following conclusions result from this testing.

  5. Description of waste pretreatment and interfacing systems dynamic simulation model

    International Nuclear Information System (INIS)

    Garbrick, D.J.; Zimmerman, B.D.

    1995-05-01

    The Waste Pretreatment and Interfacing Systems Dynamic Simulation Model was created to investigate the required pretreatment facility processing rates for both high level and low level waste so that the vitrification of tank waste can be completed according to the milestones defined in the Tri-Party Agreement (TPA). In order to achieve this objective, the processes upstream and downstream of the pretreatment facilities must also be included. The simulation model starts with retrieval of tank waste and ends with vitrification for both low level and high level wastes. This report describes the results of three simulation cases: one based on suggested average facility processing rates, one with facility rates determined so that approximately 6 new DSTs are required, and one with facility rates determined so that approximately no new DSTs are required. It appears, based on the simulation results, that reasonable facility processing rates can be selected so that no new DSTs are required by the TWRS program. However, this conclusion must be viewed with respect to the modeling assumptions, described in detail in the report. Also included in the report, in an appendix, are results of two sensitivity cases: one with glass plant water recycle steams recycled versus not recycled, and one employing the TPA SST retrieval schedule versus a more uniform SST retrieval schedule. Both recycling and retrieval schedule appear to have a significant impact on overall tank usage

  6. Standard test method for determining liquidus temperature of immobilized waste glasses and simulated waste glasses

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 These practices cover procedures for determining the liquidus temperature (TL) of nuclear waste, mixed nuclear waste, simulated nuclear waste, or hazardous waste glass in the temperature range from 600°C to 1600°C. This method differs from Practice C829 in that it employs additional methods to determine TL. TL is useful in waste glass plant operation, glass formulation, and melter design to determine the minimum temperature that must be maintained in a waste glass melt to make sure that crystallization does not occur or is below a particular constraint, for example, 1 volume % crystallinity or T1%. As of now, many institutions studying waste and simulated waste vitrification are not in agreement regarding this constraint (1). 1.2 Three methods are included, differing in (1) the type of equipment available to the analyst (that is, type of furnace and characterization equipment), (2) the quantity of glass available to the analyst, (3) the precision and accuracy desired for the measurement, and (4) candi...

  7. 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.

  8. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    Energy Technology Data Exchange (ETDEWEB)

    Schultz, Peter Andrew

    2011-12-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  9. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC)

    International Nuclear Information System (INIS)

    Schultz, Peter Andrew

    2011-01-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M and S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V and V) is required throughout the system to establish evidence-based metrics for the level of confidence in M and S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V and V challenge at the subcontinuum scale, an approach to incorporate V and V concepts into subcontinuum scale modeling and simulation (M and S), and a plan to incrementally incorporate effective V and V into subcontinuum scale M and S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  10. Leaching characteristics of actinides from simulated reactor waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W.; Schweiger, J.S.

    1979-01-18

    Two methods for measuring the leach rates of simulated high level waste glass are compared. One is a modification of the standard IAEA method and the other is a one-pass method in which fresh leachant solution is pumped over the sample at a controlled flow rate and temperature. For times up to 3 days, there is close agreement between results from the two methods at 25.0/sup 0/C. Leach rates from the one-pass method show a correlation with flow rate only on day 1 at 25.0/sup 0/C, whereas they show a correlation with flow rate for all three days at 75.0/sup 0/C. /sup 237/Np rates at 75.0/sup 0/C are greater than those at 25.0/sup 0/C, but /sup 239/Pu rates at 75.0/sup 0/C are less than or equal to those at 25.0/sup 0/C.

  11. Siting simulation for low-level waste disposal facilities

    International Nuclear Information System (INIS)

    Roop, R.D.; Rope, R.C.

    1985-01-01

    The Mock Site Licensing Demonstration Project has developed the Low-Level Radioactive Waste Siting Simulation, a role-playing exercise designed to facilitate the process of siting and licensing disposal facilities for low-level waste (LLW). This paper describes the development, content, and usefulness of the siting simulation. The simulation can be conducted at a workshop or conference, involves 14 or more participants, and requires about eight hours to complete. The simulation consists of two sessions; in the first, participants negotiate the selection of siting criteria, and in the second, a preferred disposal site is chosen from three candidate sites. The project has sponsored two workshops (in Boston, Massachusetts and Richmond, Virginia) in which the simulation has been conducted for persons concerned with LLW management issues. It is concluded that the simulation can be valuable as a tool for disseminating information about LLW management; a vehicle that can foster communication; and a step toward consensus building and conflict resolution. The DOE National Low-Level Waste Management Program is now making the siting simulation available for use by states, regional compacts, and other organizations involved in development of LLW disposal facilities

  12. Effects of simulant mixed waste on EPDM and butyl rubber

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Dickens, T.G.

    1998-01-01

    We have developed a Chemical Compatibility Testing Program for the evaluation of plastic packaging components which may be used in transporting mixed waste forms. In this program, we have screened 10 plastic materials in four liquid mixed waste simulants. These plastics were butadiene-acrylonitrile copolymer (Nitrile) rubber, cross-linked polyethylene, epi-chloro-hydrin rubber, ethylene-propylene (EPDM) rubber, fluorocarbons (Viton and Kel-F), poly-tetrafluoroethylene (Teflon), high-density polyethylene, isobutylene-isoprene copolymer (Butyl) rubber, polypropylene, and styrene-butadiene (SBR) rubber. The selected simulant mixed wastes were (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. The screening testing protocol involved exposing the respective materials to ∼3 kGy of gamma radiation followed by 14-day exposures to the waste simulants at 60 deg. C. The rubber materials or elastomers were tested using VTR measurements while the liner materials were tested using specific gravity as a metric. For these tests, screening criteria of ∼1 g/hr/m 2 for VTR and specific gravity change of 10% were used. Those materials that failed to meet these criteria were judged to have failed the screening tests and were excluded from the next phase of this experimental program. We have completed the comprehensive testing phase of liner materials in a simulant Hanford Tank waste consisting of an aqueous alkaline mixture of sodium nitrate and sodium nitrite. From the data analyses performed, we have identified the chloro-fluorocarbon Kel-F as having the greatest chemical durability after having been exposed to gamma radiation followed by exposure to the aqueous alkaline simulant mixed waste. The most striking observation from this study was the extremely poor performance of Teflon under these conditions. We have also completed the comprehensive

  13. Solute transport in fractured rock - applications to radionuclide waste repositories

    International Nuclear Information System (INIS)

    Neretnieks, I.

    1990-12-01

    Flow and solute transport in fractured rocks has been intensively studied in the last decade. The increased interest is mainly due to the plans in many countries to site repositories for high level nuclear waste in deep geologic formations. All investigated crystalline rocks have been found to be fractured and most of the water flows in the fractures and fracture zones. The water transports dissolved species and radionuclides. It is thus of interest to be able to understand and to do predictive modelling of the flowrate of water, the flowpaths and the residence times of the water and of the nuclides. The dissolved species including the nuclides will interact with the surrounding rock in different ways and will in many cases be strongly retarded relative to the water velocity. Ionic species may be ion exchanged or sorbed in the mineral surfaces. Charges and neutral species may diffuse into the stagnant waters in the rock matrix and thus be withdrawn from the mobile water. These effects will be strongly dependent on how much rock surface is in contact with the flowing water. It has been found in a set of field experiments and by other observations that not all fractures conduct water. Furthermore it is found that conductive fractures only conduct the water in a small part of the fracture in what is called channels or preferential flowpaths. This report summarizes the present concepts of water flow and solute transport in fractured rocks. The data needs for predictive modelling are discussed and both field and laboratory measurement which have been used to obtain data are described. Several large scale field experiments which have been specially designed to study flow and tracer transport in crystalline rocks are described. In many of the field experients new techniques have been developed and used. (81 refs.) (author)

  14. 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.

  15. Reaction chemistry of nitrogen species in hydrothermal systems: Simple reactions, waste simulants, and actual wastes

    International Nuclear Information System (INIS)

    Dell'Orco, P.; Luan, L.; Proesmans, P.; Wilmanns, E.

    1995-01-01

    Results are presented from hydrothermal reaction systems containing organic components, nitrogen components, and an oxidant. Reaction chemistry observed in simple systems and in simple waste simulants is used to develop a model which presents global nitrogen chemistry in these reactive systems. The global reaction path suggested is then compared with results obtained for the treatment of an actual waste stream containing only C-N-0-H species

  16. Pitfall in quantum mechanical/molecular mechanical molecular dynamics simulation of small solutes in solution.

    Science.gov (United States)

    Hu, Hao; Liu, Haiyan

    2013-05-30

    Developments in computing hardware and algorithms have made direct molecular dynamics simulation with the combined quantum mechanical/molecular mechanical methods affordable for small solute molecules in solution, in which much improved accuracy can be obtained via the quantum mechanical treatment of the solute molecule and even sometimes water molecules in the first solvation shell. However, unlike the conventional molecular mechanical simulations of large molecules, e.g., proteins, in solutions, special care must be taken in the technical details of the simulation, including the thermostat of the solute/solvent system, so that the conformational space of the solute molecules can be properly sampled. We show here that the common setup for classical molecular mechanical molecular dynamics simulations, such as the Berendsen or single Nose-Hoover thermostat, and/or rigid water models could lead to pathological sampling of the solutes' conformation. In the extreme example of a methanol molecule in aqueous solution, improper and sluggish setups could generate two peaks in the distribution of the O-H bond length. We discuss the factors responsible for this somewhat unexpected result and evoke a simple and ancient technical fix-up to resolve this problem.

  17. Removal of palladium precipitate from a simulated high-level radioactive liquid waste by reduction by ascorbic acid

    International Nuclear Information System (INIS)

    Kim, Eung Ho; Yoo, Jae Hyung; Choi, Cheong Song

    1998-01-01

    A study of the selective removal of Palladium from a simulated solution of high-level radioactive liquid waste (HLLW) was carried out. The simulated solution contained 7 representative elements (Pd 2+ , Cs + , Sr 2+ , Fe 3+ , MoO 2 2+ , Ru 4+ , and Nd 3+ ) typical of HLLW, ascorbic acid was added to the solution at room temperature. Pd 2+ in the simulated solution was easily reduced to Pd metal by the ascorbic acid and then the metal precipitate could be removed from the solution, whereas other elements remained mainly in solution. When the resulting Pd metal was left in solution, it was reoxidized to Pb 2+ ion and redissolved in a nitric acid medium. The oxidation rate of Pd 2+ depended on the presence of a transition metal such as ferric ion, and was also in proportion to the concentration of nitric acid and in inverse proportion to the concentration of ascrobic acid. (orig.)

  18. Caustic-Side Solvent Extraction: Prediction of Cesium Extraction from Actual Wastes and Actual Waste Simulants

    International Nuclear Information System (INIS)

    Delmau, L.H.; Haverlock, T.J.; Sloop, F.V. Jr.; Moyer, B.A.

    2003-01-01

    This report presents the work that followed the CSSX model development completed in FY2002. The developed cesium and potassium extraction model was based on extraction data obtained from simple aqueous media. It was tested to ensure the validity of the prediction for the cesium extraction from actual waste. Compositions of the actual tank waste were obtained from the Savannah River Site personnel and were used to prepare defined simulants and to predict cesium distribution ratios using the model. It was therefore possible to compare the cesium distribution ratios obtained from the actual waste, the simulant, and the predicted values. It was determined that the predicted values agree with the measured values for the simulants. Predicted values also agreed, with three exceptions, with measured values for the tank wastes. Discrepancies were attributed in part to the uncertainty in the cation/anion balance in the actual waste composition, but likely more so to the uncertainty in the potassium concentration in the waste, given the demonstrated large competing effect of this metal on cesium extraction. It was demonstrated that the upper limit for the potassium concentration in the feed ought to not exceed 0.05 M in order to maintain suitable cesium distribution ratios

  19. TRANSPORT OF WASTE SIMULANTS IN PJM VENT LINES

    Energy Technology Data Exchange (ETDEWEB)

    Qureshi, Z

    2007-02-21

    The experimental work was conducted to determine whether there is a potential for waste simulant to transport or 'creep' up the air link line and contaminate the pulse jet vent system, and possibly cause long term restriction of the air link line. Additionally, if simulant creep occurred, establish operating parameters for washing down the line. The amount of the addition of flush fluids and mixer downtime must be quantified.

  20. Simulation used to qualify nuclear waste glass for disposal

    International Nuclear Information System (INIS)

    Reimus, T.W.; Kuhn, W.L.

    1987-07-01

    A hypothetical vitrification system was simulated errors associated with controlling and predicting the composition of the nuclear waste glass produced in the system. The composition of the glass must fall within certain limits to qualify for permanent geologic disposal. The estimated error in predicting the concentrations of various constituents in the glass was 2% to 8%, depending on the strategy for sampling and analyzing the feed and on the assumed magnitudes of the process uncertainties. The estimated error in controlling the glass composition was 2% to 9%, depending on the strategy for sampling and analyzing the waste and on the assumed magnitudes of the uncertainties. This work demonstrates that simulation techniques can be used to assist in qualifying nuclear waste glass for disposal. 3 refs., 2 figs., 4 tabs

  1. Incineration of simulated plutonium-contaminated waste

    International Nuclear Information System (INIS)

    Ford, L.H.; Jenkins, M.J.

    1984-01-01

    Pyrolysis rate data are presented which will enable larger pyrolyser furnaces to be made for processing solid plutonium-contaminated materials at throughputs of up to 20 kg/h using either 1 or 2.5 kg packages as feed. The influence of liquids, such as water, kerosene or oil, on the pyrolysis process has also been assessed. The products of pyrolysis for a range of individual materials and their mixtures have been defined. The oxidation rates for both static and stirred beds of char have been obtained. The implications of both the pyrolysis and char-oxidation processes for plant design are discussed. This work has been commissioned by the Department of the Environment as part of its radioactive waste management programme. The results will be used in the formulation of government policy, but as this stage they do not necessarily represent that policy

  2. Couplex1 test case nuclear - Waste disposal far field simulation

    International Nuclear Information System (INIS)

    2001-01-01

    This first COUPLEX test case is to compute a simplified Far Field model used in nuclear waste management simulation. From the mathematical point of view the problem is of convection diffusion type but the parameters are highly varying from one layer to another. Another particularity is the very concentrated nature of the source, both in space and in time. (author)

  3. ''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

  4. Chemical activation of tea waste and use for the removal of chromium (Vi) from aqueous solution

    International Nuclear Information System (INIS)

    Qureshi, K.; Bhatti, I.; Ansari, A.K.

    2009-01-01

    Tea waste is the residue left after the preparation of tea. At present the tea waste is regarded as a waste product having no use. In this study, tea waste is converted into an adsorbent. Tea waste is chemically activated with phosphoric acid at low temperature 450 degree C. This activated carbon is then utilized as an adsorbent for the removal of Chromium (VI) from aqueous solution. The various sorption parameters i.e pH, sorbent dose sorbate concentration, shaking time and shaking speed are first optimized. 75% of chromium from aqueous solution is effectively removed at pH 2. The best optimum conditions were obtained when 1 gm of sorbent was agitated at 100 rpm with 60 mg/l of sorbate for 50 minutes. Better results were obtained when low concentrations of sorbates were used. Hence tea waste could also be successfully used for the sorption of Chromium (VI), from industrial waste water. (author)

  5. WASTES II: Waste System Transportation and Economic Simulation. Version II. User's guide

    International Nuclear Information System (INIS)

    Shay, M.R.; Buxbaum, M.E.

    1986-02-01

    The WASTES II model was developed to provide detailed analyses beyond the capabilities of other available models. WASTES uses discrete event simulation techniques to model the generation of commercial spent nuclear fuel, the buildup of spent fuel inventories within the system, and the transportation requirements for the movement of radioactive waste throughout the system. The model is written in FORTRAN 77 as an extension to the SLAM commercial simulation language package. In addition to the pool storage and dry storage located at the reactors, the WASTES model provides a choice of up to ten other storage facilities of four different types. The simulation performed by WASTES may be controlled by a combination of source- and/or destination-controlled transfers that are requested by the code user. The user supplies shipping cask characteristics for truck or rail shipment casks. As part of the facility description, the user specifies which casks the facility can use. Shipments within the system can be user specified to occur optimally, or proximally. Optimized shipping can be used when exactly two destination facilities of the same facility type are open for receipt of fuel. Optimized shipping selects source/destination pairs so that the total shipping distance or total shipping costs in a given year are minimized when both facilities are fully utilized. Proximity shipping sequentially fills the closest facility to the source according to the shipment priorities without regard for the total annual shipments. This results in sub-optimal routing of waste material but can be used to approximate an optimal shipping strategy when more than two facilities of the same type are available to receive waste. WASTES is currently able to analyze each of the commercial spent fuel logistics scenarios specified in the 1985 DOE Mission Plan

  6. Chemical and Nuclear Waste Disposal: Problems and Solutions

    OpenAIRE

    James P. Murray; Joseph J. Harrington; Richard Wilson

    1982-01-01

    The problems of waste disposal have always been with us. In biblical times, the residents of Jerusalem always burnt their wastes inthehideousValeofGehenna.Thisgavewaytoburialofwasteor sometimes dumping it in shallow oceans. All too often the sewage pipes of the seaside towns did not even take the waste to the low tide mark; and the use of the deep oceans as a disposal site has been almost unknown...

  7. Municipal solid waste management. Strategies and technologies for sustainable solutions

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, C.; Hellweg, S.; Stucki, S. (eds.)

    2002-10-01

    The way municipal solid waste is handled greatly determines its impact on the local as well as the global environment. New technologies habe emerged for the treatment of waste, for the recovery of raw materials and energy, and for safe final disposal. The environmental performance of technologies, their social acceptance and their economic viability are key issues to be considered in sustainable waste management. This book provides an overview of current practices in waste management and a synthesis of new developments achieved through interdisciplinary discussions of recent research results. (orig.)

  8. Hanford Waste Simulants Created to Support the Research and Development on the River Protection Project - Waste Treatment Plant

    Energy Technology Data Exchange (ETDEWEB)

    Eibling, R.E.

    2001-07-26

    The development of nonradioactive waste simulants to support the River Protection Project - Waste Treatment Plant bench and pilot-scale testing is crucial to the design of the facility. The report documents the simulants development to support the SRTC programs and the strategies used to produce the simulants.

  9. Defense Waste Processing Facility Process Simulation Package Life Cycle

    International Nuclear Information System (INIS)

    Reuter, K.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) will be used to immobilize high level liquid radioactive waste into safe, stable, and manageable solid form. The complexity and classification of the facility requires that a performance based operator training to satisfy Department of Energy orders and guidelines. A major portion of the training program will be the application and utilization of Process Simulation Packages to assist in training the Control Room Operators on the fluctionality of the process and the application of the Distribution Control System (DCS) in operating and managing the DWPF process. The packages are being developed by the DWPF Computer and Information Systems Simulation Group. This paper will describe the DWPF Process Simulation Package Life Cycle. The areas of package scope, development, validation, and configuration management will be reviewed and discussed in detail

  10. Electrochemical processing of nitrate waste solutions. Phase 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Genders, D.; Weinberg, N.; Hartsough, D. [Electrosynthesis Co., Inc., Cheektowaga, NY (US)

    1992-10-07

    The second phase of research performed at The Electrosynthesis Co., Inc. has demonstrated the successful removal of nitrite and nitrate from a synthetic effluent stream via a direct electrochemical reduction at a cathode. It was shown that direct reduction occurs at good current efficiencies in 1,000 hour studies. The membrane separation process is not readily achievable for the removal of nitrites and nitrates due to poor current efficiencies and membrane stability problems. A direct reduction process was studied at various cathode materials in a flow cell using the complete synthetic mix. Lead was found to be the cathode material of choice, displaying good current efficiencies and stability in short and long term tests under conditions of high temperature and high current density. Several anode materials were studied in both undivided and divided cell configurations. A divided cell configuration was preferable because it would prevent re-oxidation of nitrite by the anode. The technical objective of eliminating electrode fouling and solids formation was achieved although anode materials which had demonstrated good stability in short term divided cell tests corroded in 1,000 hour experiments. The cause for corrosion is thought to be F{sup {minus}} ions from the synthetic mix migrating across the cation exchange membrane and forming HF in the acid anolyte. Other possibilities for anode materials were explored. A membrane separation process was investigated which employs an anion and cation exchange membrane to remove nitrite and nitrate, recovering caustic and nitric acid. Present research has shown poor current efficiencies for nitrite and nitrate transport across the anion exchange membrane due to co-migration of hydroxide anions. Precipitates form within the anion exchange membranes which would eventually result in the failure of the membranes. Electrochemical processing offers a highly promising and viable method for the treatment of nitrate waste solutions.

  11. CHARACTERIZATION OF ACTINIDES IN SIMULATED ALKALINE TANK WASTE SLUDGES AND LEACHATES

    Energy Technology Data Exchange (ETDEWEB)

    Nash, Kenneth L.

    2008-11-20

    In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

  12. Characterization Of Actinides In Simulated Alkaline Tank Waste Sludges And Leachates

    International Nuclear Information System (INIS)

    Nash, Kenneth L.

    2008-01-01

    In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

  13. Modeling unsteady-state VOC transport in simulated waste drums

    International Nuclear Information System (INIS)

    Liekhus, K.J.; Gresham, G.L.; Peterson, E.S.; Rae, C.; Hotz, N.J.; Connolly, M.J.

    1994-01-01

    This report is a revision of an EG ampersand G Idaho informal report originally titled Modeling VOC Transport in Simulated Waste Drums. A volatile organic compound (VOC) transport model has been developed to describe unsteady-state VOC permeation and diffusion within a waste drum. Model equations account for three primary mechanisms for VOC transport from a void volume within the drum. These mechanisms are VOC permeation across a polymer boundary, VOC diffusion across an opening in a volume boundary, and VOC solubilization in a polymer boundary. A series of lab-scale experiments was performed in which the VOC concentration was measured in simulated waste drums under different conditions. A lab-scale simulated waste drum consisted of a sized-down 55-gal metal drum containing a modified rigid polyethylene drum liner. Four polyethylene bags were sealed inside a large polyethylene bag, supported by a wire cage, and placed inside the drum liner. The small bags were filled with VOC-air gas mixture and the VOC concentration was measured throughout the drum over a period of time. Test variables included the type of VOC-air gas mixtures introduced into the small bags, the small bag closure type, and the presence or absence of a variable external heat source. Model results were calculated for those trials where the permeability had been measured

  14. System Planning With The Hanford Waste Operations Simulator

    International Nuclear Information System (INIS)

    Crawford, T.W.; Certa, P.J.; Wells, M.N.

    2010-01-01

    At the U. S. Department of Energy's Hanford Site in southeastern Washington State, 216 million liters (57 million gallons) of nuclear waste is currently stored in aging underground tanks, threatening the Columbia River. The River Protection Project (RPP), a fully integrated system of waste storage, retrieval, treatment, and disposal facilities, is in varying stages of design, construction, operation, and future planning. These facilities face many overlapping technical, regulatory, and financial hurdles to achieve site cleanup and closure. Program execution is ongoing, but completion is currently expected to take approximately 40 more years. Strategic planning for the treatment of Hanford tank waste is by nature a multi-faceted, complex and iterative process. To help manage the planning, a report referred to as the RPP System Plan is prepared to provide a basis for aligning the program scope with the cost and schedule, from upper-tier contracts to individual facility operating plans. The Hanford Tank Waste Operations Simulator (HTWOS), a dynamic flowsheet simulation and mass balance computer model, is used to simulate the current planned RPP mission, evaluate the impacts of changes to the mission, and assist in planning near-term facility operations. Development of additional modeling tools, including an operations research model and a cost model, will further improve long-term planning confidence. The most recent RPP System Plan, Revision 4, was published in September 2009.

  15. Organic analyses of an actual and simulated mixed waste. Hanford's organic complexant waste revisited

    International Nuclear Information System (INIS)

    Toste, A.P.; Osborn, B.C.; Polach, K.J.; Lechner-Fish, T.J.

    1995-01-01

    Reanalysis of the organics in a mixed waste, an organic complexant waste, from the U.S. Department of Energy's Hanford Site, has yielded an 80.4% accounting of the waste's total organic content. In addition to several complexing and chelating agents (citrate, EDTA, HEDTA and NTA), 38 chelator/complexor fragments have been identified, compared to only 11 in the original analysis, all presumably formed via organic degradation. Moreover, a mis identification, methanetricarboxylic acid, has been re-identified as the chelator fragment N-(methylamine)imino-diacetic acid (MAIDA). A nonradioactive simulant of the actual waste, containing the parent organics (citrate, EDTA, HEDTA and NTA), was formulated and stored in the dark at ambient temperature for 90 days. Twenty chelator and complexor fragments were identified in the simulant, along with several carboxylic acids, confirming that myriad chelator and complexor fragments are formed via degradation of the parent organics. Moreover, their abundance in the simulant (60.9% of the organics identified) argues that the harsh chemistries of mixed wastes like Hanford's organic degradation, even in the absence of radiation. (author). 26 refs., 2 tabs

  16. Operating Range for High Temperature Borosilicate Waste Glasses: (Simulated Hanford Enveloped)

    International Nuclear Information System (INIS)

    Mohammad, J.; Ramsey, W. G.; Toghiani, R. K.

    2003-01-01

    The following results are a part of an independent thesis study conducted at Diagnostic Instrumentation and Analysis Laboratory-Mississippi State University. A series of small-scale borosilicate glass melts from high-level waste simulant were produced with waste loadings ranging from 20% to 55% (by mass). Crushed glass was allowed to react in an aqueous environment under static conditions for 7 days. The data obtained from the chemical analysis of the leachate solutions were used to test the durability of the resulting glasses. Studies were performed to determine the qualitative effects of increasing the B2O3 content on the overall waste glass leaching behavior. Structural changes in a glass arising due to B2O3 were detected indirectly by its chemical durability, which is a strong function of composition and structure. Modeling was performed to predict glass durability quantitatively in an aqueous environment as a direct function of oxide composition

  17. Simplified analytical model to simulate radionuclide release from radioactive waste trenches

    International Nuclear Information System (INIS)

    Sa, Bernardete Lemes Vieira de

    2001-01-01

    In order to evaluate postclosure off-site doses from low-level radioactive waste disposal facilities, a computer code was developed to simulate the radionuclide released from waste form, transport through vadose zone and transport in the saturated zone. This paper describes the methodology used to model these process. The radionuclide released from the waste is calculated using a model based on first order kinetics and the transport through porous media was determined using semi-analytical solution of the mass transport equation, considering the limiting case of unidirectional convective transport with three-dimensional dispersion in an isotropic medium. The results obtained in this work were compared with other codes, showing good agreement. (author)

  18. Simulations for the transmutation of nuclear wastes with hybrid reactors

    International Nuclear Information System (INIS)

    Vuillier, St.

    1998-06-01

    A Monte Carlo simulation, devoted to the spallation, has been built in the framework of the hybrid systems proposed for the nuclear wastes incineration. This system GSPARTE, described the reactions evolution. It takes into account and improves the nuclear codes and the low and high energy particles transport in the GEANT code environment, adapted to the geometry of the hybrid reactors. Many applications and abacus useful for the wastes transmutation, have been realized with this system: production of thick target neutrons, source definition, material damages. (A.L.B.)

  19. Distributions of 12 elements on 64 absorbers from simulated Hanford Neutralized Current Acid Waste (NCAW)

    International Nuclear Information System (INIS)

    Svitra, Z.V.; Bowen, S.M.; Marsh, S.F.

    1994-12-01

    As part of the Hanford Tank Waste Remediation System program at Los Alamos, we evaluated 64 commercially available or experimental absorber materials for their ability to remove hazardous components from high-level waste. These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. We tested these absorbers with a solution that simulates Hanford neutralized current acid waste (NCAW) (pH 14.2). To this simulant solution we added the appropriate radionuclides and used gamma spectrometry to measure fission products (Cs, Sr, Tc, and Y) and matrix elements (Cr, Co, Fe, Mn, Ni, V, Zn, and Zr). For each of 768 element/absorber combinations, we measured distribution coefficients for dynamic contact periods of 30 min, 2 h, and 6 h to obtain information about sorption kinetics. On the basis of these 2304 measured distribution coefficients, we determined that many of the tested absorbers may be suitable for processing NCAW solutions

  20. Sorption Potentials of Waste Tyre for Some Heavy Metals (Pb Cd in Aqueous Solution

    Directory of Open Access Journals (Sweden)

    Austin Kanayo ASIAGWU

    2009-07-01

    Full Text Available An investigation into the adsorption potential of activated and inactivated waste tyre powders for some heavy metals (Pb2+ and Cd2+ in their aqueous solution has been studied. The result indicated that inactivated waste tyre is a good non-conventional adsorbent for the removal of Cd from aqueous solution. A total of 93.3% of Cadmium contents was removed. The inactivated waste type proved a good adsorbent for the removal of Pb2+ 5g of 500mm activated tyre removed over 86.66% of Pb2+ from solution.

  1. Alternative solutions for the disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Ramsey, R.W. Jr.

    1976-01-01

    Besides outlining the possibility of dispatching concentrated highly radioactive waste by rockets into space, or of transmuting long-lived isotopes by nuclear reactions into short-lived ones, the author discusses further alternatives for the disposal of radioactive wastes, especially the storage in geologic formations. (HR/LN) [de

  2. Material resources, energy, and nutrient recovery from waste: are waste refineries the solution for the future?

    DEFF Research Database (Denmark)

    Tonini, Davide; Martinez-Sanchez, Veronica; Astrup, Thomas Fruergaard

    2013-01-01

    Waste refineries focusing on multiple outputs of material resources, energy carriers, and nutrients may potentially provide more sustainable utilization of waste resources than traditional waste technologies. This consequential life cycle assessment (LCA) evaluated the environmental performance....... Overall, the waste refinery provided global warming (GW) savings comparable with efficient incineration, MBT, and bioreactor landfilling technologies. The main environmental benefits from waste refining were a potential for improved phosphorus recovery (about 85%) and increased electricity production (by...

  3. Simulation applied to innovative waste management options

    International Nuclear Information System (INIS)

    2001-10-01

    The aim of this workshop is to share experience and methods in the domains of neutronics, spallation, thermo-hydraulics, corrosion, materials, mechanics; to define the needs for the other domains in order to improve the evaluation of concepts: ADS, MSR, Gas cooled reactors; and to determine the necessity or not to couple codes and tools for the concepts studied in the frame of GEDEON activities. This document groups together the transparencies of 6 presentations given at this workshop: design and performances of CEA spallation targets; simulation of spallation; the neutronic benchmark on the Megapie spallation target; the core physics of fast spectrum gas cooled reactors; the study and modeling of the thermal-mechanical behaviour of composite fuel in reactor. (J.S.)

  4. Cross-flow filtration during the washing of a simulated radioactive waste stream

    International Nuclear Information System (INIS)

    MARK R., DUIGNAN

    2005-01-01

    Bechtel National, Inc. has been contracted by the Department of Energy to design a Waste Treatment and Immobilization Plant (WTP) to stabilize liquid radioactive waste that is stored at the Hanford Site as part of the River Protection Project (RPP). Because of its experience with radioactive waste stabilization, the Savannah River National Laboratory (SRNL) of the Westinghouse Savannah River Company is working with Bechtel and Washington Group International, to help design and test certain parts of the waste treatment facility. One part of the process is the separation of radioactive solids from the liquid wastes by cross-flow ultrafiltration. To test this process a cross-flow filter was used that was prototypic in porosity, length, and diameter, along with a simulated radioactive waste slurry, made to prototypically represent the chemical and physical characteristics of a Hanford waste in tank 241-AY-102/C-106. To mimic the filtration process the waste slurry undergoes several steps, including dewatering and washing. During dewatering the concentration of undissolved solids (UDS) of the simulated AY102/C106 waste is increased from 12 wt percent to at least 20 wt percent. Once at the higher concentration the waste must be washed to prepare for its eventual receipt in a High Level Radioactive Waste Melter to be vitrified. This paper describes the process of washing and filtering a batch of concentrated simulated waste in two cycles, which each containing 22 washing steps that used approximately 7.7 liters of a solution of 0.01 M NaOH per step. This will be the method used by the full-scale WTP to prepare the waste for vitrification. The first washing cycle started with the simulated waste that had a solids concentration of 20 wt percent UDS. This cycle began with a permeate filter flux of 0.015 gpm/ft2 (3.68 cm/hr) at 19.6 wt percent UDS with a density of 1.33 kg/L, and yield stress of 8.5 Pa. At the end of the 22 washing steps the permeate filter flux increased to

  5. Thermodynamic Models from Fluctuation Solution Theory Analysis of Molecular Simulations

    DEFF Research Database (Denmark)

    Christensen, Steen; Peters, Günther H.j.; Hansen, Flemming Yssing

    2007-01-01

    Fluctuation solution theory (FST) is employed to analyze results of molecular dynamics (MD) simulations of liquid mixtures. The objective is to generate parameters for macroscopic GE-models, here the modified Margules model. We present a strategy for choosing the number of parameters included...

  6. Solution of partial differential equations by agent-based simulation

    International Nuclear Information System (INIS)

    Szilagyi, Miklos N

    2014-01-01

    The purpose of this short note is to demonstrate that partial differential equations can be quickly solved by agent-based simulation with high accuracy. There is no need for the solution of large systems of algebraic equations. This method is especially useful for quick determination of potential distributions and demonstration purposes in teaching electromagnetism. (letters and comments)

  7. Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, Duane J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilmarth, William R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-01-01

    (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that

  8. Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    International Nuclear Information System (INIS)

    Adamson, Duane J.; Nash, Charles A.; McCabe, Daniel J.; Crawford, Charles L.; Wilmarth, William R.

    2014-01-01

    (chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that

  9. Cesium uptake capacity of simulated ferrocyanide tank waste. Interim report FY 1994, Ferrocyanide Safety Project

    International Nuclear Information System (INIS)

    Burgeson, I.E.; Bryan, S.A.; Burger, L.E.

    1994-09-01

    The objective of this project is to determine the capacity for 137 CS uptake by mixed metal ferrocyanides present in Hanford waste tanks, and to assess the potential for aggregation of these 137 CS exchanged materials to form tank ''hot-spots.'' This research, performed at the Pacific Northwest Laboratory (PNL) for the Westinghouse Hanford Company (WHC), stems from concerns of possible localized radiolytic heating within the tanks. If radioactive cesium is exchanged and concentrated by the remaining nickel ferrocyanide present in the tanks, this heating could cause temperatures to rise above the safety limits specified for the ferrocyanide tanks. For the purposes of this study, two simulants, In-Farm-2 and U-Plant-2, were chosen to represent the wastes generated by the scavenging processes. These simulants were formulated using protocols from the original cesium scavenging campaign. Later additions of cesium-rich wastes from various processes also were considered. The simulants were prepared and centrifuged to obtain a moist ferrocyanide sludge. The centrifuged sludges were treated with the original supernate spiked with a known amount of cesium nitrate. After analysis by flame atomic absorption spectrometry, distribution coefficients (K d ) were calculated. The capacity of solid waste simulants to exchange radioactive cesium from solution was examined. Initial results showed that the greater the molar ratio of cesium to cesium nickel ferrocyanide, the less effective the exchange of cesium from solution. The theoretical capacity of 2 mol cesium per mol of nickel ferrocyanide was not observed. The maximum capacity under experimental conditions was 0.35 mol cesium per mol nickel ferrocyanide. Future work on this project will examine the layering tendency of the cesium nickel ferrocyanide species

  10. Potential use of maize waste for the removal of Pb(II) from aqueous solution

    CSIR Research Space (South Africa)

    Okonkwo, J

    2006-09-01

    Full Text Available batch adsorption procedures. The utilization of tassels for the removal of toxic heavy metals from effluent solutions would, however, attach some economic value to this waste material. Tassel flowers were collected just prior to harvest, dried under...

  11. Determination of plutonium 241 in solutions of nuclear wastes

    International Nuclear Information System (INIS)

    Raymond, A.; Bilcot, J.B.; Poletiko, C.

    1990-09-01

    Determination of plutonium 241 in nuclear wastes is important because of long period and high energy of some daughter products. In this report are presented two quantitative analysis methods using both scintillation techniques: A complete method, in any case, by selective extraction of plutonium on an anionic resin allowing simultaneous determination of Pu 241 and the sum of other plutonium isotopes; a simplified method when alpha activity is higher than beta/gamma activity by liquid extraction with TTA. These methods are applied for analysis of 4 waste types: cement encapsulated wastes, bitumen encapsulated wastes, incineration ashes, leaching of encapsulated incineration ashes. In these 4 examples, Pu 241 activity is equal or higher than the sum of alpha plutonium isotope activity. Separation efficiency, measured from Pu 239 or with Pu 236 as tracer, is between 90 and 99% [fr

  12. Rheological properties of kaolin and chemically simulated waste

    International Nuclear Information System (INIS)

    Selby, C.L.

    1981-12-01

    The Savannah River Laboratory is conducting tests to determine the best operating conditions of pumps used to transfer insoluble radioactive sludges from old to new waste tanks. Because it is not feasible to conduct these tests with real or chemically simulated sludges, kaolin clay is being used as a stand-in for the solid waste. The rheology tests described herein were conducted to determine whether the properties of kaolin were sufficiently similar to those of real sludge to permit meaningful pump tests. The rheology study showed that kaolin can be substituted for real waste to accurately determine pump performance. Once adequately sheared, kaolin properties were found to remain constant. Test results determined that kaolin should not be allowed to settle more than two weeks between pump tests. Water or supernate from the waste tanks can be used to dilute sludge on an equal volume basis because they identically affect the rheological properties of sludge. It was further found that the fluid properties of kaolin and waste are insensitive to temperature

  13. Powder technological vitrification of simulated high-level waste

    International Nuclear Information System (INIS)

    Gahlert, S.

    1988-03-01

    High-level waste simulate from the reprocessing of light water reactor and fast breeder fuel was vitrified by powder technology. After denitration with formaldehyde, the simulated HLW is mixed with glass frit and simultaneously dried in an oil-heated mixer. After 'in-can calcination' for at least 24 hours at 850 or 950 K (depending on the type of waste and glass), the mixture is hot-pressed in-can for several hours at 920 or 1020 K respectively, at pressures between 0.4 and 1.0 MPa. The technology has been demonstrated inactively up to diameters of 30 cm. Leach resistance is significantly enhanced when compared to common borosilicate glasses by the utilization of glasses with higher silicon and aluminium content and lower sodium content. (orig.) [de

  14. Leaching behavior of simulated high-level waste glass

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi

    1987-03-01

    The author's work in the study on the leaching behavior of simulated high-level waste (HLW) glass were summarized. The subjects described are (1) leach rates at high temperatures, (2) effects of cracks on leach rates, (3) effects of flow rate on leach rates, and (4) an in-situ burial test in natural groundwater. In the following section, the leach rates obtained by various experiments were summarized and discussed. (author)

  15. Rheological evaluation of simulated neutralized current acid waste

    International Nuclear Information System (INIS)

    Fow, C.L.; McCarthy, D.; Thornton, G.T.

