WorldWideScience

Sample records for waste atw aqueous-based

  1. Neutronics-processing interface analyses for the Accelerator Transmutation of Waste (ATW) aqueous-based blanket system

    International Nuclear Information System (INIS)

    Davidson, J.W.; Battat, M.E.

    1993-01-01

    Neutronics-processing interface parameters have large impacts on the neutron economy and transmutation performance of an aqueous-based Accelerator Transmutation of Waste (ATW) system. A detailed assessment of the interdependence of these blanket neutronic and chemical processing parameters has been performed. Neutronic performance analyses require that neutron transport calculations for the ATW blanket systems be fully coupled with the blanket processing and include all neutron absorptions in candidate waste nuclides as well as in fission and transmutation products. The effects of processing rates, flux levels, flux spectra, and external-to-blanket inventories on blanket neutronic performance were determined. In addition, the inventories and isotopics in the various subsystems were also calculated for various actinide and long-lived fission product transmutation strategies

  2. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group

    International Nuclear Information System (INIS)

    Collins, E.; Duguid, J.; Henry, R.; Karell, E.J.; Laidler, J.J.; McDeavitt, S.M.; Thompson, M.; Toth, L.M.; Williamson, M.; Willit, J.L.

    1999-01-01

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD and D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years

  3. Preparation of a technology development roadmap for the Accelerator Transmutation of Waste (ATW) System : report of the ATW separations technologies and waste forms technical working group.

    Energy Technology Data Exchange (ETDEWEB)

    Collins, E.; Duguid, J.; Henry, R.; Karell, E.; Laidler, J.; McDeavitt, S.; Thompson, M.; Toth, M.; Williamson, M.; Willit, J.

    1999-08-12

    In response to a Congressional mandate to prepare a roadmap for the development of Accelerator Transmutation of Waste (ATW) technology, a Technical Working Group comprised of members from various DOE laboratories was convened in March 1999 for the purpose of preparing that part of the technology development roadmap dealing with the separation of certain radionuclides for transmutation and the disposal of residual radioactive wastes from these partitioning operations. The Technical Working Group for ATW Separations Technologies and Waste Forms completed its work in June 1999, having carefully considered the technology options available. A baseline process flowsheet and backup process were identified for initial emphasis in a future research, development and demonstration program. The baseline process combines aqueous and pyrochemical processes to permit the efficient separation of the uranium, technetium, iodine and transuranic elements from the light water reactor (LWR) fuel in the head-end step. The backup process is an all- pyrochemical system. In conjunction with the aqueous process, the baseline flowsheet includes a pyrochemical process to prepare the transuranic material for fabrication of the ATW fuel assemblies. For the internal ATW fuel cycle the baseline process specifies another pyrochemical process to extract the transuranic elements, Tc and 1 from the ATW fuel. Fission products not separated for transmutation and trace amounts of actinide elements would be directed to two high-level waste forms, one a zirconium-based alloy and the other a glass/sodalite composite. Baseline cost and schedule estimates are provided for a RD&D program that would provide a full-scale demonstration of the complete separations and waste production flowsheet within 20 years.

  4. ATW economics

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1993-01-01

    A parametric systems model of the ATW [Accelerator Transmutation of (Nuclear) Waste] has been used to examine key system tradeoffs and design drivers on the basis of unit costs. This model has been applied primarily to the aqueous-slurry blanket concept for an ATW that generates net-electric power from the fissioning of spent reactor fuel. An important goal of this study is the development of essential parametric tradeoff studies to aid in any eventual engineering design of an ATW that would burn and generate net- electric power from spent reactor fuel

  5. Performance estimates for waste treatment pyroprocesses in ATW

    International Nuclear Information System (INIS)

    Li, N.

    1997-01-01

    The author has identified several pyrometallurgical processes for the conceptual ATW waste treatment cycle. These processes include reductive extraction, electrowinning and electrorefining, which constitute some versatile treatment cycles for liquid-metal based and molten-salt based waste forms when they are properly integrated. This paper examines the implementation of these processes and the achievable separations for some typical species. The author also presents a simple analysis of the processing rates limited by mass diffusion through a thin hydrodynamic boundary layer. It is shown that these processes can be realized with compact and efficient devices to meet the ATW demand for the periodic feeding and cleaning of the waste

  6. Accelerator transmutation of wastes (ATW) - Prospects and safety

    International Nuclear Information System (INIS)

    Gudowski, W.; Pettersson, Kjell; Thedeen, T.

    1993-11-01

    Accelerator transmutation of nuclear waste (ATW) has during last years gained interest as a technologically possible method to transform radioactive wastes into short-lived or stable isotopes. Different ATW-projects are described from the physical and technical point of view. The principal sketch of the safety analysis of the ATW-idea is given. Due to the very limited technical data for existing ATW-projects the safety analysis can cause some risks for the health and environmental safety for the closest environment. General public should not be affected. 35 refs, 22 figs, 4 tabs

  7. ATW system impact on high-level waste

    International Nuclear Information System (INIS)

    Arthur, E.D.

    1992-01-01

    This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products

  8. The Los Alamos accelerator-driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

    International Nuclear Information System (INIS)

    Venneri, F.; Williamson, M.A.; Ning, L.

    1997-01-01

    In the past several years, the Los Alamos ADTT program has conducted studies of an innovative technology for solving the nuclear waste problem and building a new generation of safer and non-proliferant nuclear power plants. The ATW concept destroys higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. In this paper special attention is given to the basic design of the ATW Molten Salt concept and the safety perspective. 40 refs., 11 figs

  9. Induced structural radioactivity inventory analysis of the base case aqueous ATW reactor concept

    International Nuclear Information System (INIS)

    Bezdecny, J.A.; Henderson, D.L.; Sailor, W.C.

    1993-01-01

    The purpose of the Los Alamos National Laboratory Accelerator Transmutation of Nuclear Waste (ATW) project is the substantial reduction in volume of this country's long-lived high-level radioactive waste in a safe and energy efficient manner. An evaluation of the Accelerator Transmutation of Nuclear Waste concept has four aspects; material balance, energy balance, performance and cost. An evaluation of the material balance compares the amount of long-lived high-level waste transmuted with the amount and type of waste created in the process. One component of the material balance is the activation of structural materials over the lifetime of the transmutation reactor. An activation analysis has been performed on four structure regions of the reaction vessel: the tungsten target; the lead target and annulus; the Zircalloy and aluminum tubing carrying the actinide slurry and; the stainless steel tank

  10. A Los Alamos concept for accelerator transmutation of waste and energy production (ATW)

    International Nuclear Information System (INIS)

    1990-01-01

    This document contains the diagrams presented at the ATW (Accelerator Transmutation of Waste and Energy Production) External Review, December 10-12, 1990, held at Los Alamos National Laboratory. Included are the charge to the committee and the presentations for the committee's review. Topics of the presentations included an overview of the concept, LINAC technology, near-term application -- high-level defense wastes (intense thermal neutron source, chemistry and materials), advanced application of the ATW concept -- fission energy without a high-level waste stream (overview, advanced technology, and advanced chemistry), and a summary of the research issues

  11. Target/blanket conceptual design for the Los Alamos ATW concept

    International Nuclear Information System (INIS)

    Ames, K.; Cappiello, M.; Ireland, J.; Sapir, J.; Farnum, G.

    1992-01-01

    The Los Alamos Accelerator Transmutation of Waste (ATW) concept has many potential applications that include defense waste transmutation, defense material production (i.e., tritium and 238 Pu), and the transmutation of hazardous nuclear wastes from commercial nuclear reactors (fission products and actinides). A more advanced long-term Los Alamos effort is investigating the potential of an accelerator- driven system to produce fission energy with a minimal nuclear waste stream. All applications employ a high-energy (800- to 1600-MeV), high-current (25--250 mA) proton linear accelerator as the driver. In this report, we discuss only the target/blanket conceptual design for the commercial nuclear waste application. A conceptual design for the target/blanket of the Los Alamos ATW concept has been presented. The neutronics, mechanical design, and heat transfer have been investigated in some detail for the base-case design. Much more work needs to be done, but at this point it appears that the design is feasible and will approach the design goal of supporting two commercial power reactors with each target/blanket module

  12. Preliminary analysis of the induced structural radioactivity inventory of the base-case aqueous accelerator transmutation of waste reactor concept

    International Nuclear Information System (INIS)

    Bezdecny, J.A.; Vance, K.M.; Henderson, D.L.