    1986-06-01

    A byproduct of the Purex process is an aqueous waste stream that contains fission products. This waste stream, called current acid waste, is chemically neutralized and stored in double shell tanks on the Hanford Site. This neutralized current acid waste (NCAW) will be transported by pipe to B-Plant, a processing plant on the Hanford Site. Rheological and transport properties of NCAW slurry were evaluated. First, researchers conducted lab rheological evaluations of simulated NCAW. The results of these evaluations were then correlated with classical rheological models and scaled up to predict the performance that is likely to occur in the full-scale system. The NCAW in the tank will either be retrieved as is, i.e., no change in the concentration presently in the tank, or will be slightly concentrated before retrieval. Sluicing may be required to retrieve the solids. Three concentrations of simulated NCAW were evaluated that would simulate the different retrieval options: NCAW in the concentration that is presently in the tank; a slightly concentrated NCAW, called NCAW5.5; and equal parts of NCAW settled solids and water (simulating the sluicing stage), called NCAW1:1. The physical and rheological properties of three samples of each concentration at 25 and 100 0 C were evaluated in the laboratory. The properties displayed by NCAW and NCAW5.5 at 25 and 100 0 C allowed it to be classified as a pseudoplastic non-Newtonian fluid. NCAW1:1 at 25 and 100 0 C displayed properties of a yield-pseudoplastic non-Newtonian fluid. The classical non-Newtonian models for pseudoplastic and yield-pseudoplastic fluids were used with the laboratory data to predict the full-scale pump-pipe network parameters

  16. FRACTIONAL CRYSTALLIZATION LABORATORY TESTS WITH SIMULATED TANK WASTE

    International Nuclear Information System (INIS)

    HERTING DL

    2007-01-01

    Results are presented for several simulated waste tests related to development of the fractional crystallization process. Product salt dissolution rates were measured to support pilot plant equipment design. Evaporation tests were performed to evaluate the effects of organics on slurry behavior and to determine optimum antifoam addition levels. A loss-of-power test was performed to support pilot plant accident scenario analysis. Envelope limit tests were done to address variations in feed composition

  17. Decomposition Technology Development of Organic Component in a Decontamination Waste Solution

    International Nuclear Information System (INIS)

    Jung, Chong Hun; Oh, W. Z.; Won, H. J.; Choi, W. K.; Kim, G. N.; Moon, J. K.

    2007-11-01

    Through the project of 'Decomposition Technology Development of Organic Component in a Decontamination Waste Solution', the followings were studied. 1. Investigation of decontamination characteristics of chemical decontamination process 2. Analysis of COD, ferrous ion concentration, hydrogen peroxide concentration 3. Decomposition tests of hardly decomposable organic compounds 4. Improvement of organic acid decomposition process by ultrasonic wave and UV light 5. Optimization of decomposition process using a surrogate decontamination waste solution

  18. Decomposition Technology Development of Organic Component in a Decontamination Waste Solution

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chong Hun; Oh, W. Z.; Won, H. J.; Choi, W. K.; Kim, G. N.; Moon, J. K

    2007-11-15

    Through the project of 'Decomposition Technology Development of Organic Component in a Decontamination Waste Solution', the followings were studied. 1. Investigation of decontamination characteristics of chemical decontamination process 2. Analysis of COD, ferrous ion concentration, hydrogen peroxide concentration 3. Decomposition tests of hardly decomposable organic compounds 4. Improvement of organic acid decomposition process by ultrasonic wave and UV light 5. Optimization of decomposition process using a surrogate decontamination waste solution.

  19. 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

  20. Water hyacinth for phytoremediation of radioactive waste simulate contaminated with cesium and cobalt radionuclides

    International Nuclear Information System (INIS)

    Saleh, H.M.

    2012-01-01

    Highlights: ► Phytoremediation of radioactive wastes containing 137 Cs and 60 Co radionuclides. ► Using water hyacinth for radioactive waste treatment. ► Bioaccumulation of radionuclides from radioactive waste streams. ► Factors affecting bioaccumulation of 137 Cs and 60 Co using floating plants. - Abstract: Phytoremediation is based on the capability of plants to remove hazardous contaminants present in the environment. This study aimed to demonstrate some factors controlling the phytoremediation efficiency of live floating plant, water hyacinth (Eichhornia crassipes), towards the effluents contaminated with 137 Cs and/or 60 Co. Cesium has unknown vital biological role for plant while cobalt is one of the essential trace elements required for plant. The main idea of this work i.e. using undesirable species, water hyacinth, in purification of radiocontaminated aqueous solutions has been receiving much attention. The controlling factors such as radioactivity concentration, pH values, the amount of biomass and the light were studied. The uptake rate of radiocesium from the simulated waste solution is inversely proportional to the initial activity content and directly proportional to the increase in mass of plant and sunlight exposure. A spiked solution of pH ≈ 4.9 was found to be the suitable medium for the treatment process. The uptake efficiency of 137 Cs present with 60 Co in mixed solution was higher than if it was present separately. On the contrary, uptake of 60 Co is affected negatively by the presence of 137 Cs in their mixed solution. Sunlight is the most required factor for the plant vitality and radiation resistance. The results of the present study indicated that water hyacinth may be a potential candidate plant of high concentration ratios (CR) for phytoremediation of radionuclides such as 137 Cs and 60 Co.

  1. Water hyacinth for phytoremediation of radioactive waste simulate contaminated with cesium and cobalt radionuclides

    Energy Technology Data Exchange (ETDEWEB)

    Saleh, H.M., E-mail: hosamsaleh70@yahoo.com [Radioisotope Department, Nuclear Research Center, Atomic Energy Authority, Dokki 12311, Giza (Egypt)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer Phytoremediation of radioactive wastes containing {sup 137}Cs and {sup 60}Co radionuclides. Black-Right-Pointing-Pointer Using water hyacinth for radioactive waste treatment. Black-Right-Pointing-Pointer Bioaccumulation of radionuclides from radioactive waste streams. Black-Right-Pointing-Pointer Factors affecting bioaccumulation of {sup 137}Cs and {sup 60}Co using floating plants. - Abstract: Phytoremediation is based on the capability of plants to remove hazardous contaminants present in the environment. This study aimed to demonstrate some factors controlling the phytoremediation efficiency of live floating plant, water hyacinth (Eichhornia crassipes), towards the effluents contaminated with {sup 137}Cs and/or {sup 60}Co. Cesium has unknown vital biological role for plant while cobalt is one of the essential trace elements required for plant. The main idea of this work i.e. using undesirable species, water hyacinth, in purification of radiocontaminated aqueous solutions has been receiving much attention. The controlling factors such as radioactivity concentration, pH values, the amount of biomass and the light were studied. The uptake rate of radiocesium from the simulated waste solution is inversely proportional to the initial activity content and directly proportional to the increase in mass of plant and sunlight exposure. A spiked solution of pH Almost-Equal-To 4.9 was found to be the suitable medium for the treatment process. The uptake efficiency of {sup 137}Cs present with {sup 60}Co in mixed solution was higher than if it was present separately. On the contrary, uptake of {sup 60}Co is affected negatively by the presence of {sup 137}Cs in their mixed solution. Sunlight is the most required factor for the plant vitality and radiation resistance. The results of the present study indicated that water hyacinth may be a potential candidate plant of high concentration ratios (CR) for phytoremediation of radionuclides

  2. GENERAL REQUIREMENTS FOR SIMULATION MODELS IN WASTE MANAGEMENT

    International Nuclear Information System (INIS)

    Miller, Ian; Kossik, Rick; Voss, Charlie

    2003-01-01

    Most waste management activities are decided upon and carried out in a public or semi-public arena, typically involving the waste management organization, one or more regulators, and often other stakeholders and members of the public. In these environments, simulation modeling can be a powerful tool in reaching a consensus on the best path forward, but only if the models that are developed are understood and accepted by all of the parties involved. These requirements for understanding and acceptance of the models constrain the appropriate software and model development procedures that are employed. This paper discusses requirements for both simulation software and for the models that are developed using the software. Requirements for the software include transparency, accessibility, flexibility, extensibility, quality assurance, ability to do discrete and/or continuous simulation, and efficiency. Requirements for the models that are developed include traceability, transparency, credibility/validity, and quality control. The paper discusses these requirements with specific reference to the requirements for performance assessment models that are used for predicting the long-term safety of waste disposal facilities, such as the proposed Yucca Mountain repository

  3. 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

  4. 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.)

  5. 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)

  6. WASTES: Wastes system transportation and economic simulation: Version 2, Programmer's reference manual

    International Nuclear Information System (INIS)

    Buxbaum, M.E.; Shay, M.R.

    1986-11-01

    The WASTES Version II (WASTES II) Programmer's Reference Manual was written to document code development activities performed under the Monitored Retrievable Storage (MRS) Program at Pacific Northwest Laboratory (PNL). The manual will also serve as a valuable tool for programmers involved in maintenance of and updates to the WASTES II code. The intended audience for this manual are experienced FORTRAN programmers who have only a limited knowledge of nuclear reactor operation, the nuclear fuel cycle, or nuclear waste management practices. It is assumed that the readers of this manual have previously reviewed the WASTES II Users Guide published as PNL Report 5714. The WASTES II code is written in FORTRAN 77 as an extension to the SLAM commercial simulation package. The model is predominately a FORTRAN based model that makes extensive use of the SLAM file maintenance and time management routines. This manual documents the general manner in which the code is constructed and the interactions between SLAM and the WASTES subroutines. The functionality of each of the major WASTES subroutines is illustrated with ''block flow'' diagrams. The basic function of each of these subroutines, the algorithms used in them, and a discussion of items of particular note in the subroutine are reviewed in this manual. The items of note may include an assumption, a coding practice that particularly applies to a subroutine, or sections of the code that are particularly intricate or whose mastery may be difficult. The appendices to the manual provide extensive detail on the use of arrays, subroutines, included common blocks, parameters, variables, and files

  7. 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

  8. Leaching of vitrified high-level-active-waste in a near reality simulated repository system

    International Nuclear Information System (INIS)

    Froeschen, W.; Wolf, G.K.

    1987-01-01

    In the FRG it is planned to vitrify the high level waste from spent fuel reprocessing and to dispose of in a salt-mine. If water penetrates into the repository a highly corrosive brine (Q-brine) will be formed and radioactive material may be leached from the glasses and transported to human environment. The corrosion system of brine, corroded steel containers of the vitrified waste, and waste-glasses was investigated under near reality conditions. Experiments in hydrothermal environment were carried out including gamma radiation of the waste-glasses and ceramic In Can Lining between glasses and metallic containments. Screening experiments by application of external cobalt-gamma-radiation showed no principal changes in leaching behaviour of simulate glasses compared to leaching without radiation. Radiation effects result in pH changes mainly which are diminished by buffer capacity of Q-brine. Lining of steel containments with ceramic fleece does not reduce leaching but retards solution of Mo and Sr into brine. Decreasing of elements Sr, Cs and Mo in the near surface area of the glass and increasing of Zr and Ti has been found to be enhanced considerably in presence of canister corrosion products in Q-brine as well as in NaCl-leaching solution. (orig.) With 13 refs., 22 figs [de

  9. removal of hazardous pollutants from industrial waste solutions using membrane techniques

    International Nuclear Information System (INIS)

    Selim, Y.T.M.

    2001-01-01

    the removal of hazardous pollutants from industrial waste solutions is of essential demand field for both scientific and industrial work. the present work includes detailed studies on the possible use of membrane technology especially liquid emulsion membrane for the removal of hazardous pollutants such as; cadmium , cobalt , lead, copper and uranium from different industrial waste solution . this research can be applied for mixed waste problems. the work carried out in this thesis is presented in three main chapters, namely introduction, experimental and results and discussion

  10. Removal of radionuclides from partitioning waste solutions by adsorption and catalytic oxidation methods

    Energy Technology Data Exchange (ETDEWEB)

    Yamagishi, Isao; Yamaguchi, Isoo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kubota, Masumitsu [Research Organization for Information Science and Technology (RIST), Tokai, Ibaraki (Japan)

    2000-09-01

    Adsorption of radionuclides with inorganic ion exchangers and catalytic oxidation of a complexant were studied for the decontamination of waste solutions generated in past partitioning tests with high-level liquid waste. Granulated ferrocyanide and titanic acid were used for adsorption of Cs and Sr, respectively, from an alkaline solution resulting from direct neutralization of an acidic waste solution. Both Na and Ba inhibited adsorption of Sr but Na did not that of Cs. These exchangers adsorbed Cs and Sr at low concentration with distribution coefficients of more than 10{sup 4}ml/g from 2M Na solution of pH11. Overall decontamination factors (DFs) of Cs and total {beta} nuclides exceeded 10{sup 5} and 10{sup 3}, respectively, at the neutralization-adsorption step of actual waste solutions free from a complexant. The DF of total {alpha} nuclides was less than 10{sup 3} for a waste solution containing diethylenetriaminepentaacetic acid (DTPA). DTPA was rapidly oxidized by nitric acid in the presence of a platinum catalyst, and radionuclides were removed as precipitates by neutralization of the resultant solution. The DF of {alpha} nuclides increased to 8x10{sup 4} by addition of the oxidation step. The DFs of Sb and Co were quite low through the adsorption step. A synthesized Ti-base exchanger (PTC) could remove Sb with the DF of more than 4x10{sup 3}. (author)

  11. Chemical hazards from decontamination solutions in low level waste

    International Nuclear Information System (INIS)

    Leventhal, L.; Miller, A.; Turney, J.; Naughton, M.; IMPELL Corp., Walnut Creek, CA; Electric Power Research Inst., Palo Alto, CA)

    1985-01-01

    Recent regulations are focussing more attention on the non-radioactive matrix materials associated with radioactive wastes. Decontamination of operating facilities is becoming a more significant source of low-level waste. This study reviewed the chemical and biological hazards of over 50 decontamination processes. Seventeen of the most prominent hard and soft decontamination processes were examined in detail. The chemical and biological hazards of these seventeen are presented in this paper. These hazards influence the choice of radwaste processing and packaging operations and methods. Federal, state and local regulations further impact on operations and waste disposal. Hazards to personnel, in plant and off-site, resulting from the decontamination cycle are evaluated. 1 fig., 5 tabs

  12. Rheological evaluation of simulated neutralized current acid waste - transuranics

    International Nuclear Information System (INIS)

    Fow, C.L.; McCarthy, D.; Thornton, G.T.; Scott, P.A.; Bray, L.A.

    1986-09-01

    At the Hanford Plutonium and Uranium Extraction Plant (PUREX), in Richland, Washington, plutonium and uranium products are recovered from irradiated fuel by a solvent extraction process. A byproduct of this process is an aqueous waste stream that contains fission products. This waste stream, called current acid waste (CAW), is chemically neutralized and stored in double shell tanks (DSTs) on the Hanford Site. This neutralized current acid waste (NCAW) will be transported by pipe to B-Plant, a processing plant located nearby. In B-Plant, the transuranic (TRU) elements in NCAW are separated from the non-TRU elements. The majority of the TRU elements in NCAW are in the solids. Therefore, the primary processing operation is to separate the NCAW solids (NCAW-TRU) from the NCAW liquid. These two waste streams will be pumped to suitable holding tanks before being further processed for permanent disposal. To ensure that the retrieval and transportation of NCAW and NCAW-TRU are successful, researchers at Pacific Northwest Laboratory (PNL) evaluated the rheological and transport properties of the slurries. This evaluation had two phases. First, researchers conducted laboratory rheological evaluations of simulated NCAW and NCAW-TRU. The results of these evaluations were then correlated with classical rheological models and scaled up to predict the performance that is likely to occur in the full-scale system. This scale-up procedure has already been successfully used to predict the critical transport properties of a slurry (Neutralized Cladding Removal Waste) with rheological properties similar to those displayed by NCAW and NCAW-TRU

  13. Molecular Dynamics Simulation Studies of Caffeine Aggregation in Aqueous Solution

    OpenAIRE

    Tavagnacco, Letizia; Schnupf, Udo; Mason, Philip E.; Saboungi, Marie-Louise; Cesàro, Attilio; Brady, John W.

    2011-01-01

    Molecular dynamics simulations were carried out on a system of eight independent caffeine molecules in a periodic box of water at 300 K, representing a solution near the solubility limit for caffeine at room temperature, using a newly-developed CHARMM-type force field for caffeine in water. Simulations were also conducted for single caffeine molecules in water using two different water models (TIP3P and TIP4P). Water was found to structure in a complex fashion around the planar caffeine molec...

  14. Pitting growth rate in carbon steel exposed to simulated radioactive waste

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1995-01-01

    Dilute high-level radioactive waste slurries can induce pitting corrosion in carbon steel tanks in which such waste is stored and processed. The waste is normally maintained with closely monitored nitrite and hydroxide concentrations known to prevent the initiation of pitting. Coupon immersion are being conducted in laboratory simulants of waste to determine the probability and growth rate of pitting in steel in the event of below-limits nitrite concentrations. Sets of about 36 carbon steel coupons have been immersed in known corrosive conditions (nitrite < 5% of the established limit) at a temperature of 50 C. Three sets have been removed from testing after 64, 150, and 350 days of immersion. The long immersion times introduced variability in the exposure conditions due to the evaporation and replenishment of solution. The deepest corrosive attack was measured one each coupon by optical microscopy. The deepest pits were ranked and analyzed as a type 1 extreme value distribution to extrapolate from the coupon population to the maximum pit depths in a waste tank structure. The data were compared to a power law for pit growth, although the deepest pits did not increase monotonically with time in the limited data set

  15. Pitting growth rate in carbon steel exposed to simulated radioactive waste

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1996-06-01

    Dilute high-level radioactive waste slurries can induce pitting corrosion in carbon steel tanks in which such waste is stored and processed. The waste is normally maintained with closely monitored nitrite and hydroxide concentrations known to prevent the initiation of pitting. Coupon immersion tests are being conducted in laboratory simulants of waste to determine the probability and growth rate of pitting in steel in the event of out-of-limits nitrite concentrations. Sets of about 36 carbon steel coupons have been immersed in known corrosive conditions (nitrite < 5 per cent of the established limit) at a temperature of 50 degrees C. Three sets have been removed from testing after 64, 150, and 350 days of immersion. The long immersion times introduced variability in the exposure conditions due to the evaporation and replenishment of solution. The deepest corrosive attack was measured on each coupon by optical microscopy. The deepest pits were ranked and analyzed as a type 1 extreme value distribution to extrapolate from the coupon population to the maximum expected pit depths in a waste tank structure. The data were compared to a power law for pit growth, although the deepest pits did not increase monotonically with time in the limited data set

  16. Waste-to-energy plants - a solution for a cleaner future

    International Nuclear Information System (INIS)

    Pfeiffer, J.

    2007-01-01

    Waste-to-energy plants reduce the municipal solid waste volume by about 80% and convert it into residue. The residue quality naturally depends on the burned waste quality and also on the combustion parameters. Hence, tighter control of the plant can improve the residue quality. The generated combustion energy is regarded as renewable energy and is typically used to feed a turbine to generate electricity. Waste-to-energy furnaces react slowly on changing waste charge, so they are not used for peak load generation. The generated electrical power is a plant by product and is sold as base load generation. Usually the waste is burned on a grate which limits the plant size to about 160,000 tons of waste per year or 20 tons of waste per hour or about 28 MW. More recent technology utilizes fluidized bed combustion, which allows larger plant sizes up to 50 MW. Due to the unknown waste composition and stringent environmental standards involved, waste-to-energy plants employ sophisticated flue gas cleaning devices for emission control. ABB's Performance Monitoring continuously compares actual plant and equipment performance to expected performance. This includes the on-line calculation of the waste calorific heat allowing operator decision support and automated control system responses. Dedicated reports offer detailed data on operations, maintenance and emissions to plant management staff. ABB combustion optimization solutions use model based predictive control techniques to reliably find the most suitable set-points for improving the heat rate and reducing emissions like NO x . (author)

  17. 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

  18. Molecular-dynamics simulations of urea nucleation from aqueous solution.

    Science.gov (United States)

    Salvalaglio, Matteo; Perego, Claudio; Giberti, Federico; Mazzotti, Marco; Parrinello, Michele

    2015-01-06

    Despite its ubiquitous character and relevance in many branches of science and engineering, nucleation from solution remains elusive. In this framework, molecular simulations represent a powerful tool to provide insight into nucleation at the molecular scale. In this work, we combine theory and molecular simulations to describe urea nucleation from aqueous solution. Taking advantage of well-tempered metadynamics, we compute the free-energy change associated to the phase transition. We find that such a free-energy profile is characterized by significant finite-size effects that can, however, be accounted for. The description of the nucleation process emerging from our analysis differs from classical nucleation theory. Nucleation of crystal-like clusters is in fact preceded by large concentration fluctuations, indicating a predominant two-step process, whereby embryonic crystal nuclei emerge from dense, disordered urea clusters. Furthermore, in the early stages of nucleation, two different polymorphs are seen to compete.

  19. Molecular-dynamics simulations of urea nucleation from aqueous solution

    Science.gov (United States)

    Salvalaglio, Matteo; Perego, Claudio; Giberti, Federico; Mazzotti, Marco; Parrinello, Michele

    2015-01-01

    Despite its ubiquitous character and relevance in many branches of science and engineering, nucleation from solution remains elusive. In this framework, molecular simulations represent a powerful tool to provide insight into nucleation at the molecular scale. In this work, we combine theory and molecular simulations to describe urea nucleation from aqueous solution. Taking advantage of well-tempered metadynamics, we compute the free-energy change associated to the phase transition. We find that such a free-energy profile is characterized by significant finite-size effects that can, however, be accounted for. The description of the nucleation process emerging from our analysis differs from classical nucleation theory. Nucleation of crystal-like clusters is in fact preceded by large concentration fluctuations, indicating a predominant two-step process, whereby embryonic crystal nuclei emerge from dense, disordered urea clusters. Furthermore, in the early stages of nucleation, two different polymorphs are seen to compete. PMID:25492932

  20. Process simulation and economic analysis of biodiesel production from waste cooking oil with membrane bioreactor

    Science.gov (United States)

    Abdurakhman, Yuanita Budiman; Putra, Zulfan Adi; Bilad, Muhammad Roil

    2017-10-01

    Pollution and shortage of clean energy supply are among major problems that are caused by rapid population growth. Due to this growth, waste cooking oil is one of the pollution sources. On the other hand, biodiesel appears to be one of the most promising and feasible energy sources as it emits less toxic pollutants and greenhouse gases than petroleum diesel. Thus, biodiesel production using waste cooking oil offers a two-in-one solution to cater pollution and energy issues. However, the conventional biodiesel production process using homogeneous base catalyst and stirred tank reactor is unable to produce high purity of biodiesel from waste cooking oil. It is due its sensitivity to free fatty acid (FFA) content in waste cooking oil and purification difficulties. Therefore, biodiesel production using heterogeneous acid catalyst in membrane reactor is suggested. The product of this process is fatty acid methyl esters (FAME) or biodiesel with glycerol as by-product. This project is aimed to study techno-economic feasibility of biodiesel production from waste cooking oil via heterogeneous acid catalyst in membrane reactor. Aspen HYSYS is used to accomplish this aim. Several cases, such as considering different residence times and the production of pharmaceutical (USP) grade glycerol, are evaluated and compared. Economic potential of these cases is calculated by considering capital expenditure, utilities cost, product and by-product sales, as well as raw material costs. Waste cooking oil, inorganic pressure-driven membrane and WAl is used as raw material, type of membrane and heterogeneous acid catalyst respectively. Based on literature data, FAME yield formulation is developed and used in the reactor simulation. Simulation results shows that economic potential increases by 30% if pharmaceutical (USP) grade glycerol is produced regardless the residence time of the reactor. In addition, there is no significant effect of residence time on the economic potential.

  1. Hanford tank waste simulants specification and their applicability for the retrieval, pretreatment, and vitrification processes

    Energy Technology Data Exchange (ETDEWEB)

    GR Golcar; NG Colton; JG Darab; HD Smith

    2000-04-04

    A wide variety of waste simulants were developed over the past few years to test various retrieval, pretreatment and waste immobilization technologies and unit operations. Experiments can be performed cost-effectively using non-radioactive waste simulants in open laboratories. This document reviews the composition of many previously used waste simulants for remediation of tank wastes at the Hanford reservation. In this review, the simulants used in testing for the retrieval, pretreatment, and vitrification processes are compiled, and the representative chemical and physical characteristics of each simulant are specified. The retrieval and transport simulants may be useful for testing in-plant fluidic devices and in some cases for filtration technologies. The pretreatment simulants will be useful for filtration, Sr/TRU removal, and ion exchange testing. The vitrification simulants will be useful for testing melter, melter feed preparation technologies, and for waste form evaluations.

  2. Hanford tank waste simulants specification and their applicability for the retrieval, pretreatment, and vitrification processes

    International Nuclear Information System (INIS)

    GR Golcar; NG Colton; JG Darab; HD Smith

    2000-01-01

    A wide variety of waste simulants were developed over the past few years to test various retrieval, pretreatment and waste immobilization technologies and unit operations. Experiments can be performed cost-effectively using non-radioactive waste simulants in open laboratories. This document reviews the composition of many previously used waste simulants for remediation of tank wastes at the Hanford reservation. In this review, the simulants used in testing for the retrieval, pretreatment, and vitrification processes are compiled, and the representative chemical and physical characteristics of each simulant are specified. The retrieval and transport simulants may be useful for testing in-plant fluidic devices and in some cases for filtration technologies. The pretreatment simulants will be useful for filtration, Sr/TRU removal, and ion exchange testing. The vitrification simulants will be useful for testing melter, melter feed preparation technologies, and for waste form evaluations

  3. Waste Isolation Pilot Plant simulated RH TRU waste experiments: Data and interpretation pilot

    International Nuclear Information System (INIS)

    Molecke, M.A.; Argueello, G.J.; Beraun, R.

    1993-04-01

    The simulated, i.e., nonradioactive remote-handled transuranic waste (RH TRU) experiments being conducted underground in the Waste Isolation Pilot Plant (WIPP) were emplaced in mid-1986 and have been in heated test operation since 9/23/86. These experiments involve the in situ, waste package performance testing of eight full-size, reference RH TRU containers emplaced in horizontal, unlined test holes in the rock salt ribs (walls) of WIPP Room T. All of the test containers have internal electrical heaters; four of the test emplacements were filled with bentonite and silica sand backfill materials. We designed test conditions to be ''near-reference'' with respect to anticipated thermal outputs of RH TRU canisters and their geometrical spacing or layout in WIPP repository rooms, with RH TRU waste reference conditions current as of the start date of this test program. We also conducted some thermal overtest evaluations. This paper provides a: detailed test overview; comprehensive data update for the first 5 years of test operations; summary of experiment observations; initial data interpretations; and, several status; experimental objectives -- how these tests support WIPP TRU waste acceptance, performance assessment studies, underground operations, and the overall WIPP mission; and, in situ performance evaluations of RH TRU waste package materials plus design details and options. We provide instrument data and results for in situ waste container and borehole temperatures, pressures exerted on test containers through the backfill materials, and vertical and horizontal borehole-closure measurements and rates. The effects of heat on borehole closure, fracturing, and near-field materials (metals, backfills, rock salt, and intruding brine) interactions were closely monitored and are summarized, as are assorted test observations. Predictive 3-dimensional thermal and structural modeling studies of borehole and room closures and temperature fields were also performed

  4. Results from simulated contact-handled transuranic waste experiments at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Molecke, M.A.; Sorensen, N.R.; Krumhansl, J.L.

    1993-01-01

    We conducted in situ experiments with nonradioactive, contact-handled transuranic (CH TRU) waste drums at the Waste Isolation Pilot Plant (WIPP) facility for about four years. We performed these tests in two rooms in rock salt, at WIPP, with drums surrounded by crushed salt or 70 wt % salt/30 wt % bentonite clay backfills, or partially submerged in a NaCl brine pool. Air and brine temperatures were maintained at ∼40C. These full-scale (210-L drum) experiments provided in situ data on: backfill material moisture-sorption and physical properties in the presence of brine; waste container corrosion adequacy; and, migration of chemical tracers (nonradioactive actinide and fission product simulants) in the near-field vicinity, all as a function of time. Individual drums, backfill, and brine samples were removed periodically for laboratory evaluations. Waste container testing in the presence of brine and brine-moistened backfill materials served as a severe overtest of long-term conditions that could be anticipated in an actual salt waste repository. We also obtained relevant operational-test emplacement and retrieval experience. All test results are intended to support both the acceptance of actual TRU wastes at the WIPP and performance assessment data needs. We provide an overview and technical data summary focusing on the WIPP CH TRU envirorunental overtests involving 174 waste drums in the presence of backfill materials and the brine pool, with posttest laboratory materials analyses of backfill sorbed-moisture content, CH TRU drum corrosion, tracer migration, and associated test observations

  5. Household Hazardous Waste: Everyone's Problem--Everyone's Solution.

    Science.gov (United States)

    Evenson, Linda

    1985-01-01

    Examines the household hazardous waste problem, addressing several areas related to regulation, disposal, and control. Also gives a list of safer alternatives for household cleaners/disinfectants, paint products, and pesticides. Indicates that individuals can collectively make a difference in public exposure by changing purchases and practices.…

  6. Construction materials as a waste management solution for cellulose sludge.

    Science.gov (United States)

    Modolo, R; Ferreira, V M; Machado, L M; Rodrigues, M; Coelho, I

    2011-02-01

    Sustainable waste management system for effluents treatment sludge has been a pressing issue for pulp and paper sector. Recycling is always recommended in terms of environmental sustainability. Following an approach of waste valorisation, this work aims to demonstrate the technical viability of producing fiber-cement roof sheets incorporating cellulose primary sludge generated on paper and pulp mills. From the results obtained with preliminary studies it was possible to verify the possibility of producing fiber-cement sheets by replacing 25% of the conventional used virgin long fiber by primary effluent treatment cellulose sludge. This amount of incorporation was tested on an industrial scale. Environmental parameters related to water and waste, as well as tests for checking the quality of the final product was performed. These control parameters involved total solids in suspension, dissolved salts, chlorides, sulphates, COD, metals content. In the product, parameters like moisture, density and strength were controlled. The results showed that it is possible to replace the virgin long fibers pulp by primary sludge without impacts in final product characteristics and on the environment. This work ensures the elimination of significant waste amounts, which are nowadays sent to landfill, as well as reduces costs associated with the standard raw materials use in the fiber-cement industrial sector. Copyright © 2010 Elsevier Ltd. All rights reserved.

  7. Construction materials as a waste management solution for cellulose sludge

    International Nuclear Information System (INIS)

    Modolo, R.; Ferreira, V.M.; Machado, L.M.; Rodrigues, M.; Coelho, I.

    2011-01-01

    Sustainable waste management system for effluents treatment sludge has been a pressing issue for pulp and paper sector. Recycling is always recommended in terms of environmental sustainability. Following an approach of waste valorisation, this work aims to demonstrate the technical viability of producing fiber-cement roof sheets incorporating cellulose primary sludge generated on paper and pulp mills. From the results obtained with preliminary studies it was possible to verify the possibility of producing fiber-cement sheets by replacing 25% of the conventional used virgin long fiber by primary effluent treatment cellulose sludge. This amount of incorporation was tested on an industrial scale. Environmental parameters related to water and waste, as well as tests for checking the quality of the final product was performed. These control parameters involved total solids in suspension, dissolved salts, chlorides, sulphates, COD, metals content. In the product, parameters like moisture, density and strength were controlled. The results showed that it is possible to replace the virgin long fibers pulp by primary sludge without impacts in final product characteristics and on the environment. This work ensures the elimination of significant waste amounts, which are nowadays sent to landfill, as well as reduces costs associated with the standard raw materials use in the fiber-cement industrial sector.

  8. Selective partitioning of mercury from co-extracted actinides in a simulated acidic ICPP waste stream

    International Nuclear Information System (INIS)

    Brewer, K.N.; Herbst, R.S.; Tranter, T.J.

    1995-01-01

    The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/l. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, 203 Hg, was spiked into a simulated solution of SBW containing 1 g/l mercury. Successive extraction batch contacts with the mercury spiked waste simulant and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl 2 from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HNO 3 and 0.077 with 0.25 M Na 2 CO 3 . An experimental flowsheet was designed from the batch contact tests and tested counter-currently using 5.5 cm centrifugal contactors. Results from the counter-current test show that mercury can be removed from the acidic mixed SBW simulant and recovered separately from the actinides

  9. A dynamic simulation model of the Savannah River Site high level waste complex

    International Nuclear Information System (INIS)

    Gregory, M.V.; Aull, J.E.; Dimenna, R.A.

    1994-01-01

    A detailed, dynamic simulation entire high level radioactive waste complex at the Savannah River Site has been developed using SPEEDUP(tm) software. The model represents mass transfer, evaporation, precipitation, sludge washing, effluent treatment, and vitrification unit operation processes through the solution of 7800 coupled differential and algebraic equations. Twenty-seven discrete chemical constituents are tracked through the unit operations. The simultaneous simultaneous simulation of concurrent batch and continuous processes is achieved by several novel, customized SPEEDUP(tm) algorithms. Due to the model's computational burden, a high-end work station is required: simulation of a years operation of the complex requires approximately three CPU hours on an IBM RS/6000 Model 590 processor. The model will be used to develop optimal high level waste (HLW) processing strategies over a thirty year time horizon. It will be employed to better understand the dynamic inter-relationships between different HLW unit operations, and to suggest strategies that will maximize available working tank space during the early years of operation and minimize overall waste processing cost over the long-term history of the complex. Model validation runs are currently underway with comparisons against actual plant operating data providing an excellent match

  10. Behavior of mercury and iodine during vitrification of simulated alkaline Purex waste

    International Nuclear Information System (INIS)

    Holton, L.K.

    1981-09-01

    Current plans indicate that the high-level wastes stored at the Savannah River Plant will be solidified by vitrification. The behavior of mercury and iodine during the vitrification process is of concern because: mercury is present in the waste in high concentrations (0.1 to 2.8 wt%); mercury will react with iodine and the other halogens present in the waste during vitrification and; the mercury compounds formed will be volatilized from the vitrification process placing a high particulate load in the vitrification system off-gas. Twelve experiments were completed to study the behavior of mercury during vitrification of simulated SRP Purex waste. The mercury was completely volatized from the vitrification system in all experiments. The mercury reacted with iodine, chlorine and oxygen to form a fine particulate solid. Quantitative recovery of mercury compounds formed in the vitrification system off-gas was not possible due to high (37 to 90%) deposition of solids in the off-gas piping. The behavior of mercury and iodine was most strongly influenced by the vitrification system atmosphere. During experiments performed in which the oxygen content of the vitrification system atmosphere was low (< 1 vol%); iodine retention in the glass product was 27 to 55%, the mercury composition of the solids recovered from the off-gas scrub solutions was 75 to 85 wt%, and a small quantity of metallic mercury was recovered from the off-gas scrub solution. During experiments performed in which the oxygen content of the vitrification system atmosphere was high (20 vol%), iodide retention in the glass product was 3 to 15%, the mercury composition of the solids recovered from the off-gas scrub solutions was 60 to 80 wt%, and very little metallic mercury was recovered from the off-gas scrub solution

  11. Steel corrosion resistance in model solutions and reinforced mortar containing wastes

    NARCIS (Netherlands)

    Koleva, D.A.; Van Breugel, K.

    2012-01-01

    This work reports on the corrosion resistance of steel in alkaline model solutions and in cement-based materials (mortar). The model solutions and the mortar specimens were Ordinary Portland Cement (OPC) based. Further, hereby discussed is the implementation of an eco-friendly approach of waste

  12. Durability of cemented waste in repository and under simulated conditions

    International Nuclear Information System (INIS)

    Dragolici, F.; Nicu, M.; Lungu, L.; Turcanu, C.; Rotarescu, Gh.

    2000-01-01

    The Romanian Radioactive Waste National Repository for low level and intermediate level radioactive waste was built in Baita - Bihor county, in an extinct uranium exploitation. The site is at 840 m above sea level and the host rock is crystalline with a low porosity, a good chemical homogeneity and impermeability, keeping these qualities over a considerable horizontal and vertical spans. To obtain the experimental data necessary for the waste form and package characterization together with the back-filling material behaviour in the repository environment, a medium term research programme (1996 - 2010) was implemented. The purpose of this experimental programme is to obtain a part of the data base necessary for the approach of medium and long term assessment of the safety and performance of Baita - Bihor Repository. The programme will provide: a deeper knowledge of the chemical species and reaction mechanisms, the structure, properties and performances of the final products. For safety reasons the behaviour of waste package, which is a main barrier, must be properly known in terms of long term durability in real repository conditions. Characterization of the behaviour includes many interactions between the waste package itself and the surrounding near field conditions such as mineralogy, hydrogeology and groundwater chemistry. To obtain a more deeper knowledge of the species and physical-chemical reactions participating in the matrix formation, as well as their future behaviour during the disposal period, a thorough XRD study started in 1998. For Romanian Radioactive Waste National Repository (DNDR) Baita - Bihor the following steps are planned for the conditioned waste matrix characterization in simulated and real conditions: - preparation and characterization of normal reference matrices based on different cement formulations; - preparation of reference simulated sludge cemented matrices containing iron hydroxide and iron phosphate; - selection of real and

  13. Simulated HLLW compositions for cold test of waste management development

    International Nuclear Information System (INIS)

    Banba, Tsunetaka; Kimura, Hideo; Kamizono, Hiroshi; Tashiro, Shingo

    1982-07-01

    Three grades of simulated high-level liquid waste (HLLW)-JW-A, JW-B, and JW-C - were proposed to be used respectively according to stages of various cold tests for safety assessment of HLW management. The composition of HLLW was estimated taking into account the spectrum of fission products and actinides, waste volume, corrosion products, and chemical additives. One of conditions, the spectrum of fission products and actinides of LWR spent fuels, was calculated by DCHAIN-code. Fuel burn-up of 28,000 MWD/tUO 2 and 33,000 MWD/tUO 2 were adopted as normal and maximum values of Japanese LWR power plants. The other conditions were estimated using the data obtained at Marcoule plant in France. (author)

  14. XPS and ion beam scattering studies of leaching in simulated waste glass containing uranium

    International Nuclear Information System (INIS)

    Karim, D.P.; Pronko, P.P.; Marcuso, T.L.M.; Lam, D.J.; Paulikas, A.P.

    1980-01-01

    Glass samples (consisting of 2 mole % UO 3 dissolved in a number of complex borosilicate simulated waste glasses including Battelle 76-68) were leached for varying times in distilled water at 75 0 C. The glass surfaces were examined before and after leaching using x-ray photoemission spectroscopy and back-scattered ion beam profiling. Leached samples showed enhanced surface layer concentrations of several elements including uranium, titanium, zinc, iron and rare earths. An experiment involving the leaching of two glasses in the same vessel showed that the uranium surface enhancement is probably not due to redeposition from solution

  15. Segregation of the elements of the platinum group in a simulated high-level waste glass

    International Nuclear Information System (INIS)

    Mitamura, H.; Banba, T.; Kamizono, H.; Kiriyama, Y.; Kumata, M.; Murakami, T.; Tashiro, S.

    1983-01-01

    Segregation of the elements of the platinum group occurred during vitrification of the borosilicate glass containing 20 wt% simulated high-level waste oxides. The segregated materials were composed of two crystalline phases: one was the solid solution of ruthenium and rhodium dioxides and the other was that of palladium and rhodium metals also with tellurium. The segregated materials were not distributed homogeneously throughout the glass: (i) on the surface of the glass, there occurred palladium, rhodium and tellurium alloy alone; and (ii) at the inner part of the glass, the agglomerates of the two phases were concentrated in one part and dispersed in the other

  16. 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

  17. Compatibility of packaging components with simulant mixed waste

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Dickens, T.G.

    1996-01-01

    The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the US have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (US DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). Based on these national requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program provides a basis to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, the authors present the results of the second phase of this testing program. The first phase screened five liner materials and six seal materials towards four simulant mixed wastes. This phase involved the comprehensive testing of five candidate liner materials to an aqueous Hanford Tank simulant mixed waste. The comprehensive testing protocol involved exposing the respective materials a matrix of four gamma radiation doses (∼ 1, 3, 6, and 40 kGy), three temperatures (18, 50, and 60 C), and four exposure times (7, 14, 28, and 180 days). Following their exposure to these combinations of conditions, the materials were evaluated by measuring five material properties. These properties were specific gravity, dimensional changes, hardness, stress cracking, and mechanical properties

  18. The incineration of solid radioactive waste: a centralized solution

    International Nuclear Information System (INIS)

    Hernborg, G.; Broden, K.; Oehrn, G.

    1985-01-01

    Almost all the combustible low-level β- and γ-radioactive waste from Sweden, and even some waste from German nuclear power plants, is treated in an incineration plant at Studsvik. To date most of the ash has been put into 100-litre drums, which in turn have been put in 200-litre drums with concrete in between. Recently, methods have been developed and equipment installed for homogeneous solidification of the ash into concrete. Over the years since the start-up of the plant in 1976 the incinerator has worked with a high availability factor. Personnel doses and activity releases to the environment are well below limits set by regulatory authorities. (orig.)

  19. Reduction of waste solution volume generated on electrokinetic remediation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gye-Nam; Koo, Dae-Seo; Kim, Seung-Soo; Jeong, Jung-Whan; Han, Gyu-Seong; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this study, for the reduction of volume of metal oxides generated in cathode chamber, the optimum pH of waste electrolyte in cathode chamber were drawn out through several experiments with the manufactured electrokinetic decontamination equipment. Also, the required time to reach to below the clearance concentration level for self- disposal was estimated through experiments using the manufactured electrokinetic decontamination equipment. A diagram of soil decontamination process for the removal of uranium from contaminated soil was drawn out. The optimum pH of waste electrolyte in cathode chamber for the reduction of volume of metal oxides was below 2.35. Also, when the initial uranium concentration of the soils were 7-20 Bq/g, the required times to reach to below the clearance concentration level for self- disposal were 25-40 days. A diagram of soil decontamination process for the removal of uranium from contaminated soil was drawn out.

  20. Density of simulated americium/curium melter feed solution

    International Nuclear Information System (INIS)

    Rudisill, T.S.

    1997-01-01

    Vitrification will be used to stabilize an americium/curium (Am/Cm) solution presently stored in F-Canyon for eventual transport to Oak Ridge National Laboratory and use in heavy isotope production programs. Prior to vitrification, a series of in-tank oxalate precipitation and nitric/oxalic acid washes will be used to separate these elements and lanthanide fission products from the bulk of the uranium and metal impurities present in the solution. Following nitric acid dissolution and oxalate destruction, the solution will be denitrated and evaporated to a dissolved solids concentration of approximately 100 g/l (on an oxide basis). During the Am/Cm vitrification, an airlift will be used to supply the concentrated feed solution to a constant head tank which drains through a filter and an in-line orifice to the melter. Since the delivery system is sensitive to the physical properties of the feed, a simulated solution was prepared and used to measure the density as a function of temperature between 20 to 70 degrees C. The measured density decreased linearly at a rate of 0.0007 g/cm3/degree C from an average value of 1.2326 g/cm 3 at 20 degrees C to an average value of 1.1973g/cm 3 at 70 degrees C

  1. Density of simulated americium/curium melter feed solution

    Energy Technology Data Exchange (ETDEWEB)

    Rudisill, T.S.

    1997-09-22

    Vitrification will be used to stabilize an americium/curium (Am/Cm) solution presently stored in F-Canyon for eventual transport to Oak Ridge National Laboratory and use in heavy isotope production programs. Prior to vitrification, a series of in-tank oxalate precipitation and nitric/oxalic acid washes will be used to separate these elements and lanthanide fission products from the bulk of the uranium and metal impurities present in the solution. Following nitric acid dissolution and oxalate destruction, the solution will be denitrated and evaporated to a dissolved solids concentration of approximately 100 g/l (on an oxide basis). During the Am/Cm vitrification, an airlift will be used to supply the concentrated feed solution to a constant head tank which drains through a filter and an in-line orifice to the melter. Since the delivery system is sensitive to the physical properties of the feed, a simulated solution was prepared and used to measure the density as a function of temperature between 20 to 70{degrees} C. The measured density decreased linearly at a rate of 0.0007 g/cm3/{degree} C from an average value of 1.2326 g/cm{sup 3} at 20{degrees} C to an average value of 1.1973g/cm{sup 3} at 70{degrees} C.

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

    OpenAIRE

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

    2012-01-01

    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 precipit...

  3. Community Solutions for Solid Waste Pollution, Level 6. Teacher Guide. Operation Waste Watch.

    Science.gov (United States)

    Virginia State Dept. of Waste Management, Richmond. Div. of Litter & Recycling.

    Operation Waste Watch is a series of seven sequential learning units which addresses the subject of litter control and solid waste management. Each unit may be used in a variety of ways, depending on the needs and schedules of individual schools, and may be incorporated into various social studies, science, language arts, health, mathematics, and…

  4. Cementation of the solid radioactive waste with polymer-cement solutions using the method of impregnation

    International Nuclear Information System (INIS)

    Gorbunova, O.

    2015-01-01

    Cementation of solid radioactive waste (SRW), i.e. inclusion of solid radioactive waste into cement matrix without cavities - is one of the main technological processes used for conditioning low and intermediate level radioactive waste. At FSUE 'Radon' the industrialized method of impregnation has been developed and since 2003 has been using for cementation of solid radioactive waste. The technology is that the polymer-cement solution, having high penetrating properties, is supplied under pressure through a tube to the bottom of the container in which solid radioactive waste has preliminarily been placed. The polymer-cement solution is evenly moving upwards through the channels between the particles of solid radioactive waste, fills the voids in the bulk volume of the waste and hardens, forming a cement compound, the amount of which is equal to the original volume. The aim of the investigation was a selection of a cement solution suitable for SRW impregnation (including fine particles) without solution depletion and bottom layers stuffing. It has been chosen a polymer: PHMG (polyhexamethylene-guanidine), which is a stabilizing and water-retaining component of the cement solution. The experiments confirm that the polymer increases the permeability of the cement solution by a 2-2.5 factor, the viscosity by a 1.2 factor, the stability of the consistency by a 1.5-1.7 factor, and extends the operating range of the W/C ratio to 0.5-1.1. So it is possible to penetrate a volume of SRW bigger by a 1.5-2.0 factor. It has been proved, that PHMG polymer increases strength and frost-resistance of the final compounds by a 1.8-2.7 factor, and contributes to fast strength development at the beginning of hardening and it decreases Cs-137 leashing rate by a 1.5-2 factor

  5. Laboratory plant for the separation of cesium from waste solutions of the PUREX process

    International Nuclear Information System (INIS)

    Richter, M.; Eckert, B.; Riemenschneider, J.; Mallon, C.; Mann, D.

    1983-01-01

    A laboratory plant for the separation of cesium from a fission product waste solution of the fuel reprocessing is described. The plant consists of two stages. In the first stage cesium is adsorbed on ammonium molybdatophosphate (AMP). Then the adsorbent is dissolved. From the solution cesium is adsorbed on a cationic ion exchanger in the second stage. Then AMP can be reproduced from this solution. For the elution of cesium in the second stage a NH 4 NO 3 solution (3 m) is used. Flow sheet, construction and the control device of the plant are described and the results of tests with a model solution are given. (author)

  6. A novel Canadian solution for processing and disposal of mixed liquid wastes

    International Nuclear Information System (INIS)

    Suryanarayan, S.; Husain, A.; Husain, S.; Grey, M.; Elwood, C.; White, T.; Wigle, K.

    2011-01-01

    In 2009, Bruce Power contracted with Kinectrics for the disposal of its accumulated mixed liquid waste (MLW) inventory. The waste consists of solvent, PCB (Poly Chlorinated Biphenyls) and non-PCB contaminated oils and aqueous waste drums. The radioactivity in the wastes is principally due to cobalt-60, cesium-137 and tritium. Historically, MLW drums originating from Canadian utilities were shipped to a licensed US facility for destruction via incineration. This option is relatively expensive considering the significant logistics and destruction costs involved. In addition, restrictions now apply on importation of PCB containing wastes in to the US. Because of this, Kinectrics developed a wholly Canadian solution for the disposal of the MLW. Disposal of Bruce Power's MLW was conceived to be carried out in three phases. Phase 1: Develop an overall plan for disposal of the accumulated wastes, Phase 2: Dispose the PCB oil waste drums (highest priority), and Phase 3: Dispose all other waste drums. Phases 1 & 2 have been completed and Phase 3 is currently underway with 17 drums having been disposed so far. A description of the key activities undertaken to date are described in this paper. This work sets the stage for the future management of MLW based exclusively or largely on disposal within Canada. All key technical, regulatory and logistical issues pertaining to the receipt, handling, processing and shipment of the wastes were addressed. Equipment was installed for basic processing of the incoming wastes. Based on Pathways methodology, it was shown that the wastes can be shipped to unlicensed facilities within Canada without exceeding the 10 μSv per annum exposure to the critical individual. Despite this and for compliance with ALARA, wastes exceeding self-imposed threshold levels of radioactivity will be solidified and shipped for storage as radioactive waste. (author)

  7. A novel Canadian solution for processing and disposal of mixed liquid wastes

    Energy Technology Data Exchange (ETDEWEB)

    Suryanarayan, S.; Husain, A. [Kinectrics Inc., Toronto, ON (Canada); Husain, S.; Grey, M. [Candesco, Toronto, ON (Canada); Elwood, C.; White, T.; Wigle, K. [Bruce Power, Tiverton, ON (Canada)

    2011-07-01

    In 2009, Bruce Power contracted with Kinectrics for the disposal of its accumulated mixed liquid waste (MLW) inventory. The waste consists of solvent, PCB (Poly Chlorinated Biphenyls) and non-PCB contaminated oils and aqueous waste drums. The radioactivity in the wastes is principally due to cobalt-60, cesium-137 and tritium. Historically, MLW drums originating from Canadian utilities were shipped to a licensed US facility for destruction via incineration. This option is relatively expensive considering the significant logistics and destruction costs involved. In addition, restrictions now apply on importation of PCB containing wastes in to the US. Because of this, Kinectrics developed a wholly Canadian solution for the disposal of the MLW. Disposal of Bruce Power's MLW was conceived to be carried out in three phases. Phase 1: Develop an overall plan for disposal of the accumulated wastes, Phase 2: Dispose the PCB oil waste drums (highest priority), and Phase 3: Dispose all other waste drums. Phases 1 & 2 have been completed and Phase 3 is currently underway with 17 drums having been disposed so far. A description of the key activities undertaken to date are described in this paper. This work sets the stage for the future management of MLW based exclusively or largely on disposal within Canada. All key technical, regulatory and logistical issues pertaining to the receipt, handling, processing and shipment of the wastes were addressed. Equipment was installed for basic processing of the incoming wastes. Based on Pathways methodology, it was shown that the wastes can be shipped to unlicensed facilities within Canada without exceeding the 10 μSv per annum exposure to the critical individual. Despite this and for compliance with ALARA, wastes exceeding self-imposed threshold levels of radioactivity will be solidified and shipped for storage as radioactive waste. (author)

  8. HANFORD TANK WASTE OPERATIONS SIMULATOR VERSION DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    ALLEN, G.K.

    2003-01-01

    This document describes the software version controls established for the Hanford Tank Waste Operations Simulator (HTWOS). It defines: the methods employed to control the configuration of HTWOS; the version of each of the 26 separate modules for the version 1.0 of HTWOS; the numbering rules for incrementing the version number of each module; and a requirement to include module version numbers in each case results documentation. Version 1.0 of HTWOS is the first version under formal software version control. HTWOS contains separate revision numbers for each of its 26 modules. Individual module version numbers do not reflect the major release HTWOS configured version number

  9. Stream-simulation experiments for waste-repository investigations

    International Nuclear Information System (INIS)

    Seitz, M.G.

    1980-01-01

    The potential for radionuclide migration by groundwater flow from a breached-water repository depends on the leaching process and on chemical changes that might occur as the radionuclide moves away from the repository. Therefore, migration involves the interactions of leached species with (1) the waste and canister, (2) the engineered barrier, and (3) the geologic materials surrounding the repository. Rather than attempt to synthesize each species and study it individually, another approach is to integrate all species and interactions using stream-simulation experiments. Interactions identified in these studies can then be investigated in detail in simpler experiments

  10. Molecular dynamics simulations of solutions at constant chemical potential

    Science.gov (United States)

    Perego, C.; Salvalaglio, M.; Parrinello, M.

    2015-04-01

    Molecular dynamics studies of chemical processes in solution are of great value in a wide spectrum of applications, which range from nano-technology to pharmaceutical chemistry. However, these calculations are affected by severe finite-size effects, such as the solution being depleted as the chemical process proceeds, which influence the outcome of the simulations. To overcome these limitations, one must allow the system to exchange molecules with a macroscopic reservoir, thus sampling a grand-canonical ensemble. Despite the fact that different remedies have been proposed, this still represents a key challenge in molecular simulations. In the present work, we propose the Constant Chemical Potential Molecular Dynamics (CμMD) method, which introduces an external force that controls the environment of the chemical process of interest. This external force, drawing molecules from a finite reservoir, maintains the chemical potential constant in the region where the process takes place. We have applied the CμMD method to the paradigmatic case of urea crystallization in aqueous solution. As a result, we have been able to study crystal growth dynamics under constant supersaturation conditions and to extract growth rates and free-energy barriers.

  11. Molecular dynamics simulation studies of caffeine aggregation in aqueous solution.

    Science.gov (United States)

    Tavagnacco, Letizia; Schnupf, Udo; Mason, Philip E; Saboungi, Marie-Louise; Cesàro, Attilio; Brady, John W

    2011-09-22

    Molecular dynamics simulations were carried out on a system of eight independent caffeine molecules in a periodic box of water at 300 K, representing a solution near the solubility limit for caffeine at room temperature, using a newly developed CHARMM-type force field for caffeine in water. Simulations were also conducted for single caffeine molecules in water using two different water models (TIP3P and TIP4P). Water was found to structure in a complex fashion around the planar caffeine molecules, which was not sensitive to the water model used. As expected, extensive aggregation of the caffeine molecules was observed, with the molecules stacking their flat faces against one another like coins, with their methylene groups staggered to avoid steric clashes. A dynamic equilibrum was observed between large n-mers, including stacks with all eight solute molecules, and smaller clusters, with the calculated osmotic coefficient being in acceptable agreement with the experimental value. The insensitivity of the results to water model and the congruence with experimental thermodynamic data suggest that the observed stacking interactions are a realistic representation of the actual association mechanism in aqueous caffeine solutions.

  12. Spectrophotometric determination of nitrite in simulated Purex Process solutions

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, I.daC. de; Matsuda, H T; Araujo, B.F. de; Araujo, J.A. de

    1984-01-01

    A spectrophotometric method for nitrite determination in simulated Purex Process solutions is presented, utilizing the Griess reagent for the formation of the coloured azocompound with an absorption maximum at 525 nm. Molar absortivity was 36,262 and the sensitivity of the method 10/sup -6/M for nitrite. The calibration curve is linear in the range of 2 to 30..mu..g NO/sup -//sub 2//25 ml in cells of 1 cm optical path. The method can be used in the presence of uranium up to limits of an U/NO/sup -//sub 2/ ratio of 150. Test solutions were prepared to simulate composition and concentrations as obtained by irradiating standard fuel with a neutro flux of 3.2 x 10/sup 13/ n.s/sup -1/.cm/sup -2/, with a burn-up value of 33,000 Mwd/T and cooling time of two years. Nitrite determinations in these solutions were accurate within limits of 5%.

  13. Partitioning of actinides from high active waste solution of Purex origin counter-current extraction studies using TBP and CMPO

    International Nuclear Information System (INIS)

    Chitnis, R.R.; Dhami, P.S.; Gopalkrishnan, V.; Wattal, P.K.; Ramanujam, A.; Murali, M.S.; Mathur, J.N.; Bauri, A.K.; Chattopadhyay, S.

    2000-10-01

    A solvent extraction scheme has been formulated for the partitioning of actinides from Purex high level waste (HLW). The scheme is based on the results of earlier studies carried out with simulated waste solutions. In the present studies, the scheme was tested with high active waste (HAW) solution generated during the reprocessing of spent fuel from research reactors using laboratory scale mixer-settlers. The proposed process involved two-step extraction using tri-n-butyl phosphate (TBP) and octyl (phenyl)-N,N-diisobutylcarbamolylmethylphosphine oxide (CMPO). In the first step, uranium, neptunium and plutonium were removed from the waste using TBP as extractant. The minor actinides left in the raffinate were extracted using a mixture of CMPO and TBP in the second step. The results showed complete extraction of actinides from the waste solution. Plutonium and neptunium extracted in TBP, were stripped together using a mixture of hydrogen peroxide and ascorbic acid in 2 M nitric acid medium, leaving uranium in the organic phase. Uranium can later be stripped using dilute nitric acid. Actinides extracted in CMPO-TBP phase were stripped using a mixture of formic acid, hydrazine, hydrate and citric acid. The stripping was quantitative in both the stripping runs. An additional extraction step for the preferential recovery of uranium, neptunium and plutonium from the waste solution using TBP is a modification over the conventional Truex process. Selective stripping of neptunium and plutonium from large quantities of uranium. The extraction of uranium using TBP eliminates the possibility of third phase and undesired loading of CMPO-TBP in the following step. Use of citrate-containing strippant allows the recovery of actinides from loaded CMPO-TBP mixture without causing any reflux of the actinides during stripping. The process has been developed with due consideration to minimising the generation of secondary wastes. The proposed strippants are effective even in presence of

  14. Synthesis of magnetic nanoparticles as a draw solute in forward osmosis membrane process for the treatment of radioactive liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Heeman; Lee, Kune Woo; Moon, Jei Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    These wastes contain about 0.3 ∼ 0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40 ∼ 90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. As an emerging technology forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination because FO operates at low or no hydraulic pressures. 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, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe{sub 3}O{sub 4} nanoparticles can be separated from water by an external magnet field easily. If Fe{sub 3}O{sub 4} nanoparticles are coated with highly soluble organic substances, thus they can be used as a draw solute by concurrently generating high osmotic pressure and easy separation. The carboxylated polyglycerol coated Fe{sub 3}O{sub 4} nanoparticles have been successfully synthesized. The nanoparticles were about 50 nm in diameter and showed the good colloidal stability in aqueous solution. The osmolality and osmotic pressure were enough high to be used as a draw solute in FO. For the future work, we will investigate the performance of our magnetic draw solute in FO to remove boron in the simulated liquid waste.

  15. Synthesis of magnetic nanoparticles as a draw solute in forward osmosis membrane process for the treatment of radioactive liquid waste

    International Nuclear Information System (INIS)

    Yang, Heeman; Lee, Kune Woo; Moon, Jei Kwon

    2013-01-01

    These wastes contain about 0.3 ∼ 0.8 wt% of boric acid. It is known that reverse osmosis (RO) membrane can eliminate boron at high pH and boron of 40 ∼ 90% can be removed by RO membrane in pH condition. RO uses hydraulic pressure to oppose, and exceed, the osmotic pressure of an aqueous feed solution containing boric acid. As an emerging technology forward osmosis (FO) has attracted growing interest in wastewater treatment and desalination because FO operates at low or no hydraulic pressures. 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, the challenges of FO still lie in the fabrication of eligible FO membranes and the readily separable draw solutes of high osmotic pressures. Superparamagnetic Fe 3 O 4 nanoparticles can be separated from water by an external magnet field easily. If Fe 3 O 4 nanoparticles are coated with highly soluble organic substances, thus they can be used as a draw solute by concurrently generating high osmotic pressure and easy separation. The carboxylated polyglycerol coated Fe 3 O 4 nanoparticles have been successfully synthesized. The nanoparticles were about 50 nm in diameter and showed the good colloidal stability in aqueous solution. The osmolality and osmotic pressure were enough high to be used as a draw solute in FO. For the future work, we will investigate the performance of our magnetic draw solute in FO to remove boron in the simulated liquid waste

  16. Engineering solutions of environmental problems in organic waste handling

    Science.gov (United States)

    Briukhanov, A. Y.; Vasilev, E. V.; Shalavina, E. V.; Kucheruk, O. N.

    2017-10-01

    This study shows the urgent need to consider modernization of agricultural production in terms of sustainable development, which takes into account environmental implications of intensive technologies in livestock farming. Some science-based approaches are offered to address related environmental challenges. High-end technologies of organic livestock waste processing were substantiated by the feasibility study and nutrient balance calculation. The technologies were assessed on the basis of best available techniques criteria, including measures such as specific capital and operational costs associated with nutrient conservation and their delivery to the plants.

  17. Laboratory-Scale SuperLig 639 Column Tests With Hanford Waste Simulants

    International Nuclear Information System (INIS)

    King, William D.; Spencer, William A.; Bussey, Myra Pettis

    2003-01-01

    This report describes the results of SuperLig 639 column tests conducted at the Savannah River Technology Center (SRTC) in support of the Hanford River Protection Project - Waste Treatment Plant (RPP-WTP). The RPP-WTP contract was awarded to Bechtel National Inc. (BNI) for the design, construction, and initial operation of a plant for the treatment and vitrification of millions of gallons of radioactive waste currently stored in tanks at Hanford, WA. Part of the current treatment process involves the removal of technetium from tank supernate solutions using columns containing SuperLig 639 resin. This report is part of a body of work intended to quantify and optimize the operation of the technetium removal columns with regard to various parameters (such as liquid flow rate, column aspect ratio, resin particle size, loading and elution temperature, etc.). The tests were conducted using nonradioactive simulants of the actual tank waste samples containing rhenium as a surrogate for the technetium in the actual waste. A previous report focused on the impacts of liquid flow rate and column aspect ratio upon performance. More recent studies have focused on the impacts of resin particle size, solution composition, and temperature. This report describes column loading experiments conducted varying temperature and solution composition. Each loading experiment was followed by high temperature elution of the sorbed rhenium. Results from limited testing are also described which were intended to evaluate the physical stability of SuperLig 639 resin during exposure to repeated temperature cycles covering the range of potential processing extremes

  18. Hot-wall corrosion testing of simulated high level nuclear waste

    International Nuclear Information System (INIS)

    Chandler, G.T.; Zapp, P.E.; Mickalonis, J.I.

    1995-01-01

    Three materials of construction for steam tubes used in the evaporation of high level radioactive waste were tested under heat flux conditions, referred to as hot-wall tests. The materials were type 304L stainless steel alloy C276, and alloy G3. Non-radioactive acidic and alkaline salt solutions containing halides and mercury simulated different high level waste solutions stored or processed at the United States Department of Energy's Savannah River Site. Alloy C276 was also tested for corrosion susceptibility under steady-state conditions. The nickel-based alloys C276 and G3 exhibited excellent corrosion resistance under the conditions studied. Alloy C276 was not susceptible to localized corrosion and had a corrosion rate of 0.01 mpy (0.25 μm/y) when exposed to acidic waste sludge and precipitate slurry at a hot-wall temperature of 150 degrees C. Type 304L was susceptible to localized corrosion under the same conditions. Alloy G3 had a corrosion rate of 0.1 mpy (2.5 μm/y) when exposed to caustic high level waste evaporator solution at a hot-wall temperature of 220 degrees C compared to 1.1 mpy (28.0 μ/y) for type 304L. Under extreme caustic conditions (45 weight percent sodium hydroxide) G3 had a corrosion rate of 0.1 mpy (2.5 μm/y) at a hot-wall temperature of 180 degrees C while type 304L had a high corrosion rate of 69.4 mpy (1.8 mm/y)

  19. Separation of technetium from nuclear waste stream simulants. Final report

    International Nuclear Information System (INIS)

    Strauss, S.H.

    1995-01-01

    The author studied liquid anion exchangers, such as Aliquat-336 nitrate, various pyridinium nitrates, and related salts, so that they may be applied toward a specific process for extracting (partitioning) and recovering 99 TcO 4 - from nuclear waste streams. Many of the waste streams are caustic and contain a variety of other ions. For this reason, the author studied waste stream simulants that are caustic and contain appropriate concentrations of selected, relevant ions. Methods of measuring the performance of the exchangers and extractant systems included contact experiments. Batch contact experiments were used to determine the forward and reverse extraction parameters as a function of temperature, contact time, phase ratio, concentration, solvent (diluent), and other physical properties. They were also used for stability and competition studies. Specifically, the author investigated the solvent extraction behavior of salts of perrhenate (ReO 4 - ), a stable (non-radioactive) chemical surrogate for 99 TcO 4 - . Results are discussed for alternate organic solvents; metalloporphyrins, ferrocenes, and N-cetyl pyridium nitrate as alternate extractant salts; electroactive polymers; and recovery of ReO 4 - and TcO 4 -

  20. Separation of technetium from nuclear waste stream simulants. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Strauss, S.H. [Colorado State Univ., Fort Collins, CO (United States). Dept. of Chemistry

    1995-09-11

    The author studied liquid anion exchangers, such as Aliquat-336 nitrate, various pyridinium nitrates, and related salts, so that they may be applied toward a specific process for extracting (partitioning) and recovering {sup 99}TcO{sub 4}{sup {minus}} from nuclear waste streams. Many of the waste streams are caustic and contain a variety of other ions. For this reason, the author studied waste stream simulants that are caustic and contain appropriate concentrations of selected, relevant ions. Methods of measuring the performance of the exchangers and extractant systems included contact experiments. Batch contact experiments were used to determine the forward and reverse extraction parameters as a function of temperature, contact time, phase ratio, concentration, solvent (diluent), and other physical properties. They were also used for stability and competition studies. Specifically, the author investigated the solvent extraction behavior of salts of perrhenate (ReO{sub 4}{sup {minus}}), a stable (non-radioactive) chemical surrogate for {sup 99}TcO{sub 4}{sup {minus}}. Results are discussed for alternate organic solvents; metalloporphyrins, ferrocenes, and N-cetyl pyridium nitrate as alternate extractant salts; electroactive polymers; and recovery of ReO{sub 4}{sup {minus}} and TcO{sub 4}{sup {minus}}.

  1. 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.)

  2. Study of physical properties, gas generation and gas retention in simulated Hanford waste

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pederson, L.R.; Scheele, R.D.

    1993-04-01

    The purpose of this study was to establish the chemical and physical processes responsible for the generation and retention of gases within high-level waste from Tank 101-SY on the Hanford Site. This research, conducted using simulated waste on a laboratory scale, supports the development of mitigation/remediation strategies for Tank 101-SY. Simulated waste formulations are based on actual waste compositions. Selected physical properties of the simulated waste are compared to properties of actual Tank 101-SY waste samples. Laboratory studies using aged simulated waste show that significant gas generation occurs thermally at current tank temperatures (∼60 degrees C). Gas compositions include the same gases produced in actual tank waste, primarily N 2 , N 2 O, and H 2 . Gas stoichiometries have been shown to be greatly influenced by several organic and inorganic constituents within the simulated waste. Retention of gases in the simulated waste is in the form of bubble attachment to solid particles. This attachment phenomenon is related to the presence of organic constituents (HEDTA, EDTA, and citrate) of the simulated waste. A mechanism is discussed that relates the gas bubble/particle interactions to the partially hydrophobic surface produced on the solids by the organic constituents

  3. Development of a freeze-drying process of waste-solution, 2

    International Nuclear Information System (INIS)

    Kondo, Isao; Kawasaki, Takeshi

    1988-01-01

    The waste solution treatment process in Plutonium Conversion Development Facility (PCDF) consists of Evaporation-Condensation and Neutrazation-Agglometation-Precipitation process, which produces the distillate as recovered acid at first step and separates Pu-U element from condenced solution at second step. This process needs many stages to get high decontamination efficiency and then the Evaporator is in very corrosive state because the nitric acid solution is heated over 100 degrees C to be evaporated. So, in PCDF, it was started the development of Freeze-Drying process to waste solution treatment. This process is suitable for a little quantity of the solution including nitric acid as produced in the Microwave Heating method. Moreover the process has high decontamination efficiency and has good performance of equipment. The result of the cold test of Freeze-Drying process with nitric acid is discribed in this paper. (author)

  4. Depleted Hydrocarbon Reservoirs Present a Safe and Practical Burial Solution for Graphite Waste

    International Nuclear Information System (INIS)

    Rahmani, L.

    2016-01-01

    A solution for graphite waste is proposed that combines reliance on thick impermeable host rock that is needed to confine the long-life radioactivity content of most irradiated graphite with low capitalistic and operational unit volume costs that are required to render this bulky waste form manageable. The solution, uniquely applicable to irradiated graphite due to its low dose rates, moderate mechanical strength and light density, consists in three steps: first, graphite is fine-crushed under water; second, it is made in an aqueous suspension; third, the suspension is injected into a deep, disused hydrocarbon reservoir. Each of these steps only involves well mastered techniques. Regulatory changes that may allow this solution to be added to the gamut of available waste routes, geochemical issues, availability of depleted reservoirs and cost projections are presented. (author)

  5. 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.

  6. Rippled shock front solutions for testing hydrodynamic stability simulations

    International Nuclear Information System (INIS)

    Munro, D.H.

    1989-01-01

    The response of a shock front to arbitrary small perturbations can be calculated analytically. Such rippled shock front solutions are useful for determining the accuracy of hydrodynamic simulation codes such as LASNEX [Comments Plasma Phys. Controlled Fusion 2, 51 (1977)], which are used to compute perturbation growth in inertial fusion targets. The LASNEX fractional errors are of order κ 2 L 2 , where κ is the transverse wavenumber of the perturbation, and L is the largest zone dimension. Numerical errors are about 25% for a calculation using 26 zones per transverse wavelength

  7. Simulation of transportation of low enriched uranium solutions

    International Nuclear Information System (INIS)

    Hope, E.P.; Ades, M.J.

    1996-01-01

    A simulation of the transportation by truck of low enriched uranium solutions has been completed for NEPA purposes at the Savannah River Site. The analysis involves three distinct source terms, and establishes the radiological risks of shipment to three possible destinations. Additionally, loading accidents were analyzed to determine the radiological consequences of mishaps during handling and delivery. Source terms were developed from laboratory measurements of chemical samples from low enriched uranium feed materials being stored at SRS facilities, and from manufacturer data on transport containers. The transportation simulations were accomplished over the INTERNET using the DOE TRANSNET system at Sandia National Laboratory. The HIGHWAY 3.3 code was used to analyze routing scenarios, and the RADTRAN 4 code was used to analyze incident free and accident risks of transporting radiological materials. Loading accidents were assessed using the Savannah River Site AXAIR89Q and RELEASE 2 codes

  8. 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

  9. 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

  10. Complexant Identification in Hanford Waste Simulant Sr/TRU Filtrate

    International Nuclear Information System (INIS)

    Bannochie, C.J.

    2003-01-01

    This project was designed to characterize the available multidentate ligand species and metal ion complexes of iron, strontium and manganese formed with the parent chelators, complexing agents and their fragment products. Complex identification was applied to AN-102 and AN-107 filtrate simulants for Hanford waste after an oxidation reaction with sodium permanganate to create a freshly precipitated manganese dioxide solid for adsorption of transuranic elements. Separation efficiency of different ligands was investigated based on the exchange capability of different ion exchange and ion exclusion analytical columns including Dionex IonPac AS-5A, AS-10, AS-11 and AS-6. The elution programs developed with different mobile phase concentrations were based on the change in the effective charge of the anionic species and therefore the retention on the stationary phase. In the present work, qualitative and quantitative assessments of multidentate ligands were investigated. Identification methods for the metal ion complexes responsible for solubilizing Fe, Mn and Sr were applied to aged and fresh simulant waste filtrates. Although concentration measurements of both fresh and 3-week aged filtrates showed that the degradation process occurs mainly due to the harsh chemical environment, it was found that the concentration of iron and manganese did not increase, within the error of the analytical measurements, after three weeks when compared with fresh filtrate

  11. Processing results of 1,800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    The mercury-contaminated rinse solution (INEL waste ID number-sign 123; File 8 waste) was successfully treated at the Idaho National Engineering Laboratory (INEL). This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 (HTRE-3) reactor shield tank. Approximately 1,800 gal of waste was generated and was placed into 33 drums. Each drum contained precipitated sludge material ranging from 1--10 in. in depth, with the average depth of about 2.5 in. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act (RCRA) limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/ml, while the average sludge contamination was about 13,800 pci/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. Because of difficulties in processing, three trials were required to reduce the mercury levels to below the RCRA limit. In the first trial, insufficient filtration of the waste allowed solid particulate produced during pH adjustment to enter into the ion exchange columns and ultimately the waste storage tank. In the second trial, the waste was filtered down to 0.1 μ to remove all solid mercury compounds. However, before filtration could take place, a solid mercury complex dissolved and mercury levels exceeded the RCRA limit after filtration. In the third trial, the waste was filtered through 0.3-A filters and then passed through the S-920 resin to remove the dissolved mercury. The resulting solut

  12. Fundamental Science-Based Simulation of Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin; Khaleel, Mohammad A.

    2010-10-04

    This report presents a hierarchical multiscale modeling scheme based on two-way information exchange. To account for all essential phenomena in waste forms over geological time scales, the models have to span length scales from nanometer to kilometer and time scales from picoseconds to millenia. A single model cannot cover this wide range and a multi-scale approach that integrates a number of different at-scale models is called for. The approach outlined here involves integration of quantum mechanical calculations, classical molecular dynamics simulations, kinetic Monte Carlo and phase field methods at the mesoscale, and continuum models. The ultimate aim is to provide science-based input in the form of constitutive equations to integrated codes. The atomistic component of this scheme is demonstrated in the promising waste form xenotime. Density functional theory calculations have yielded valuable information about defect formation energies. This data can be used to develop interatomic potentials for molecular dynamics simulations of radiation damage. Potentials developed in the present work show a good match for the equilibrium lattice constants, elastic constants and thermal expansion of xenotime. In novel waste forms, such as xenotime, a considerable amount of data needed to validate the models is not available. Integration of multiscale modeling with experimental work is essential to generate missing data needed to validate the modeling scheme and the individual models. Density functional theory can also be used to fill knowledge gaps. Key challenges lie in the areas of uncertainty quantification, verification and validation, which must be performed at each level of the multiscale model and across scales. The approach used to exchange information between different levels must also be rigorously validated. The outlook for multiscale modeling of wasteforms is quite promising.

  13. 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)

  14. Durability of simulated waste glass: effects of pressure and formation of surface layers

    International Nuclear Information System (INIS)

    Wicks, G.G.; Mosley, W.C.; Whitkop, P.G.; Saturday, K.A.

    1981-01-01

    The leaching behavior of simulated Savannah River Plant (SRP) waste glass was studied at elevated pressures and anticipated storage temperatures. An integrated approach, which combined leachate solution analyses with both bulk and surface studies, was used to study the corrosion process. Compositions of leachates were evaluated by colorimetry and atomic absorption. Used in the bulk and surface analyses were optical microscopy, scanning electron microscopy, x-ray energy spectroscopy, wide-angle x-ray, diffraction, electron microprobe analysis, infrared reflectance spectroscopy, electron spectroscopy for chemical analysis, and Auger electron spectroscopy. Results from this study show that there is no significant adverse effect of pressure, up to 1500 psi and 90 0 C, on the chemical durability of simulated SPR waste glass leached for one month in deionized water. In addition, the leached glass surface layer was characterized by an adsorbed film rich in minor constituents from the glass. This film remained on the glass surface even after leaching in relatively alkaline solutions at elevated pressures at 90 0 C for one month. The sample surface area to volume of leachant ratios (SA/V) was 10:1 cm -1 and 1:10 cm -1 . The corrosion mechanisms and surface and subsurface layers produced will be discussed along with the potential importance of these results to repository storage

  15. DEMONSTRATION OF THE NEXT-GENERATION CAUSTIC-SIDE SOLVENT EXTRACTION SOLVENT WITH 2-CM CENTRIFUGAL CONTRACTORS USING TANK 49H WASTE AND WASTE SIMULANT

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R.; Peters, T.; Crowder, M.; Caldwell, T.; Pak, D; Fink, S.; Blessing, R.; Washington, A.

    2011-09-27

    Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. Conclusions from this work include the following. The CSSX process is capable of reducing {sup 137}Cs in high level radioactive waste by a factor of more than 40,000 using five extraction, two scrub, and five strip stages. Tests demonstrated extraction and strip section stage efficiencies of greater than 93% for the Tank 49H waste test and greater than 88% for the simulant waste test. During a test with HLW, researchers processed 39 liters of Tank 49H solution and the waste raffinate had an average decontamination factor (DF) of 6.78E+04, with a maximum of 1.08E+05. A simulant waste solution ({approx}34.5 liters) with an initial Cs concentration of 83.1 mg/L was processed and had an average DF greater than 5.9E+03, with a maximum DF of greater than 6.6E+03. The difference may be attributable to differences in contactor stage efficiencies. Test results showed the solvent can be stripped of cesium and recycled for {approx}25 solvent turnovers without the occurrence of any measurable solvent degradation or negative effects from minor components. Based on the performance of the 12-stage 2-cm apparatus with the Tank 49H HLW, the projected DF for MCU with seven extraction, two scrub, and seven strip stages operating at a nominal efficiency of 90% is {approx}388,000. At 95% stage efficiency, the DF in MCU would be {approx}3.2 million. Carryover of organic solvent in aqueous streams (and aqueous in organic

  16. Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

    Energy Technology Data Exchange (ETDEWEB)

    Kinker, M., E-mail: M.Kinker@iaea.org [International Atomic Energy Agency (IAEA), Vienna (Austria); Avila, R.; Hofman, D., E-mail: rodolfo@facilia.se [FACILIA AB, Stockholm (Sweden); Jova Sed, L., E-mail: jovaluis@gmail.com [Centro Nacional de Seguridad Nuclear (CNSN), La Habana (Cuba); Ledroit, F., E-mail: frederic.ledroit@irsn.fr [IRSN PSN-EXP/SSRD/BTE, (France)

    2013-07-01

    In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

  17. Management of radioactive waste in France-policy, issues, and solutions

    International Nuclear Information System (INIS)

    Tamborini, J.