    1995-01-01

    The purpose of the Los Alamos National Laboratory Accelerator Transmutation of (Nuclear) Waste (ATW) project is the substantial reduction in volume of long-lived high-level radioactive waste of the US in a safe and energy-efficient manner. An evaluation of the ATW concept has four aspects: material balance, energy balance, performance, and cost. An evaluation of the material balance compares the amount of long-lived high-level waste transmuted with the amount and type, of waste created in the process. One component of the material balance is the activation of structural materials over the lifetime of the transmutation reactor. A preliminary radioactivity and radioactive mass balance analysis has been performed on four structure regions of the reaction chamber: the tungsten target, the lead annulus, six tubing materials carrying the actinide slurry, and five reaction vessel structural materials. The amount of radioactive material remaining after a 100-yr cooling period for the base-case ATW was found to be 338 kg of radionuclides. The bulk of this material (313 kg) was generated in the zirconium-niobium (Zr-Nb) actinide tubing material. Replacement of the Zr-Nb tubing material with one of the alternative tubing materials analyzed would significantly reduce the short- and long-term radioactive mass produced. The alternative vessel material Al-6061 alloys, Tenelon, HT-9, and 2 1/4 Cr-1 Mo and the alternative actinide tubing materials Al-6061 alloy, carbon-carbon matrix, silicon carbide, and Ti-6 Al-4 V qualify for shallow land burial. Alternative disposal options for the base-case structural material Type 304L stainless steel and the actinide tubing material Zr-Nb will need to be considered as neither qualifies for shallow land burial

  13. Separations technologies supporting the development of a deployable ATW system

    International Nuclear Information System (INIS)

    Laidler, J. J.

    2000-01-01

    A program has been initiated for the purpose of developing the chemical separations technologies necessary to support a large Accelerator Transmutation of Waste (ATW) system capable of dealing with the projected inventory of spent fuel from the commercial nuclear power stations in the United States. The first several years of the program will be directed toward an elucidation of related technical issues and to the establishment, by means of comprehensive trade studies, of an optimum configuration of the elements of the chemical processing infrastructure required for support of the total ATW system. By adopting this sort of disciplined systems engineering approach, it is expected that development and demonstration costs can be minimized and that it will be possible to deploy an ATW system that is an environmentally sound and economically viable venture

  14. Proposal on the accelerator driven molten-salt reactor (ATW concept) benchmark calculations. (STAGE 1 - without an external neutron source)

    International Nuclear Information System (INIS)

    Svarny, J.; Mikolas, P.

    1999-01-01

    The first stage of ATW neutronic benchmark (without an external source), based on the simple modelling of two component concept is presented. The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark is not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (author)

  15. TRACE Assessment for BWR ATWS Analysis

    International Nuclear Information System (INIS)

    Cheng, L.Y.; Diamond, D.; Cuadra, Arantxa; Raitses, Gilad; Aronson, Arnold

    2010-01-01

    A TRACE/PARCS input model has been developed in order to be able to analyze anticipated transients without scram (ATWS) in a boiling water reactor. The model is based on one developed previously for the Browns Ferry reactor for doing loss-of-coolant accident analysis. This model was updated by adding the control systems needed for ATWS and a core model using PARCS. The control systems were based on models previously developed for the TRAC-B code. The PARCS model is based on information (e.g., exposure and moderator density (void) history distributions) obtained from General Electric Hitachi and cross sections for GE14 fuel obtained from an independent source. The model is able to calculate an ATWS, initiated by the closure of main steam isolation valves, with recirculation pump trip, water level control, injection of borated water from the standby liquid control system and actuation of the automatic depressurization system. The model is not considered complete and recommendations are made on how it should be improved.

  16. Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century

    International Nuclear Information System (INIS)

    Beller, D.E.; Sailor, W.C.; Venneri, F.

    1998-01-01

    A future nuclear energy scenario with a closed, thorium-uranium-oxide (ThUOX) fuel cycle and new light water reactors (TULWRs) supported by Accelerator Transmutation of Waste (ATW) systems could provide several improvements beyond today's once-through, UO 2 -fueled nuclear technology. A deployment scenario with TULWRs plus ATWs to burn the actinides produced by these LWRs and to close the back-end of the ThUOX fuel cycle was modeled to satisfy a US demand that increases linearly from 80 GWe in 2020 to 200 GWe by 2100. During the first 20 years of the scenario (2000-2020), nuclear energy production in the US declines from today's 100 GWe to about 80 GWe, in accordance with forecasts of the US DOE's Energy Information Administration. No new nuclear systems are added during this declining nuclear energy period, and all existing LWRs are shut down by 2045. Beginning in 2020, ATWs that transmute the actinides from existing LWRs are deployed, along with TULWRs and additional ATWs with a support ratio of 1 ATW to 7 TULWRs to meet the energy demand scenario. A final mix of 174 GWe from TULWRs and 26 GWe from ATWs provides the 200 GWe demand in 2100. Compared to a once-through LWR scenario that meets the same energy demand, the TULWR/ATW concept could result in the following improvements: depletion of natural uranium resources would be reduced by 50%; inventories of Pu which may result in weapons proliferation will be reduced in quantity by more than 98% and in quality because of higher neutron emissions and 50 times the alpha-decay heating of weapons-grade plutonium; actinides (and possibly fission products) for final disposal in nuclear waste would be substantially reduced; and the cost of fuel and the fuel cycle may be 20-30% less than the once-through UO 2 fuel cycle

  17. ILK statement about ATWS requirements

    International Nuclear Information System (INIS)

    2005-01-01

    A controversial debate is going on in Germany about the management of operating transients in case of the failure, additionally assumed, of the scram system (ATWS=Anticipated Transients without Scram). It was triggered by a recommendation by the German Advisory Committee on Reactor Safeguards (RSK) in a statement of May 3, 2001 according to which the demonstration that ATWS events were under control was to deviate from requirements in the RSK Guidelines for pressurized water reactors of 1981 (last amended in 1996) and not to take credit of the effects of one-off measures initiated actively, especially shutdown of the main coolant pumps. ILK therefore expresses its opinion in this Statement about the criteria to be met in demonstrations that ATWS is under control in pressurized water reactors. Also in boiling water reactors, studies of ATWS transients are part of the licensing procedure. However, the assumptions to be made there in demonstrating effective pressure limitation have been unchanged and uncontested long since. ILK included in its considerations especially also practices in the United States, France and Finland. In doing so, the Committee found the basic approach in dealing with ATWS to be the same in Germany, the United States and in France, namely to show that the consequences remain tolerable without the application of aggravating postulates. ILK feels that the approach so far employed in demonstrating safety in ATWS events results in balanced risk mitigation. The initiating event already has a very low probability of occurrence. Reliable measures are in place to manage it. (orig.)

  18. Effects of buffer thickness on ATW blanket performances

    International Nuclear Information System (INIS)

    Yang, Won Sik

    2001-01-01

    This paper presents the preliminary results of target and buffer design studies for a lead-bismuth eutectic (LBE) cooled accelerator transmutation of waste (ATW) system, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using an 840 MWt LBE cooled ATW design, the effects of buffer thickness on the blanket performances have been studied. Varying the buffer thickness for a given blanket configuration, system performances have been estimated by a series of calculations using MCNPX and REBUS-3 codes. The effects of source importance change are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. As the irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. The results show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable

  19. ATW neutron spectrum measurements at LAMPF

    Energy Technology Data Exchange (ETDEWEB)

    Butler, G.W.; Littleton, P.E.; Morgan, G.L. [Los Alamos National Laboratory, NM (United States)] [and others

    1995-10-01

    Accelerator transmutation of waste (ATW) is a proposal to use a high flux of accelerator-produced thermalized neutrons to transmute both fission product and higher actinide commercial nuclear waste into stable or short-lived radioactive species in order to avoid long-term storage of nuclear waste. At LAMPF the authors recently performed experiments that were designed to measure the spectrum of neutrons produced per incident proton for full-scale proposed ATW targets of lead and lithium. The neutrons produced in such targets have a spectrum of energies that extends up to the energy of the incident proton beam, but the distribution peaks between 1 and 5 MeV. Transmutation reactions and fission of actinides are most efficient when the neutron energy is below a few eV, so the target must be surrounded by a non-absorbing material (blanket) to produce additional neutrons and reduce the energy of high energy neutrons without loss. The experiments with the lead target, 25 cm diameter by 40 cm long, were conducted with 800 MeV protons, while those with the lithium target, 25 cm diameter by 175 cm long, were conducted with 400 MeV protons. The blanket in both sets of experiments was a 60 cm diameter by 200 cm long annulus of lead that surrounded the target. Surrounding the blanket was a steel water tank with dimensions of 250 cm diameter by 300 cm long that simulated the transmutation region. A small sample pipe penetrated the length of the lead blanket and other sample pipes penetrated the length of the water tank at different radii from the beam axis so that the neutron spectra at different locations could be measured by foil activation. After irradiation the activated foil sets were extracted and counted with calibrated high resolution germanium gamma ray detectors at the Los Alamos nuclear chemistry counting facility.

  20. Preliminary ATWS analysis for the IRIS PRA

    International Nuclear Information System (INIS)

    Maddalena Barra; Marco S Ghisu; David J Finnicum; Luca Oriani

    2005-01-01

    to be a non dominant contribution to the plant CDF. During this second phase of the IRIS PRA development, the original model has been updated with the insertion of a study of ATWS, with the purpose of confirming the assumptions previously made. Within this framework, the most significant ATWS sequences have been identified and a first set of analyses has been run. The thermal-hydraulic computer code RELAP5 is used for the accident analyses, using a conservative evaluation model based on the Westinghouse ATWS evaluation model used during the analyses that supported the development of the ATWS rule (10CFR50.62). The key figure of merit used in these ATWS analyses is the pressure peak resulting from the imbalance between the rates of energy deposition into the core and removal from the reactor coolant that could cause damage to RCS components necessary for safe plant shutdown. Core damage is assumed if the peak pressure exceeds a limiting value. To support the evaluation of the ATWS contribution to the CDF, analyses with different values of critical parameters (mainly the moderator temperature coefficient and the number of operative safety valves in the pressurizer) were performed. The use of these analyses on the ATWS event tree development and on the calculation of the ATWS to the plant CDF is discussed in this paper. (authors)

  1. Strategy and Economic Prospect of Back-end Cycle through ATW

    International Nuclear Information System (INIS)

    Hendri Firman Windarto; Siti Alimah

    2003-01-01

    Strategy and economic prospect of back-end cycle through ATW has been studied. Nuclear fuel cycle through ATW is a single stratum of back-end cycle. By ATW, volume of spent fuel which should be disposed in long term can be reduced from 70,000 MHTM to 3,000 MHTM and half-life of spent fuel can be reduced from 15,700,000 years to 300 years. Strategic values of the ATW cycle are to prevent proliferation risk and to reduce the uncertainty of long term dispose. Economic prospect of the ATW cycle will give some advantages on reducing of spent fuel volume and its disposal period, and producing electricity. (author)

  2. Pyrochemical separations technologies envisioned for the U.S. accelerator transmutation of waste system

    International Nuclear Information System (INIS)

    Laidler, J. J.

    2000-01-01

    A program has been initiated for the purpose of developing the chemical separations technologies necessary to support a large Accelerator Transmutation of Waste (ATW) system capable of dealing with the projected inventory of spent fuel from the commercial nuclear power stations in the United States. The baseline process selected combines aqueous and pyrochemical processes to enable the efficient separation of uranium, technetium, iodine, and the transuranic elements from LWR spent fuel. The diversity of processing methods was chosen for both technical and economic factors. A six-year technology evaluation and development program is foreseen, by the end of which an informed decision can be made on proceeding with demonstration of the ATW system

  3. New insights regarding ATWS for BWRS

    International Nuclear Information System (INIS)

    Drouin, M.T.; Kolaczkowski, A.M.; LaChance, J.L.; Ferrell, W.L.

    1987-01-01

    Anticipated transients without scram (ATWS) accident sequences have been found in past studies to have a relatively high core damage frequency (ranging from 5.4E-6 to 3E-4 per year) that represents a significant contribution to the total core damage frequency (ranging from 7-to-33%). Results of analyses for the two boiling water reactors (BWRs) analyzed as part of NUREG/CR-4550 indicate both a lower core damage frequency (ranging from 2E-7 to 1E-6 per year) and a lower contribution to the total core damage frequency (ranging from <1-to-10%). Based on these updated analyses, newer insights on the effects of reactor power equilibration, recirculation pump trip, high and low pressure injection and high pressure seal failure coupled with a detailed accident sequence analysis have resulted in lowering the significance of ATWS to core damage frequency

  4. Plutonium scrap waste processing based on aqueous nitrate and chloride media

    International Nuclear Information System (INIS)

    Navratil, J.D.

    1985-01-01

    A brief review of plutonium scrap aqueous waste processing technology at Rocky Flats is given. Nitric acid unit operations include dissolution and leaching, anion exchange purification and precipitation. Chloride waste processing consists of cation exchange and carbonate precipitation. Ferrite and carrier precipitation waste treatment processes are also described. 3 figs

  5. Processing flowsheet for the accelerator transmutation of waste (ATW) program

    International Nuclear Information System (INIS)

    Dewey, H.; Walker, R.; Yarbro, S.

    1992-01-01

    At Los Alamos, an innovative approach to transmuting long-lived radioactive waste is under investigation. The concept is to use a linear proton accelerator coupled to a solid target to produce an intense neutron flux. The intense stream of neutrons can then be used to fission or transmute long-lived radionuclides to either stable or shorter-lived isotopes. For the program to be successful, robust chemical separations with high efficiencies (>10 5 ) are required. The actual mission, either defense or commercial, will determine what suite of unit operations will be needed. If the mission is to process commercial spent fuel, there are several options available for feed preparation and blanket processing. The baseline option would be an improved PUREX system with the main alternative being the current ATW actinide blanket processing flowsheet. 99 Tc and 129 I are more likely to reach the biosphere than the actinides. Many models have been developed for predicting how the radionuclides will behave in a repository over long time periods. The general conclusion is that the actinides will be sorbed by the soil. Therefore, over a long time period, e.g., a million years their hazard will be lessened because of radioactive decay and dispersion. However, some of the long-lived fission products are not sorbed and could potentially reach the environment over a few thousand year period. Hence, they could present a significant safety hazard. Because of limited resources, most of the priority has been focused on the actinide and technetium blanket assemblies

  6. BWR ATWS mitigation by Fine Motion Control Rod

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

    1994-01-01

    Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

  7. Effects of buffer thickness on ATW blanket performance

    International Nuclear Information System (INIS)

    Yang, W. S.; Mercatali, L.; Taiwo, T. A.; Hill, R. N.

    2001-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy ( and lt; 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level

  8. Effects of Buffer Thickness on ATW Blanket Performance

    International Nuclear Information System (INIS)

    Yang, W.S.; Mercatali, L.; Taiwo, T.A.; Hill, R.N.

    2002-01-01

    This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level. (authors)

  9. ATWS sensitivity studies to support PSA success criteria

    International Nuclear Information System (INIS)

    Zheng Yaoyao; Xu Zhen; Ke Xiao

    2010-01-01

    The limiting anticipated transient without scram (ATWS) event is the heatup transient caused by a reduction of heat removal capability by the secondary side of the plant. In order to evaluate the AP1000 plant behavior following an ATWS, loss of normal feedwater ATWS event has been analyzed using the LOFTRAN code. Several sensitivity studies are also performed to address some key issues, such as steam dump capacity, core makeup tank (CMT) characteristic and boron coefficient, reactor coolant pumps (RCPs) availability, startup feedwater system (STS) availability and steam generator (SG) heat flux. The results of the analysis show that in order to mitigate the consequence of such an accident, the steam dump should be isolated and RCP should trip on CMT actuation signal. (authors)

  10. Aqueous Corrosion Rates for Waste Package Materials

    Energy Technology Data Exchange (ETDEWEB)

    S. Arthur

    2004-10-08

    The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports.