    1996-01-01

    The French nuclear industry has conducted a study to define a policy and an organization to deal with the waste generated from nuclear power plants, the fuel cycle industries, and medicine, research, and other industrial nuclear applications. This has resulted in the introduction of an organization which, by appropriate and responsible management, can guarantee to protect people and the environment while ensuring industrial effectiveness. The body in charge of waste management in France is the National Radioactive Waste Management Agency (ANDRA) created in 1979. The French policy is based on waste classification and the related solutions for the evacuation of these wastes. High-level and long-lived waste management is regulated by a law passed Dec 30, 1991. The law outlines the research program to be conducted. Three main research objectives are prescribed: 1. reduction of the waste volumes and toxicity (partitioning and transmutation); 2. assessment of the waste isolation properties of deep geologic formations by underground research laboratories; 3. development of solidification processes and storage techniques for long-term interim storage in near-surface facilities. This research will be implemented within a 15 yr period. At present, applications are submitted to the authorities for the construction of underground research laboratories. At the end of this period, reports will be submitted to parliament. It will have to choose among various options. The construction of a deep geologic repository, if this option is chosen, will need the passage of a new law

  18. Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

    International Nuclear Information System (INIS)

    Kinker, M.; Avila, R.; Hofman, D.; Jova Sed, L.; Ledroit, F.

    2013-01-01

    In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

  19. Effect of nitrite concentration on pit depth in carbon steel exposed to simulated radioactive waste

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1997-01-01

    The growth of pits in carbon steel exposed to dilute (0.055 M nitrate-bearing) alkaline salt solutions that simulate radioactive waste was investigated in coupon immersion tests. Most coupons were tested in the as-received condition, with the remainder having been heat treated to produce an oxide film. Nitrite, which is an established pitting inhibitor in these solutions, was present in concentrations from 0 to 0.031 M to 0.16 M; the last concentration is known to prevent pitting initiation in the test solution at the 50 degrees C test temperature. The depths of the deepest pits on coupons of particular exposure conditions were measure microscopically and were analyzed as simple, type 1 extreme value statistical distributions, to predict the deepest expected pit in a radioactive waste tank subject to the test conditions. While the growth rate of pits could not be established from these tests, the absolute value of the deepest pits predicted is of the order of 100 mils after 448 days of exposure. The data indicate that even nitrite concentrations insufficient to prevent pitting have a beneficial effect on limiting the growth of deepest pits

  20. Modeling and simulation of ventilation devices in nuclear waste storage

    International Nuclear Information System (INIS)

    Zhang, Yumeng

    2015-01-01

    The objective of this thesis is to develop models and algorithms to simulate efficiently the mass exchanges occurring at the interface between the nuclear waste deep geological repositories and the ventilation excavated galleries. To model such physical processes, one needs to account in the porous medium for the flow of the liquid and gas phases including the vaporization of the water component in the gas phase and the dissolution of the gaseous components in the liquid phase. In the free flow region, a single phase gas free flow is considered assuming that the liquid phase is instantaneously vaporized at the interface. This gas free flow has to be compositional to account for the change of the relative humidity in the free flow region which has a strong feedback on the liquid flow rate at the interface. In chapter 1, three formulations of the gas liquid compositional Darcy flow are studied. Their equivalence from the point of phase transitions is shown and they are compared numerically on 1D and 3D test cases including gas appearance and liquid disappearance. The 3D discretization is based on the Vertex Approximate Gradient (VAG) scheme and takes into account discontinuous capillary pressures. In chapter 2, a reduced model coupling a 3D gas liquid compositional Darcy flow in a fractured porous medium, and a 1D compositional free gas flow is introduced. The VAG discretization is extended to such models taking into account the coupling between the 3D matrix, the 2D network of fractures and the 1D gallery. Its convergence is studied both for the linear single phase stationary model and for a non linear model coupling the Richards equation to a single phase 1D flow or a 1D tracer equation in the gallery. Different test cases with Andra data sets are presented. In Chapter 3, a splitting algorithm to solve the coupling between the gas liquid compositional Darcy flow in the porous medium and the gas compositional free flow in the gallery is developed. The idea is to

  1. Potential agents for removal of actinides from waste solutions

    International Nuclear Information System (INIS)

    Romanovski, V.V.; Whisenhunt, D.W.; Veeck, A.C.; Andersen, W.A.; Hoffman, D.C.; Jide, X.; White, D.; Raymond, K.N.

    1996-01-01

    The uptake of Th(IV) from nitric acid and hydrochloric acid solutions by chelating ion exchange resins containing catechol, 1,2- hydroxypyridinone (1,2-HOPO) and 3,4-hydroxypyridinone (3,4-HOPO) functional groups, has been investigated. These polystyrene based materials show excellent kinetics for uptake of Th(IV) and have a high loading capacity. Liquid/liquid extractants have also been synthesized by addition of lipophilic side chains to the chelating groups (1,2-HOPO; 3,4-HOPO; 3,2-HOPO; catecholamide; terephthalamide). The initial evaluation of the extraction properties has been carried out

  2. Porous materials based on foaming solutions obtained from industrial waste

    Science.gov (United States)

    Starostina, I. V.; Antipova, A. N.; Ovcharova, I. V.; Starostina, Yu L.

    2018-03-01

    This study analyzes foam concrete production efficiency. Research has shown the possibility of using a newly-designed protein-based foaming agent to produce porous materials using gypsum and cement binders. The protein foaming agent is obtained by alkaline hydrolysis of a raw mixture consisting of industrial waste in an electromagnetic field. The mixture consists of spent biomass of the Aspergillus niger fungus and dust from burning furnaces used in cement production. Varying the content of the foaming agent allows obtaining gypsum binder-based foam concretes with the density of 200-500 kg/m3 and compressive strength of 0.1-1.0 MPa, which can be used for thermal and sound insulation of building interiors. Cement binders were used to obtain structural and thermal insulation materials with the density of 300-950 kg/m3 and compressive strength of 0.9-9.0 MPa. The maximum operating temperature of cement-based foam concretes is 500°C because it provides the shrinkage of less than 2%.

  3. Volumetric determination of hydroxide, aluminate, and carbonate in alkaline solutions of nuclear waste

    International Nuclear Information System (INIS)

    Baumann, E.W.

    1975-06-01

    An integrated procedure was developed for determining OH - , Al(OH) 4 - , and CO 3 2- in alkaline nuclear waste. The free alkali, the hydroxide released when Al(OH) 3 is complexed with oxalate, and the precipitated BaCO 3 were determined by acidimetric titration. With a 50-μl sample, the relative standard deviations were 1 to 2 percent for nonradioactive test solutions and 2 to 5 percent for radioactive process solutions. (U.S.)

  4. Bioremediation of {sup 60}Co from simulated spent decontamination solutions

    Energy Technology Data Exchange (ETDEWEB)

    Rashmi, K.; Naga Sowjanya, T.; Maruthi Mohan, P.; Balaji, V.; Venkateswaran, G

    2004-07-26

    Bioremediation of {sup 60}Co from simulated spent decontamination solutions by utilizing different biomass of (Neurospora crassa, Trichoderma viridae, Mucor recemosus, Rhizopus chinensis, Penicillium citrinum, Aspergillus niger and, Aspergillus flavus) fungi is reported. Various fungal species were screened to evaluate their potential for removing cobalt from very low concentrations (0.03-0.16 {mu}M) in presence of a high background of iron (9.33 mM) and nickel (0.93 mM) complexed with EDTA (10.3 mM). The different fungal isolates employed in this study showed a pickup of cobalt in the range 8-500 ng/g of dry biomass. The [Fe]/[Co] and [Ni]/[Co] ratios in the solutions before and after exposure to the fungi were also determined. At micromolar level the cobalt pickup by many fungi especially the mutants of N. crassa is seen to be proportional to the initial cobalt concentration taken in the solution. However, R. chinensis exhibits a low but iron concentration dependent cobalt pickup. Prior saturating the fungi with excess of iron during their growth showed the presence of selective cobalt pickup sites. The existence of cobalt specific sorption sites is shown by a model experiment with R. chinensis wherein at a constant cobalt concentration (0.034 {mu}M) and varying iron concentrations so as to yield [Fe/Co]{sub initial} ratios in solution of 10, 100, 1000 and 287 000 have all yielded a definite Co pickup capacity in the range 8-47 ng/g. The presence of Cr(III)EDTA (3 mM) in solution along with complexed Fe and Ni has not influenced the cobalt removal. The significant feature of this study is that even when cobalt is present in trace level (sub-micromolar) in a matrix of high concentration (millimolar levels) of iron, nickel and chromium, a situation typically encountered in spent decontamination solutions arising from stainless steel based primary systems of nuclear reactors, a number of fungi studied in this work showed a good sensitivity for cobalt pickup.

  5. Bioremediation of 60Co from simulated spent decontamination solutions

    International Nuclear Information System (INIS)

    Rashmi, K.; Naga Sowjanya, T.; Maruthi Mohan, P.; Balaji, V.; Venkateswaran, G.

    2004-01-01

    Bioremediation of 60 Co from simulated spent decontamination solutions by utilizing different biomass of (Neurospora crassa, Trichoderma viridae, Mucor recemosus, Rhizopus chinensis, Penicillium citrinum, Aspergillus niger and, Aspergillus flavus) fungi is reported. Various fungal species were screened to evaluate their potential for removing cobalt from very low concentrations (0.03-0.16 μM) in presence of a high background of iron (9.33 mM) and nickel (0.93 mM) complexed with EDTA (10.3 mM). The different fungal isolates employed in this study showed a pickup of cobalt in the range 8-500 ng/g of dry biomass. The [Fe]/[Co] and [Ni]/[Co] ratios in the solutions before and after exposure to the fungi were also determined. At micromolar level the cobalt pickup by many fungi especially the mutants of N. crassa is seen to be proportional to the initial cobalt concentration taken in the solution. However, R. chinensis exhibits a low but iron concentration dependent cobalt pickup. Prior saturating the fungi with excess of iron during their growth showed the presence of selective cobalt pickup sites. The existence of cobalt specific sorption sites is shown by a model experiment with R. chinensis wherein at a constant cobalt concentration (0.034 μM) and varying iron concentrations so as to yield [Fe/Co] initial ratios in solution of 10, 100, 1000 and 287 000 have all yielded a definite Co pickup capacity in the range 8-47 ng/g. The presence of Cr(III)EDTA (3 mM) in solution along with complexed Fe and Ni has not influenced the cobalt removal. The significant feature of this study is that even when cobalt is present in trace level (sub-micromolar) in a matrix of high concentration (millimolar levels) of iron, nickel and chromium, a situation typically encountered in spent decontamination solutions arising from stainless steel based primary systems of nuclear reactors, a number of fungi studied in this work showed a good sensitivity for cobalt pickup

  6. Effects of mixed waste simulants on transportation packaging plastic components

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Dickens, T.G.

    1994-01-01

    The purpose of hazardous and radioactive materials packaging is to, enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified. The design requirements for both hazardous and radioactive materials packaging specify packaging compatibility, i.e., that the materials of the packaging and any contents be chemically compatible with each other. Furthermore, Type A and Type B packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program, supported by the US Department of Energy's (DOE) Transportation Management Division, EM-261 provides the means to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, we describe the general elements of the testing program and the experimental results of the screening tests. The implications of the results of this testing are discussed in the general context of packaging development. Additionally, we present the results of the first phase of this experimental program. This phase involved the screening of five candidate liner and six seal materials against four simulant mixed wastes

  7. The Waste Isolation Pilot Plant: a potential solution for the disposal of transuranic waste

    National Research Council Canada - National Science Library

    National Research Council Staff; Commission on Geosciences, Environment and Resources; Division on Earth and Life Studies; National Research Council; National Academy of Sciences

    ... Isolation Pilot Plant Board on Radioactive Waste Management Commission on Geosciences, Environment, and Resources National Research Council NATIONAL ACADEMY PRESS Washington, D.C. 1996 i Copyrighttrue Please breaks inserted. are Page files. accidentally typesetting been have may original from the errors not typographic original retained, and from the c...

  8. The corrosion resistance of Nitinol alloy in simulated physiological solutions

    International Nuclear Information System (INIS)

    Milošev, Ingrid; Kapun, Barbara

    2012-01-01

    The corrosion behaviour of Nitinol alloy containing nearly equi-atomic composition of nickel and titanium and its constituent metals (nickel and titanium) was investigated in simulated Hanks physiological solution (pH value 7.5) and pH modified simulated Hanks physiological solution (pH values 4.5 and 6.5) and by electrochemical method of anodic potentiodynamic polarization at 37 °C. In this chloride-rich medium the corrosion stability of Nitinol is limited by the susceptibility to localized corrosion and is in that sense more similar to nickel than to titanium. The corrosion stability of Nitinol is strongly dependent on the surface preparation—grinding, polishing or chemical etching. Whereas a ground surface is not resistant to localized corrosion, polished and chemically etched surfaces are resistant to this type of corrosion attack. The reasons for this behaviour were investigated through metallurgical, topographical and chemical properties of the surface as a function of surface preparation. For that purpose, scanning electron microscopy combined with chemical analysis, confocal microscopy and X-ray photoelectron spectroscopy were used. The surface roughness decreased in the following order: chemically etched > ground > polished surface. Besides differences in topography, distinct differences in the chemical composition of the outermost surface are observed. Ground, rough surfaces comprised mainly titanium oxides and small amounts of nickel metal. Chemically etched and, especially, polished surfaces are composed of a mixture of titanium, nickel and titanium oxides, as studied by angle resolved X-ray photoelectron spectroscopy. These results emphasize the importance of detailed investigation of the metal surface since small differences in surface preparation may induce large differences in corrosion stability of material when exposed to corrosive environments. - Highlights: ► The corrosion resistance of Nitinol is dependent on the surface preparation.

  9. Nuclear waste disposal: Can there be a resolution? Past problems and future solutions

    Energy Technology Data Exchange (ETDEWEB)

    Ahearne, J [Scientific Research Society, Sigma Xi, Research Triangle Park, NC (United States)

    1990-07-01

    Why does the high level waste problem have to be solved now? There are perhaps three answers to that question. First, to have a recovery of nuclear power. But a lack of resolution of the high level waste problem is not the principal reason that nuclear power has foundered and, consequently, solving it will not automatically revive nuclear power. However, if the nuclear industry is adamantly convinced that this is the key to reviving nuclear power, then the nuclear industry should demonstrate its conviction by putting much greater effort into resolving the high level waste problem technically, not through public relations. For example, a substantial effort on the actinide burning approach might demonstrate, in the old American phrase, 'putting your money where your mouth is'. Second, the high level waste problem must be solved now because it is a devil's brew. However, chemical wastes last longer, as we all know, than do the radioactive wastes. As one expert has noted: 'There is real risk in nuclear power, just as there is real risk in coal power.... For some of [these risks], like the greenhouse effect, the potential damage is devastating. While for others, like nuclear accidents, the risk is limited, but imaginations are not. For still others, like the risk posed by a high-level waste repository, there is essentially nothing outside the imagination of the gullible.' Furthermore, any technical solution or any solution to a risky problem requires one to think carefully. It is often better to do it right than quickly. A third reason for requiring it to be solved right now is that HLW disposal is a major technical problem blocking a potentially valuable energy source. But we need a new solution. The current solutions are not working. I believe that we ought to recognize the failure of the geologic repository approach. I believe the federal government should identify, with industry's assistance, the best techniques for surface storage. Some federal locations should be

  10. Nuclear waste disposal: Can there be a resolution? Past problems and future solutions

    International Nuclear Information System (INIS)

    Ahearne, J.

    1990-01-01

    Why does the high level waste problem have to be solved now? There are perhaps three answers to that question. First, to have a recovery of nuclear power. But a lack of resolution of the high level waste problem is not the principal reason that nuclear power has foundered and, consequently, solving it will not automatically revive nuclear power. However, if the nuclear industry is adamantly convinced that this is the key to reviving nuclear power, then the nuclear industry should demonstrate its conviction by putting much greater effort into resolving the high level waste problem technically, not through public relations. For example, a substantial effort on the actinide burning approach might demonstrate, in the old American phrase, 'putting your money where your mouth is'. Second, the high level waste problem must be solved now because it is a devil's brew. However, chemical wastes last longer, as we all know, than do the radioactive wastes. As one expert has noted: 'There is real risk in nuclear power, just as there is real risk in coal power.... For some of [these risks], like the greenhouse effect, the potential damage is devastating. While for others, like nuclear accidents, the risk is limited, but imaginations are not. For still others, like the risk posed by a high-level waste repository, there is essentially nothing outside the imagination of the gullible.' Furthermore, any technical solution or any solution to a risky problem requires one to think carefully. It is often better to do it right than quickly. A third reason for requiring it to be solved right now is that HLW disposal is a major technical problem blocking a potentially valuable energy source. But we need a new solution. The current solutions are not working. I believe that we ought to recognize the failure of the geologic repository approach. I believe the federal government should identify, with industry's assistance, the best techniques for surface storage. Some federal locations should be

  11. MODELING AN ION EXCHANGE PROCESS FOR CESIUM REMOVAL FROM ALKALINE RADIOACTIVE WASTE SOLUTIONS

    International Nuclear Information System (INIS)

    Smith, F.; Hamm, Luther; Aleman, Sebastian; Michael, Johnston

    2008-01-01

    The performance of spherical Resorcinol-Formaldehyde ion-exchange resin for the removal of cesium from alkaline radioactive waste solutions has been investigated through computer modeling. Cesium adsorption isotherms were obtained by fitting experimental data using a thermodynamic framework. Results show that ion-exchange is an efficient method for cesium removal from highly alkaline radioactive waste solutions. On average, two 1300 liter columns operating in series are able to treat 690,000 liters of waste with an initial cesium concentration of 0.09 mM in 11 days achieving a decontamination factor of over 50,000. The study also tested the sensitivity of ion-exchange column performance to variations in flow rate, temperature and column dimensions. Modeling results can be used to optimize design of the ion exchange system

  12. Analysis of a waste-heat boiler by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yongziang; Jokilaakso, A. [Helsinki Univ. of Technology, Otaniemi (Finland)

    1996-12-31

    Waste-heat boilers play important roles in the continuous operation of a smelter and in the conservation of energy. However, the fluid flow and heat transfer behaviour has not been well studied, concerning the boiler performance and design. This presentation describes simulated gas flow and heat transfer of a waste-heat boiler in the Outokumpu copper flash smelting process. The governing transport equations for the conservation of mass, momentum and enthalpy were solved with a commercial CFD-code PHOENICS. The standard k-{epsilon} turbulence model and a composite-flux radiation model were used in the computations. The computational results show that the flow is strongly recirculating and distinctly three-dimensional in most part of the boiler, particularly in the radiation section. The predicted flow pattern and temperature distribution were in a good agreement with laboratory models and industrial measurements. The results provide detailed information of flow pattern, the temperature distribution and gas cooling efficiency. The CFD proved to be a useful tool in analysing the boiler operation. (author)

  13. Analysis of a waste-heat boiler by CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yongziang; Jokilaakso, A [Helsinki Univ. of Technology, Otaniemi (Finland)

    1997-12-31

    Waste-heat boilers play important roles in the continuous operation of a smelter and in the conservation of energy. However, the fluid flow and heat transfer behaviour has not been well studied, concerning the boiler performance and design. This presentation describes simulated gas flow and heat transfer of a waste-heat boiler in the Outokumpu copper flash smelting process. The governing transport equations for the conservation of mass, momentum and enthalpy were solved with a commercial CFD-code PHOENICS. The standard k-{epsilon} turbulence model and a composite-flux radiation model were used in the computations. The computational results show that the flow is strongly recirculating and distinctly three-dimensional in most part of the boiler, particularly in the radiation section. The predicted flow pattern and temperature distribution were in a good agreement with laboratory models and industrial measurements. The results provide detailed information of flow pattern, the temperature distribution and gas cooling efficiency. The CFD proved to be a useful tool in analysing the boiler operation. (author)

  14. Selective Recovery Of Copper From Solutions After Bioleaching Electronic Waste

    Directory of Open Access Journals (Sweden)

    Willner Joanna

    2015-06-01

    Full Text Available Research on selective extraction of copper from solution after bioleaching grounded printed circuit boards (PCBs using LIX 860N-IC were conducted. The effect of LIX 860N-IC concentration, phase ratio and influence of initial pH value of aqueous phase on the extraction of copper and iron was examined. It was found that the extraction rate of copper increases with the LIX 860N-IC concentration. Best results of Cu extraction (98 % were achieved with extractant concentration of 5 % and pH 1.9. Higher pH value of aqueous phase (pH=2.4 is conducive to the simultaneous effect of Fe co-extraction.

  15. Malonamides as new extractants for nuclear waste solutions

    International Nuclear Information System (INIS)

    Cuillerdier, C.; Musikas, C.; Hoel, P.; Nigond, L.; Vitart, X.

    1989-01-01

    Substituted malonamides are able to extract α emitters from radioactive solutions in nitric acid, all the actinides (III, IV, VI) are well extracted and can be easily back extracted. Some problems remain with neptunium and technetium. These solvents are not expensive. For an industrial purpose, synthesis has been optimized, and a proper choice of commercial basic products can decrease the cost. The solvent obtained on a pilot scale (1 kg) was found to be pure enough, it didn't need any additional treatment. Degradation under hydrolysis or radiolysis is not important in the conditions of practical experiment (t 0 C). Degradation products can be washed with NaOH (carboxylic acids) they don't give precipitates or emulsions. Efficiency of the solvent is good compared to CMPO, taking into account the lack of extensive industrial development. Further researches are undertaken in two main directions: optimizing the synthesis and use of aliphatic diluents

  16. Treatment of nuclear waste solutions using a new class of extractants: pentaalkyl propane diamides

    International Nuclear Information System (INIS)

    Cuillerdier, C.; Musikas, C.; Hoel, P.

    1990-01-01

    A new class of bifunctional extractants pentaalkyl propane diamides is studied in order to extract trivalent cations (Am 3+ , Cm 3+ ...) and other actinides contained in waste solutions of nuclear industry. These solvents are completely incinerable and don't produce harmfull degradation products. Their main chemicals properties are reviewed. The results of a mixer-settler battery experiment are given

  17. Rapid immobilization of simulated radioactive soil waste by microwave sintering.

    Science.gov (United States)

    Zhang, Shuai; Shu, Xiaoyan; Chen, Shunzhang; Yang, Huimin; Hou, Chenxi; Mao, Xueli; Chi, Fangting; Song, Mianxin; Lu, Xirui

    2017-09-05

    A rapid and efficient method is particularly necessary in the timely disposal of seriously radioactive contaminated soil. In this paper, a series of simulated radioactive soil waste containing different contents of neodymium oxide (3-25wt.%) has been successfully vitrified by microwave sintering at 1300°C for 30min. The microstructures, morphology, element distribution, density and chemical durability of as obtained vitrified forms have been analyzed. The results show that the amorphous structure, homogeneous element distribution, and regular density improvement are well kept, except slight cracks emerge on the magnified surface for the 25wt.% Nd 2 O 3 -containing sample. Moreover, all the vitrified forms exhibit excellent chemical durability, and the leaching rates of Nd are kept as ∼10 -4 -10 -6 g/(m 2 day) within 42days. This demonstrates a potential application of microwave sintering in radioactive contaminated soil disposal. Copyright © 2017 Elsevier B.V. All rights reserved.

  18. Physical and Liquid Chemical Simulant Formulations for Transuranic Waste in Hanford Single-Shell Tanks

    International Nuclear Information System (INIS)

    Rassat, Scot D.; Bagaasen, Larry M.; Mahoney, Lenna A.; Russell, Renee L.; Caldwell, Dustin D.; Mendoza, Donaldo P.

    2003-01-01

    CH2M HILL Hanford Group, Inc. (CH2M HILL) is in the process of identifying and developing supplemental process technologies to accelerate the tank waste cleanup mission. A range of technologies is being evaluated to allow disposal of Hanford waste types, including transuranic (TRU) process wastes. Ten Hanford single-shell tanks (SSTs) have been identified whose contents may meet the criteria for designation as TRU waste: the B-200 series (241-B-201, -B-202, -B 203, and B 204), the T-200 series (241-T-201, T 202, -T-203, and -T-204), and Tanks 241-T-110 and -T-111. CH2M HILL has requested vendor proposals to develop a system to transfer and package the contact-handled TRU (CH-TRU) waste retrieved from the SSTs for subsequent disposal at the Waste Isolation Pilot Plant (WIPP). Current plans call for a modified ''dry'' retrieval process in which a liquid stream is used to help mobilize the waste for retrieval and transfer through lines and vessels. This retrieval approach requires that a significant portion of the liquid be removed from the mobilized waste sludge in a ''dewatering'' process such as centrifugation prior to transferring to waste packages in a form suitable for acceptance at WIPP. In support of CH2M HILL's effort to procure a TRU waste handling and packaging process, Pacific Northwest National Laboratory (PNNL) developed waste simulant formulations to be used in evaluating the vendor's system. For the SST CH-TRU wastes, the suite of simulants includes (1) nonradioactive chemical simulants of the liquid fraction of the waste, (2) physical simulants that reproduce the important dewatering properties of the waste, and (3) physical simulants that can be used to mimic important rheological properties of the waste at different points in the TRU waste handling and packaging process. To validate the simulant formulations, their measured properties were compared with the limited data for actual TRU waste samples. PNNL developed the final simulant formulations

  19. Molecular dynamics simulations of lysozyme in water/sugar solutions

    Energy Technology Data Exchange (ETDEWEB)

    Lerbret, A. [Department of Food Science, Cornell University, 101 Stocking Hall, Ithaca, NY 14853 (United States); Affouard, F. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires, UMR CNRS 8024, Universite Lille I, 59655 Villeneuve d' Ascq Cedex (France)], E-mail: frederic.affouard@univ-lille1.fr; Bordat, P. [Laboratoire de Chimie Theorique et de Physico-Chimie Moleculaire, UMR 5624, Universite de Pau et des Pays de l' Adour, 64000 Pau (France); Hedoux, A.; Guinet, Y.; Descamps, M. [Laboratoire de Dynamique et Structure des Materiaux Moleculaires, UMR CNRS 8024, Universite Lille I, 59655 Villeneuve d' Ascq Cedex (France)

    2008-04-18

    Structural and dynamical properties of the solvent at the protein/solvent interface have been investigated by molecular dynamics simulations of lysozyme in trehalose, maltose and sucrose solutions. Results are discussed in the framework of the bioprotection phenomena. The analysis of the relative concentration of water oxygen atoms around lysozyme suggests that lysozyme is preferentially hydrated. When comparing the three sugars, trehalose is seen more excluded than maltose and sucrose. The preferential exclusion of sugars from the protein surface induces some differences in the behavior of trehalose and maltose, particularly at 50 and 60 wt% concentrations, that are not observed experimentally in binary sugar/mixtures. The dynamical slowing down of the solvent is suggested to mainly arise from the homogeneity of the water/sugar matrices controlled by the percolation of the sugar hydrogen bonds networks. Furthermore, lysozyme strongly increases relaxation times of solvent molecules at the protein/solvent interface.

  20. Bioleaching of fly ash from municipal solid waste incineration using kitchen waste saccharified solution as culture medium

    International Nuclear Information System (INIS)

    Wei, S.; Juan, W.; Qunhui, W.

    2013-01-01

    Summary: Reduced sugar in saccharified solution from kitchen waste was used as the carbon source. Domesticated A. niger AS 3.879C , which can withstand 20% of kitchen waste, was used as the inoculum in the bioleaching process of municipal solid waste incineration fly ash. The effect of reduced sugar concentration, fly ash concentration, and medium volume on the heavy metal extraction and yield of fly ash as well as the optimum bioleaching conditions; the inoculation amount of AS 3 .879C 1% (v/v), reduced sugar concentration of 80 g/l, fly ash concentration of 20 g/l, medium volume of 200 ml, and the addition of fly ash (20 g/l) after culturing for 4 days at 30 degree C and 140 r/min were obtained. Under the optimum condition, the extraction yield of the seven tested heavy metals are in the order of Cd > Zn > Cu > Mn > Pb > Cr > Fe; the extraction yield of Cd and Zn reached 88.7% and 73.1% respectively. Fly ash satisfied the Standard for Pollution Control on the Security Landfill Site for Hazardous Wastes (GB 18598-2001) after heavy metal extraction. (author)

  1. Modeling long-term leaching experiments of full scale cemented wastes: effect of solution composition on diffusion

    International Nuclear Information System (INIS)

    Borkel, C.; Montoya, V.; Kienzler, B.

    2015-01-01

    The code PHREECQ V3.1 has been used to simulate leaching experiments performed with cemented simulated waste products in tap water for more than 30 years. In this work the main focus is related with the leaching of Cs explained by diffusion processes. A simplifying model using the code PHREECQ V3.1 was used to investigate the influence of different parameters on the release of Cs from the cement solid to the leaching solution. The model setup bases on four main assumptions: a) the solid as well as the distribution of Cs is homogeneous and of isotropic texture, b) there is no preferential direction regarding cement degradation or water intrusion into the solid, c) the pore space is entirely connected and d) Cs adsorption to the cement or container is negligible. In the modeling the constraint of charge balance was stressed. Effective diffusion coefficients (D e ) were obtained analytically and from modeling the diffusive release of Cs from cemented waste simulates. The obtained values D e for Cs leaching are in perfect agreement with the values published in literature. Contradictory results to diffusive release were obtained from XRD analysis of the solids, suggesting that water may not have penetrated the cement monoliths entirely, but only to some centimeters depth. XRD analysis have been done to determine the solid phases present in cement and are used to help outlining strength and weaknesses of the different models

  2. Corrosion of construction steel in pore simulated solution

    International Nuclear Information System (INIS)

    Valdes, Matias; Vasquez, Marcela

    2004-01-01

    The corrosion of steel for reinforcing reinforced cement structures is a common problem particularly in structures that are exposed to a marine environment. Loosened masonry originating by the diametrical stress that iron oxides place on the cement is not unusual. These situations involve risk to people and goods and make it necessary to repair the structure to prolong its useful service life. Some preliminary results are presented from the reproduction of the corrosive process with the use of a solution that simulates the chemical surroundings in the concrete pores. These results will help to evaluate the incidence of contaminants (CO 2 , chloride ions), inhibitors and coatings, among others, in the following stages by conveniently adjusting the solution's composition. The composition of the chosen solution is: 0.01 mol NaOH - 0.002 mol/l Ca(OH) 2 . The effect was evaluated of a passive film generated on the surface of the steel of the reinforcements at 100 mV for 14 minutes and for 12 hours. This potential corresponds to the passive region, as determined by recording tests with cyclic volt amperometry and in accordance with the Pourbaix diagram for steel. The corrosion current was defined by recording the resistance to polarization using different electrochemical methods: potential sweep, potentiostatic jump and sweep electrochemical impedance. The results show that neither of the two times selected are enough to generate the metal's passive state and that the potential of 100 mV used to generate the passive film may be too low to produce a compact and long lasting layer, considering that the passive zone interval comes to 700 mV, according to the volt amperometry readings (CW)

  3. Recovery of cyanide in gold leach waste solution by volatilization and absorption.

    Science.gov (United States)

    Gönen, N; Kabasakal, O S; Ozdil, G

    2004-09-10

    In this study, the effects of pH, time and temperature in regeneration of cyanide in the leaching waste solution of gold production from disseminated gold ore by cyanidation process were investigated and the optimum conditions, consumptions and cyanide recovery values were determined. The sample of waste solution containing 156 mg/l free CN- and 358 mg/l total CN-, that was obtained from Gümüşhane-Mastra/Turkey disseminated gold ores by cyanidation and carbon-in-pulp (CIP) process under laboratory conditions was used in the experiments. Acidification with H2SO4, volatilization of hydrogen cyanide (HCN) with air stripping and absorption of HCN in a basic solution stages were applied and under optimum conditions, 100% of free cyanide and 48% of complex cyanide and consequently 70% of the total cyanide in the liquid phase of gold leach effluent are recovered.

  4. Analysis by simulation of the disposition of nuclear fuel waste

    International Nuclear Information System (INIS)

    Turek, J.L.

    1980-09-01

    A descriptive simulation model is developed which includes all aspects of nuclear waste disposition. The model is comprised of two systems, the second system orchestrated by GASP IV. A spent fuel generation prediction module is interfaced with the AFR Program Management Information System and a repository scheduling information module. The user is permitted a wide range of options with which to tailor the simulation to any desired storage scenario. The model projects storage requirements through the year 2020. The outputs are evaluations of the impact that alternative decision policies and milestone date changes have on the demand for, the availability of, and the utilization of spent fuel storage capacities. Both graphs and detailed listings are available. These outputs give a comprehensive view of the particular scenario under observation, including the tracking, by year, of each discharge from every reactor. Included within the work is a review of the status of spent fuel disposition based on input data accurate as of August 1980. The results indicate that some temporary storage techniques (e.g., transshipment of fuel and/or additional at-reactor storage pools) must be utilized to prevent reactor shutdowns. These techniques will be required until the 1990's when several AFR facilities, and possibly one repository, can become operational

  5. Characterization and reaction behavior of ferrocyanide simulants and Hanford Site high-level ferrocyanide waste

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Simpson, B.C.

    1994-02-01

    Nonradioactive waste simulants and initial ferrocyanide tank waste samples were characterized to assess potential safety concerns associated with ferrocyanide high-level radioactive waste stored at the Hanford Site in underground single-shell tanks (SSTs). Chemical, physical, thermodynamic, and reaction properties of the waste simulants were determined and compared to properties of initial samples of actual ferrocyanide wastes presently in the tanks. The simulants were shown to not support propagating reactions when subjected to a strong ignition source. The simulant with the greatest ferrocyanide concentration was shown to not support a propagating reaction that would involve surrounding waste because of its high water content. Evaluation of dried simulants indicated a concentration limit of about 14 wt% disodium mononickel ferrocyanide, below which propagating reactions could not occur in the ambient temperature bulk tank waste. For postulated localized hot spots where dried waste is postulated to be at an initial temperature of 130 C, a concentration limit of about 13 wt% disodium mononickel ferrocyanide was determined, below which propagating reactions could not occur. Analyses of initial samples of the presently stored ferrocyanide waste indicate that the waste tank ferrocyanide concentrations are considerably lower than the limit for propagation for dry waste and that the water content is near that of the as-prepared simulants. If the initial trend continues, it will be possible to show that runaway ferrocyanide reactions are not possible under present tank conditions. The lower ferrocyanide concentrations in actual tank waste may be due to tank waste mixing and/or degradation from radiolysis and/or hydrolysis, which may have occurred over approximately 35 years of storage

  6. Nitrate-cancrinite precipitation on quartz sand in simulated Hanford tank solutions.

    Science.gov (United States)

    Bickmore, B R; Nagy, K L; Young, J S; Drexler, J W

    2001-11-15

    Caustic NaNO3 solutions containing dissolved Al were reacted with quartz sand at 89 degrees C to simulate possible reactions between leaked nuclear waste and primary subsurface minerals at the U.S. Department of Energy's Hanford site in Washington. Nitrate-cancrinite began to precipitate onto the quartz after 2-10 days, cementing the grains together. Estimates of the equilibrium constant for the precipitation reaction differ for solutions with 0.1 or 1.0 m OH- (log Keq = 30.4 +/- 0.8 and 36.2 +/- 0.6, respectively). The difference in solubility may be attributable to more perfect crystallinity (i.e., fewer stacking faults) in the higher-pH cancrinite structure. This is supported by electron micrographs of crystal morphology and measured rates of Na volatilization under an electron beam. Precipitate crystallinity may affect radionuclide mobility, because stacking faults in the cancrinite structure can diminish its zeolitic cation exchange properties. The precipitation rate near the onset of nucleation depends on the total Al and Si concentrations in solution. The evolution of experimental Si concentrations was modeled by considering the dependence of quartz dissolution rate on AI(OH)4- activity, cancrinite precipitation, and the reduction of reactive surface area of quartz due to coverage by cancrinite.

  7. Chemical compatibility screening results of plastic packaging to mixed waste simulants

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Dickens, T.G.

    1995-01-01

    We have developed a chemical compatibility program for evaluating transportation packaging components for transporting mixed waste forms. We have performed the first phase of this experimental program to determine the effects of simulant mixed wastes on packaging materials. This effort involved the screening of 10 plastic materials in four liquid mixed waste simulants. The testing protocol involved exposing the respective materials to ∼3 kGy of gamma radiation followed by 14 day exposures to the waste simulants of 60 C. The seal materials or rubbers were tested using VTR (vapor transport rate) measurements while the liner materials were tested using specific gravity as a metric. For these tests, a screening criteria of ∼1 g/m 2 /hr for VTR and a specific gravity change of 10% was used. It was concluded that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only VITON passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture simulant mixed waste, none of the seal materials met the screening criteria. It is anticipated that those materials with the lowest VTRs will be evaluated in the comprehensive phase of the program. For specific gravity testing of liner materials the data showed that while all materials with the exception of polypropylene passed the screening criteria, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals

  8. Simulating Radionuclide Migrations of Low-level Wastes in Nearshore Environment

    Science.gov (United States)

    Lu, C. C.; Li, M. H.; Chen, J. S.; Yeh, G. T.

    2016-12-01

    Tunnel disposal into nearshore mountains was tentatively selected as one of final disposal sites for low-level wastes in Taiwan. Safety assessment on radionuclide migrations in far-filed may involve geosphere processes under coastal environments and into nearshore ocean. In this study the 3-D HYDROFEOCHE5.6 numerical model was used to perform simulations of groundwater flow and radionuclide transport with decay chains. Domain of interest on the surface includes nearby watersheds delineated by digital elevation models and nearshore seabed. As deep as 800 m below the surface and 400 m below sea bed were considered for simulations. The disposal site was located at 200m below the surface. Release rates of radionuclides from near-field was estimated by analytical solutions of radionuclide diffusion with decay out of engineered barriers. Far-field safety assessments were performed starting from the release of radionuclides out of engineered barriers to a time scale of 10,000 years. Sensitivity analyses of geosphere and transport parameters were performed to improve our understanding of safety on final disposal of low-level waste in nearshore environments.

  9. Evaluation of the transport and resuspension of a simulated nuclear waste slurry: Nuclear Waste Treatment Program

    International Nuclear Information System (INIS)

    Carleson, T.E.; Drown, D.C.; Hart, R.E.; Peterson, M.E.