  11. Aqueous Corrosion Rates for Waste Package Materials

    International Nuclear Information System (INIS)

    Arthur, S.

    2004-01-01

    The purpose of this analysis, as directed by ''Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]), is to compile applicable corrosion data from the literature (journal articles, engineering documents, materials handbooks, or standards, and national laboratory reports), evaluate the quality of these data, and use these to perform statistical analyses and distributions for aqueous corrosion rates of waste package materials. The purpose of this report is not to describe the performance of engineered barriers for the TSPA-LA. Instead, the analysis provides simple statistics on aqueous corrosion rates of steels and alloys. These rates are limited by various aqueous parameters such as temperature (up to 100 C), water type (i.e., fresh versus saline), and pH. Corrosion data of materials at pH extremes (below 4 and above 9) are not included in this analysis, as materials commonly display different corrosion behaviors under these conditions. The exception is highly corrosion-resistant materials (Inconel Alloys) for which rate data from corrosion tests at a pH of approximately 3 were included. The waste package materials investigated are those from the long and short 5-DHLW waste packages, 2-MCO/2-DHLW waste package, and the 21-PWR commercial waste package. This analysis also contains rate data for some of the materials present inside the fuel canisters for the following fuel types: U-Mo (Fermi U-10%Mo), MOX (FFTF), Thorium Carbide and Th/U Carbide (Fort Saint Vrain [FSVR]), Th/U Oxide (Shippingport LWBR), U-metal (N Reactor), Intact U-Oxide (Shippingport PWR, Commercial), aluminum-based, and U-Zr-H (TRIGA). Analysis of corrosion rates for Alloy 22, spent nuclear fuel, defense high level waste (DHLW) glass, and Titanium Grade 7 can be found in other analysis or model reports

  12. Audit Calculations of ATWS for Ulchin Unit 1 and 2 Power Uprate

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jun Soo; Huh, Byung Gil; Choi, Yong Seog; Seul, Kwang Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, the regulatory audit calculation for ATWS of Ulchin Unit 1 and 2 with 4.5% power uprate was performed to support the licensing review and to confirm the validity of licensee's calculation. In order to simulate the transient behavior of ATWS initiated by a loss of feed water, the systems of Ulchin Unit 1 and 2 was modeled with MARS-KS 1.3. In this study, the regulatory audit calculation of ATWS for Ulchin 1 and 2 with 4.5% power uprating and 99% MTC in the specific cycle designs was performed. It is conformed that the analysis results of ATWS for Ulchin 1 and 2 power uprate meets the RCS pressure acceptance criteria. An anticipated transient accompanied by a failure in the Reactor Trip System (RTS) to shut down the reactor is defined as an Anticipated Transient Without Scram (ATWS). Under certain postulated conditions, the ATWS could lead to Reactor Coolant system (RCS) pressure boundary fracture and/or core damage. For a conventional pressurized water reactor (PWR), the temperature corresponding to the NSSC notice No.2013.09(Performance Criteria for ECCS of the Pressurized Water Reactor Nuclear Power Plants), 1204 .deg. C and the pressure corresponding to the ASME Boiler and Pressure Vessel Code service level C stress, 221.5 bar is assumed to be an unacceptable plant condition against ATWS, above which the RCS pressure boundary could deform to the point of inoperability and the safe shutdown by injection of borated water could be challenged. Such potentially excessive RCS overpressure may occur in the ATWS initiated from a loss of heat sink. Currently, the modification of Ulchin 1 and 2 operating license for 4.5% power uprate is under review.

  13. Estimation of the development possibility of the ABC/ATW fuel cycle based on LiF-BeF2 fuel salt. Part 2

    International Nuclear Information System (INIS)

    Bychkov, A.V.; Naumov, V.S.

    1994-01-01

    The aim of the first chapter was generalization of data on solubility and equilibrium states of fission product and actinide fluorides in fluoride salt melts-solvents and fuel composition melts based on LiF-BeF 2 mixture which was proposed as fuel basis for ABC/ATW facility. The second chapter is devoted to description of processes proposed for the chemical-technological complex of the ABC/ATW facility and their physico-chemical peculiarities. The complex is responsible for the removal of fission products and actinides from irradiated fuel salt

  14. Mercury separation from aqueous wastes

    International Nuclear Information System (INIS)

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

    1995-07-01

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

  15. Quantification of operator actions during ATWS following MSIV closure

    International Nuclear Information System (INIS)

    Luckas, W.J. Jr.; O'Brien, J.N.; Perline, R.K.; Spettell, C.M.

    1986-01-01

    Brookhaven National Laboratory (BNL) assisted the Accident Sequence Evaluation Program (ASEP) by performing a Human Reliability Analysis (HRA) of the operations crew tasks during the Anticipated Transient Without Scram (ATWS) accident sequence with Main Steam Isolation Valve (MSIV) closure at the Peach Bottom Atomic Power Station, Unit 2. A detailed task analysis was performed based on consideration of staffing, team interaction, and control room layout at Peach Bottom. ATWS scenarios developed by Oak Ridge National Laboratory (ORNL) and Idaho National Engineering Laboratory (INEL) were reviewed. Discussions were held with thermal-hydrodynamic/core neutronics engineers at BNL to determine the success criterion for tasks. Five major operator tasks were identified. After reviewing a computerized data base of human error probabilities (HEPs) from 19 probabilistic risk assessments (PRAs) for tasks similar to those above to establish the historic range of HEPs for such errors, consensus opinion and structured expert judgment was used to quantify each of these tasks at each branch point in the event tree within that range

  16. ATWS analyses for Krsko Full Scope Simulator verification

    Energy Technology Data Exchange (ETDEWEB)

    Cerne, G; Tiselj, I; Parzer, I [Reactor Engineering Div., Inst. Jozef Stefan, Ljubljana (Slovenia)

    2000-07-01

    The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for verification of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD2 code and the input card deck for NPP Krsko was used. The analyses for ATWS were performed to assess the influence and benefit of ATWS Mitigation System Actuation Circuitry (AMSAC). In the presented paper the most severe ATWS scenarios have been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied regarding the AMSAC availability. (author)

  17. Accelerator-driven destruction of long-lived radioactive waste and energy production

    International Nuclear Information System (INIS)

    Schriber, S.O.

    1997-01-01

    Nuclear waste management involves many issues. ATW is an option that can assist a repository by enhancing its capability and thereby assist nuclear waste management. Technology advances and the recent release of liquid metal coolant information from Russia has had an enormous impact on the viability of an ATW system. It now appears economic with many repository enhancing attributes. In time, an ATW option added to present repository activities will provide the public with a nuclear fuel cycle that is acceptable from economic and environmental points of view

  18. Aqueous radioactive waste bituminization

    International Nuclear Information System (INIS)

    Williamson, A.S.

    1980-08-01

    The bituminzation of decontamination and ion exchange resin stripping wastes with four grades of asphalt was investigated to determine the effects of asphalt type on the properties of the final products. All waste forms deformed readily under light loads indicating they would flow if not restrained. It was observed in all cases that product leaching rates increased as the hardness of the asphalt used to treat the waste increased. If bituminization is adopted for any Ontario Hydro aqueous radioactive wastes they should be treated with soft asphalt to obtain optimum leaching resistance and mechanical stability during interim storage should be provided by a corrosion resistant container

  19. Method for aqueous radioactive waste treatment

    Science.gov (United States)

    Bray, L.A.; Burger, L.L.

    1994-03-29

    Plutonium, strontium, and cesium found in aqueous waste solutions resulting from nuclear fuel processing are removed by contacting the waste solutions with synthetic zeolite incorporating up to about 5 wt % titanium as sodium titanate in an ion exchange system. More than 99.9% of the plutonium, strontium, and cesium are removed from the waste solutions. 3 figures.

  20. Proposal on the accelerator driven molten-salt reactor (ATW-concept) benchmark calculation (stage-1 without an external neutron sources)

    International Nuclear Information System (INIS)

    Svarny, J.; Mikolas, P.

    1999-01-01

    The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark will be not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (Authors)

  1. Direct irradiation of long-lived fission products in an ATW system

    Energy Technology Data Exchange (ETDEWEB)

    Carter, T.F. [Univ. of Tennessee, Knoxville, TN (United States); Henderson, D. [Univ. of Wisconsin, Madison, WI (United States); Sailor, W.C. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    The feasibility of directly irradiating five long-lived fission products (LLFPs: {sup 79}Se, {sup 93}Zr, {sup 107}Pd, {sup 126}Sn, and {sup 135}Cs, each with a half-life greater than 10,000 years), by incorporating them into the target of an Accelerator Transmutation of Waste (ATW) system is discussed. The important parameters used to judge the feasibility of a direct irradiation system were the target`s neutron spallation yield (given in neutrons produced per incident proton), and the removal rate of the LLFP, with the baseline incineration rate set at two light water reactors (LWRs) worth of the LLFP waste per year. A target was constructed which consisted of a LLFP cylindrical {open_quotes}plug{close_quotes} inserted into the top (where the proton beam strikes) of a 30 cm radius, 100 cm length lead target. {sup 126}Sn and {sup 79}Se were each found to have high enough removal rates to support two LWR`s production of the LLFP per year of ATW operation. For the baseline plug geometry (5 cm radius, 30 cm length) containing {sup 126}Sn, 3.5 LWRs could be supported per year (at 75% beam availability). Furthermore, the addition of a {sup 126}Sn plus had a slightly positive effect on the target`s neutron yield. The neutron production was 36.83 {plus_minus}.0039 neutrons per proton with a pure lead target having a yield of 36.29 {plus_minus}.0038. It was also found that a plug composed of a tin-selenide compound (SnSe) had high enough removal rates to burn two or more reactor years of both LLFPs simultaneously.

  2. Accelerator transmutation of waste economics

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1995-01-01

    A parametric systems model of the accelerator transmutation of (nuclear) waste (ATW) is used to examine key system trade-offs and design drivers on the basis of unit costs. This model is applied primarily to a fluid-fuel blanket concept for an ATW that generates net electric power from the fissioning of spent commercial reactor fuel. An important goal of this study is the development of essential parametric trade-offs to aid in any future conceptual engineering design of an ATW that would burn spent commercial fuel and generate net electric power. As such, costing procedures and methodologies used to estimate and compare advanced nuclear power generation systems are applied. The cost of electricity required by an electrical power-generating ATW fueled with spent commercial fuels is generally found to be above that projected for other advanced fission power plants. The accelerator and the chemical plant equipment cost accounts are quantitatively identified as main cost drivers, with the capital cost of radio-frequency power dominating the former. Significant reductions of this cost differential are possible by increased blanket neutron multiplication, increased plant capacity, or increased thermal-to-electric conversion efficiency. The benefits of reduced long-lived fission products and spent commercial fuel actinides provided by the ATW approach translate into a less tangible source of revenue to be provided by a charge that must be levied on the client fission power plants being serviced. The main goal of this study, however, is not a direct cost comparison but is instead a quantitative determination of cost-based sensitivity of key cost drivers and operational modes for an ATW concept that would address the growing spent commercial fuel problem; parametric results presented focus on this goal, and a specific ATW ''straw man'' is given to achieve this main objective

  3. ATWS: a reappraisal. Part III. Frequency of anticipated transients. Interim report

    International Nuclear Information System (INIS)

    Leverenz, F.L. Jr.; Koren, J.M.; Erdmann, R.C.; Lellouche, G.S.

    1978-07-01

    The document is Part III of the Institute study of the ATWS question. The frequencies of the various events which have led to a reactor scram are documented from the nuclear power plant records. Some of these events, in the absence of scram, could lead to undesirable system response and are the ''transients of significance'' which comprise the anticipated transients of the ATWS question

  4. Treatment of low and intermediate aqueous waste containing Cs-137 by chemical precipitation

    International Nuclear Information System (INIS)

    Valdezco, E.M.; Marcelo, E.A.; Alamares, A.L.; Junio, J.B.; Dela Cruz, J.M.

    1996-01-01

    The use of radioactive materials in various applications has been increasing since its introduction in the early sixties. The Philippine Nuclear Research Institute has established a centralized facility for treating radioactive wastes i.e. aqueous wastes with assistance from the International Atomic Energy Agency - Technical Cooperation Programme. Liquid wastes containing Cs-137 are generated from aqueous wastes containing Cs-137 by nickel ferrocyanide precipitation will be presented. The aim of this study is to investigate the efficiency treatment in removing Cs-137 from an aqueous effluent. Actual aqueous wastes known to contain Cs-137 were used in the experiments. Low cost and simple nickel ferrocyanide precipitation method with the aid of a flocculant has been selected for the separation of Cs-137 from low and intermediate aqueous waste. By varying the chemical dosage added into the aqueous waste, different decontamination factors were obtained. Hence, the optimum dosage of the chemicals that give the highest decontamination factor can be determined. (author)

  5. Technetium removal from aqueous wastes

    International Nuclear Information System (INIS)

    Fletcher, P.A.; Jones, C.P.; Junkison, A.R.; Turner, A.D.; Kavanagh, P.R.

    1992-03-01

    The research discussed in this report has compared several ''state of the art'' techniques for the removal of traces of the radionuclide, technetium, from aqueous wastes. The techniques investigated were: electrochemical reduction to an insoluble oxide, electrochemical ion exchange, seeded ultrafiltration and chemical reduction followed by filtration. Each technique was examined using a simulant based upon the waste generated by the Enhanced Actinide Removal Plant (EARP) at Sellafield. The technique selected for further investigation was direct electrochemical reduction which offers an ideal route for the removal of technetium from the stream (DFs 10-100) and can be operated continuously with a low power consumption 25 kW for the waste generated by EARP. Cell designs for scale up have been suggested to treat the 1000m 3 of waste produced every day. Future work is proposed to investigate the simultaneous removal of other key radionuclides, such as ruthenium, plutonium and cobalt as well as scale up of the resulting process and to investigate the effect of these other radionuclides on the efficiency of the electrochemical reduction technique for the removal of technetium. Total development and full scale plant costs are estimated to be of the order of 5 pounds - 10M, with a time scale of 5 -8 years to realisation. (author)

  6. Analysis of SBO ATWS for Maanshan PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Che-Hao; Chen, Shao-Wen [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Shih, Chunkuan [National Tsing Hua Univ., Hsinchu, Taiwan (China). Inst. of Nuclear Engineering and Science; Nuclear and New Energy Education and Research Foundation, Hsinchu, Taiwan (China); Lin, Hao-Tzu [Atomic Energy Council, Taoyuan, Taiwan (China). Inst. of Nuclear Energy Research

    2015-11-15

    Station blackout anticipated transient without scram (SBO ATWS) is considered as loss of off-site and on-site power but no credit for automatic reactor trip. SBO ATWS causes reactor coolant pump (RCP) trip, loss of all main feedwater pumps and turbine trip, then the reactor coolant system (RCS) pressure rises rapidly due to loss of heat removal paths. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) is assumed to be an unacceptable plant condition in SECY-83-293. The simulation is performed by TRACE which is a thermal-hydraulic code developed by U.S. NRC. Three different AFW flows are modeled to ensure the pressures will not be beyond the criteria. RCP seal-leakage is concerned as a SBLOCA due to loss of RCP seal-cooling. Four possible leakage flows are modeled to examine the reactor core water level and temperature variation.

  7. Investigations of anticipated transients without scram (ATWS) for the high temperature reactor

    International Nuclear Information System (INIS)

    Heckhoff, H.D.