    1987-09-01

    The Department of Chemical Engineering at the University of Idaho conducted research on the transport and resuspension of a simulated high-level nuclear waste slurry. In the United States, the reference process for treating both defense and civilian HLLW is vitrification using the liquid-fed ceramic melter process. The non-Newtonian behavior of the slurry complicates the evaluation of the transport and resuspension characteristics of the slurry. The resuspension of a simulated (nonradioactive) melter feed slurry was evaluated using a slurry designated as WV-205. The simulated slurry was developed for the West Valley Demonstration Project and was used during a pilot-scale ceramic melter (PSCM) experiment conducted at PNL in July 1985 (PSCM-21). This study involved determining the transport characteristics of a fully suspended slurry and the resuspension characteristics of settled solids in a pilot-scale pipe loop. The goal was to predict the transport and resuspension of a full-scale system based on rheological data for a specific slurry. The rheological behavior of the slurry was evaluated using a concentric cylinder rotational viscometer, a capillary tube viscometer, and the pilot-scale pipe loop. The results obtained from the three approaches were compared. 40 refs., 74 figs., 15 tabs

  10. Investigating the effect of compression on solute transport through degrading municipal solid waste

    Energy Technology Data Exchange (ETDEWEB)

    Woodman, N.D., E-mail: n.d.woodman@soton.ac.uk; Rees-White, T.C.; Stringfellow, A.M.; Beaven, R.P.; Hudson, A.P.

    2014-11-15

    Highlights: • The influence of compression on MSW flushing was evaluated using 13 tracer tests. • Compression has little effect on solute diffusion times in MSW. • Lithium tracer was conservative in non-degrading waste but not in degrading waste. • Bromide tracer was conservative, but deuterium was not. - Abstract: The effect of applied compression on the nature of liquid flow and hence the movement of contaminants within municipal solid waste was examined by means of thirteen tracer tests conducted on five separate waste samples. The conservative nature of bromide, lithium and deuterium tracers was evaluated and linked to the presence of degradation in the sample. Lithium and deuterium tracers were non-conservative in the presence of degradation, whereas the bromide remained effectively conservative under all conditions. Solute diffusion times into and out of less mobile blocks of waste were compared for each test under the assumption of dominantly dual-porosity flow. Despite the fact that hydraulic conductivity changed strongly with applied stress, the block diffusion times were found to be much less sensitive to compression. A simple conceptual model, whereby flow is dominated by sub-parallel low permeability obstructions which define predominantly horizontally aligned less mobile zones, is able to explain this result. Compression tends to narrow the gap between the obstructions, but not significantly alter the horizontal length scale. Irrespective of knowledge of the true flow pattern, these results show that simple models of solute flushing from landfill which do not include depth dependent changes in solute transport parameters are justified.

  11. Contribution to the understanding and to the simulation of processes occurring at the vicinity of a radioactive waste repository

    International Nuclear Information System (INIS)

    Trotignon, L.

    2004-04-01

    The author gives an overview of his research activities between 1986 and 2004. These activities were focused on the observation, analysis and simulation of solid-solution interactions, with application to radioactive waste storage in deep geologic formations. More precisely, these works dealt with the evolution of rock porosity (dissolution-crystallization under stress), the aqueous corrosion of nuclear glasses, the redox transient (how and at which rate a disturbance related to dissolved oxygen intrusion will be resorbed), and the transport-chemistry simulation and natural analogues

  12. Pu speciation in actual and simulated aged wastes

    Energy Technology Data Exchange (ETDEWEB)

    Lezama-pacheco, Juan S [Los Alamos National Laboratory; Conradson, Steven D [Los Alamos National Laboratory

    2008-01-01

    X-ray Absorption Fine Structure Spectroscopy (XAFS) at the Pu L{sub II/III} edge was used to determine the speciation of this element in (1) Hanford Z-9 Pu crib samples, (2) deteriorated waste resins from a chloride process ion-exchange purification line, and (3) the sediments from two Waste Isolation Pilot Plant Liter Scale simulant brine systems. The Pu speciation in all of these samples except one is within the range previously displayed by PuO{sub 2+x-2y}(OH){sub y}{center_dot}zH{sub 2}O compounds, which is expected based on the putative thermodynamic stability of this system for Pu equilibrated with excess H{sub 2}O and O{sub 2} under environmental conditions. The primary exception was a near neutral brine experiment that displayed evidence for partial substitution of the normal O-based ligands with Cl{sup -} and a concomitant expansion of the Pu-Pu distance relative to the much more highly ordered Pu near neighbor shell in PuO{sub 2}. However, although the Pu speciation was not necessarily unusual, the Pu chemistry identified via the history of these samples did exhibit unexpected patterns, the most significant of which may be that the presence of the Pu(V)-oxo species may decrease rather than increase the overall solubility of these compounds. Several additional aspects of the Pu speciation have also not been previously observed in laboratory-based samples. The molecular environmental chemistry of Pu is therefore likely to be more complicated than would be predicted based solely on the behavior of PuO{sub 2} under laboratory conditions.

  13. Applying fluid dynamics simulations to improve processing and remediation of nuclear waste - 59172

    International Nuclear Information System (INIS)

    Knight, Kelly J.; Peltier, Joel; Berkoe, Jon; Rosendall, Brigette; Kennedy, Chris

    2012-01-01

    Transport and processing of nuclear waste for treatment and storage can involve unique and complex thermal and fluid dynamic conditions that pose potential for safety risk and/or design uncertainty and also are likely to be subjected to more precise performance requirements than in other industries. From an engineering analysis perspective, certainty of outcome is essential. Advanced robust methods for engineering analysis and simulation of critical processes can help reduce risk of design uncertainty and help mitigate or reduce the amount of expensive full-scale demonstration testing. This paper will discuss experience gained in applying computational fluid dynamics models to key processes for mixing, transporting, and thermal treatment of nuclear waste as part of designing a massive vitrification process plant that will convert high and low level nuclear waste into glass for permanent storage. Examples from industrial scale simulations will be presented. The computational models have shown promise in replicating several complex physical processes such as solid-liquid flows in suspension, blending of slurries, and cooling of materials at extremely high temperature. Knowledge gained from applying simulation has provided detailed insight into determining the most critical aspects of these complex processes that can ultimately be used to help guide the optimum design of waste handling equipment based on credible calculations while ensuring risk of design uncertainty is minimized. The WTP Project is faced with complex technical challenges that must have solutions that enable the successful operation of the plant for its 30+ year operating life. The Project chose to reduce those risks by employing an experienced team that applied CFD in a disciplined manner and adhered to an established guideline with the following benefits: - Gained an improvement in accuracy of predictions for complex physical situations; - Gained an improvement of the quality of experimental

  14. Plasma processing of compacted drums of simulated radioactive waste

    International Nuclear Information System (INIS)

    Geimer, R.; Batdorf, J.; Larsen, M.M.

    1991-01-01

    The charter of the Department of Energy (DOE) Office of Technology Development (OTD) is to identify and develop technologies that have potential application in the treatment of DOE wastes. One particular waste of concern within the DOE is transuranic (TRU) waste, which is generated and stored at several DOE sites. High temperature DC arc generated plasma technology is an emerging treatment method for TRU waste, and its use has the potential to provide many benefits in the management of TRU. This paper begins by discussing the need for development of a treatment process for TRU waste, and the potential benefits that a plasma waste treatment system can provide in treating TRU waste. This is followed by a discussion of the results of a project conducted for the DOE to demonstrate the effectiveness of a plasma process for treating supercompacted TRU waste. 1 fig., 1 tab

  15. A system dynamics-based environmental performance simulation of construction waste reduction management in China.

    Science.gov (United States)

    Ding, Zhikun; Yi, Guizhen; Tam, Vivian W Y; Huang, Tengyue

    2016-05-01

    A huge amount of construction waste has been generated from increasingly higher number of construction activities than in the past, which has significant negative impacts on the environment if they are not properly managed. Therefore, effective construction waste management is of primary importance for future sustainable development. Based on the theory of planned behaviors, this paper develops a system dynamic model of construction waste reduction management at the construction phase to simulate the environmental benefits of construction waste reduction management. The application of the proposed model is shown using a case study in Shenzhen, China. Vensim is applied to simulate and analyze the model. The simulation results indicate that source reduction is an effective waste reduction measure which can reduce 27.05% of the total waste generation. Sorting behaviors are a premise for improving the construction waste recycling and reuse rates which account for 15.49% of the total waste generated. The environmental benefits of source reduction outweigh those of sorting behaviors. Therefore, to achieve better environmental performance of the construction waste reduction management, attention should be paid to source reduction such as low waste technologies and on-site management performance. In the meantime, sorting behaviors encouragement such as improving stakeholders' waste awareness, refining regulations, strengthening government supervision and controlling illegal dumping should be emphasized. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. Analytic solution for one-dimensional diffusion of radionuclides from a waste package

    International Nuclear Information System (INIS)

    Oliver, D.L.

    1985-01-01

    This work implements an analytical solution for diffusion of radionuclides from a cylindrical waste form through the packing material into the surrounding host rock. Recent interest in predicting the performance of a proposed geological repository for nuclear waste has led to the development of several computer programs to predict the performance of such a repository for the next several millenia. These numerical codes are generally designed to accommodate a broad spectrum of geometrical configurations and repository conditions in order to accurately predict the behavior of the radionuclides in the repository environment. Confidence in such general purpose codes is gained by verifying the numerical modeling and the software through comparison of the numerical predictions generated by these computer codes with analytical solutions to reasonably complex problems. The analysis discussed herein implements the analytic solution, proposed by J.C. Jaeger in 1941 for radial diffusion through two concentric circular cylinders. Jaeger's solution was applied to the problem of diffusional mass transfer from a long cylindrical waste form and subsequently into the surrounding geological formation. Analytic predictions of fractional release rates, including the effects of sorption, were generated

  17. Corrosivity of solutions from evaporation of radioactive liquid wastes. Final report

    International Nuclear Information System (INIS)

    Payer, H.; Kolic, E.S.; Boyd, W.K.

    1977-01-01

    New double-shell storage tanks are constructed with ASTM A-516 Grade 65 steel. This study had two main objectives: To characterize the corrosivity of synthetic nonradioactive terminal waste solutions to ASTM A-516 Grade 65 steel and to determine the severity of stress-corrosion cracking of carbon steel in terminal waste solutions. The information developed provides guidance in the characterization of the aggressiveness of actual terminal liquors and in the design and operation of fail-safe tanks. Corrosion behavior was measured over a range of oxidizing conditions by the potentiodynamic polarization technique. Oxidizing conditions in a solution likely to promote general corrosion, pitting or stress-corrosion cracking (SCC) were identified. Absolute stress-corrosion cracking susceptibility was determined by constant strain rate procedure for ASTM A-516 Grade 65 steel for conditions identified by polarization experiments as likely to promote SCC. Based on the results of this study, terminal waste storage tanks are safe from stress-corrosion cracking under freely corroding conditions. Corrosion potential of steel in solutions within anticipated compositions is at the positive end of the critical range for stress-corrosion cracking, and no conditions were observed which would lower the potential to more negative values within the cracking range under freely corroding conditions. Measurement of corrosion potential and hydroxide concentration provides a means to extend these results to compositions outside of the composition range studied

  18. An analytic solution for one-dimensional diffusion of radionuclides from a waste package

    International Nuclear Information System (INIS)

    1985-01-01

    This work implements an analytical solution for diffusion of radionuclides from a cylindrical waste form through the packing material into the surrounding host rock. Recent interest in predicting the performance of a proposed geological repository for nuclear waste has led to the development of several computer programs to predict the performance of such a repository for the next several millenia. These numerical codes are generally designed to accommodate a broad spectrum of geometrical configurations and repository conditions in order to accurately predict the behavior of the radionuclides in the repository environment. Confidence in such general purpose codes is gained by verifying the numerical modeling and the software through comparison of the numerical predictions generated by these computer codes with analytical solutions to reasonably complex problems. The analysis discussed herein implements the analytic solution, proposed by J.C. Jaeger in 1941 for radial diffusion through two concentric circular cylinders. Jaeger's solution was applied to the problem of diffusional mass transfer from a long cylindrical waste form and subsequently into the surrounding geological formation. Analytic predictions of fractional release rates, including the effects of sorption, were generated. 6 refs., 2 figs., 2 tabs

  19. Regulatory dilemmas of a trans-solutional problem: Spatial and temporal isolation of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Johnsrud, J.H. [State College, PA (United States). Sierra Club National Nuclear Waste Task Force

    1999-04-01

    This paper explores the `trans-solutional` nature of nuclear waste control - that it is in essence beyond human solution. To protect present and future human health, radioactive wastes require effective isolation from the biosphere for the full hazardous life of the wastes. Waste sequestration is essential to protect human beings, other forms of life in the bio-system, and the environment from adverse mutational impacts of exposures to ionizing radiation experienced in excess of those received from naturally-occurring background sources. The linear hypothesis of dose-response describes the relationship of radiation exposures to health at least with respect to cancer induction and hereditary genetic effects. The issue of risks of low-level radiation effects remains in controversy, with pressures exerted on regulators to ignore or minimize those impacts. However, recent research indicates that chronic low-dose exposures via inhalation and ingestion pathways may also give rise to non-fatal non-cancer deleterious health effects. Fatal cancers, now the primary measure of radiation injury in setting standards, may be less significant to a population in the long run than more subtle low-level impacts affecting genetic material. The latter, hard to identify or measure, may reduce developmental and reproductive capability. Given the hazardous longevity of high-level wastes, it is imperative that both protective standards and waste regulation be framed within an ethic of species responsibility. In our half century we have generated vast amounts of long-lived waste, with more promised in the coming millennium. The regulatory obligation is to isolate all nuclear wastes to best prevent any releases to the biosphere now but also to assure future generations an equal opportunity when our `disposal` methods inevitably fail over future time, to be able to retrieve and continue to isolate the wastes that we have caused to be produced. lt follows that the standards must not calculate

  20. Regulatory dilemmas of a trans-solutional problem: Spatial and temporal isolation of radioactive wastes

    International Nuclear Information System (INIS)

    Johnsrud, J.H.

    1999-01-01

    This paper explores the 'trans-solutional' nature of nuclear waste control - that it is in essence beyond human solution. To protect present and future human health, radioactive wastes require effective isolation from the biosphere for the full hazardous life of the wastes. Waste sequestration is essential to protect human beings, other forms of life in the bio-system, and the environment from adverse mutational impacts of exposures to ionizing radiation experienced in excess of those received from naturally-occurring background sources. The linear hypothesis of dose-response describes the relationship of radiation exposures to health at least with respect to cancer induction and hereditary genetic effects. The issue of risks of low-level radiation effects remains in controversy, with pressures exerted on regulators to ignore or minimize those impacts. However, recent research indicates that chronic low-dose exposures via inhalation and ingestion pathways may also give rise to non-fatal non-cancer deleterious health effects. Fatal cancers, now the primary measure of radiation injury in setting standards, may be less significant to a population in the long run than more subtle low-level impacts affecting genetic material. The latter, hard to identify or measure, may reduce developmental and reproductive capability. Given the hazardous longevity of high-level wastes, it is imperative that both protective standards and waste regulation be framed within an ethic of species responsibility. In our half century we have generated vast amounts of long-lived waste, with more promised in the coming millennium. The regulatory obligation is to isolate all nuclear wastes to best prevent any releases to the biosphere now but also to assure future generations an equal opportunity when our 'disposal' methods inevitably fail over future time, to be able to retrieve and continue to isolate the wastes that we have caused to be produced. lt follows that the standards must not calculate

  1. Removal of fission products from waste solutions using 16 different soil samples

    International Nuclear Information System (INIS)

    Bangash, M.A.; Hanif, J.

    1997-01-01

    Most of the nuclear sites use pits in the surrounding soils for the storage/disposal of low active waste (LAW) solutions. The characteristics of the soil if not suitable for the fixation or adsorption of the radioactive nuclides, may cause migration of these nuclides to hydrosphere. The phenomenon has the risk of radio toxic pollution for the living bodies therefore minerals composing the soil and their adsorption properties need to be investigated. For this purpose 16 different soil samples were collected from all over Pakistan. Mineralogical composition of the soils was determined by X-ray diffraction analysis. It was found that most of the samples contained clay minerals, illite, kaolinite and montmorillonite. Studies for the removal of fission products like, /sup 137/Cs. /sup 60/Sr and activation product /sup 60/CO from solution were carried out on these samples. The sorption experiments were performed by batch technique using radioactive as tracers. Distribution co-efficient were determined by mixing he element solution at pH 3 with the soil at soil solution ratios of 1 to 20. It is revealed from the experimental data that efficient removal of fission products from solutions is achieved by soil samples containing clay mineral montmorillonite, followed by little and kaolinite. These soils thus can be effectively used for the disposal of low level radioactive waste solutions without causing any environmental hazard. (author)

  2. Effect of phosphate ion on filtration characteristics of solids generated in simulated high level liquid waste

    International Nuclear Information System (INIS)

    Kondo, Y.

    1998-01-01

    The effect of phosphate ion on the filtration characteristics of solids generated in a high level liquid waste was experimentally examined. Addition of phosphate ion into the simulated HLLW induced the formation of phosphate such as zirconium phosphate and phosphomolybdic acid. The filtration rate of zirconium phosphate abruptly dropped in the midst of filtration because of a gel-cake formation on the filter surface. The denitration of the simulated HLLW contained zirconium phosphate improved the filterability of this gelatinous solid. The filtration rates of denitrated HLLW decreased with increase of the phosphate ion concentration, since the solids formed by denitration had irregular particle size and configuration in the simulated HLLW with phosphate ion. To increase the filtration rate of denitrated HLLW, a solid suspension filtration tester was designed. The solid-suspension accelerated the filtration rate only in the simulated HLLW with more than 1500 ppm phosphate ion concentration. Under this condition, the simple agitation can easily suspend the constituent solids of filter cake in the solution and a much higher filtration rate can be obtained because the filter cake is continuously swept from the filter surface by rotation of propellers. (authors)

  3. Hydration products and mechanical properties of hydroceramics solidified waste for simulated Non-alpha low and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    Wang Jin; Hong Ming; Wang Junxia; Li Yuxiang; Teng Yuancheng; Wu Xiuling

    2011-01-01

    In this paper, simulated non-alpha low and intermediate level radioactive wastes was handled as curing object and that of 'alkali-slag-coal fly ash-metakaolin' hydroceramics waste forms were prepared by hydrothermal synthesis method. The hydration products were analyzed by X ray diffraction. The composition of hydrates and the compressive strength of waste forms were determined and measured. The results indicate that the main crystalline phase of hydration products were analcite when the temperature was 150 to 180 degree C and the salt content ratio was 0.10 to 0.30. Analcite diffraction peaks in hydration products is increasing when the temperature was raised and the reaction time prolonged. Strength test results show that the solidified waste forms have superior compressive strength. The compressive strength gradually decreased with the increase in salt content ratio in waste forms. (authors)

  4. FFSM, Long-Term Nuclear Waste Repository Site Simulation by Monte-Carlo

    International Nuclear Information System (INIS)

    Hadlock, L.R.; Hellstrom, D.I.; Mikulis, M.J.B.; Little, A.D.; Golis, M.J.

    1988-01-01

    1 - Description of program or function: FFSM (Far Field State Model) predicts the approximate geologic and climatic state of a site for a nuclear waste repository over relatively long periods of time. The purpose of FFSM is to represent quantitatively certain events and processes that could alter the effectiveness of one or more natural barriers in a waste isolation system. The barriers treated by the model are primarily components of the geologic environment surrounding the repository, although biosphere components (e.g. climate parameters) that could affect the impact of radionuclide releases are also considered. These components are treated outside the realm of wastes or repository-induced effects, which is indicated by use of the term f ar field . The model treats both natural and man-induced changes in these barriers within a probabilistic framework, and it accounts for cumulative and interactive effects of multiple phenomena. 2 - Method of solution: Fifteen submodels are included in FFSM to account for phenomena that may be of importance individually or in combination in evaluating sites for repositories. These submodels include: undetected features, climate, worldwide glaciation, local glaciation, folding, salt dispersion, magmatic events, faulting, biosphere state, regional deformation, geomorphic processes, dissolution fronts, localized dissolution (breccia pipes), solution mining, and drilling. FFSM can be used in both a deterministic mode, to evaluate interactions or to calculate point values, and a probabilistic mode, to make statistical estimates of future changes. In the probabilistic mode, Monte Carlo simulation is used to generate output probabilities, based on user-supplied input, largely in the form of probability density functions for variable or uncertain parameters

  5. Hydrogeochemistry and simulated solute transport, Piceance Basin, northwestern Colorado

    Science.gov (United States)

    Robson, S.G.; Saulnier, G.J.

    1981-01-01

    Oil-shale mining activities in Piceance basin in northwestern Colorado could adversely affect the ground- and surface-water quality in the basin. This study of the hydrology and geochemistry of the area used ground-water solute-transport-modeling techniques to investigate the possible impact of the mines on water quality. Maps of the extent and structure of the aquifer were prepared and show that a saturated thickness of 2,000 feet occurs in the northeast part of the basin. Ground-water recharge in the upland areas in the east, south, and west parts of the basin moves down into deeper zones in the aquifer and laterally to the discharge areas along Piceance and Yellow Creeks. The saline zone and the unsaturated zone provide the majority of the dissolved solids found in the ground water. Precipitation, ion-exchange, and oxidation-reduction reactions are also occuring in the aquifer. Model simulations of ground-water pumpage in tracts C-a and C-b indicate that the altered direction of ground-water movement near the pumped mines will cause an improvement in ground-water quality near the mines and a degradation of water quality downgradient from the tracts. Model simulations of mine leaching in tract C-a and C-b indicate that equal rates of mine leaching in the tracts will produce much different effects on the water quality in the basin. Tract C-a, by virtue of its remote location from perennial streams, will primarily degrade the ground-water quality over a large area to the northeast of the tract. Tract C-b, by contrast, will primarily degrade the surface-water quality in Piceance Creek, with only localized effects on the ground-water quality. (USGS)

  6. MANAGEMENT OF SOLID WASTE GENERATED BY THE INTEGRATED STEELWORKS ACTIVITY AND SOLUTIONS TO REDUCE THE ENVIRONMENTAL IMPACT

    Directory of Open Access Journals (Sweden)

    Anişoara CIOCAN

    2010-05-01

    Full Text Available The development of steel industry is subject to solve major problems arising from industry-nature relationship, strictly targeted on pollution control and protection of natural resources and energy. In this paper we discussed about the management of solid waste generated by an integrated steelwork located near a major urban area and the adopted solutions for the reduction of environmental impact. There are summarized technical solutions that are currently applied and were proposed some solutions that can be applied in accordance with the environmental legislations. The new solutions are proposed for integrated management of solid wastes in accordance with: the exact quantification (quantitative, qualitative and the generation sources of emissions and solid wastes; controlled storage; minimization of the wastes and its harmfulness; transformation of the wastes into valuable by-products used directly by the company in a subsequent process, or by external down-stream user.

  7. Adsorption of Ag (I) from aqueous solution by waste yeast: kinetic, equilibrium and mechanism studies.

    Science.gov (United States)

    Zhao, Yufeng; Wang, Dongfang; Xie, Hezhen; Won, Sung Wook; Cui, Longzhe; Wu, Guiping

    2015-01-01

    One type of biosorbents, brewer fermentation industry waste yeast, was developed to adsorb the Ag (I) in aqueous solution. The result of FTIR analysis of waste yeast indicated that the ion exchange, chelating and reduction were the main binding mechanisms between the silver ions and the binding sites on the surface of the biomass. Furthermore, TEM, XRD and XPS results suggested that Ag(0) nanoparticles were deposited on the surface of yeast. The kinetic experiments revealed that sorption equilibrium could reach within 60 min, and the removal efficiency of Ag (I) could be still over 93 % when the initial concentration of Ag (I) was below 100 mg/L. Thermodynamic parameters of the adsorption process (ΔG, ΔH and ΔS) identified that the adsorption was a spontaneous and exothermic process. The waste yeast, playing a significant role in the adsorption of the silver ions, is useful to fast adsorb Ag (I) from low concentration.

  8. A nuclear waste deposit in space - the ultimate solution for low-cost and safe disposal

    International Nuclear Information System (INIS)

    Ruppe, H.O.; Hayn, D.; Braitinger, M.; Schmucker, R.H.

    1980-01-01

    The disposal of nuclear high-active waste (HAW) is representative for the problem of burdening the environment with highly active or toxic waste products at present and in the future. Safe disposal methods on Earth are technically very difficult to achieve and the costs of establishment and maintenance of such plants are extremely high. Furthermore the emotionally based rejection by a wide sector of the population gives sufficient reason to look for new solutions. Here, space technology can offer a real alternative - a waste deposit in space. With the Space Transportation System, which shall soon be operative, and the resulting high flight frequencies it will be possible to transport all future HAW into space at economical casts. (orig.) [de

  9. Treatment and conditioning of radioactive waste solution by natural clay minerals

    International Nuclear Information System (INIS)

    El-Dessouky, M.I.; El-Massry, E.H.; Khalifa, S.M.; Aly, H.F.

    1999-01-01

    Natural inorganic exchangers. Was used to remove caesium, cobalt and europium using zinc sulfate as coagulant also different clay minerals. These calys include, feldrspare, aswanly, bentionite, hematite, mud, calcite, basalt, magnetite, kaoline sand stone, limonite and sand. The factros affecting the removal process namely PH, particle size, temperature and weight of the clay have been studied. Highest removal for Cs-137, Co-60 and Eu-152 and 154 was achived by asswanly and bentonite. Sand stone is more effective than the other clays. Removal of Cs-137 from low level waste solution is in the order the sequence, aswanly (85.5%)> bentonite (82.2%)> sandstone (65.4%). Solidified cement products have been evaluated to determine optimum conditions of mixing most sludges contained clays by testing mechanical strength and leaching rates of the waste products. The solidified waste forms were found more acceptable for handing, storage and ultimate disposal

  10. Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

    Science.gov (United States)

    Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

    2013-11-01

    A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl11O18 and Ce2SiO5. The leaching rate of cerium over a period of 28 days was 10-5-10-6 g/(m2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products.

  11. Alpha spectrum profiling of plutonium in leached simulated high-level radioactive waste-glass

    International Nuclear Information System (INIS)

    Diamond, H.; Friedman, A.M.

    1981-01-01

    Low-geometry X-ray spectra from /sup 239/Pu and /sup 237/Np, incorporated into simulated high-level radioactive waste-glass, were transformed into depth distributions for these elements. Changes in the depth profiles were observed for a series of static leachings in 75/degree/C water. Radiochemical assay of the leach solutions revealed that little neptunium or plutonium was leached, and that the amount leached was independent of leaching time. The depth profiles of the leached specimens showed that there was selective leaching of nonradioactive components of the glass, concentrating the remaining neptunium and plutonium in a broad zone near (but not at) the glass surface. Eventual redeposition of nonradioactive material onto the glass surface inhibited further leaching

  12. Immobilization of simulated radioactive soil waste containing cerium by self-propagating high-temperature synthesis

    International Nuclear Information System (INIS)

    Mao, Xianhe; Qin, Zhigui; Yuan, Xiaoning; Wang, Chunming; Cai, Xinan; Zhao, Weixia; Zhao, Kang; Yang, Ping; Fan, Xiaoling

    2013-01-01

    A simulated radioactive soil waste containing cerium as an imitator element has been immobilized by a thermite self-propagating high-temperature synthesis (SHS) process. The compositions, structures, and element leaching rates of products with different cerium contents have been characterized. To investigate the influence of iron on the chemical stability of the immobilized products, leaching tests of samples with different iron contents with different leaching solutions were carried out. The results showed that the imitator element cerium mainly forms the crystalline phases CeAl 11 O 18 and Ce 2 SiO 5 . The leaching rate of cerium over a period of 28 days was 10 −5 –10 −6 g/(m 2 day). Iron in the reactants, the reaction products, and the environment has no significant effect on the chemical stability of the immobilized SHS products

  13. Stress corrosion cracking tests on high-level-waste container materials in simulated tuff repository environments

    International Nuclear Information System (INIS)

    Abraham, T.; Jain, H.; Soo, P.

    1986-06-01

    Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-level waste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented

  14. Simulation of Industrial Wastewater Treatment from the Suspended Impurities into the Flooded Waste Mining Workings

    Science.gov (United States)

    Bondareva, L.; Zakharov, Yu; Goudov, A.

    2017-04-01

    The paper is dedicated to the mathematical model of slurry wastewater treatment and disposal in a flooded mine working. The goal of the research is to develop and analyze the mathematical model of suspended impurities flow and distribution. Impurity sedimentation model is under consideration. Due to the sediment compaction problem solution domain can be modified. The model allows making a forecast whether volley emission is possible. Numerical simulation results for “Kolchuginskaya” coal mine presented. Impurity concentration diagrams in outflow corresponding to the real full-scale data obtained. Safely operation time mine workings like a wastewater treatment facility are estimated. The carried out calculations demonstrate that the method of industrial wastewater treatment in flooded waste mine workings can be put into practice but it is very important to observe all the processes going on to avoid volley emission of accumulated impurities.

  15. Pyrolysis of plastic packaging waste: A comparison of plastic residuals from material recovery facilities with simulated plastic waste

    International Nuclear Information System (INIS)

    Adrados, A.; Marco, I. de; Caballero, B.M.; López, A.; Laresgoiti, M.F.; Torres, A.

    2012-01-01

    Highlights: ► Pyrolysis of plastic waste. ► Comparison of different samples: real waste, simulated and real waste + catalyst. ► Study of the effects of inorganic components in the pyrolysis products. - Abstract: Pyrolysis may be an alternative for the reclamation of rejected streams of waste from sorting plants where packing and packaging plastic waste is separated and classified. These rejected streams consist of many different materials (e.g., polyethylene (PE), polypropylene (PP), polystyrene (PS), polyvinyl chloride (PVC), polyethylene terephthalate (PET), acrylonitrile butadiene styrene (ABS), aluminum, tetra-brik, and film) for which an attempt at complete separation is not technically possible or economically viable, and they are typically sent to landfills or incinerators. For this study, a simulated plastic mixture and a real waste sample from a sorting plant were pyrolyzed using a non-stirred semi-batch reactor. Red mud, a byproduct of the aluminum industry, was used as a catalyst. Despite the fact that the samples had a similar volume of material, there were noteworthy differences in the pyrolysis yields. The real waste sample resulted, after pyrolysis, in higher gas and solid yields and consequently produced less liquid. There were also significant differences noted in the compositions of the compared pyrolysis products.

  16. Processing of high-temperature simulated waste glass in a continuous ceramic melter

    International Nuclear Information System (INIS)

    Barnes, S.M.; Brouns, R.A.; Hanson, M.S.

    1980-01-01

    Recent operations have demonstrated that high-melting-point glasses and glass-ceramics can be successfully processed in joule-heated, ceramic-lined melters with minor modifications to the existing technology. Over 500 kg of simulated waste glasses have been processed at temperatures up to 1410 0 C. The processability of the two high-temperature waste forms tested is similar to existing borosilicate waste glasses. High-temperature waste glass formulations produced in the bench-scale melter exhibit quality comparing favorably to standard waste glass formulations

  17. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    International Nuclear Information System (INIS)

    Klein, Steven Karl; Day, Christy M.; Determan, John C.

    2015-01-01

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  18. Simulator for an Accelerator-Driven Subcritical Fissile Solution System

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Day, Christy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Determan, John C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-14

    LANL has developed a process to generate a progressive family of system models for a fissile solution system. This family includes a dynamic system simulation comprised of coupled nonlinear differential equations describing the time evolution of the system. Neutron kinetics, radiolytic gas generation and transport, and core thermal hydraulics are included in the DSS. Extensions to explicit operation of cooling loops and radiolytic gas handling are embedded in these systems as is a stability model. The DSS may then be converted to an implementation in Visual Studio to provide a design team the ability to rapidly estimate system performance impacts from a variety of design decisions. This provides a method to assist in optimization of the system design. Once design has been generated in some detail the C++ version of the system model may then be implemented in a LabVIEW user interface to evaluate operator controls and instrumentation and operator recognition and response to off-normal events. Taken as a set of system models the DSS, Visual Studio, and LabVIEW progression provides a comprehensive set of design support tools.

  19. Textile Dye Removal from Aqueous Solution using Modified Graphite Waste/Lanthanum/Chitosan Composite

    Science.gov (United States)

    Kusrini, E.; Wicaksono, B.; Yulizar, Y.; Prasetyanto, EA; Gunawan, C.

    2018-03-01

    We investigated various pre-treatment processes of graphite waste using thermal, mechanical and chemical methods. The aim of this work is to study the performance of modified graphite waste/lanthanum/chitosan composite (MG) as adsorbent for textile dye removal from aqueous solution. Effect of graphite waste resources, adsorbent size and lanthanum concentration on the dye removal were studied in batch experiments. Selectivity of MG was also investigated. Pre-heated graphite waste (NMG) was conducted at 80°C for 1 h, followed by mechanical crushing of the resultant graphite to 75 μm particle size, giving adsorption performance of ˜58%, ˜67%, ˜93% and ˜98% of the model dye rhodamine B (concentration determined by UV-vis spectroscopy at 554 nm), methyl orange (464 nm), methylene blue (664 nm) and methyl violet (580 nm), respectively from aqueous solution. For this process, the system required less than ˜5 min for adsorbent material to be completely saturated with the adsorbate. Further chemical modification of the pre-treated graphite waste (MG) with lanthanum (0.01 – V 0.03 M) and chitosan (0.5% w/w) did not improve the performance of dye adsorption. Under comparable experimental conditions, as those of the ‘thermal-mechanical-pre-treated-only’ (NMG), modification of graphite waste (MG) with 0.03 M lanthanum and 0.5% w/w chitosan resulted in ˜14%, ˜47%, ˜72% and ˜85% adsorption of rhodamine B, methyl orange, methylene blue and methyl violet, respectively. Selective adsorption of methylene blue at most to ˜79%, followed by methyl orange, methyl violet and rhodamine B with adsorption efficiency ˜67, ˜38, and ˜9% sequentially using MG with 0.03 M lanthanum and 0.5% w/w chitosan.

  20. Cyclic deformation-induced solute transport in tissue scaffolds with computer designed, interconnected, pore networks: experiments and simulations.

    Science.gov (United States)

    Den Buijs, Jorn Op; Dragomir-Daescu, Dan; Ritman, Erik L

    2009-08-01

    Nutrient supply and waste removal in porous tissue engineering scaffolds decrease from the periphery to the center, leading to limited depth of ingrowth of new tissue into the scaffold. However, as many tissues experience cyclic physiological strains, this may provide a mechanism to enhance solute transport in vivo before vascularization of the scaffold. The hypothesis of this study was that pore cross-sectional geometry and interconnectivity are of major importance for the effectiveness of cyclic deformation-induced solute transport. Transparent elastic polyurethane scaffolds, with computer-programmed design of pore networks in the form of interconnected channels, were fabricated using a 3D printing and injection molding technique. The scaffold pores were loaded with a colored tracer for optical contrast, cyclically compressed with deformations of 10 and 15% of the original undeformed height at 1.0 Hz. Digital imaging was used to quantify the spatial distribution of the tracer concentration within the pores. Numerical simulations of a fluid-structure interaction model of deformation-induced solute transport were compared to the experimental data. The results of experiments and modeling agreed well and showed that pore interconnectivity heavily influences deformation-induced solute transport. Pore cross-sectional geometry appears to be of less relative importance in interconnected pore networks. Validated computer models of solute transport can be used to design optimal scaffold pore geometries that will enhance the convective transport of nutrients inside the scaffold and the removal of waste, thus improving the cell survivability deep inside the scaffold.

  1. Disposal of by-products in olive oil industry: waste-to-energy solutions

    International Nuclear Information System (INIS)

    Caputo, Antonio C.; Scacchia, Federica; Pelagagge, Pacifico M.

    2003-01-01

    Olive oil production industry is characterized by relevant amounts of liquid and solid by-products [olive mill wastewater (OMW) and olive husk (OH)], and by economical, technical and organizational constraints that make difficult the adoption of environmentally sustainable waste disposal approaches. In this context, waste treatment technologies aimed at energy recovery represent an interesting alternative. In the paper, a technical and economical analysis of thermal disposal plant solutions with energy recovery has been carried out. The considered plants enable the combined treatment of OMW and OH which, although penalizes the energy recovery, proves to be feasible and profitable in a future legislative scenario when stricter limitation on OMW disposal will force oil producers to bear high disposal costs. Results are compared by using economic performance measures, including revenues from produced energy and avoided disposal costs. A sensitivity and risk analysis is also performed in order to assess the economic profitability of the proposed solutions

  2. Recovery of thorium along with uranium 233 from Thorex waste solution employing Chitosan

    International Nuclear Information System (INIS)

    Priya, S.; Reghuram, D.; Kumaraguru, K.; Vijayan, K.; Jambunathan, U.

    2003-01-01

    The low level waste solution, generated from Thorex process during the processing of U 233 , contains thorium along with traces of Th 228 and U 233 . Chitosan, a natural bio-polymer derived from Chitin, was earlier used to recover the uranium and americium. The studies were extended to find out its thorium sorption characteristics. Chitosan exhibited very good absorption of thorium (350 mg/g). Chitosan was equilibrated directly with the low level waste solution at different pH after adjusting its pH, for 60 minutes with a Chitosan to aqueous ratio of 1:100 and the raffinates were filtered and analysed. The results showed more than 99% of thorium and U 233 could be recovered by Chitosan between pH 4 and 5. Loaded thorium and uranium could be eluted from the Chitosan by 1M HNO 3 quantitatively. (author)

  3. Evaluation of extractant-coated magnetic microparticles for the recovery of hazardous metals from waste solution

    International Nuclear Information System (INIS)

    Kaminski, M. D.

    1998-01-01

    A magnetically assisted chemical separation (MACS) process was developed earlier at Argonne National Laboratory (ANL). This compact process was designed for the separation of transuranics (TRU) and radionuclides from the liquid waste streams that exist at many DOE sites, with an overall reduction in waste volume requiring disposal. The MACS process combines the selectivity afforded by solvent extractant/ion exchange materials with magnetic separation to provide an efficient chemical separation. Recently, the MACS process has been evaluated with acidic organophosphorus extractants for hazardous metal recovery from waste solutions. Moreover, process scale-up design issues have been addressed with respect to particle filtration and recovery. Two acidic organophosphorus compounds have been investigated for hazardous metal recovery, bis(2,4,4-trimethylpentyl) phosphinic acid (Cyanexreg-sign 272) and bis(2,4,4-trimethylpentyl) dithiophosphinic acid (Cyanexreg-sign 301). Coated onto magnetic microparticles, these extractants demonstrated superior recovery of hazardous metals from solution, relative to what was expected on the basis of results from solvent extraction experiments. The results illustrate the diverse applications of MACS technology for dilute waste streams. Preliminary process scale-up experiments with a high-gradient magnetic separator at Oak Ridge National Laboratory have revealed that very low microparticle loss rates are possible

  4. Applying Lean Techniques to Reduce Intravenous Waste Through Premixed Solutions and Increasing Production Frequency.

    Science.gov (United States)

    Lin, Alex C; Penm, Jonathan; Ivey, Marianne F; Deng, Yihong; Commins, Monica

    This study aims to use lean techniques and evaluate the impact of increasing the use of premixed IV solutions and increased IV production frequency on IV waste. Study was conducted at a tertiary hospital pharmacy department in three phases. Phase I included evaluation of IV waste when IV production occurred three times a day and eight premixed IV products were used. Phase II increased the number of premixed IV products to 16. Phase III then increased IV production to five times a day. During Phase I, an estimate of 2,673 IV doses were wasted monthly, accounting for 6.14% of overall IV doses. This accounted for 688 L that cost $60,135. During Phase II, the average monthly IV wastage reduced significantly to 1,069 doses (2.84%), accounting for 447 L and $34,003. During Phase III, the average monthly IV wastage was further decreased to 675 doses (1.69%), accounting for 78 L and $3,431. Hence, a potential annual saving of $449,208 could result from these changes. IV waste was reduced through the increased use of premixed solutions and increasing IV production frequency.