    1981-10-01

    In this study anticipated transients without scram (ATWS) are investigated for the high temperature reactor, especially for the thorium high temperature reactor (THTR) 300 MWe as an example. It is shown that the two ATWS 'feedwater flow reduction from full power' and 'positive reactivity insertion of 1 mNile/s from 40 per cent power' are the most important transients for the THTR. The additional load caused by the ATWS can be reduced sufficiently by some small modifications of the afterheat removal system. Supplementary precautions are not necessary. In the last part of this study some possibilities to improve the behaviour of the power plant are shown with regard to high temperature reactors of the future, the partial scram as well as some modifications of heating and cooling of the steam generator. (orig.) [de

  8. Core dynamics of HTR under ATWS and accident conditions

    International Nuclear Information System (INIS)

    Nabbi, R.

    1988-05-01

    The systematic classification of the ATWS has been undertaken by analogy to the considerations made for LWR. The initiating events of ATWS and protection actions of safety systems resulting from monitoring of the system variables have been described. The main emphasis of this work is the analysis of the core dynamic consequences of scram failure during the anticipated transients. The investigation has shown that because of the temperature feedback mechanisms a temperature rise during the ATWS results in a self-shutdown of the reactor. Further inherent safety features of the HTR - conditioned by the high heat capacity of the core and by the compressibility of the coolant - do effectively counteract an undesirable increase of temperature and pressure in the primary circuit. In case of the long-term failure of the forced cooling and following core heatup, neutron physical phenomena appear which determine the reactivity behaviour of the HTR. They are, for instance, the decay of Xenon 135, release of the fission products and subsiding of the top reflector. The results of the computer simulations show that a recriticality has to be excluded during the first 2 days if the reactor is shutdown by the reflector rods at the beginning of the accident. (orig./HP) [de

  9. Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

    International Nuclear Information System (INIS)

    Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun

    2014-01-01

    The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available

  10. Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available.

  11. Transmutation of radioactive waste: Effect on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Rasmussen, N.C.; Pigford, T.H.

    1997-01-01

    A committee of the National Research Council reviewed three concepts for transmuting radionuclides recovered from the chemical reprocessing of commercial light-water-reactor (LWR) fuel: LWR transmutation reactors fueled with recycled actinides, advanced liquid-metal reactors (ALMRs), and accelerator-driven subcritical reactors for transmutation of waste (ATW). The concepts were evaluated in terms of: (1) the extent to which waste disposal would benefit from transmutation, (2) time required to reduce the total inventory of radionuclides in the waste and fuel cycle, (3) the complexity of the overall transmutation system, (4) the extent of new development required, and (5) institutional and economic problems of operating such systems. Transmutation could affect geologic disposal of waste by reducing the inventory of transuranics (TRUs), fission products, and other radionuclides in the waste. Reducing the inventory of transuranics does not necessarily affect radiation doses to people who use contaminated ground water if the dissolution rate of transuranics in waste is controlled by elemental solubilities. However, reducing inventories of Am and Pu would decrease potential hazards from human intrusion. The likelihood for underground nuclear criticality would also be reduced. The long-lived fission products Tc-99, I-129, Cs-135 and others typically contribute most to the long-term radiation doses to future populations who use contaminated water from the repository. Their transmutation requires thermal or epithermal neutrons, readily available in LWR and ATW transmutors. ALMR and LWR transmutors would require several hundred years to reduce the total transuranic inventory by even a factor of 10 at constant electric power, and thousands of years for a hundred-fold reduction. For the same electrical power, the ATW could reduce total transuranic inventory about tenfold more rapidly, because of its very high thermal-neutron flux. However, extremely low process losses would be

  12. ATWS analysis for total loss of feedwater sequence in UCN 3 and 4

    International Nuclear Information System (INIS)

    Park, S. H.; Song, Y. M.; Kim, D. H.; Kim, S. D.; Park, S. Y.

    1999-01-01

    ATWS is a trip-failed severe accident initiated from the transients like a turbine trip, a control bank withdrawal, and a loss of feedwater which are expected to occur comparatively often (one or two occurrences / year). In this study, an ATWS sequence in Ulchin 3 and 4 is analyzed and the effects of the important systems are studied for accident management purpose using a MIDAS/PK computer code. The MIDAS/PK code has been developed via coupling a point kinetics module with the MELCOR code. The code calculates a primary peak pressure of about 24MPa at 240 seconds for the ATWS initiated by a TLOF (Total Loss of Feedwater) transient. Along with the basic ATWS analysis, several sensitivity runs are performed. From these, the turbines and the safety depressurization system (SDS) are judged to be important. The turbine trip resulting in a loss of offsite power and a RCP trip, degrades primary heat transfer to the secondary sides, and in turn, increases primary coolant temperature which reduces the reactor power due to the negative moderator temperature coefficient. Manual operation of SDS has an effect to lower the primary peak pressure considerably via supplementary depressurization in addition to the PORVs

  13. ACCELERATOR TRANSMUTATION OF WASTE TECHNOLOGY AND IMPLEMENTATION SCENARIOS

    International Nuclear Information System (INIS)

    Beller, D.; Tuyle, G. van

    2000-01-01

    During 1999, the U.S. Department of Energy, in conjunction with its nuclear laboratories, a national steering committee, and a panel of world experts, developed a roadmap for research, development, demonstration, and deployment of Accelerator-driven Transmutation of Waste (ATW). The ATW concept that was examined in this roadmap study was based on that developed at the Los Alamos National Laboratory (LANL) during the 1990s. The reference deployment scenario in the Roadmap was developed to treat 86,300 tn (metric tonnes initial heavy metal) of spent nuclear fuel that will accumulate through 2035 from existing U.S. nuclear power plants (without license extensions). The disposition of this spent nuclear reactor fuel is an issue of national importance, as is disposition of spent fuel in other nations. The U.S. program for the disposition of this once-through fuel is focused to characterize a candidate site at Yucca Mountain, Nevada for a geological repository for spent fuel and high-level waste. The ATW concept is being examined in the U.S. because removal of plutonium minor actinides, and two very long-lived isotopes from the spent fuel can achieve some important objectives. These objectives include near-elimination of plutonium, reduction of the inventory and mobility of long-lived radionuclides in the repository, and use of the remaining energy content of the spent fuel to produce power. The long-lived radionuclides iodine and technetium have roughly one million year half-lives, and they are candidates for transport into the environment via movement of ground water. The scientists and engineers who contributed to the Roadmap Study determined that the ATW is affordable, doable, and its deployment would support all the objectives. We report the status of the U.S. ATW program describe baseline and alternate technologies, and discuss deployment scenarios to support the existing U.S. nuclear capability and/or future growth with a variety of new fuel cycles

  14. Transmutation of fission products and actinide waste at Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Daemen, L.L.; Pitcher, E.J.; Russell, G.J. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    The authors studied the neutronics of an ATW system for the transmutation of the fission products ({sup 99}Tc in particular) and the type of actinide waste stored in several tanks at Hanford. The heart of the system is a highly-efficient neutron production target. It is surrounded by a blanket containing a moderator/reflector material, as well as the products to be transmuted. The fission products are injected into the blanket in the form of an aqueous solution in heavy water, whereas an aqueous actinides slurry is circulated in the outer part of the blanket. For the sake of definiteness, the authors focussed on {sup 99}Tc (the most difficult fission product to transmute), and {sup 239}Pu, {sup 237}Np, and {sup 241}Am. Because of the low thermal neutron absorption cross-section of {sup 99}Tc, considerable care and effort must be devoted to the design of a very efficient neutron source.

  15. Accelerator-driven transmutation of high-level waste from the defense and commercial sectors

    International Nuclear Information System (INIS)

    Bowman, C.; Arthur, E.; Beard, C.

    1996-01-01

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The major goal has been to develop accelerator transmutation of waste (ATW) system designs that will thoroughly and rapidly transmute nuclear waste, including plutonium from dismantled weapons and spent reactor fuel, while generating useful electrical power and without producing a long-lived radioactive waste stream. We have identified and quantified the unique qualities of subcritical nuclear systems and their capabilities in bringing about the complete destruction of plutonium. Although the 1191 subcritical systems involved in our most effective designs radically depart from traditional nuclear reactor concepts, they are based on extrapolations of existing technologies. Overall, care was taken to retain the highly desired features that nuclear technology has developed over the years within a conservative design envelope. We believe that the ATW systems designed in this project will enable almost complete destruction of nuclear waste (conversion to stable species) at a faster rate and without many of the safety concerns associated with the possible reactor approaches

  16. Basic design of alpha aqueous waste treatment process in NUCEF

    Energy Technology Data Exchange (ETDEWEB)

    Mineo, Hideaki; Matsumura, Tatsuro; Nishizawa, Ichio; Mitsui, Takeshi; Ueki, Hiroyuki; Wada, Atsushi; Sakai, Ichita; Takeshita, Isao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nishimura, Kenji

    1996-11-01

    This paper described the basic design of Alpha Aqueous Waste Treatment Process in NUCEF. Since various experiments using the TRU (transuranium) elements are carried out in NUCEF, wastes containing TRU elements arise. The liquid wastes in NUCEF are categorized into three types. Decontamination and volume reduction of the liquid waste mainly of recovery water from acid recovery process which has lowest radioactive concentration is the most important task, because the arising rate of the waste is large. The major function of the Alpha Aqueous Waste Treatment Process is to decontaminate the radioactive concentration below the level which is allowed to discharge into sea. Prior the process design of this facility, the followings are evaluated:property and arising rate of the liquid waste, room space to install and licensing condition. Considering varieties of liquid wastes and their large volume, the very high decontamination factor was proposed by a process of multiple evaporation supported with filtration and adsorption in the head end part and reverse osmosis in the distillate part. (author)

  17. Quantitative Characterization of Aqueous Byproducts from Hydrothermal Liquefaction of Municipal Wastes, Food Industry Wastes, and Biomass Grown on Waste

    Energy Technology Data Exchange (ETDEWEB)

    Maddi, Balakrishna; Panisko, Ellen; Wietsma, Thomas; Lemmon, Teresa; Swita, Marie; Albrecht, Karl; Howe, Daniel

    2017-01-27

    Hydrothermal liquefaction (HTL) is a viable thermochemical process for converting wet solid wastes into biocrude which can be hydroprocessed to liquid transportation fuel blendstocks and specialty chemicals. The aqueous byproduct from HTL contains significant amounts (20 to 50%) of the feed carbon, which must be used to enhance economic sustainability of the process on an industrial scale. In this study, aqueous fractions produced from HTL of industrial and municipal waste were characterized using a wide variety of analytical approaches. Organic chemical compounds present in these aqueous fractions were identified using two-dimensional gas chromatography equipped with time-of-flight mass spectrometry. Identified compounds include organic acids, nitrogen compounds, alcohols, aldehydes, and ketones. Conventional gas chromatography and liquid chromatography methods were employed to quantify the identified compounds. Inorganic species, in the aqueous stream of hydrothermal liquefaction of these aqueous byproducts, also were quantified using ion chromatography and inductively coupled plasma optical emission spectroscopy. The concentrations of organic chemical compounds and inorganic species are reported, and the significance of these results is discussed in detail.

  18. Planning the research and development necessary for accelerator transmutation of waste, leading to integrated proof of performance testing

    International Nuclear Information System (INIS)

    Bennett, D.R.; Pasamehmetoglu, K.; Finck, P.; Pitcher, E.; Khalil, H.; Todosow, M.; Hill, R.; Van Tuyle, G.; Laidler, J.; Crawford, D.; Thomas, K.

    2001-01-01

    The Research and Development (R and D) Plan for the Accelerator Transmutation of Waste (ATW) Program has been developed for the Department of Energy, Office of Nuclear Energy (DOE/NE) to serve as a focus and progressional guide in developing critical transmutation technologies. It is intended that the Plan will serve as a logical reference considering all elements of an integrated accelerator-driven transmutation system, and will maximize the use of resources by identifying and prioritizing research, design, development and trade activities. The R and D Plan provides a structured framework for identifying and prioritizing activities leading to technically-justifiable integrated Proof of Performance testing within ten years and ultimate demonstration of Accelerator Transmutation of Waste (ATW). The Plan builds from the decision objectives specified for ATW, utilizes informational input from the ATW Roadmap and programmatic System Point Design efforts, and employs the knowledge and expertise provided by professionals familiar with ATW technologies. With the firm intent of understanding what, why and when information is needed, including critical interfaces, the Plan then develops a progressional strategy for developing ATW technologies with the use of a Technology Readiness Level (TRL) scale. The TRL approach is first used to develop a comprehensive, yet generic, listing of experimental, analytical and trade study activities critical to developing ATW technologies. Technology-specific and concept-specific aspects are then laid over the generic mapping to gage readiness levels. Prioritization criteria for reducing technical uncertainty, providing information to decision points, and levering off of international collaborations are then applied to focus analytical, experimental and trade activities. (author)

  19. Water reuse achieved by zero discharge of aqueous waste

    International Nuclear Information System (INIS)

    Kelchner, B.L.

    1976-01-01

    Plans for zero discharge of aqueous waste from ERDA's nuclear weapons plant near Denver are discussed. Two plants - a process waste treatment facility now under construction, and a reverse osmosis desalting plant now under design, will provide total reuse of waste water for boiler feed and cooling tower supply. Seventy million gallons of water per year will be conserved and downstream municipalities will be free of inadvertent pollution hazards

  20. ATWS: a reappraisal, part II, evaluation of societal risks due to reactor protection systems failure. Vol. 3. Pwr risk analysis. Phase report

    International Nuclear Information System (INIS)

    Lellouche, G.S.

    1976-08-01

    This document is the third volume of part 2 in a series of studies which will examine the basis for the problem of Anticipated Transients Without Scram (ATWS). The purpose of part 2 is an evaluation of societal risks due to RPS failure based on more current data and methodology than used in WASH-1270. This volume examines and documents the potential contribution to societal risk due to ATWS in the PWR. Volumes 1 and 2 described a similar analysis for the BWR

  1. Effect of reactor conditions on MSIV-ATWS power level

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1987-01-01

    In a boiling water reactor (BWR) when there is closure of the main steam isolation valves (MSIVs), the energy generated in the core will be transferred to the pressure suppression pool (PSP) via steam that flows out of the relief valves. The pool has limited capacity as a heat sink and hence, if there is no reactor trip [an anticipated transient without scram (ATWS) event], there is the possibility that the pool temperature may rise beyond acceptable limits. The present study was undertaken to determine how the initial reactor conditions affect the power level during an MSIV-ATWS event. The time of interest is the 20- to 30-min period when it is assumed that the reactor is in a quasi equilibrium condition with the water level and pressure fixed, natural circulation conditions and no control rod movement or significant boron in the core. The initial conditions of interest are the time of the cycle and the operating state

  2. Performance comparison of liquid metal and gas cooled ATW system point designs

    International Nuclear Information System (INIS)

    Yang, W.S.; Taiwo, T.A.; Hill, R.N.; Khalil, H.S.; Wade, D.C.

    2001-01-01

    As part of the Advanced Accelerator Application (AAA) program in the U.S., preliminary design studies have been performed at Argonne National Laboratory (ANL) and Los Alamos National Laboratory (LANL) to define and compare candidate Accelerator Transmutation of Waste (ATW) systems. The studies at ANL have focused primarily on the transmutation blanket component of the overall system. Lead-bismuth eutectic (LBE), sodium, and gas cooled systems are among the blanket technology options currently under consideration. This paper summarizes the results from neutronics trade studies performed at ANL. Core designs have been developed for LBE and sodium cooled 840 MWt fast spectrum accelerator driven systems employing re-cycle. Additionally, neutronics analyses have been performed for a helium-cooled 600 MWt hybrid thermal and fast spectrum system proposed by General Atomics (GA), which is operated in the critical mode for three cycles and in a subcritical accelerator driven mode for a subsequent single cycle. For these three point designs, isotopic inventories, consumption rates, and annual burnup rates are compared. The mass flows and the ultimate loss of transuranic (TRU) isotopes to the waste stream per unit of heat generated during transmutation are also compared on a consistent basis. (author)

  3. The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF2 basis

    International Nuclear Information System (INIS)

    Naumov, V.S.; Bychkov, A.V.