  5. Processing results of 1800 gallons of mercury and radioactively contaminated mixed waste rinse solution

    International Nuclear Information System (INIS)

    Thiesen, B.P.

    1993-01-01

    Mercury-contaminated rinse solution was successfully treated at the Idaho National Engineering Laboratory. This waste was generated during the decontamination of the Heat Transfer Reactor Experiment 3 reactor shield tank. Approximately 6.8 m 3 (1,800 pi) of waste was generated and placed into 33 drums. Each drum contained precipitated sludge material ranging from 2--5 cm in depth, with the average depth of about 6 cm. The pH of each drum varied from 3--11. The bulk liquid waste had a mercury level of 7.0 mg/l, which exceeded the Resource Conservation and Recovery Act limit of 0.2 mg/l. The average liquid bulk radioactivity was about 2.1 pCi/mL while the average sludge contamination was about 13,800 pCi/g. Treatment of the waste required separation of the liquid from the sludge, filtration, pH adjustment, and ion exchange. The resulting solution after treatment had mercury levels at 0.0186 mg/l and radioactivity of 0.282 pCi/ml

  6. X-RAY FLUORESCENCE ANALYSIS OF HANFORD LOW ACTIVITY WASTE SIMULANTS

    Energy Technology Data Exchange (ETDEWEB)

    Jurgensen, A; David Missimer, D; Ronny Rutherford, R

    2006-05-08

    Savannah River National Laboratory (SRNL) was requested to develop an x-ray fluorescence (XRF) spectrometry method for elemental characterization of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) pretreated low activity waste (LAW) stream to the LAW Vitrification Plant. The WTP is evaluating the potential for using XRF as a rapid turnaround technique to support LAW product compliance and glass former batching. The overall objective of this task was to develop XRF analytical methods that provide the rapid turnaround time (<8 hours) requested by the WTP, while providing sufficient accuracy and precision to determine waste composition variations. For Phase 1a, SRNL (1) evaluated, selected, and procured an XRF instrument for WTP installation, (2) investigated three XRF sample methods for preparing the LAW sub-sample for XRF analysis, and (3) initiated scoping studies on AN-105 (Envelope A) simulant to determine the instrument's capability, limitations, and optimum operating parameters. After preliminary method development on simulants and the completion of Phase 1a activities, SRNL received approval from WTP to begin Phase 1b activities with the objective of optimizing the XRF methodology. Three XRF sample methods used for preparing the LAW sub-sample for XRF analysis were studied: direct liquid analysis, dried spot, and fused glass. The direct liquid method was selected because its major advantage is that the LAW can be analyzed directly without any sample alteration that could bias the method accuracy. It also is the fastest preparation technique--a typical XRF measurement could be completed in < 1hr after sample delivery. Except for sodium, the method detection limits (MDLs) for the most important analytes in solution, the hold point elements, were achieved by this method. The XRF detection limits are generally adequate for glass former batching and product composition reporting, but may be inadequate for some species (Hg, Cd, and Ba) important

  7. X-RAY FLUORESCENCE ANALYSIS OF HANFORD LOW ACTIVITY WASTE SIMULANTS

    International Nuclear Information System (INIS)

    Jurgensen, A; David Missimer, D; Ronny Rutherford, R

    2006-01-01

    Savannah River National Laboratory (SRNL) was requested to develop an x-ray fluorescence (XRF) spectrometry method for elemental characterization of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) pretreated low activity waste (LAW) stream to the LAW Vitrification Plant. The WTP is evaluating the potential for using XRF as a rapid turnaround technique to support LAW product compliance and glass former batching. The overall objective of this task was to develop XRF analytical methods that provide the rapid turnaround time (<8 hours) requested by the WTP, while providing sufficient accuracy and precision to determine waste composition variations. For Phase 1a, SRNL (1) evaluated, selected, and procured an XRF instrument for WTP installation, (2) investigated three XRF sample methods for preparing the LAW sub-sample for XRF analysis, and (3) initiated scoping studies on AN-105 (Envelope A) simulant to determine the instrument's capability, limitations, and optimum operating parameters. After preliminary method development on simulants and the completion of Phase 1a activities, SRNL received approval from WTP to begin Phase 1b activities with the objective of optimizing the XRF methodology. Three XRF sample methods used for preparing the LAW sub-sample for XRF analysis were studied: direct liquid analysis, dried spot, and fused glass. The direct liquid method was selected because its major advantage is that the LAW can be analyzed directly without any sample alteration that could bias the method accuracy. It also is the fastest preparation technique--a typical XRF measurement could be completed in < 1hr after sample delivery. Except for sodium, the method detection limits (MDLs) for the most important analytes in solution, the hold point elements, were achieved by this method. The XRF detection limits are generally adequate for glass former batching and product composition reporting, but may be inadequate for some species (Hg, Cd, and Ba) important to

  8. Formulation Efforts for Direct Vitrification of INEEL Blend Calcine Waste Simulate: Fiscal Year 2000

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Vienna, John D.; Peeler, David K.; Reamer, I. A.

    2001-03-30

    This report documents the results of glass formulation efforts for Idaho National Engineering and Environmental Laboratory (INEEL) high level waste (HWL) calcine. Two waste compositions were used during testing. Testing started by using the Run 78 calcine composition and switched to simulated Blend calcine composition when it became available. The goal of the glass formulation efforts was to develop a frit composition that will accept higher waste loading that satisfies the glass processing and product acceptance constraints. 1. Melting temperature of 1125 ? 25?C 2. Viscosity between 2 and 10 Pa?s at the melting temperature 3. Liquidus temperature at least 100?C below the melting temperature 4. Normalized release of B, Li and Na each below 1 g/m2 (per ASTM C 1285-97) Glass formulation efforts tested several frit compositions with variable waste loadings of Run 78 calcine waste simulant. Frit 107 was selected as the primary candidate for processing since it met all process and performance criteria up to 45 mass% waste loading. When the simulated Blend calcine waste composition became available Frits 107 and 108 compositions were retested and again Frit 107 remained the primary candidate. However, both frits suffered a decrease in waste loading when switching from the Run 78 calcine to simulated Blend calcine waste composition. This was due to increase concentrations of both F and Al2O3 along with a decrease in CaO and Na2O in the simulate Blend calcine waste all of which have strong impacts on the glass properties that limit waste loading of this type of waste.

  9. 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

  10. 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.

  11. Use of almond endocarp shell in sorption of radioactive 152+154Europium from waste solutions

    International Nuclear Information System (INIS)

    Dakroury, G.A.; Khalil, T.; Abou El-Nour, F.H.

    2007-01-01

    In an attempt to remove radioactive ( 152 + 154 )Eu from waste solutions, the present study was tried to explore the possibility of using a natural by-product. Almond endocarp (AEC) shell produced from Sinai (El-Arish area) was selected as agricultural by-product in treatment of waste solutions containing ( 152 + 154 )Eu through a batch technique. The different physico-chemical characteristics of AEC such as specific surface area, total pore volume, average pore diameter, apparent density, porosity and pore size distribution were calculated. The adsorption process was described by a Freundlich type isotherm. The uptake percent of the metal ion was determined for the sorbent material as a function of contact time, pH-value, mass of the sorbent material, metal ion concentration and the effect of competing ions on the sorption process. The obtained data were analyzed and showed that almond endocarp shell powder can be considered as an efficient natural material to be used for sorption of radioactive ( 152 + 154 )Eu from their radioactive waste solutions

  12. Development of a Thermodynamic Model for the Hanford Tank Waste Operations Simulator - 12193

    Energy Technology Data Exchange (ETDEWEB)

    Carter, Robert; Seniow, Kendra [Washington River Protection Solutions, LLC, Richland, Washington (United States)

    2012-07-01

    The Hanford Tank Waste Operations Simulator (HTWOS) is the current tool used by the Hanford Tank Operations Contractor for system planning and assessment of different operational strategies. Activities such as waste retrievals in the Hanford tank farms and washing and leaching of waste in the Waste Treatment and Immobilization Plant (WTP) are currently modeled in HTWOS. To predict phase compositions during these activities, HTWOS currently uses simple wash and leach factors that were developed many years ago. To improve these predictions, a rigorous thermodynamic framework has been developed based on the multi-component Pitzer ion interaction model for use with several important chemical species in Hanford tank waste. These chemical species are those with the greatest impact on high-level waste glass production in the WTP and whose solubility depends on the processing conditions. Starting with Pitzer parameter coefficients and species chemical potential coefficients collated from open literature sources, reconciliation with published experimental data led to a self-consistent set of coefficients known as the HTWOS Pitzer database. Using Gibbs energy minimization with the Pitzer ion interaction equations in Microsoft Excel,1 a number of successful predictions were made for the solubility of simple mixtures of the chosen species. Currently, this thermodynamic framework is being programmed into HTWOS as the mechanism for determining the solid-liquid phase distributions for the chosen species, replacing their simple wash and leach factors. Starting from a variety of open literature sources, a collection of Pitzer parameters and species chemical potentials, as functions of temperature, was tested for consistency and accuracy by comparison with available experimental thermodynamic data (e.g., osmotic coefficients and solubility). Reconciliation of the initial set of parameter coefficients with the experimental data led to the development of the self-consistent set known

  13. Long-term durability experiments with concrete-based waste packages in simulated repository conditions

    International Nuclear Information System (INIS)

    Ipatti, A.

    1993-03-01

    Two extensive experiments on long-term durability of waste packages in simulated repository conditions are described. The first one is a 'half-scale experiment' comprising radioactive waste product and half-scale concrete containers in site specific groundwater conditions. The second one is 'full-scale experiment' including simulated inactive waste product and full-scale concrete container stored in slowly flowing fresh water. The scope of the experiments is to demonstrate long-term behaviour of the designed waste packages in contact with moderately concrete aggressive groundwater, and to evaluate the possible interactions between the waste product, concrete container and ground water. As the waste packages are made of high-quality concrete, provisions have been made to continue the experiments for several years

  14. Test plan for Fauske and Associates to perform tube propagation experiments with simulated Hanford tank wastes

    International Nuclear Information System (INIS)

    Carlson, C.D.; Babad, H.

    1996-05-01

    This test plan, prepared at Pacific Northwest National Laboratory for Westinghouse Hanford Company, provides guidance for performing tube propagation experiments on simulated Hanford tank wastes and on actual tank waste samples. Simulant compositions are defined and an experimental logic tree is provided for Fauske and Associates (FAI) to perform the experiments. From this guidance, methods and equipment for small-scale tube propagation experiments to be performed at the Hanford Site on actual tank samples will be developed. Propagation behavior of wastes will directly support the safety analysis (SARR) for the organic tanks. Tube propagation may be the definitive tool for determining the relative reactivity of the wastes contained in the Hanford tanks. FAI have performed tube propagation studies previously on simple two- and three-component surrogate mixtures. The simulant defined in this test plan more closely represents actual tank composition. Data will be used to support preparation of criteria for determining the relative safety of the organic bearing wastes

  15. Effect of fluidization number on the combustion of simulated municipal solid waste in a fluidized bed

    International Nuclear Information System (INIS)

    Anwar Johari; Mutahharah, M.M.; Abdul, A.; Salema, A.; Kalantarifard, A.; Rozainee, M.

    2010-01-01

    The effect of fluidization number on the combustion of simulated municipal solid was in a fluidized bed was investigated. Simulated municipal solid waste was used a sample and it was formulated from major waste composition found in Malaysia which comprised of food waste, paper, plastic and vegetable waste. Proximate and ultimate analyses of the simulated were conducted and results showed its composition was similar to the actual Malaysian municipal solid waste composition. Combustion study was carried out in a rectangular fluidized bed with sand of mean particle size of 0.34 mm as a fluidising medium. The range of fluidization numbers investigated was 3 to 11 U mf . The combustion was carried out at stoichiometric condition (Air Factor = 1). Results showed that the best fluidization number was in the range of 5 to 7 U mf with 5 U mf being the most optimum in which the bed temperature was sustained in a much longer period. (author)

  16. SPEEDUP simulation of liquid waste batch processing. Revision 1

    International Nuclear Information System (INIS)

    Shannahan, K.L.; Aull, J.E.; Dimenna, R.A.

    1994-01-01

    The Savannah River Site (SRS) has accumulated radioactive hazardous waste for over 40 years during the time SRS made nuclear materials for the United States Department of Energy (DOE) and its predecessors. This waste is being stored as caustic slurry in a large number of 1 million gallon steel tanks, some of which were initially constructed in the early 1950's. SRS and DOE intend to clean up the Site and convert this waste into stable forms which then can be safely stored. The liquid waste will be separated into a partially decontaminated low-level and radioactive high-level waste in one feed preparation operation, In-Tank Precipitation. The low-level waste will be used to make a concrete product called saltstone in the Saltstone Facility, a part of the Defense Waste Processing Facility (DWPF). The concrete will be poured into large vaults, where it will be permanently stored. The high-level waste will be added to glass-formers and waste slurry solids from another feed preparation operation, Extended Sludge Processing. The mixture will then be converted to a stable borosilicate glass by a vitrification process that is the other major part of the DWPF. This glass will be poured into stainless steel canisters and sent to a temporary storage facility prior to delivery to a permanent underground storage site

  17. Groundwater-stream-simulation experiments for the evaluation of the safety of proposed nuclear waste repositories

    International Nuclear Information System (INIS)

    Seitz, M.G.

    1981-01-01

    A bench-scale experimental design which integrates repository components to simulate a groundwater stream infiltrating a breached repository is described in this paper. An experiment performed with a nuclear waste solid and one rock core is briefly summarized. The nuclear waste solid consists of borosilicate glass containing formulated nuclear waste and is the source of the leached radionuclides. The rock core used is of granite and serves as the adsorption medium for the leached radionuclides

  18. Computational model for simulation small testing launcher, technical solution

    Energy Technology Data Exchange (ETDEWEB)

    Chelaru, Teodor-Viorel, E-mail: teodor.chelaru@upb.ro [University POLITEHNICA of Bucharest - Research Center for Aeronautics and Space, Str. Ghe Polizu, nr. 1, Bucharest, Sector 1 (Romania); Cristian, Barbu, E-mail: barbucr@mta.ro [Military Technical Academy, Romania, B-dul. George Coşbuc, nr. 81-83, Bucharest, Sector 5 (Romania); Chelaru, Adrian, E-mail: achelaru@incas.ro [INCAS -National Institute for Aerospace Research Elie Carafoli, B-dul Iuliu Maniu 220, 061126, Bucharest, Sector 6 (Romania)

    2014-12-10

    The purpose of this paper is to present some aspects regarding the computational model and technical solutions for multistage suborbital launcher for testing (SLT) used to test spatial equipment and scientific measurements. The computational model consists in numerical simulation of SLT evolution for different start conditions. The launcher model presented will be with six degrees of freedom (6DOF) and variable mass. The results analysed will be the flight parameters and ballistic performances. The discussions area will focus around the technical possibility to realize a small multi-stage launcher, by recycling military rocket motors. From technical point of view, the paper is focused on national project 'Suborbital Launcher for Testing' (SLT), which is based on hybrid propulsion and control systems, obtained through an original design. Therefore, while classical suborbital sounding rockets are unguided and they use as propulsion solid fuel motor having an uncontrolled ballistic flight, SLT project is introducing a different approach, by proposing the creation of a guided suborbital launcher, which is basically a satellite launcher at a smaller scale, containing its main subsystems. This is why the project itself can be considered an intermediary step in the development of a wider range of launching systems based on hybrid propulsion technology, which may have a major impact in the future European launchers programs. SLT project, as it is shown in the title, has two major objectives: first, a short term objective, which consists in obtaining a suborbital launching system which will be able to go into service in a predictable period of time, and a long term objective that consists in the development and testing of some unconventional sub-systems which will be integrated later in the satellite launcher as a part of the European space program. This is why the technical content of the project must be carried out beyond the range of the existing suborbital

  19. Preoperational assessment of solute release from waste rock at proposed mining operations

    International Nuclear Information System (INIS)

    Lapakko, Kim A.

    2015-01-01

    Highlights: • Modeling to estimate solute release from waste rock at proposed mines is described. • Components of the modeling process are identified and described. • Modeling inputs required are identified and described. • Examples of data generated and their application are presented. • Challenges inherent to environmental review are identified. - Abstract: Environmental assessments are conducted prior to mineral development at proposed mining operations. Among the objectives of these assessments is prediction of solute release from mine wastes projected to be generated by the proposed mining and associated operations. This paper provides guidance to those engaged in these assessments and, in more detail, provides insights on solid-phase characterization and application of kinetic test results for predicting solute release from waste rock. The logic guiding the process is consistent with general model construction practices and recent publications. Baseline conditions at the proposed site are determined and a detailed operational plan is developed and imposed upon the site. Block modeling of the mine geology is conducted to identify the mineral assemblages present, their masses and compositional variations. This information is used to select samples, representative of waste rock to be generated, that will be analyzed and tested to describe characteristics influencing waste rock drainage quality. The characterization results are used to select samples for laboratory dissolution testing (kinetic tests). These tests provide empirical data on dissolution of the various mineral assemblages present as waste rock. The data generated are used, in conjunction with environmental conditions, the proposed method of mine waste storage, and scientific and technical principles, to estimate solute release rates for the operational scale waste rock. Common concerns regarding waste rock are generation of acidic drainage and release of heavy metals and sulfate. Key solid

  20. Specific transport and storage solutions: Waste management facing current and future stakes of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Deniau, Helene; Gagner, Laurent; Gendreau, Francoise; Presta, Anne

    2006-01-01

    With major projects ongoing or being planned, and also with the daily management of radioactive waste from nuclear facilities, the role of transport and/or storage packaging has been often overlooked. Indeed, the packaging development process and transport solutions implemented are a key part of the waste management challenge: protection of people and environment. During over four decades, the AREVA Group has developed a complete and coherent system for the transport of waste produced by nuclear industries. The transport solutions integrate the factors to consider, as industrial transportation needs, various waste forms, associated hazards and current regulations. Thus, COGEMA LOGISTICS has designed, licensed and manufactured a large number of different transport, storage and dual purpose cask models for residues and all kinds of radioactive wastes. The present paper proposes to illustrate how a company acting both as a cask designer and a carrier is key to the waste management issue and how it can support the waste management policy of nuclear producers through their operational choices. We will focus on the COGEMA LOGISTICS technical solutions implemented to guarantee safe and secure transportation and storage solutions. We will describe different aspects of the cask design process, insisting on how it enables to fulfill both customer needs and regulation requirements. We will also mention the associated services developed by the AREVA Business Unit Logistics (COGEMA LOGISTICS, TRANSNUCLEAR, MAINCO, and LEMARECHAL CELESTIN) in order to manage transportation of liquid and solid waste towards interim or final storage sites. The paper has the following contents: About radioactive waste; - Radioactive waste classification; - High level activity waste and long-lived intermediate level waste; - Long-lived low level waste; - Short-lived low- and intermediate level waste; - Very low level waste; - The radioactive waste in nuclear fuel cycle; - Packaging design and

  1. Transportable Vitrification System: Operational experience gained during vitrification of simulated mixed waste

    International Nuclear Information System (INIS)

    Whitehouse, J.C.; Burket, P.R.; Crowley, D.A.; Hansen, E.K.; Jantzen, C.M.; Smith, M.E.; Singer, R.P.; Young, S.R.; Zamecnik, J.R.; Overcamp, T.J.; Pence, I.W. Jr.

    1996-01-01

    The Transportable Vitrification System (TVS) is a large-scale, fully-integrated, transportable, vitrification system for the treatment of low-level nuclear and mixed wastes in the form of sludges, soils, incinerator ash, and similar waste streams. The TVS was built to demonstrate the vitrification of actual mixed waste at U. S. Department of Energy (DOE) sites. Currently, Westinghouse Savannah River Company (WSRC) is working with Lockheed Martin Energy Systems (LMES) to apply field scale vitrification to actual mixed waste at Oak Ridge Reservation's (ORR) K-25 Site. Prior to the application of the TVS to actual mixed waste it was tested on simulated K-25 B and C Pond waste at Clemson University. This paper describes the results of that testing and preparations for the demonstration on actual mixed waste

  2. Laboratory simulation of high-level liquid waste evaporation and storage

    International Nuclear Information System (INIS)

    Anderson, P.A.

    1978-01-01

    The reprocessing of nuclear fuel generates high-level liquid wastes (HLLW) which require interim storage pending solidification. Interim storage facilities are most efficient if the HLLW is evaporated prior to or during the storage period. Laboratory evaporation and storage studies with simulated waste slurries have yielded data which are applicable to the efficient design and economical operation of actual process equipment

  3. Simulating sanitation and waste flows and their environmental impacts in East African urban centres

    NARCIS (Netherlands)

    Oyoo, R.

    2014-01-01

    Simulating Sanitation and Waste Flows and their Environmental Impacts in East African Urban Centres

    Abstract

    If improperly managed, urban waste flows can pose a significant threat to the quality of both the natural environment and public health.

  4. 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

  5. Thermal phase stability of some simulated Defense waste glasses

    International Nuclear Information System (INIS)

    May, R.P.

    1981-04-01

    Three simulated defense waste glass compositions developed by Savannah River Laboratories were studied to determine viscosity and compositional effects on the comparative thermal phase stabilities of these glasses. The glass compositions are similar except that the 411 glasses are high in lithium and low in sodium compared to the 211 glass, and the T glasses are high in iron and low in aluminum compared to the C glass. Specimens of these glasses were heat treated using isothermal anneals as short as 10 min and up to 15 days over the temperature range of 450 0 C to 1100 0 C. Additionally, a specimen of each glass was cooled at a constant cooling rate of 7 0 C/hour from an 1100 0 C melt down to 500 0 C where it was removed from the furnace. The following were observed. The slow cooling rate of 7 0 C/hour is possible as a canister centerline cooling rate for large canisters. Accordingly, it is important to note that a short range diffusion mechanism like cooperative growth phenomena can result in extensive devitrification at lower temperatures and higher yields than a long-range diffusion mechanism can; and can do it without the growth of large crystals that can fracture the glass. Refractory oxides like CeO 2 and (Ni, Mn, Fe) 2 O 4 form very rapidly at higher temperatures than silicates and significant yields can be obtained at sufficiently high temperatures that settling of these dense phases becomes a major microstructural feature during slow cooling of some glasses. These annealing studies further show that below 500 0 C there is but little devitrification occurring implying that glass canisters stored at 300 0 C may be kinetically stable despite not being thermodynamically so

  6. Thermal phase stability of some simulated Defense waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    May, R.P.

    1981-04-01

    Three simulated defense waste glass compositions developed by Savannah River Laboratories were studied to determine viscosity and compositional effects on the comparative thermal phase stabilities of these glasses. The glass compositions are similar except that the 411 glasses are high in lithium and low in sodium compared to the 211 glass, and the T glasses are high in iron and low in aluminum compared to the C glass. Specimens of these glasses were heat treated using isothermal anneals as short as 10 min and up to 15 days over the temperature range of 450/sup 0/C to 1100/sup 0/C. Additionally, a specimen of each glass was cooled at a constant cooling rate of 7/sup 0/C/hour from an 1100/sup 0/C melt down to 500/sup 0/C where it was removed from the furnace. The following were observed. The slow cooling rate of 7/sup 0/C/hour is possible as a canister centerline cooling rate for large canisters. Accordingly, it is important to note that a short range diffusion mechanism like cooperative growth phenomena can result in extensive devitrification at lower temperatures and higher yields than a long-range diffusion mechanism can; and can do it without the growth of large crystals that can fracture the glass. Refractory oxides like CeO/sub 2/ and (Ni, Mn, Fe)/sub 2/O/sub 4/ form very rapidly at higher temperatures than silicates and significant yields can be obtained at sufficiently high temperatures that settling of these dense phases becomes a major microstructural feature during slow cooling of some glasses. These annealing studies further show that below 500/sup 0/C there is but little devitrification occurring implying that glass canisters stored at 300/sup 0/C may be kinetically stable despite not being thermodynamically so.

  7. Development of simulated tank wastes for the US Department of Energy's Underground Storage Tank Integrated Demonstration

    International Nuclear Information System (INIS)

    Elmore, M.R.; Colton, N.G.; Jones, E.O.

    1992-08-01

    The purpose of the Underground Storage Tank Integrated Demonstration (USTID) is to identify and evaluate technologies that may be used to characterize, retrieve, treat, and dispose of hazardous and radioactive wastes contained in tanks on US Department of Energy sites. Simulated wastes are an essential component of the evaluation process because they provide controlled samples for technology assessment, and minimize costs and risks involved when working with radioactive wastes. Pacific Northwest Laboratory has developed a recipe to simulate Hanford single-shell tank, (SST) waste. The recipe is derived from existing process recipes, and elemental concentrations are based on characterization data from 18 SSTs. In this procedure, salt cake and metal oxide/hydroxide sludge are prepared individually, and mixed together at varying ratios depending on the specific tank, waste to be simulated or the test being conducted. Elemental and physical properties of the stimulant are comparable with analyzed tank samples, and chemical speciation in the simulant is being improved as speciation data for actual wastes become available. The nonradioactive chemical waste simulant described here is useful for testing technologies on a small scale

  8. Development of solute transport models in YMPYRÄ framework to simulate solute migration in military shooting and training areas

    Science.gov (United States)

    Warsta, L.; Karvonen, T.

    2017-12-01

    There are currently 25 shooting and training areas in Finland managed by The Finnish Defence Forces (FDF), where military activities can cause contamination of open waters and groundwater reservoirs. In the YMPYRÄ project, a computer software framework is being developed that combines existing open environmental data and proprietary information collected by FDF with computational models to investigate current and prevent future environmental problems. A data centric philosophy is followed in the development of the system, i.e. the models are updated and extended to handle available data from different areas. The results generated by the models are summarized as easily understandable flow and risk maps that can be opened in GIS programs and used in environmental assessments by experts. Substances investigated with the system include explosives and metals such as lead, and both surface and groundwater dominated areas can be simulated. The YMPYRÄ framework is composed of a three dimensional soil and groundwater flow model, several solute transport models and an uncertainty assessment system. Solute transport models in the framework include particle based, stream tube and finite volume based approaches. The models can be used to simulate solute dissolution from source area, transport in the unsaturated layers to groundwater and finally migration in groundwater to water extraction wells and springs. The models can be used to simulate advection, dispersion, equilibrium adsorption on soil particles, solubility and dissolution from solute phase and dendritic solute decay chains. Correct numerical solutions were confirmed by comparing results to analytical 1D and 2D solutions and by comparing the numerical solutions to each other. The particle based and stream tube type solute transport models were useful as they could complement the traditional finite volume based approach which in certain circumstances produced numerical dispersion due to piecewise solution of the

  9. 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.

  10. Removal of phenol from radioactive waste solutions using activated granular Carbon and activated vermiculite

    International Nuclear Information System (INIS)

    Ezz El-Din, M.R.; Atta, E.R.

    2006-01-01

    The efficiency of both activated granular carbon (AGC) and activated vermiculite (AV) in removal of phenol from aqueous waste solutions is of great interest. The aim of the present study is to compare the absorbance capacities of both AGC and AV for the removal of phenol from radioactive waste solutions and to identify the factors affecting the sorption process. The experimental results were in the form of batch sorption measurements for the removal of phenol at ambient temperature (29 ± 1 degree C) and for times up to 40 min and 180 min for AGC and AV, respectively. The results indicated that activated carbon has good efficiency to adsorb phenol. Freundlich equation has been fitted to both AGC and AV for the contaminant removal. The adsorption capacities of both AGC and AV to phenol were 17.4 mg g-1 and 4.5 mg g-1, respectively. The maximum desorption percent of phenol from both loaded AGC and loaded AV were 9 % and 0 %, respectively, and it attained within about 200 min. accordingly, it is recommended that activated carbon is preferred in the applied field for removing phenol from radioactive aqueous wastes

  11. Treatment and Conditioning of Radioactive Waste Solution by Natural Clay Minerals

    International Nuclear Information System (INIS)

    El-Dessouky, M.I.; Abdel-Raouf, M.W.; El-Massry, E.H.; Khalifa, S.M.; Aly, H.F.

    1999-01-01

    Chemical precipitation processes have been used for the treatment of radioactive elements from aqueous solution. The volume reduction is not very great and storage facilities are expensive. There are some radionuclides which are so difficult to be precipitated by this common method, so they may be precipitated by adding solid materials such as natural inorganic exchangers. In this woek, improvement the removal of caesium, cobalt and europium with zinc sulfate as coagulant and different clay minerals have been investigated. These include, Feldespare, Aswanly, Bentionite, Hematite, Mud, Calcite, Basalt, Magnetite, Kaoline, Sand stone, Limonite and Sand. The parameters affecting the precipitation process such as pH, particle size, temperature and weight of the clay have been studied. The results indicate that, the highest removal for Cs-137, Co-60 and Eu-152 and154 by Asswanly, Bentonite and Sand stone is more than the other clays. Removal of Cs-137 from low level waste solution with these three natural clays took the sequence, Aswanly (85.5%) > Bentonite (82.2%) > Sandstone (65.4%). Solidified cement products have been evaluated to determine mechanical strength and leaching rates of the waste products. The solidified waste forms were found more acceptable for handling ,storage and ultimate disposal

  12. Removal of actinides from nuclear fuel reprocessing waste solutions with bidentate organophosphorus extractants

    International Nuclear Information System (INIS)

    Schulz, W.W.; McIsaac, L.D.

    1975-08-01

    The neutral bidentate organophosphorus reagents DBDECMP (dibutyl-N,N-diethylcarbamylmethylenephosphonate) and its dihexyl analogue DHDECMP are candidate extractants for removal of actinides from certain acidic waste streams produced at the U. S. ERDA Hanford and Idaho Falls sites. Various chemical and physical properties including availability, cost, purification, alpha radiolysis, and aqueous phase solubility of DBDECMP and DHDECMP are reviewed. A conceptual flowsheet employing a 15 percent DBDECMP (or DHDECMP)--CCl 4 extractant for removal (and recovery) of Am and Pu from Hanford's Plutonium Reclamation Facility acid waste stream (CAW solution) was successfully demonstrated in laboratory-scale mixer-settler tests; this extraction scheme can be used to produce an actinide-free waste. A 30 percent DBDECMP-xylene flowsheet is being tested at the Idaho Falls site for removal of U, Np, Pu, and Am from Idaho Chemical Processing Plant first-cycle high-level raffinate to produce an actinide-free (less than 10 nCi alpha activity/gram) waste. (auth)

  13. Solid solution hardening in face centered binary alloys: Gliding statistics of a dislocation in random solid solution by atomistic simulation

    International Nuclear Information System (INIS)

    Patinet, S.

    2009-12-01

    The glide of edge and screw dislocation in solid solution is modeled through atomistic simulations in two model alloys of Ni(Al) and Al(Mg) described within the embedded atom method. Our approach is based on the study of the elementary interaction between dislocations and solutes to derive solid solution hardening of face centered cubic binary alloys. We identify the physical origins of the intensity and range of the interaction between a dislocation and a solute atom. The thermally activated crossing of a solute atom by a dislocation is studied at the atomistic scale. We show that hardening of edge and screw segments are similar. We develop a line tension model that reproduces quantitatively the atomistic calculations of the flow stress. We identify the universality class to which the dislocation depinning transition in solid solution belongs. (author)

  14. 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

  15. Digital simulation of an enrichment process for solutions by means of an advection-diffusion chamber

    International Nuclear Information System (INIS)

    Artucio, G.; Suarez, R.; Uruguay Catholic University)

    1995-01-01

    An ab-initio digital simulation of the space-time dynamics of the concentration field of a solute in an advection-diffusion chamber is done. Some questions related to the digital simulation of the concentration field using the analytical solution obtained in a previous paper are discussed

  16. Formation and filtration characteristics of solids generated in a high level liquid waste treatment process. Solids formation behavior from simulated high level liquid waste

    International Nuclear Information System (INIS)

    Kondo, Y.; Kubota, M.

    1997-01-01

    The solids formation behavior in a simulated high level liquid waste (HLLW) was experimentally examined, when the simulated HLLW was treated in the ordinary way of actual HLLW treatment process. Solids formation conditions and mechanism were closely discussed. The solids formation during a concentration step can be explained by considering the formation of zirconium phosphate, phosphomolybdic acid and precipitation of strontium and barium nitrates and their solubilities. For the solids formation during the denitration step, at least four courses were observed; formation of an undissolved material by a chemical reaction with each other of solute elements (zirconium, molybdenum, tellurium) precipitation by reduction (platinum group metals) formation of hydroxide or carbonate compounds (chromium, neodymium, iron, nickel, strontium, barium) and a physical adsorption to stable solid such as zirconium molybdate (nickel, strontium, barium). (author)

  17. Hanford Waste Vitrification Plant full-scale feed preparation testing with water and process simulant slurries

    International Nuclear Information System (INIS)

    Gaskill, J.R.; Larson, D.E.; Abrigo, G.P.

    1996-03-01

    The Hanford Waste Vitrification Plant was intended to convert selected, pretreated defense high-level waste and transuranic waste from the Hanford Site into a borosilicate glass. A full-scale testing program was conducted with nonradioactive waste simulants to develop information for process and equipment design of the feed-preparation system. The equipment systems tested included the Slurry Receipt and Adjustment Tank, Slurry Mix Evaporator, and Melter-Feed Tank. The areas of data generation included heat transfer (boiling, heating, and cooling), slurry mixing, slurry pumping and transport, slurry sampling, and process chemistry. 13 refs., 129 figs., 68 tabs

  18. Volumetric change of simulated radioactive waste glass irradiated by electron accelerator. [Silica glass

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Seichi; Furuya, Hirotaka; Inagaki, Yaohiro; Kozaka, Tetsuo; Sugisaki, Masayasu

    1987-11-01

    Density changes of simulated radioactive waste glasses, silica glass and Pyrex glass irradiated by an electron accelerator were measured by a ''sink-float'' technique. The density changes of the waste and silica glasses were less than 0.05 %, irradiated at 2.0 MeV up to the fluence of 1.7 x 10/sup 17/ ecm/sup 2/, while were remarkably smaller than that of Pyrex glass of 0.18 % shrinkage. Precision of the measurements in the density changes of the waste glass was lower than that of Pyrex glass possibly because of the inhomogeneity of the waste glass

  19. Biosorption of Pb2+ and Cu2+ in aqueous solutions using agricultural wastes

    Directory of Open Access Journals (Sweden)

    Nieva Aileen D.

    2017-01-01

    Full Text Available This study aimed to determine and compare the adsorptive capacity of Pb2+ and Cu2+ in simulated wastewater onto three agricultural wastes The adsorption capacities of Pb2+ onto the agricultural wastes can be arranged as Litchi chinensis (4.30 mg of sorbate per g of sorbent (mg g-1, 85.68% adsorption > Bambusa vulgaris (3.83 mg g-1, 76.19% adsorption > Annona squamosa (2.70 mg g-1, 53.66% adsorption while the adsorption capacities of Cu2+ onto the same agricultural wastes can be arranged in the order: Bambusa vulgaris (3.86 mg g-1, 77.17% adsorption > Annona squamosal (3.58 mg g-1, 71.58% adsorption > Litchi chinensis (3.42 mg g-1, 68.32% adsorption. The biosorbents had relatively higher adsorptive capacities with Cu2+ as compared to that of Pb2+ except for Litchi chinensis. Although the results show lower adsorptive capacity as compared to a number of treated agricultural wastes showing 80% up to almost 100% adsorption of Pb2+ and Cu2+, the results show that Annona squamosa, Bamubusa vulgaris, and Litchi chinensis are potential biosorbents and promote sustainable treatment process.

  20. Constant extension rate testing of Type 304L stainless steel in simulated waste tank environments

    International Nuclear Information System (INIS)

    Wiersma, B.J.

    1992-01-01

    New tanks for storage of low level radioactive wastes will be constructed at the Savannah River Site (SRS) of AISI Type 304L stainless steel (304L). The presence of chlorides and fluorides in the wastes may induce Stress Corrosion Cracking (SCC) in 304L. Constant Extension Rate Tests (CERT) were performed to determine the susceptibility of 304L to SCC in simulated wastes. In five of the six tests conducted thus far 304L was not susceptible to SCC in the simulated waste environments. Conflicting results were obtained in the final test and will be resolved by further tests. For comparison purposes the CERT tests were also performed with A537 carbon steel, a material similar to that utilized for the existing nuclear waste storage tanks at SRS

  1. Preparation and Characterization of Chemical Plugs Based on Selected Hanford Waste Simulants

    International Nuclear Information System (INIS)

    Mattigod, Shas V.; Wellman, Dawn M.; Parker, Kent E.; Cordova, Elsa A.; Gunderson, Katie M.; Baum, Steven R.; Crum, Jarrod V.; Poloski, Adam P.

    2008-01-01

    This report presents the results of preparation and characterization of chemical plugs based on selected Hanford Site waste simulants. Included are the results of chemical plug bench testing conducted in support of the M1/M6 Flow Loop Chemical Plugging/Unplugging Test (TP-RPP-WTP-495 Rev A). These results support the proposed plug simulants for the chemical plugging/ unplugging tests. Based on the available simulant data, a set of simulants was identified that would likely result in chemical plugs. The three types of chemical plugs that were generated and tested in this task consisted of: 1. Aluminum hydroxide (NAH), 2. Sodium aluminosilicate (NAS), and 3. Sodium aluminum phosphate (NAP). While both solvents, namely 2 molar (2 M) nitric acid (HNO3) and 2 M sodium hydroxide (NaOH) at 60 C, used in these tests were effective in dissolving the chemical plugs, the 2 M nitric acid was significantly more effective in dissolving the NAH and NAS plugs. The caustic was only slightly more effecting at dissolving the NAP plug. In the bench-scale dissolution tests, hot (60 C) 2 M nitric acid was the most effective solvent in that it completely dissolved both NAH and NAS chemical plugs much faster (1.5 - 2 x) than 2 M sodium hydroxide. So unless there are operational benefits for the use of caustic verses nitric acid, 2 M nitric acid heated to 60 C should be the solvent of choice for dissolving these chemical plugs. Flow-loop testing was planned to identify a combination of parameters such as pressure, flush solution, composition, and temperature that would effectively dissolve and flush each type of chemical plug from preformed chemical plugs in 3-inch-diameter and 4-feet-long pipe sections. However, based on a review of the results of the bench-top tests and technical discussions, the Waste Treatment Plant (WTP) Research and Technology (R and T), Engineering and Mechanical Systems (EMS), and Operations concluded that flow-loop testing of the chemically plugged pipe sections

  2. Application of stochastic dynamic simulation to waste form qualification for the HWVP vitrification process

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Westsik, J.H. Jr.

    1989-01-01

    Processing steps during the conversion of high-level nuclear waste into borosilicate glass in the Hanford Waste Vitrification Plant are being simulated on a computer by addressing transient mass balances. The results are being used to address the US Department of Energy's Waste Form Qualification requirements. The simulated addresses discontinuous (batch) operations and perturbations in the transient behavior of the process caused by errors in measurements and control actions. A collection of tests, based on process measurements, is continually checked and used to halt the simulated process when specified conditions are met. An associated set of control actions is then implemented in the simulation. The results for an example simulation are shown. 8 refs

  3. [Bioregeneration of the solutions obtained during the leaching of nonferrous metals from waste slag by acidophilic microorganisms].

    Science.gov (United States)

    Fomchenko, N V; Murav'ev, M I; Kondrat'eva, T F

    2014-01-01

    The bioregeneration of the solutions obtained after the leaching of copper and zinc from waste slag by sulfuric solutions of ferric sulfate is examined. For bioregeneration, associations of mesophilic and moderately thermqophilic acidophilic chemolithotrophic microorganisms were made. It has been shown that the complete oxidation of iron ions in solutions obtained after the leaching of nonferrous metals from waste slag is possible at a dilution of the pregnant solution with a nutrient medium. It has been found that the maximal rate of oxidation of iron ions is observed at the use of a mesophilic association of microorganisms at a threefold dilution of the pregnant solution with a nutrient medium. The application ofbioregeneration during the production of nonferrous metals from both waste and converter slags would make it possible to approach the technology of their processing using the closed cycle of workflows.

  4. The development of high performance numerical simulation code for transient groundwater flow and reactive solute transport problems based on local discontinuous Galerkin method

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Motoshima, Takayuki; Naemura, Yumi; Kubo, Shin; Kanie, Shunji

    2009-01-01

    The authors develop a numerical code based on Local Discontinuous Galerkin Method for transient groundwater flow and reactive solute transport problems in order to make it possible to do three dimensional performance assessment on radioactive waste repositories at the earliest stage possible. Local discontinuous Galerkin Method is one of mixed finite element methods which are more accurate ones than standard finite element methods. In this paper, the developed numerical code is applied to several problems which are provided analytical solutions in order to examine its accuracy and flexibility. The results of the simulations show the new code gives highly accurate numeric solutions. (author)

  5. Extraction of zirconium from simulated acidic nitrate waste using liquid membrane in hollow fiber contactor

    International Nuclear Information System (INIS)

    Pandey, G.; Chinchale, R.; Renjith, A.U.; Dixit, S.; Mukhopadhyay, S.; Shenoy, K.T.; Ghosh, S.K.

    2015-01-01

    The acidic waste raffinate stream of zirconium (Zr) purification plant contains about 2 gpl of Zr in about 2M free nitric acid. TBP, which is the most commonly used solvent in the nuclear industry, is not suitable for the extraction of Zr from this lean solution as its distribution coefficient is less than one. In house synthesized Mixed Alkyl Phosphine Oxide (MAPO) is a potential extractant for Zr from this lean stream. Intensification of this process for recovery of Zr has been attempted through use of efficient contactor, namely, hollow fiber module and efficient process, namely, simultaneous extraction and stripping across liquid membrane containing MAPO. Based on batch equilibrium studies selection of suitable concentration of extractant, composition of diluent, selection and concentration of strippant for the proposed liquid membrane system was made. The selected organic and strippant concentration was used to study suitability of application of Dispersion Liquid Membrane (DLM) in hollow fiber contactor for recovery Zr from solution simulated to Zr plant raffinate. Challenges related to stable operation of the liquid membrane system like stability of the organic phase in the micropores of lumen and stability of the dispersion during the pertraction were addressed through pressure balance across the lumen and choice of adequate dispersion condition respectively. (author)

  6. Fluidized-bed calcination of simulated commercial high-level radioactive wastes

    International Nuclear Information System (INIS)

    Freeby, W.A.

    1975-11-01

    Work is in progress at the Idaho Chemical Processing Plant to verify process flowsheets for converting simulated commercial high-level liquid wastes to granular solids using the fluidized-bed calcination process. Primary emphasis in the series of runs reported was to define flowsheets for calcining simulated Allied-General Nuclear Services (AGNS) waste and to evaluate product properties significant to calcination, solids storage, or post treatment. Pilot-plant studies using simulated high-level acid wastes representative of those to be produced by Nuclear Fuel Services, Inc. (NFS) are also included. Combined AGNS high-level and intermediate-level waste (0.26 M Na in blend) was successfully calcined when powdered iron was added (to result in a Na/Fe mole ratio of 1.0) to the feed to prevent particle agglomeration due to sodium nitrate. Long-term runs (approximately 100 hours) showed that calcination of the combined waste is practical. Concentrated AGNS waste containing sodium at concentrations less than 0.2 M were calcined successfully; concentrated waste containing 1.13 M Na calcined successfully when powdered iron was added to the feed to suppress sodium nitrate formation. Calcination of dilute AGNS waste by conventional fluid-bed techniques was unsuccessful due to the inability to control bed particle size--both particle size and bed level decreased. Fluid-bed solidification of AGNS dilute waste at conditions in which most of the calcined solids left the calciner vessel with the off-gas was successful. In such a concept, the steady-state composition of the bed material would be approximately 22 wt percent calcined solids deposited on inert particles. Calcination of simulated NFS acid waste indicated that solidification by the fluid-bed process is feasible

  7. Electrochemical destruction of organics and nitrates in simulated and actual radioactive Hanford tank waste

    International Nuclear Information System (INIS)

    Elmore, M.R.; Lawrence, W.E.

    1996-09-01

    Pacific Northwest National Laboratory has conducted an evaluation of electrochemical processing for use in radioactive tank waste cleanup activities. An electrochemical organic destruction (ECOD) process was evaluated, with the main focus being the destruction of organic compounds (especially organic complexants of radionuclides) in simulated and actual radioactive Hanford tank wastes. A primary reason for destroying the organic species in the complexant concentrate tank waste is to decomplex/defunctionalize species that chelate radionuclides. the separations processes required to remove the radionuclides are much less efficient when chelators are present. A second objective, the destruction of nitrates and nitrites in the wastes, was also assessed. Organic compounds, nitrates, and nitrites may affect waste management and safety considerations, not only at Hanford but at other US Department of Energy sites that maintain high- level waste storage tanks

  8. Iron phosphate glass containing simulated fast reactor waste: Characterization and comparison with pristine iron phosphate glass

    International Nuclear Information System (INIS)

    Joseph, Kitheri; Asuvathraman, R.; Venkata Krishnan, R.; Ravindran, T.R.; Govindaraj, R.; Govindan Kutty, K.V.; Vasudeva Rao, P.R.

    2014-01-01

    Detailed characterization was carried out on an iron phosphate glass waste form containing 20 wt.% of a simulated nuclear waste. High temperature viscosity measurement was carried out by the rotating spindle method. The Fe 3+ /Fe ratio and structure of this waste loaded iron phosphate glass was investigated using Mössbauer and Raman spectroscopy respectively. Specific heat measurement was carried out in the temperature range of 300–700 K using differential scanning calorimeter. Isoconversional kinetic analysis was employed to understand the crystallization behavior of the waste loaded iron phosphate glass. The glass forming ability and glass stability of the waste loaded glass were also evaluated. All the measured properties of the waste loaded glass were compared with the characteristics of pristine iron phosphate glass

  9. MCNP efficiency calculations of INEEL passive active neutron assay system for simulated TRU waste assays

    International Nuclear Information System (INIS)

    Yoon, W.Y.; Meachum, T.R.; Blackwood, L.G.; Harker, Y.D.

    2000-01-01

    The Idaho National Engineering and Environmental Laboratory Stored Waste Examination Pilot Plant (SWEPP) passive active neutron (PAN) radioassay system is used to certify transuranic (TRU) waste drums in terms of quantifying plutonium and other TRU element activities. Depending on the waste form involved, significant systematic and random errors need quantification in addition to the counting statistics. To determine the total uncertainty of the radioassay results, a statistical sampling and verification approach has been developed. In this approach, the total performance of the PAN nondestructive assay system is simulated using the computer models of the assay system, and the resultant output is compared with the known input to assess the total uncertainty. The supporting steps in performing the uncertainty analysis for the passive assay measurements in particular are as follows: (1) Create simulated waste drums and associated conditions; (2) Simulate measurements to determine the basic counting data that would be produced by the PAN assay system under the conditions specified; and (3) Apply the PAN assay system analysis algorithm to the set of counting data produced by simulating measurements to determine the measured plutonium mass. The validity of this simulation approach was verified by comparing simulated output against results from actual measurements using known plutonium sources and surrogate waste drums. The computer simulation of the PAN system performance uses the Monte Carlo N-Particle (MCNP) Code System to produce a neutron transport calculation for a simulated waste drum. Specifically, the passive system uses the neutron coincidence counting technique, utilizing the spontaneous fission of 240 Pu. MCNP application to the SWEPP PAN assay system uncertainty analysis has been very useful for a variety of waste types contained in 208-ell drums measured by a passive radioassay system. The application of MCNP to the active radioassay system is also feasible

  10. Bioprecipitation of uranium from alkaline waste solutions using recombinant Deinococcus radiodurans

    Energy Technology Data Exchange (ETDEWEB)

    Kulkarni, Sayali; Ballal, Anand; Apte, Shree Kumar, E-mail: aptesk@barc.gov.in

    2013-11-15

    Highlights: • Deinococcus radiodurans was genetically engineered to overexpress alkaline phosphatase (PhoK). • Deino-PhoK bioprecipitated U efficiently over a wide range of input U concentration. • A maximal loading of 10.7 g U/g of biomass at 10 mM input U was observed. • Radioresistance and U precipitation by Deino-PhoK remained unaffected by γ radiation. • Immobilization of Deino-PhoK facilitated easy separation of precipitated U. -- Abstract: Bioremediation of uranium (U) from alkaline waste solutions remains inadequately explored. We engineered the phoK gene (encoding a novel alkaline phosphatase, PhoK) from Sphingomonas sp. for overexpression in the radioresistant bacterium Deinococcus radiodurans. The recombinant strain thus obtained (Deino-PhoK) exhibited remarkably high alkaline phosphatase activity as evidenced by zymographic and enzyme activity assays. Deino-PhoK cells could efficiently precipitate uranium over a wide range of input U concentrations. At low uranyl concentrations (1 mM), the strain precipitated >90% of uranium within 2 h while a high loading capacity of around 10.7 g U/g of dry weight of cells was achieved at 10 mM U concentration. Uranium bioprecipitation by Deino-PhoK cells was not affected in the presence of Cs and Sr, commonly present in intermediate and low level liquid radioactive waste, or after exposure to very high doses of ionizing radiation. Transmission electron micrographs revealed the extracellular nature of bioprecipitated U, while X-ray diffraction and fluorescence analysis identified the precipitated uranyl phosphate species as chernikovite. When immobilized into calcium alginate beads, Deino-PhoK cells efficiently removed uranium, which remained trapped in beads, thus accomplishing physical separation of precipitated uranyl phosphate from solutions. The data demonstrate superior ability of Deino-PhoK, over earlier reported strains, in removal of uranium from alkaline solutions and its potential use in

  11. Waste treatment process for removal of contaminants from aqueous, mixed-waste solutions using sequential chemical treatment and crossflow microfiltration, followed by dewatering

    Science.gov (United States)

    Vijayan, S.; Wong, C.F.; Buckley, L.P.

    1994-11-22

    In processes of this invention aqueous waste solutions containing a variety of mixed waste contaminants are treated to remove the contaminants by a sequential addition of chemicals and adsorption/ion exchange powdered materials to remove the contaminants including lead, cadmium, uranium, cesium-137, strontium-85/90, trichloroethylene and benzene, and impurities including iron and calcium. Staged conditioning of the waste solution produces a polydisperse system of size enlarged complexes of the contaminants in three distinct configurations: water-soluble metal complexes, insoluble metal precipitation complexes, and contaminant-bearing particles of ion exchange and adsorbent materials. The volume of the waste is reduced by separation of the polydisperse system by cross-flow microfiltration, followed by low-temperature evaporation and/or filter pressing. The water produced as filtrate is discharged if it meets a specified target water quality, or else the filtrate is recycled until the target is achieved. 1 fig.

  12. A PC-based discrete event simulation model of the Civilian Radioactive Waste Management System

    International Nuclear Information System (INIS)

    Airth, G.L.; Joy, D.S.; Nehls, J.W.

    1991-01-01

    A System Simulation Model has been developed for the Department of Energy to simulate the movement of individual waste packages (spent fuel assemblies and fuel containers) through the Civilian Radioactive Waste Management System (CRWMS). A discrete event simulation language, GPSS/PC, which runs on an IBM/PC and operates under DOS 5.0, mathematically represents the movement and processing of radioactive waste packages through the CRWMS and the interaction of these packages with the equipment in the various facilities. This model can be used to quantify the impacts of different operating schedules, operational rules, system configurations, and equipment reliability and availability considerations on the performance of processes comprising the CRWMS and how these factors combine to determine overall system performance for the purpose of making system design decisions. The major features of the System Simulation Model are: the ability to reference characteristics of the different types of radioactive waste (age, burnup, etc.) in order to make operational and/or system design decisions, the ability to place stochastic variations on operational parameters such as processing time and equipment outages, and the ability to include a rigorous simulation of the transportation system. Output from the model includes the numbers, types, and characteristics of waste packages at selected points in the CRWMS and the extent to which various resources will be utilized in order to transport, process, and emplace the waste

  13. A PC-based discrete event simulation model of the civilian radioactive waste management system

    International Nuclear Information System (INIS)

    Airth, G.L.; Joy, D.S.; Nehls, J.W.

    1992-01-01

    This paper discusses a System Simulation Model which has been developed for the Department of Energy to simulate the movement of individual waste packages (spent fuel assemblies and fuel containers) through the Civilian Radioactive Waste Management System (CRWMS). A discrete event simulation language, GPSS/PC, which runs on an IBM/PC and operates under DOS 5.0, mathematically represents the movement and processing of radioactive waste packages through the CRWMS and the interaction of these packages with the equipment in the various facilities. The major features of the System Simulation Model are: the ability to reference characteristics of the different types of radioactive waste (age, burnup, etc.) in order to make operational and/or system design decisions, the ability to place stochastic variations on operational parameters such as processing time and equipment outages, and the ability to include a rigorous simulation of the transportation system. Output from the model includes the numbers, types, and characteristics of waste packages at selected points in the CRWMS and the extent to which various resources will be utilized in order to transport, process, and emplace the waste

  14. Steady-State Simulation of Steam Reforming of INEEL Tank Farm Waste

    International Nuclear Information System (INIS)

    Nichols, T.T.; Taylor, D.D.; Wood, R.A.; Barnes, C.M. email toddn@inel.gov

    2002-01-01

    A steady-state model of the Sodium-Bearing Waste steam reforming process at the Idaho National Engineering and Environmental Laboratory has been performed using the commercial ASPEN Plus process simulator. The preliminary process configuration and its representation in ASPEN are described. As assessment of the capability of the model to mechanistically predict product stream compositions was made, and fidelity gaps and opportunities for model enhancement were identified, resulting in the following conclusions: (1) Appreciable benefit is derived from using an activity coefficient model for electrolyte solution thermodynamics rather than assuming ideality (unity assumed for all activity coefficients). The concentrations of fifteen percent of the species present in the primary output stream were changed by more than 50%, relative to Electrolyte NRTL, when ideality was assumed; (2) The current baseline model provides a good start for estimating mass balances and performing integrated process optimization because it contains several key species, uses a mechanistic electrolyte thermodynamic model, and is based on a reasonable process configuration; and (3) Appreciable improvement to model fidelity can be realized by expanding the species list and the list of chemical and phase transformations. A path forward is proposed focusing on the use of an improved electrolyte thermodynamic property method, addition of chemical and phase transformations for key species currently absent from the model, and the combination of RGibbs and Flash blocks to simulate simultaneous phase and chemical equilibria in the off-gas treatment train

  15. Denitration of simulated high-level liquid wastes and selective removal of cesium with zeolites

    Energy Technology Data Exchange (ETDEWEB)

    Mimura, Hitoshi; Kanno, Takuji [Tohoku Univ., Sendai (Japan). Research Inst. of Mineral Dressing and Metallurgy; Kimura, Toshiya

    1982-03-01

    Denitration of high-level liquid wastes (HLW) from nuclear fuel reprocessing has been studied. Selective removal of Cs has been also examined with various types of zeolites. The following zeolites were used in this study; Na-synthetic mordenite (NaSM), Na-natural mordenite (NaNM), Na-natural clinoptilolite (NaCP) and H-synthetic mordenites (HSM). The effective denitration is found in the simulated HLW (15 components, 2N HNO/sub 3/ soln.) containing platinum group elements in the case of the addition of formic acid, and the pH of the solution shows the value of 5.4 when the excess formic acid ((HCOOH)/(HNO/sub 3/) = 2.0) was added. Platinum group elements may react as a catalyst for the decomposition of nitric acid and the excess formic acid. The break-through properties of NaSM column are poor for the simulated HLW, and the selective removal of Cs appears to be difficult. On the other hand, good results are obtained in the denitrated HLW, i.e., break-through capacity, total capacity and column utilization are 59.4 (meq./100 g zeolite), 147 (meq./100 g zeolite) and 40.4 (%), respectively. The break-through properties of NaSM and NaNM are superior to those of HSM. The break-through capacity and column utilization increase with an increase in column temperature.

  16. Denitration of simulated high-level liquid wastes and selective removal of cesium with zeolites

    International Nuclear Information System (INIS)

    Mimura, Hitoshi; Kanno, Takuji; Kimura, Toshiya.

    1982-01-01

    Denitration of high-level liquid wastes (HLW) from nuclear fuel reprocessing has been studied. Selective removal of Cs has been also examined with various types of zeolites. The following zeolites were used in this study; Na-synthetic mordenite (NaSM), Na-natural mordenite (NaNM), Na-natural clinoptilolite (NaCP) and H-synthetic mordenites (HSM). The effective denitration is found in the simulated HLW (15 components, 2N HNO 3 soln.) containing platinum group elements in the case of the addition of formic acid, and the pH of the solution shows the value of 5.4 when the excess formic acid ([HCOOH]/[HNO 3 ] = 2.0) was added. Platinum group elements may react as a catalyst for the decomposition of nitric acid and the excess formic acid. The break-through properties of NaSM column are poor for the simulated HLW, and the selective removal of Cs appears to be difficult. On the other hand, good results are obtained in the denitrated HLW, i.e., break-through capacity, total capacity and column utilization are 59.4 (meq./100 g zeolite), 147 (meq./100 g zeolite) and 40.4 (%), respectively. The break-through properties of NaSM and NaNM are superior to those of HSM. The break-through capacity and column utilization increase with an increase in column temperature. (author)

  17. 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

  18. Cyclohexanone solvent extraction of 99TcO4 from alkaline nuclear waste solutions

    International Nuclear Information System (INIS)

    Schulz, W.W.

    1980-01-01

    Although the 99 Tc cyclohexanone solvent extraction process is still in the bench-scale development stage, the process appears well suited for engineering-scale removal of 99 Tc from alkaline Hanford waste solutions. The most pressing process development need is to resolve the phase disengaging problems encountered during water stripping operations. Stripping tests in pulse columns and/or centrifugal contactors are particularly needed to determine the magnitude of the phase disengaging problem in engineering-scale equipment and to find suitable remedies. 5 figures, 7 tables

  19. New Engineering Solutions in Creation of Mini-BOF for Metallic Waste Recycling

    Science.gov (United States)

    Eronko, S. P.; Gorbatyuk, S. M.; Oshovskaya, E. V.; Starodubtsev, B. I.

    2017-12-01

    New engineering solutions used in design of the mini melting unit capable of recycling industrial and domestic metallic waste with high content of harmful impurities are provided. High efficiency of the process technology implemented with its use is achieved due to the possibility of the heat and mass transfer intensification in the molten metal bath, controlled charge into it of large amounts of reagents in lumps and in fines, and cut-off of remaining process slag during metal tapping into the teeming ladle.

  20. An adaptive nonlinear solution scheme for reservoir simulation

    Energy Technology Data Exchange (ETDEWEB)

    Lett, G.S. [Scientific Software - Intercomp, Inc., Denver, CO (United States)

    1996-12-31

    Numerical reservoir simulation involves solving large, nonlinear systems of PDE with strongly discontinuous coefficients. Because of the large demands on computer memory and CPU, most users must perform simulations on very coarse grids. The average properties of the fluids and rocks must be estimated on these grids. These coarse grid {open_quotes}effective{close_quotes} properties are costly to determine, and risky to use, since their optimal values depend on the fluid flow being simulated. Thus, they must be found by trial-and-error techniques, and the more coarse the grid, the poorer the results. This paper describes a numerical reservoir simulator which accepts fine scale properties and automatically generates multiple levels of coarse grid rock and fluid properties. The fine grid properties and the coarse grid simulation results are used to estimate discretization errors with multilevel error expansions. These expansions are local, and identify areas requiring local grid refinement. These refinements are added adoptively by the simulator, and the resulting composite grid equations are solved by a nonlinear Fast Adaptive Composite (FAC) Grid method, with a damped Newton algorithm being used on each local grid. The nonsymmetric linear system of equations resulting from Newton`s method are in turn solved by a preconditioned Conjugate Gradients-like algorithm. The scheme is demonstrated by performing fine and coarse grid simulations of several multiphase reservoirs from around the world.

  1. A testing program to evaluate the effects of simulant mixed wastes on plastic transportation packaging components

    International Nuclear Information System (INIS)

    Nigrey, P.J.; Dickens, T.G.

    1997-01-01

    Based on regulatory requirements for Type A and B radioactive material packaging, a Testing Program was developed to evaluate the effects of mixed wastes on plastic materials which could be used as liners and seals in transportation containers. The plastics evaluated in this program were butadiene-acrylonitrile copolymer (Nitrile rubber), cross-linked polyethylene, epichlorohydrin, ethylene-propylene rubber (EPDM), fluorocarbons, high-density polyethylene (HDPE), butyl rubber, polypropylene, polytetrafluoroethylene, and styrene-butadiene rubber (SBR). These plastics were first screened in four simulant mixed wastes. The liner materials were screened using specific gravity measurements and seal materials by vapor transport rate (VTR) measurements. For the screening of liner materials, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals. The tests also indicated that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only Viton passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture waste, none of the seal materials met the screening criteria. Those materials which passed the screening tests were subjected to further comprehensive testing in each of the simulant wastes. The materials were exposed to four different radiation doses followed by exposure to a simulant mixed waste at three temperatures and four different exposure times (7, 14, 28, 180 days). Materials were tested by measuring specific gravity, dimensional, hardness, stress cracking, VTR, compression set, and tensile properties. The second phase of this Testing Program involving the comprehensive testing of plastic liner has been completed and for seal materials is currently in progress

  2. Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilmarth, William R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-01-21

    compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in high concentration in this LAW Off-Gas Condensate stream is Technetium-99 (99Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are also expected to be in appreciable concentration in the LAW Off-Gas Condensate are 129I, 90Sr, 137Cs, and {sup 241}Am. This report discusses results of preliminary radionuclide decontamination testing of the simulant. Testing examined use of Monosodium Titanate (MST) to remove 90Sr and actinides, inorganic reducing agents for 99Tc, and zeolites for 137Cs. Test results indicate that excellent removal of 99Tc was achieved using Sn(II)Cl2 as a reductant, coupled with sorption onto hydroxyapatite, even in the presence of air and at room temperature. This process was very effective at neutral pH, with a Decontamination

  3. Production of furfural from waste aqueous hemicellulose solution of hardwood over ZSM-5 zeolite.

    Science.gov (United States)

    Gao, Hongling; Liu, Haitang; Pang, Bo; Yu, Guang; Du, Jian; Zhang, Yuedong; Wang, Haisong; Mu, Xindong

    2014-11-01

    This study aimed to produce furfural from waste aqueous hemicellulose solution of a hardwood kraft-based dissolving pulp production processing in a green method. The maximum furfural yield of 82.4% and the xylose conversion of 96.8% were achieved at 463K, 1.0g ZSM-5, 1.05g NaCl and organic solvent-to-aqueous phase ratio of 30:15 (V/V) for 3h. The furfural yield was just 51.5% when the same concentration of pure xylose solution was used. Under the optimized condition, furfural yield was still up to 67.1% even after the fifth reused of catalyst. Catalyst recycling study showed that ZSM-5 has a certain stability and can be efficiently reused. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. Fluorescent Lamp Glass Waste Incorporation into Clay Ceramic: A Perfect Solution

    Science.gov (United States)

    Morais, Alline Sardinha Cordeiro; Vieira, Carlos Maurício Fontes; Rodriguez, Rubén Jesus Sanchez; Monteiro, Sergio Neves; Candido, Veronica Scarpini; Ferreira, Carlos Luiz

    2016-09-01

    The mandatory use of fluorescent lamps as part of a Brazilian energy-saving program generates a huge number of spent fluorescent lamps (SFLs). After operational life, SFLs cannot be disposed as common garbage owing to mercury and lead contamination. Recycling methods separate contaminated glass tubes and promote cleaning for reuse. In this work, glass from decontaminated SFLs was incorporated into clay ceramics, not only as an environmental solution for such glass wastes and clay mining reduction but also due to technical and economical advantages. Up to 30 wt.% of incorporation, a significant improvement in fired ceramic flexural strength and a decrease in water absorption was observed. A prospective analysis showed clay ceramic incorporation as an environmentally correct and technical alternative for recycling the enormous amount of SFLs disposed of in Brazil. This could also be a solution for other world clay ceramic producers, such as US, China and some European countries.

  5. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    Energy Technology Data Exchange (ETDEWEB)

    Digby Macdonald; Brian Marx; Balaji Soundararajan; Morgan Smith

    2005-07-28

    The different tasks that have been carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA), which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals, and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples in order to exactly predict the corrosion mechanisms; (7) Wavelet analysis of EC noise data from steel samples undergoing corrosion in an environment similar to that of the high level waste storage containers, to extract data pertaining to general, pitting and stress corrosion processes, from the overall data. The work has yielded a number of important findings, including an unequivocal demonstration of the role of chloride ion in passivity breakdown on nickel in terms of cation vacancy generation within the passive film, the first detection and characterization of individual micro fracture

  6. LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, K.; Nash, C.; McCabe, D.

    2014-09-29

    compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in greatest abundance in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are low but are also expected to be in measurable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, {sup 241}Pu, and {sup 241}Am. These are present due to their partial volatility and some entrainment in the off-gas system. This report discusses results of optimized {sup 99}Tc decontamination testing of the simulant. Testing examined use of inorganic reducing agents for {sup 99}Tc. Testing focused on minimizing the quantity of sorbents/reactants added, and minimizing mixing time to reach the decontamination targets in this simulant formulation. Stannous chloride and ferrous sulfate were tested as reducing agents to determine the minimum needed to convert soluble pertechnetate

  7. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions

    International Nuclear Information System (INIS)

    Tunsu, Cristian; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-01-01

    Highlights: • A wet-based decontamination process for fluorescent lamp waste is proposed. • Mercury can be leached using iodine in potassium iodide solution. • The efficiency of the process increases with an increase in leachant concentration. • Selective leaching of mercury from rare earth elements is achieved. • Mercury is furthered recovered using ion exchange, reduction or solvent extraction. - Abstract: With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent’s concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I 2 /KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5 M I 2 /KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe 4 BTBP showed good removal of mercury, with an

  8. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions

    Energy Technology Data Exchange (ETDEWEB)

    Tunsu, Cristian, E-mail: tunsu@chalmers.se; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-02-15

    Highlights: • A wet-based decontamination process for fluorescent lamp waste is proposed. • Mercury can be leached using iodine in potassium iodide solution. • The efficiency of the process increases with an increase in leachant concentration. • Selective leaching of mercury from rare earth elements is achieved. • Mercury is furthered recovered using ion exchange, reduction or solvent extraction. - Abstract: With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent’s concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I{sub 2}/KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5 M I{sub 2}/KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe{sub 4}BTBP showed good removal of mercury

  9. Simulation and characterization of a Hanford high-level waste slurry

    International Nuclear Information System (INIS)

    Russell, R.L.; Smith, H.D.

    1996-09-01

    The baseline waste used for this simulant is a blend of wastes from tanks 101-AZ, 102-AZ, 106-C, and 102-AY that have been through water washing. However, the simulant used in this study represents a combination of tank waste slurries and should be viewed as an example of the slurries that might be produced by blending waste from various tanks. It does not imply that this is representative of the actual waste that will be delivered to the privatization contractor(s). This blended waste sludge simulant was analyzed for grain size distribution, theological properties both as a function of concentration and aging, and calcining characteristics. The grain size distribution allows a comparison with actual waste with respect to theological properties. Slurries with similar grain size distributions of the same phases are expected to exhibit similar theological properties. Rheological properties may also change because of changes in the slurry's particulate supernate chemistry due to aging. Low temperature calcination allows the potential for hazardous gas generation to be investigated

  10. Investigations regarding the wet decontamination of fluorescent lamp waste using iodine in potassium iodide solutions.

    Science.gov (United States)

    Tunsu, Cristian; Ekberg, Christian; Foreman, Mark; Retegan, Teodora

    2015-02-01

    With the rising popularity of fluorescent lighting, simple and efficient methods for the decontamination of discarded lamps are needed. Due to their mercury content end-of-life fluorescent lamps are classified as hazardous waste, requiring special treatment for disposal. A simple wet-based decontamination process is required, especially for streams where thermal desorption, a commonly used but energy demanding method, cannot be applied. In this study the potential of a wet-based process using iodine in potassium iodide solution was studied for the recovery of mercury from fluorescent lamp waste. The influence of the leaching agent's concentration and solid/liquid ratio on the decontamination efficiency was investigated. The leaching behaviour of mercury was studied over time, as well as its recovery from the obtained leachates by means of anion exchange, reduction, and solvent extraction. Dissolution of more than 90% of the contained mercury was achieved using 0.025/0.05 M I2/KI solution at 21 °C for two hours. The efficiency of the process increased with an increase in leachant concentration. 97.3 ± 0.6% of the mercury contained was dissolved at 21 °C, in two hours, using a 0.25/0.5M I2/KI solution and a solid to liquid ratio of 10% w/v. Iodine and mercury can be efficiently removed from the leachates using Dowex 1X8 anion exchange resin or reducing agents such as sodium hydrosulphite, allowing the disposal of the obtained solution as non-hazardous industrial wastewater. The extractant CyMe4BTBP showed good removal of mercury, with an extraction efficiency of 97.5 ± 0.7% being achieved in a single stage. Better removal of mercury was achieved in a single stage using the extractants Cyanex 302 and Cyanex 923 in kerosene, respectively. Copyright © 2014 Elsevier Ltd. All rights reserved.

  11. Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-17

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

  12. Removal of cesium and strontium from low active waste solutions by zeolites

    International Nuclear Information System (INIS)

    Jain, Savita; Ramaswamy, M.; Theyyunni, T.K.

    1994-01-01

    Ion exchange, crystallographic and thermal characteristics of sodium, cesium and strontium forms of locally available synthetic zeolites have been investigated. X-ray and differential thermal analyses have confirmed that the synthetic materials AR1 and 4A belonged to the mordenite and A type families of zeolites respectively. Equilibrium uptake of cesium and strontium ions by sodium forms of zeolite was studied as a function of time, pH and sodium concentration. It was found that the rate of sorption by AR1 was higher than that by 4A. In regard to pH, distribution of nuclides on zeolites was found to pass through maxima at a pH value of around 9. Sodium ion interfered with the sorption of cesium and strontium by zeolites. However, at sodium concentration ≤ 0.01 M, distribution coefficient values for these nuclides were sufficiently high to merit consideration of these zeolites for low level waste treatment. Lab-scale column runs using 5 ml beds of materials showed that the zeolites AR1 and 4A were very effective in removing cesium and strontium nuclides respectively from large volumes (a decontamination factor of 50 for a throughput of 6000 bed volumes) of actual low level waste solutions. Thus, the zeolite system has a potential future for large scale application in the treatment of low level wastes. (author). 6 refs., 5 figs., 6 tabs

  13. Uranium extraction from aqueous solution using dried and pyrolyzed tea and coffee wastes

    International Nuclear Information System (INIS)

    Zaynab Aly

    2013-01-01

    The adsorption of U(VI) onto dried and pyrolyzed tea and coffee wastes was investigated. The adsorption properties of the materials were characterized by measuring uranium uptake as a function of solution pH, kinetics and adsorption isotherms. pH profile of uranium adsorption where UO 2 2+ is expected to be the predominant species was measured between pH 0 and 4. Both Langmuir and Freundlich adsorption models were used to describe adsorption equilibria, and corresponding constants evaluated. Using the Langmuir model, the maximum adsorption capacity of uranium by dried tea and coffee wastes was 59.5 and 34.8 mg/g, respectively at 291 K. Adsorption thermodynamic constants, ΔHdeg ΔSdeg and ΔGdeg were also calculated from adsorption data obtained at three different temperatures. Adsorption thermodynamics of uranyl ions on dried tea and coffee systems indicated spontaneous and endothermic processes. Additionally, a Lagergren pseudo-second-order kinetic model was used to fit the kinetic experimental data for both adsorbents and the constants evaluated. Dried tea and coffee wastes proved to be effective adsorbents with high capacities and significant advantage of a very low cost. (author)

  14. Reuse of waste beer yeast sludge for biosorptive decolorization of reactive blue 49 from aqueous solution.

    Science.gov (United States)

    Wang, Baoe; Guo, Xiu

    2011-06-01

    Reactive blue 49 was removed from aqueous solution by biosorption using powder waste sludge composed of Saccharomyces cerevisiae from the beer-brewing industry. The effect of initial pH, temperature and the biosorption thermodynamics, equilibrium, kinetics was investigated in this study. It was found that the biosorption capacity was at maximum at initial pH 3, that the effect of temperature on biosorption of reactive blue 49 was only slight in relation to the large biosorption capacity (25°C, 361 mg g(-1)) according as the biosorption capacity decreased only 43 mg g(-1) at the temperature increased from 25 to 50°C. The biosorption was spontaneous, exothermic in nature and the dye molecules movements decreased slightly in random at the solid/liquid interface during the biosorption of dye on biosorbents. The biosorption equilibrium data could be described by Freundich isotherm model. The biosorption rates were found to be consistent with a pseudo-second-order kinetics model. The functional group interaction analysis between waste beer yeast sludge and reactive blue 49 by the aid of Fourier transform infrared (abbr. FTIR) spectroscopy indicated that amino components involved in protein participated in the biosorption process, which may be achieved by the mutual electrostatic adsorption process between the positively charged amino groups in waste beer yeast sludge with negatively charged sulfonic groups in reactive blue 49.

  15. Fracturing of simulated high-level waste glass in canisters

    International Nuclear Information System (INIS)

    Peters, R.D.; Slate, S.C.

    1981-09-01

    Waste-glass castings generated from engineering-scale developmental processes at the Pacific Northwest Laboratory are generally found to have significant levels of cracks. The causes and extent of fracturing in full-scale canisters of waste glass as a result of cooling and accidental impact are discussed. Although the effects of cracking on waste-form performance in a repository are not well understood, cracks in waste forms can potentially increase leaching surface area. If cracks are minimized or absent in the waste-glass canisters, the potential for radionuclide release from the canister package can be reduced. Additional work on the effects of cracks on leaching of glass is needed. In addition to investigating the extent of fracturing of glass in waste-glass canisters, methods to reduce cracking by controlling cooling conditions were explored. Overall, the study shows that the extent of glass cracking in full-scale, passively-cooled, continuous melting-produced canisters is strongly dependent on the cooling rate. This observation agrees with results of previously reported Pacific Northwest Laboratory experiments on bench-scale annealed canisters. Thus, the cause of cracking is principally bulk thermal stresses. Fracture damage resulting from shearing at the glass/metal interface also contributes to cracking, more so in stainless steel canisters than in carbon steel canisters. This effect can be reduced or eliminated with a graphite coating applied to the inside of the canister. Thermal fracturing can be controlled by using a fixed amount of insulation for filling and cooling of canisters. In order to maintain production rates, a small amount of additional facility space is needed to accomodate slow-cooling canisters. Alternatively, faster cooling can be achieved using the multi-staged approach. Additional development is needed before this approach can be used on full-scale (60-cm) canisters

  16. Mathematical simulation of the behaviour of the spent organic extractive solution near the injection well area in the case of underground disposal

    International Nuclear Information System (INIS)

    Istomin, A.D.; Noskov, M.D.; Balakhonov, V.G.; Zubkov, A.A.; Egorov, G.F.

    2005-01-01

    A mathematical model is presented of the processes in the collector seam under combined disposal of organic and radioactive wastes in porous geological strata of deep bedding. The model describes filtration, mass transfer, sorption and desorption of radionuclides, radioactive decay, decomposition of organic components and heat transfer. The computer software is developed. The results of simulating the thermal field dynamics, behaviour of the components of the spent organic extractive solution and water radioactive wastes in the collector seam of deep bedding are presented [ru

  17. Reaction of water with a simulated high-level nuclear waste glass at 3000C, 300 bars

    International Nuclear Information System (INIS)

    McCarthy, G.J.; Scheetz, B.E.; Komarneni, S.; Smith, D.K.

    1978-01-01

    The hydrothermal stability of high-level nuclear wastes is an important consideration in establishing waste form acceptance criteria for a geological repository in basalt. A detailed examination of the stability of a typical simulated high-level waste glass and pressurized water at 300 0 C in a closed system has shown that extensive reaction occurred within a few weeks. The water acted first as a catalyst-solvent in devitrification of the glass and in dissolution, transport, and recrystallization of some of its constituents, and, second, as a reactant in forming hydrated and hydroxylated phases. This reaction with water resulted in the conversion of a solid shard of glass into a fragmented and partially dispersed mass of crystalline and noncrystalline material plus dissolved species within two weeks. The major crystalline reaction products were found to be analogs of naturally occurring minerals: (Cs,Na,Rb) 2 (UO 2 ) 2 .(Si 2 O 5 ) 3 .4H 2 O (weeksite) and a series of pyroxene-structure phases, (Na,Ca) (Fe,Zn,Ti)Si 2 O 6 (acmite, acmite--augites). Weeksite, however, is not expected to have long-term stability in the basalt environment. Much of the Na and Mo, and almost all of the B, in the original glass was identified in the product solutions. Of the elements or analogs of long-lived, hazardous radionuclides studied in this work, only Cs was observed in these solutions in substantial amounts. Although the comparatively rapid and extensive reactions at 300 0 C would appear to require that an acceptable glass would have low waste and heat loading, it is suggested that there is good potential for favorable glass--basalt--water hydrothermal interactions. Favorable interactions would mean that, in the event of a hydrothermal incident, the interaction products would be more stable than the original waste form and would remain in the immediate repository

  18. Phase field simulations of ice crystal growth in sugar solutions

    NARCIS (Netherlands)

    Sman, Van Der R.G.M.

    2016-01-01

    We present the first model ever, that describes explicitly ice crystal growth in a sugar solution during freezing. This 2-D model uses the phase field method, supplemented with realistic, and predictive theories on the thermodynamics and (diffusion) kinetics of this food system. We have to make

  19. A Study of the applicability of biomineralization process to the treatment of radioactive waste solutions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Heon; Choi, Kwang Soon; Kim, Yong Bok

    2004-01-01

    A study has been carried out on the removal of uranium and radionuclides like {sup 137}Cs, {sup 60}Co and {sup 90}Sr from radioactive waste solutions using a bioreactor of Serratia sp.-biofilm which is capable of producing an acid type phosphatase enzyme that liberates HPO{sub 4}{sup 2-} from a suitable organic phosphate donor and of stoichiometrically forming precipitates of the metal ions (M{sup 2+}) such as insoluble MHPO{sub 4}(HUP) at the cell surface. Sorption behaviours of the Serratia sp.-biofilm for UO{sub 2}{sup 2+}, Cs{sup +}, Sr{sup 2+} and Co{sup 2+} were investigated using synthetic radioactive waste solutions containing {sup 137}Cs, {sup 60}Co and {sup 85}Sr. Uranium was effectively removed as hydrogen uranyl phosphate (HUP) in the presence of glycerol-2-phosphate. Cs{sup +}, Sr{sup 2+} and Co{sup 2+} were removed via an intercalation by either continuous co-crystal(precipitate) growth along with U or by a pre-formed HUP host crystal as an inorganic ion exchanger. Effect of sodium ion on the sorption behaviour of serratia sp.-HUP-bioreactor and its stability were investigated. In addition, Sorption behaviours of the serratia sp.-HUP-bioreactor were compared with those of Amberlite IRN 77,cation exchanger and ammoniumphosphotungstate/silical gel which shows selective sorption property for Cs{sup +}.

  20. A Study of the applicability of biomineralization process to the treatment of radioactive waste solutions

    International Nuclear Information System (INIS)

    Lee, Chang Heon; Choi, Kwang Soon; Kim, Yong Bok

    2004-01-01

    A study has been carried out on the removal of uranium and radionuclides like 137 Cs, 60 Co and 90 Sr from radioactive waste solutions using a bioreactor of Serratia sp.-biofilm which is capable of producing an acid type phosphatase enzyme that liberates HPO 4 2- from a suitable organic phosphate donor and of stoichiometrically forming precipitates of the metal ions (M 2+ ) such as insoluble MHPO 4 (HUP) at the cell surface. Sorption behaviours of the Serratia sp.-biofilm for UO 2 2+ , Cs + , Sr 2+ and Co 2+ were investigated using synthetic radioactive waste solutions containing 137 Cs, 60 Co and 85 Sr. Uranium was effectively removed as hydrogen uranyl phosphate (HUP) in the presence of glycerol-2-phosphate. Cs + , Sr 2+ and Co 2+ were removed via an intercalation by either continuous co-crystal(precipitate) growth along with U or by a pre-formed HUP host crystal as an inorganic ion exchanger. Effect of sodium ion on the sorption behaviour of serratia sp.-HUP-bioreactor and its stability were investigated. In addition, Sorption behaviours of the serratia sp.-HUP-bioreactor were compared with those of Amberlite IRN 77,cation exchanger and ammoniumphosphotungstate/silical gel which shows selective sorption property for Cs +

  1. Development and applications of the channel network model for simulations of flow and solute transport in fractured rock

    International Nuclear Information System (INIS)

    Gylling, B.

    1997-01-01

    The Channel Network model and its computer implementation, the code CHAN3D, for simulations of fluid flow and transport of solutes have been developed. The tool may be used for performance and safety assessments of deep lying repositories in fractured rocks for nuclear and other hazardous wastes, e.g. chemical wastes. It may also be used to simulate and interpret field experiments of flow and transport in large or small scale. Fluid flow and solute transport in fractured media are of interest in the performance assessment of a repository for hazardous waste, located at depth in crystalline rock, with potential release of solutes. Fluid flow in fractured rock is found to be very unevenly distributed due to the heterogeneity of the medium. The water will seek the easiest path, channels, under a prevailing pressure gradient. Solutes in the flowing water may be transported through preferential paths and migrate from the water in the fractures into the stagnant water in the rock matrix. There, sorbing solutes may be sorbed on the micro surfaces within the matrix. The diffusion into the matrix and the sorption process may significantly retard the transport of species and increase the time available for radionuclide decay. Channelling and matrix diffusion contribute to the dispersion of solutes in the water. Important for performance assessment is that channeling may cause a portion of the solutes to arrive much faster than the rest of the solutes. Simulations of field experiments at the Aespoe Hard Rock Laboratory using the Channel Network model have been performed. The application of the model to the site and the simulation results of the pumping and tracer tests are presented. The results show that the model is capable of describing the hydraulic gradient and of predicting flow rates and tracer transport obtained in the experiments. The data requirements for the Channel Network model have been investigated to determine which data are the most important for predictions

  2. Inorganic sorbents for radiostrontium removal from waste solutions: selectivity and role of calixarenes

    International Nuclear Information System (INIS)

    Vijayan, S.; Belikov, K.; Drapailo, A.

    2011-01-01

    The challenge in the remediation of 90 Sr-contaminated waters arises from the need to achieve very high removal efficiencies to meet discharge targets from waste effluents containing relatively high concentrations of non-radioactive cations. Low-cost natural zeolites are not selective for strontium over other divalent cations, notably such ions as calcium; and produce low 90 Sr removal performance, and large volumes of spent sorbent waste. The synthesis and use of selective, synthetic inorganic sorbents could prove to be a feasible approach for high 90 Sr removal efficiencies, and much smaller volumes of secondary solid waste generation. The essential advantages of inorganic sorbents include their stability and resistance to radiation, and the potential for producing stable waste forms such as vitrified glass or ceramics for disposal. However, the cost of strontium-specific sorbents is prohibitive for large-scale applications at present. This paper is a review of the reported information on removal mechanisms and performance of Sr-specific inorganic sorbents. The analysis has revealed promising performance, efficiency and selectivity for strontium removal from solutions containing low and high concentrations of salts. The leading sorbents are crystalline silicotitanate and oxides of metals such as titanium. An initial assessment has also been made of the performance of calixarene-based macrocyclic compounds. These are known for their selectivity for strontium in solvent extraction processes. From the initial strontium removal results in bench-scale tests using different solid substrates, impregnated with calixarene derivatives, only sodium-mordenite impregnated with calyx[8]arene octamide gave an overall strontium removal efficiency in the range of 90 to 95% in the presence of 3.5 ppm calcium. There was no improvement observed for strontium-removal efficiency or selectivity over calcium in the calixarene-impregnated inorganic sorbent matrix. In several tests, the

  3. Performance of aged cement - polymer composite immobilizing borate waste simulates during flooding scenarios

    International Nuclear Information System (INIS)

    Eskander, S.B.; Bayoumi, T.A.; Saleh, H.M.

    2012-01-01

    An advanced composite of cement and water extended polyester based on the recycled Poly(ethylene terephthalate) waste was developed to incorporate the borate waste. Previous studies have reported the characterizations of the waste composite (cement-polymer composite immobilizing borate waste simulates) after 28 days of curing time. The current work studied the performance of waste composite aged for seven years and subjected to flooding scenario during 260 days using three types of water. The state of waste composite was assessed at the end of each definite interval of the water infiltration through visual examination and mechanical measurement. Scanning electron microscopy, infrared spectroscopy, X-ray diffraction and thermal analyses were used to investigate the changes that may occur in the microstructure of the waste composite under aging and flooding effects. The actual experimental results indicated reasonable evidence for the waste composite. Acceptable consistency was confirmed for the waste composite even after aging seven years and exposure to flooding scenario for 260 days.

  4. The comparison of DYNA3D to approximate solutions for a partially- full waste storage tank subjected to seismic loading

    International Nuclear Information System (INIS)

    Zaslawsky, M.; Kennedy, W.N.

    1992-01-01

    Mathematical solutions to the problem consisting of a partially-full waste tank subjected to seismic loading, embedded in soil, is classically difficult in that one has to address: soil-structure interaction, fluid-structure interaction, non-linear behavior of material, dynamic effects. Separating the problem and applying numerous assumptions will yield approximate solutions. This paper explores methods for generating these solutions accurately

  5. Direct oxidation of strong waste waters, simulating combined wastes in extended-mission space cabins

    Science.gov (United States)

    Ross, L. W.

    1973-01-01

    The applications of modern technology to the resolution of the problem of solid wastes in space cabin environments was studied with emphasis on the exploration of operating conditions that would permit lowering of process temperatures in wet oxidation of combined human wastes. It was found that the ultimate degree of degradation is not enhanced by use of a catalyst. However, the rate of oxidation is increased, and the temperature of oxidation is reduced to 400 F.

  6. G189A analytical simulation of the RITE Integrated Waste Management-Water System

    Science.gov (United States)

    Coggi, J. V.; Clonts, S. E.

    1974-01-01

    This paper discusses the computer simulation of the Integrated Waste Management-Water System Using Radioisotopes for Thermal Energy (RITE) and applications of the simulation. Variations in the system temperature and flows due to particular operating conditions and variations in equipment heating loads imposed on the system were investigated with the computer program. The results were assessed from the standpoint of the computed dynamic characteristics of the system and the potential applications of the simulation to system development and vehicle integration.

  7. Comments on ''Use of conditional simulation in nuclear waste site performance assessment'' by Carol Gotway

    International Nuclear Information System (INIS)

    Downing, D.J.

    1993-01-01

    This paper discusses Carol Gotway's paper, ''The Use of Conditional Simulation in Nuclear Waste Site Performance Assessment.'' The paper centers on the use of conditional simulation and the use of geostatistical methods to simulate an entire field of values for subsequent use in a complex computer model. The issues of sampling designs for geostatistics, semivariogram estimation and anisotropy, turning bands method for random field generation, and estimation of the comulative distribution function are brought out

  8. Specifications of the International Atomic Energy Agency's international project on safety assessment driven radioactive waste management solutions

    International Nuclear Information System (INIS)

    Ghannadi, M.; Asgharizadeh, F.; Assadi, M. R.

    2008-01-01

    Radioactive waste is produced in the generation of nuclear power and the production and use of radioactive materials in the industry, research, and medicine. The nuclear waste management facilities need to perform a safety assessment in order to ensure the safety of a facility. Nuclear safety assessment is a structured and systematic way of examining a proposed facility, process, operation and activity. In nuclear waste management point of view, safety assessment is a process which is used to evaluate the safety of radioactive waste management and disposal facilities. In this regard the International Atomic Energy Agency is planed to implement an international project with cooperation of some member states. The Safety Assessment Driving Radioactive Waste Management Solutions Project is an international programme of work to examine international approaches to safety assessment in aspects of p redisposal r adioactive waste management, including waste conditioning and storage. This study is described the rationale, common aspects, scope, objectives, work plan and anticipated outcomes of the project with refer to International Atomic Energy Agency's documents, such as International Atomic Energy Agency's Safety Standards, as well as the Safety Assessment Driving Radioactive Waste Management Solutions project reports

  9. Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

    International Nuclear Information System (INIS)

    Deshingkar, D.S.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Nair, M.K.T.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Rao, M.K.; Mathur, J.N.; Murali, M.S.; Iyer, R.H.; Badheka, L.P.; Banerji, A.

    1994-01-01

    High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs

  10. Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

    Energy Technology Data Exchange (ETDEWEB)

    Deshingkar, D S; Chitnis, R R; Wattal, P K; Theyyunni, T K; Nair, M K.T. [Bhabha Atomic Research Centre, Bombay (India). Process Engineering and Systems Development Div.; Ramanujam, A; Dhami, P S; Gopalakrishnan, V; Rao, M K [Bhabha Atomic Research Centre, Bombay (India). Fuel Reprocessing Group; Mathur, J N; Murali, M S; Iyer, R H [Bhabha Atomic Research Centre, Bombay (India). Radiochemistry Div.; Badheka, L P; Banerji, A [Bhabha Atomic Research Centre, Bombay (India). Bio-organic Div.

    1994-12-31

    High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs.

  11. Distributions of 14 elements on 60 selected absorbers from two simulant solutions (acid-dissolved sludge and alkaline supernate) for Hanford HLW Tank 102-SY

    International Nuclear Information System (INIS)

    Marsh, S.F.; Svitra, Z.V.; Bowen, S.M.

    1993-10-01

    Sixty commercially available or experimental absorber materials were evaluated for partitioning high-level radioactive waste. These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. The distributions of 14 elements onto each absorber were measured from simulated solutions that represent acid-dissolved sludge and alkaline supernate solutions from Hanford high-level waste (HLW) Tank 102-SY. The selected elements, which represent fission products (Ce, Cs, Sr, Tc, and Y); actinides (U, Pu, and Am); and matrix elements (Cr, Co, Fe, Mn, Zn, and Zr), were traced by radionuclides and assayed by gamma spectrometry. Distribution coefficients for each of the 1680 element/absorber/solution combinations were measured for dynamic contact periods of 30 min, 2 h, and 6 h to provide sorption kinetics information for the specified elements from these complex media. More than 5000 measured distribution coefficients are tabulated

  12. Characterization of cement and bitumen waste forms containing simulated low-level waste incinerator ash

    International Nuclear Information System (INIS)

    Westsik, J.H. Jr.

    1984-08-01

    Incinerator ash from the combustion of general trash and ion exchange resins was immobilized in cement and bitumen. Tests were conducted on the resulting waste forms to provide a data base for the acceptability of actual low-level waste forms. The testing was done in accordance with the US Nuclear Regulatory Commission Technical Position on Waste Form. Bitumen had a measured compressive strength of 130 psi and a leachability index of 13 as measured with the ANS 16.1 leach test procedure. Cement demonstrated a compressive strength of 1400 psi and a leachability index of 7. Both waste forms easily exceed the minimum compressive strength of 50 psi and leachability index of 6 specified in the Technical Position. Irradiation to 10 8 Rad and exposure to 31 thermal cycles ranging from +60 0 ) to -30 0 C did not significantly impact these properties. Neither waste form supported bacterial or fungal growth as measured with ASTM G21 and G22 procedures. However, there is some indication of biodegradation due to co-metabolic processes. Concentration of organic complexants in leachates of the ash, cement and bitumen were too low to significantly affect the release of radionuclides from the waste forms. Neither bitumen nor cement containing incinerator ash caused any corrosion or degradation of potential container materials including steel, polyethylene and fiberglass. However, moist ash did cause corrosion of the steel

  13. Brown & Smith Communication Solutions: A Staffing System Simulation

    Science.gov (United States)

    Small, Erika E.; Doll, Jessica L.; Bergman, Shawn M.; Heggestad, Eric D.

    2018-01-01

    Developing students' practical skills in strategic staffing and selection within the classroom can be challenging. This article describes a staffing system simulation designed to engage students and develop applied skills in strategic recruiting, assessment, and evaluation of job applicants. Instructors looking for a multifaceted team project…

  14. HVAC system simulation: overview, issues and some solutions

    NARCIS (Netherlands)

    Trcka, M.; Hensen, J.L.M.

    2011-01-01

    Integrated performance simulation of buildings’ heating, ventilation and air-conditioning (HVAC) systems can help in reducing energy consumption and increasing occupant comfort. Recognizing this fact, in the last forty years many tools have been developed to help achieving this goal. In this paper

  15. Kinetic Rate Law Parameter Measurements on a Borosilicate Waste Glass: Effect of Temperature, pH, and Solution Composition on Alkali Ion Exchange

    International Nuclear Information System (INIS)

    Pierce, Eric M.; McGrail, B PETER.; Icenhower, J P.; Rodriguez, Elsa A.; Steele, Jackie L.; Baum, Steven R.

    2004-01-01

    The reaction kinetics of glass is controlled by matrix dissolution and ion exchange (IEX). Dissolution of an alkali-rich simulated borosilicate waste glass was investigated using single-pass flow-through (SPFT) experiments. Experiments were conducted as a function of temperature, pH, and solution composition by varying the SiO 2 (aq) activity in the influent solution. Results showed that under dilute conditions matrix dissolution increased with increasing pH and temperature, and decreased with increasing SiO 2 (aq) activity. IEX rates decreased with increasing pH and temperature, and increased with increasing SiO 2 (aq) activity. Over the solution composition range interrogated in this study the dominant dissolution mechanism changed from matrix dissolution to IEX. These results suggest that ''secondary'' reactions may become dominant under certain environmental conditions and emphasize the need to incorporate these reactions into dissolution rate models

  16. Fixing of various simulated radioactive wastes in urea-formaldehyde resin

    International Nuclear Information System (INIS)

    Du Dahai; Wei Peng

    1986-01-01

    This paper outlines the results of the fixing of a variety of simulated radioactive wastes in the urea-formaldehyde resin. The radioactive waste materials fixed include spent ion exchange resin, concentrates of NaNO 3 -NaBO 2 as well as NaBO 2 and sludge. The performance of the fixed products has been improved by means of selecting the synthetic conditions of resin, a suitable hardener and an inorganic additive

  17. In Vitro Studies Evaluating Leaching of Mercury from Mine Waste Calcine Using Simulated Human Body Fluids

    OpenAIRE

    Gray, John E.; Plumlee, Geoffrey S.; Morman, Suzette A.; Higueras, Pablo L.; Crock, James G.; Lowers, Heather A.; Witten, Mark L.

    2010-01-01

    In vitro bioaccessibility (IVBA) studies were carried out on samples of mercury (Hg) mine-waste calcine (roasted Hg ore) by leaching with simulated human body fluids. The objective was to estimate potential human exposure to Hg due to inhalation of airborne calcine particulates and hand-to-mouth ingestion of Hg-bearing calcines. Mine waste calcines collected from Hg mines at Almad?n, Spain, and Terlingua, Texas, contain Hg sulfide, elemental Hg, and soluble Hg compounds, which constitute prim...

  18. Equilibrium leach tests with cobalt in the system cemented waste form/container material/aqueous solution

    International Nuclear Information System (INIS)

    Vejmelka, P.; Koester, R.; Lee, M. J.; Han, K. W.

    1991-01-01

    The equilibrium concentrations of Co in the system of cemented waste form/aqueous solutions were determined including the effect of the container material and its corrosion products under the respective conditions. The chemical conditions in the near field of the waste form were characterized by measurement of the pH and E h value. As disposal relevant solutions, saturated sodium chloride, Q-brine (main constituent MgCl 2 ) and a granitic type groundwater were used. For comparison, also experiments using deionized water were performed. In all systems investigated the cemented waste form itself has a strong influence on the chemical conditions in the near field. The pH and E h values are affected in all cases by the addition of the cemented waste form. There is no or only a slight difference between the E h values if iron powder or iron hydroxide is added to the cemented waste form/solution systems, but the E h is markedly decreased when iron powder is added to the solution free of cement. The Co concentration is decreased in all solutions by the addition of the cemented waste form, the largest effect is observed in Q-brine and this can be attributed either to the sorption of the Co-ions on the corrosion products of the cement or to the coprecipitation of Co-hydroxide and Mg-hydroxide. In the other solutions the Co concentration is decreased by precipitation of Co-hydroxide due to the high pH value of 12.5, and the concentrations are comparable for the different solutions

  19. Open Source AV solution supporting In Situ Simulation

    DEFF Research Database (Denmark)

    Krogh, Kristian; Pociunas, Gintas; Dahl, Mads Ronald

    the software to meet our expectations for a portable AV system for VAD. The system would make use of “off the shelf” hardware components which are widely available and easily replaced or expanded. The developed AV software and coding is contracted to be available as Copyleft Open Source to ensure low cost...... a stable AV software that has be developed and implemented for an in situ simulation initiative. This version (1.3) is the first on released as Open Source (Copyleft) software (see QR tag). We have found that it is possible to deliver multi-camera video assisted debriefing in a mobile, in situ simulation...... environment using an AV system constructed from “off the shelf” components and Open Source software....

  20. Quantitative measurement of cyanide complexes in simulated and actual Hanford ferrocyanide wastes

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pool, K.H.; Sell, R.L.; Bryan, S.L.

    1994-01-01

    Cyanide-containing radioactive waste from radiocesium scavenging processes conducted during the 1950's at Hanford is currently stored in 24 single shell tanks. As part of ongoing tank characterization efforts, the quantity and chemical form of cyanide in these tanks need to be determined. This report summarizes the results of studies conducted at Pacific Northwest Laboratory (PNL) under contract to Westinghouse Hanford Company (WHC) to develop methods for the quantification of total cyanide and identification of major cyanide-containing species in Ferrocyanide Tank Waste. Results from the application of FTIR, IC, and microdistillation procedures to simulated and actual Hanford waste are presented and compared where applicable

  1. The KS-KT-100 plant for two-stage vitrification of radioactive waste: results of tests with simulators

    International Nuclear Information System (INIS)

    Davydov, V.I.; Dobrygin, P.G.; Dolgov, V.V.; Sergeev, G.A.

    1976-01-01

    The Soviet Union has developed a two-stage process for phosphate vitrification of liquid radioactive waste involving the use, at the initial stage, of calcination in the pseudo-liquefied layer, followed by melting of the calcinate in a ceramic crucible (second stage). On the basis of the laboratory studies and bench tests using experimental equipment, the authors have developed and tried out an enlarged plant - the KS-KT-100. The plant includes units for preparing the solution, evaporation, calcination, melting and gas purification. The initial solution containing 240 g/litre of aluminium nitrate, 125 g/litre of sodium nitrate, 120 to 130 g/litre of orthophosphoric acid, and 90 to 150 g/litre of industrial molasses simulated fluxed nitrate waste. The tests have shown that the various units operate satisfactorily. The authors have determined the technological parameters for evaporation, calcination of the solution and melting of the calcinate. The presence of molasses in the solution (150 g/litre) makes it possible to decompose and distil 40% of the nitrate ion during evaporation. The calcination temperature is 350 to 400 0 C, and the fluidization rate 1.5 m/s. The capacity of the plant for the initial solution is 100 litres/h, for the evaporated solution 65 litres/h, and for the glass 20 kg/h. The efficiency of the gas purification system ranges between 10 7 and 10 9 . The test results show the feasibility of the two-stage method of vitrification in actual practice. (author)

  2. Metals and polybrominated diphenyl ethers leaching from electronic waste in simulated landfills

    Energy Technology Data Exchange (ETDEWEB)

    Kiddee, Peeranart [Centre for Environmental Risk Assessment and Remediation, University of South Australia, Mawson Lakes Campus, Adelaide, 5095 (Australia); Cooperative Research Centre for Contamination Assessment and Remediation of the Environment, Mawson Lakes Campus, Adelaide, 5095 (Australia); Naidu, Ravi, E-mail: ravi.naidu@crccare.com [Centre for Environmental Risk Assessment and Remediation, University of South Australia, Mawson Lakes Campus, Adelaide, 5095 (Australia); Cooperative Research Centre for Contamination Assessment and Remediation of the Environment, Mawson Lakes Campus, Adelaide, 5095 (Australia); Wong, Ming H. [Croucher Institute for Environmental Sciences, and Department of Biology, Hong Kong Baptist University, Kowloon Tong (China)

    2013-05-15

    Highlights: • Simulated landfill columns provided realistic results than lab based column study. • Column leachates showed significant seasonal effect on toxic substances. • Toxic substances in the landfill leachates pose environmental and health hazards. • A better management of e-waste is urgently needed. -- Abstract: Landfills established prior to the recognition of potential impacts from the leaching of heavy metals and toxic organic compounds often lack appropriate barriers and pose significant risks of contamination of groundwater. In this study, bioavailable metal(oids) and polybrominated diphenyl ethers (PBDEs) in leachates from landfill columns that contained intact or broken e-waste were studied under conditions that simulate landfills in terms of waste components and methods of disposal of e-wastes, and with realistic rainfall. Fourteen elements and PBDEs were analysed in leachates over a period of 21 months. The results demonstrate that the average concentrations of Al, Ba, Be, Cd, Co, Cr, Cu, Ni, Pb, Sb and V in leachates from the column that contained broken e-waste items were significantly higher than the column without e-waste. BDE-153 was the highest average PBDEs congener in all columns but the average of ∑PBDEs levels in columns that contained intact e-waste were (3.7 ng/l) and were not significantly higher than that in the leachates from the control column.

  3. Modeling by GASP-IV simulation of high-level nuclear waste disposal

    International Nuclear Information System (INIS)

    Kurstedt, H.A. Jr.; DePorter, E.L.; Turek, J.L.; Funk, S.K.; Rasbach, C.E.

    1981-01-01

    High-level nuclear waste generated by defense-oriented and commercial nuclear energy activities are to be stored ultimately in underground repositories. Research continues on the waste-form and waste-form processing. DOE managers must coordinate the results of this research, the capacities and availability times of the permanent geologic storage repositories, and the capacities and availability times of interim storage facilities (pending availability of permanent repositories). Comprehensive and active DOE program-management information systems contain predicted generation of nuclear wastes from defense and commercial activities; milestones on research on waste-forms; and milestones on research and development, design, acquisition, and construction of facilities and repositories. A GASP IV simulation model is presented which interfaces all of these data. The model accepts alternate management decisions; relates all critical milestones, all research and development data, and the generation of waste nuclear materials; simulates the passage of time; then, predicts the impact of those alternate decisions on the availability of storage capacity for waste nuclear materials. 3 references, 3 figures

  4. Adsorption of lanthanides in aqueous solution aiming to study of nuclear wastes

    International Nuclear Information System (INIS)

    Belline, Jean de Brito

    2009-01-01

    The problem of radioactive wastes is a concern of world-wide scope, a time that does not still have a defined local for the construction of a repository for radioactive wastes of high level. One of the preliminary stages for the choice of the place more appropriate is the geologic study associated to the experimental studies of adsorption of the involved chemical species in the process. In this work, a sample of basaltic rock was used, of the South Region of the Formation Serra Geral, collected in Frederico Westphalen Town (RS), that it will be probably a candidate to the rock hostess for location of radioactive wastes. Two experiments have been carried out through, namely: 'Test Batch' and Percolating, both under atmospheric pressure, at the ambient temperature of 25 deg C, with the purpose to study the capacity of sorption of the rare earth elements - REE. The REE are used in this work in function of its analogy with the actinides, aiming at to investigate the chemistry behavior and the speciation of the same in natural waters, searching the possibility of geologic storage of radioactive wastes, a time that the adsorption of the REE depends on variables of the environment as pH, ionic strength, temperature and presence of ligands, as carbonates and constituent of surfaces of minerals. Experiment of percolating of the REE was carried through, 100ppb, in the basalt (with 80 mesh) in solutions with ionic strength 1= 0,025 M and 1=0,5 M of NaCl. pH was controlled in a range of 5,6 the 7,6 with HNO 3 addition. The concentrations were analyzed by ICP-MS. The 'Batch Test' is an efficient form of studying sorption/desorption isotherms, beyond values of the reason between the distributions solid/solution and estimation of the solubility. The percolating experiment, was carried through under pH controlled around 6, and allowed to verify the behaviour of heavy REE in comparison with the light REE. (author)

  5. Adsorption behavior and mechanism of Cr(VI) using Sakura waste from aqueous solution

    International Nuclear Information System (INIS)

    Qi, Wenfang; Zhao, Yingxin; Zheng, Xinyi; Ji, Min; Zhang, Zhenya

    2016-01-01

    Graphical abstract: The main chemical components of Sakura leaves are cellulose 16.6%, hemicellulose 10.4%, lignin 18.3%, ash 11.4%, and others 43.3%. The adsorption capacity of Cr(VI) onto Sakura leaves can achieve 435.25 mg g"−"1, much higher than other similar agroforestry wastes. - Highlights: • Sakura leaves were prepared to remove Cr(VI) from aqueous solution. • The maximum adsorption capacity of Cr(VI) reached 435.25 mg g"−"1. • Cr(VI) adsorption fitted pseudo-second-order kinetic model. • Isotherm models indicated Cr(VI) adsorption occurred on a monolayer surface. • The influence order of coexisting ions followed PO_4"3"− > SO_4"2"− > Cl"−. - Abstract: A forestall waste, Sakura leave, has been studied for the adsorption of Cr(VI) from aqueous solution. The materials before and after adsorption were characterized by X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectroscopy (FTIR). To investigate the adsorption performance of Sakura waste, batch experiments were conducted under different adsorbent dosage, contact time, initial concentration of Cr(VI), and co-existing ions. Results showed the data fitted pseudo-second-order better than pseudo-first-order kinetic model. Equilibrium data was analyzed with Langmuir, Freundlich and Redlich–Peterson isotherm models at temperature ranges from 25 °C to 45 °C. The maximum adsorption capacity from the Langmuir model was 435.25 mg g"−"1 at pH 1.0. The presence of Cl"−, SO_4"2"− and PO_4"3"− would lead to an obvious negative effect on Cr(VI) adsorption, and their influence order follows PO_4"3"− > SO_4"2"− > Cl"−. The study developed a new way to reutilize wastes and showed a great potential for resource recycling.

  6. Adsorption behavior and mechanism of Cr(VI) using Sakura waste from aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Qi, Wenfang [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Zhao, Yingxin, E-mail: yingxinzhao@tju.edu.cn [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Tianjin Engineering Center of Urban River Eco-Purification Technology, Tianjin 300072 (China); Zheng, Xinyi [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Ji, Min [School of Environmental Science and Engineering, Tianjin University, Tianjin 300072 (China); Tianjin Engineering Center of Urban River Eco-Purification Technology, Tianjin 300072 (China); Zhang, Zhenya [Graduate School of Life and Environmental Sciences, University of Tsukuba, Tsukuba 3058572 (Japan)

    2016-01-01

    Graphical abstract: The main chemical components of Sakura leaves are cellulose 16.6%, hemicellulose 10.4%, lignin 18.3%, ash 11.4%, and others 43.3%. The adsorption capacity of Cr(VI) onto Sakura leaves can achieve 435.25 mg g{sup −1}, much higher than other similar agroforestry wastes. - Highlights: • Sakura leaves were prepared to remove Cr(VI) from aqueous solution. • The maximum adsorption capacity of Cr(VI) reached 435.25 mg g{sup −1}. • Cr(VI) adsorption fitted pseudo-second-order kinetic model. • Isotherm models indicated Cr(VI) adsorption occurred on a monolayer surface. • The influence order of coexisting ions followed PO{sub 4}{sup 3−} > SO{sub 4}{sup 2−} > Cl{sup −}. - Abstract: A forestall waste, Sakura leave, has been studied for the adsorption of Cr(VI) from aqueous solution. The materials before and after adsorption were characterized by X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectroscopy (FTIR). To investigate the adsorption performance of Sakura waste, batch experiments were conducted under different adsorbent dosage, contact time, initial concentration of Cr(VI), and co-existing ions. Results showed the data fitted pseudo-second-order better than pseudo-first-order kinetic model. Equilibrium data was analyzed with Langmuir, Freundlich and Redlich–Peterson isotherm models at temperature ranges from 25 °C to 45 °C. The maximum adsorption capacity from the Langmuir model was 435.25 mg g{sup −1} at pH 1.0. The presence of Cl{sup −}, SO{sub 4}{sup 2−} and PO{sub 4}{sup 3−} would lead to an obvious negative effect on Cr(VI) adsorption, and their influence order follows PO{sub 4}{sup 3−} > SO{sub 4}{sup 2−} > Cl{sup −}. The study developed a new way to reutilize wastes and showed a great potential for resource recycling.

  7. Radiation effects in vitreous and devitrified simulated waste glass

    International Nuclear Information System (INIS)

    Weber, W.J.; Turcotte, R.P.; Bunnell, L.R.; Roberts, F.P.; Westsik, J.H. Jr.

    1979-01-01

    The long-term radiation stability of vitreous and partially devitrified forms of high-level waste glass was investigated in accelerated experiments by 266 Cm doping. The effects of radiation on microstructure, phase behavior, density, impact strength, stored energy, and leachability are reported to a cumulative radiation dose of 5 x 10 18 α decays/cm 3 . This dose produces saturation of radiation effects in most properties. 4 figures

  8. Evaluation of Gas Retention in Waste Simulants: Tall Column Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Schonewill, Philip P.; Gauglitz, Phillip A.; Shimskey, Rick W.; Denslow, Kayte M.; Powell, Michael R.; Boeringa, Gregory K.; Bontha, Jagannadha R.; Karri, Naveen K.; Fifield, Leonard S.; Tran, Diana N.; Sande, Susan; Heldebrant, David J.; Meacham, Joseph E.; Smet, Dave; Bryan, Wesley E.; Calmus, Ronald B.

    2014-05-16

    Gas generation in Hanford’s underground waste storage tanks can lead to gas accumulation within the layer of settled solids (sludge) at the tank bottom. The gas, which typically has hydrogen as the major component together with other flammable species, is formed principally by radiation-driven chemical reactions. Accumulation of these gases within the sludge in a waste tank is undesirable and limits the amount of tank volume for waste storage. Further, accumulation of large amounts of gas in the sludge may potentially result in an unacceptable release of the accumulated gas if the sludge-layer density is reduced to less than that of the overlying sludge or that of the supernatant liquid. Rapid release of large amounts of flammable gases could endanger personnel and equipment near the tank. For this reason, a thorough understanding of the circumstances that can lead to a potentially problematic gas accumulation in sludge layers is needed. To respond to this need, the Deep Sludge Gas Release Event Program (DSGREP) was commissioned to examine gas release behavior in sludges.

  9. Mathematical simulation for safety assessment of nuclear waste repositories

    International Nuclear Information System (INIS)

    Brandstetter, A.; Raymond, J.R.; Benson, G.L.

    1979-01-01

    Mathematical models are being developed as part of the Waste Isolation Safety Assessment Program (WISAP) for assessing the post-closure safety of nuclear waste storage in geologic formations. The objective of this program is to develop the methods and data necessary to determine potential events that might disrupt the integrity of a waste repository and provide pathways for radionuclides to reach the bioshpere, primarily through groundwater transport. Four categories of mathematical models are being developed to assist in the analysis of potential release scenarios and consequences: (1) release scenario analysis models; (2) groundwater flow models; (3) contaminant transport models; and (4) radiation dose models. The development of the release scenario models is in a preliminary stage; the last three categories of models are fully operational. The release scenario models determine the bounds of potential future hydrogeologic changes, including potentially disruptive events. The groundwater flow and contaminant transport models compute the flowpaths, travel times, and concentrations of radionuclides that might migrate from a repository in the event of a breach and potentially reach the biosphere. The dose models compute the radiation doses to future populations. Reference site analyses are in progress to test the models for application to different geologies, including salt domes, bedded salt, and basalt

  10. 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

  11. 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

  12. Emerging concepts and requirements for the long-term management of non-radioactive hazardous wastes - would geological disposal be an appropriate solution for some of these wastes

    International Nuclear Information System (INIS)

    Rein, K. von

    1994-01-01

    This work deals with the emerging concepts and requirements for the long-term management of non-radioactive hazardous wastes. After some generalities on the pollution of natural environment and the legislations taken by the swedish government the author tries to answer to the question : would geological disposal be an appropriate solution for the non-radioactive hazardous wastes? Then is given the general discussion of the last three articles concerning the background to current environmental policies and their implementation and more particularly the evolution and current thoughts about environmental policies, the managing hazardous activities and substances and the emerging concepts and requirements for the long-term management of non-radioactive hazardous wastes. Comments and questions concerning the similarity or otherwise between the present position of radioactive waste disposal and the background to current environmental policies are indicated. (O.L.)

  13. Response of ethylene propylene diene monomer rubber (EPDM) to simulant Hanford tank waste

    Energy Technology Data Exchange (ETDEWEB)

    NIGREY,PAUL J.

    2000-02-01

    This report presents the findings of the Chemical Compatibility Program developed to evaluate plastic packaging components that may be incorporated in packaging mixed-waste forms for transportation. Consistent with the methodology outlined in this report, the author performed the second phase of this experimental program to determine the effects of simulant Hanford tank mixed wastes on packaging seal materials. That effort involved the comprehensive testing of five plastic liner materials in an aqueous mixed-waste simulant. The testing protocol involved exposing the materials to {approximately}143, 286, 571, and 3,670 krad of gamma radiation and was followed by 7-, 14-, 28-, 180-day exposures to the waste simulant at 18, 50, and 60 C. Ethylene propylene diene monomer (EPDM) rubber samples subjected to the same protocol were then evaluated by measuring seven material properties: specific gravity, dimensional changes, mass changes, hardness, compression set, vapor transport rates, and tensile properties. The author has determined that EPDM rubber has excellent resistance to radiation, this simulant, and a combination of these factors. These results suggest that EPDM is an excellent seal material to withstand aqueous mixed wastes having similar composition to the one used in this study.

  14. Regulatory off-gas analysis from the evaporation of Hanford simulated waste spiked with organic compounds.

    Science.gov (United States)

    Saito, Hiroshi H; Calloway, T Bond; Ferrara, Daro M; Choi, Alexander S; White, Thomas L; Gibson, Luther V; Burdette, Mark A

    2004-10-01

    After strontium/transuranics removal by precipitation followed by cesium/technetium removal by ion exchange, the remaining low-activity waste in the Hanford River Protection Project Waste Treatment Plant is to be concentrated by evaporation before being mixed with glass formers and vitrified. To provide a technical basis to permit the waste treatment facility, a relatively organic-rich Hanford Tank 241-AN-107 waste simulant was spiked with 14 target volatile, semi-volatile, and pesticide compounds and evaporated under vacuum in a bench-scale natural circulation evaporator fitted with an industrial stack off-gas sampler at the Savannah River National Laboratory. An evaporator material balance for the target organics was calculated by combining liquid stream mass and analytical data with off-gas emissions estimates obtained using U.S. Environmental Protection Agency (EPA) SW-846 Methods. Volatile and light semi-volatile organic compounds (1 mm Hg vapor pressure) in the waste simulant were found to largely exit through the condenser vent, while heavier semi-volatiles and pesticides generally remain in the evaporator concentrate. An OLI Environmental Simulation Program (licensed by OLI Systems, Inc.) evaporator model successfully predicted operating conditions and the experimental distribution of the fed target organics exiting in the concentrate, condensate, and off-gas streams, with the exception of a few semi-volatile and pesticide compounds. Comparison with Henry's Law predictions suggests the OLI Environmental Simulation Program model is constrained by available literature data.

  15. Research and demonstration results for a new "Double-Solution" technology for municipal solid waste treatment.

    Science.gov (United States)

    Erping, Li; Haoyun, Chen; Yanyang, Shang; Jun, Pan; Qing, Hu

    2017-11-01

    In this paper, the pyrolysis characteristics of six typical components in municipal solid waste (MSW) were investigated through a TG-FTIR combined technique and it was concluded that the main pyrolysis process of the biomass components (including food residues, sawdust and paper) occurred at 150-600°C. The main volatiles were multi-component gas including H 2 O, CO 2 , and CO. The main pyrolysis temperatures of three artificial products (PP, PVC and leather) was ranged from 200to 500°C. The wavelength of small molecule gases (CH 4 , CO 2 and CO) and the the chemical bonds (CO and CC) were observed in the infrared spectrum Based on the pyrolysis tem