    1995-01-01

    The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF 2 -PuF 3,(4) - MAF n ): - continious removal of radioactive gases, volatile impurities and 'noble fission products'; - portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally

  4. The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF2 basis

    International Nuclear Information System (INIS)

    Naumov, V. S.; Bychkov, A. V.

    1995-01-01

    The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF2-PuF3,(4)-MAFn): -continious removal of radioactive gases, volatile impurities and 'noble fission products'; -portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally

  5. Generic implications of ATWS events at the Salem Nuclear Power Plant: generic implications. Vol. 1

    International Nuclear Information System (INIS)

    1983-04-01

    This report is the first of two volumes. It documents the work of an interoffice, interdisciplinary NRC Task Force established to determine the generic implications of two anticipated transients without scram (ATWS) at the Salem Nuclear Power Plant, Unit 1 on February 22 and 25, 1983. A second report will document the NRC actions to be taken based on the work of the Task Force. The Task Force was established to address three questions: (1) Is there a need for prompt action for similar equipment in other facilities. (2) Are NRC and its licensees learning the sefety-management lessons, and, (3) How should the priority and content of the ATWS rule be adjusted. A number of short-term actions were taken through Bulletins and an Information Notice. Intermediate-term actions to address the generic issues will be addressed in the separate report and implemented through appropriate regulatory mechanisms

  6. 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.

  7. Treatment plan for aqueous/organic/decontamination wastes under the Oak Ridge Reservation FFCA Development, Demonstration, Testing, and Evaluation Program

    International Nuclear Information System (INIS)

    Backus, P.M.; Benson, C.E.; Gilbert, V.P.

    1994-08-01

    The U.S. Department of Energy (DOE) Oak Ridge Operations Office and the U.S. Environmental Protection Agency (EPA)-Region IV have entered into a Federal Facility Compliance Agreement (FFCA) which seeks to facilitate the treatment of low-level mixed wastes currently stored at the Oak Ridge Reservation (ORR) in violation of the Resource, Conservation and Recovery Act Land Disposal Restrictions. The FFCA establishes schedules for DOE to identify treatment for wastes, referred to as Appendix B wastes, that current have no identified or existing capacity for treatment. A development, demonstration, testing, and evaluation (DDT ampersand E) program was established to provide the support necessary to identify treatment methods for mixed was meeting the Appendix B criteria. The Program has assembled project teams to address treatment development needs for major categories of the Appendix B wastes based on the waste characteristics and possible treatment technologies. The Aqueous, Organic, and Decontamination (A OE D) project team was established to identify pretreatment options for aqueous and organic wastes which will render the waste acceptable for treatment in existing waste treatment facilities and to identify the processes to decontaminate heterogeneous debris waste. In addition, the project must also address the treatment of secondary waste generated by other DDT ampersand E projects. This report details the activities to be performed under the A OE D Project in support of the identification, selection, and evaluation of treatment processes. The goals of this plan are (1) to determine the major aqueous and organic waste streams requiring treatment, (2) to determine the treatment steps necessary to make the aqueous and organic waste acceptable for treatment in existing treatment facilities on the ORR or off-site, and (3) to determine the processes necessary to decontaminate heterogeneous wastes that are considered debris

  8. Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

    International Nuclear Information System (INIS)

    Giachetti, R.T.

    1989-09-01

    This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

  9. Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT Facility

    International Nuclear Information System (INIS)

    Varacalle, D.J.; Giri, A.M.; Koizumi, Y.; Koske, J.E.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAPS/MOD1 computer code showed good agreement with the experimental data

  10. The possibility of fuel cycle design for ABC/ATW complex with molten fuel on LiF-BeF{sub 2} basis

    Energy Technology Data Exchange (ETDEWEB)

    Naumov, V.S.; Bychkov, A.V. [Research Institute of Atomic Reactors, Dimitrovgrad (Russian Federation)

    1995-10-01

    The experience gained in the field of the development of molten salt reactors (MSR) can be made a basis of chemical processing of the ABC/ATW liquid fuel. The following combination of two processing principles are proposed for the ABC/ATW fuel (LiF-BeF{sub 2}-PuF{sub 3,(4)} - MAF{sub n}): - continious removal of radioactive gases, volatile impurities and {open_quotes}noble fission products{close_quotes}; - portion-by-portion electrochemical processing with removal of rare earth elements and some other fission products at an autonomous plant. After processing the fuel salt is brought back to the blanket of the ABC/ATW complex. The analysis of information previously published in different countries allows for a safe assumption that the ABC/ATW fuel cycle with liquid fuel salt is feasible and can be demonstrated experimentally.

  11. 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.

  12. Polonium release from an ATW burner system with liquid lead-bismuth coolant

    International Nuclear Information System (INIS)

    Li, N.; Yefimov, E.; Pankratov, D.

    1998-04-01

    The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods

  13. 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.

  14. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Peng, C.M.; Maly, J.

    1988-01-01

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  15. Treatability study of aqueous, land disposal restricted mixed wastes

    International Nuclear Information System (INIS)

    Haefner, D.R.

    1992-12-01

    Treatment studies have been completed on two aqueous waste streams at the Mixed Waste Storage Facility that are classified as land disposal restricted. Both wastes had mercury and lead as characteristic hazardous constituents. Samples from one of these wastes, composed of mercury and lead sulfide particles along with dissolved mercury and lead, was successfully treated by decanting, filtering, and ion exchanging. The effluent water had an average level of 0.003 and 0.025 mg/L of mercury and lead, respectively. These values are well below the targeted RCRA limits of 0.2 mg/L mercury and 5.0 mg/L lead. An acidic stream, containing the same hazardous metals, was also successfully treated using a treatment process of precipitation, filtering, and then ion exchange. Treatment of another waste was not completely successful, presumably because of the interference of a chelating agent

  16. SWEPP PAN assay system uncertainty analysis: Active mode measurements of solidified aqueous sludge waste

    International Nuclear Information System (INIS)

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

    1997-12-01

    The Idaho National Engineering and Environmental Laboratory is being used as a temporary storage facility for transuranic waste generated by the US Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. This paper is one of a series of reports quantifying the results of the uncertainty analysis of the PAN system measurements for specific waste types and measurement modes. In particular this report covers active mode measurements of weapons grade plutonium-contaminated aqueous sludge waste contained in 208 liter drums (item description codes 1, 2, 7, 800, 803, and 807). Results of the uncertainty analysis for PAN active mode measurements of aqueous sludge indicate that a bias correction multiplier of 1.55 should be applied to the PAN aqueous sludge measurements. With the bias correction, the uncertainty bounds on the expected bias are 0 ± 27%. These bounds meet the Quality Assurance Program Plan requirements for radioassay systems

  17. Aqueous Waste Treatment Plant at Aldermaston

    International Nuclear Information System (INIS)

    Keene, D.; Fowler, J.; Frier, S.

    2006-01-01

    For over half a century the Pangbourne Pipeline formed part of AWE's liquid waste management system. Since 1952 the 11.5 mile pipeline carried pre-treated wastewater from the Aldermaston site for safe dispersal in the River Thames. Such discharges were in strict compliance with the exacting conditions demanded by all regulatory authorities, latterly, those of the Environment Agency. In March 2005 AWE plc closed the Pangbourne Pipeline and ceased discharges of treated active aqueous waste to the River Thames via this route. The ability to effectively eliminate active liquid discharges to the environment is thanks to an extensive programme of waste minimization on the Aldermaston site, together with the construction of a new Waste Treatment Plant (WTP). Waste minimization measures have reduced the effluent arisings by over 70% in less than four years. The new WTP has been built using best available technology (evaporation followed by reverse osmosis) to remove trace levels of radioactivity from wastewater to exceptionally stringent standards. Active operation has confirmed early pilot scale trials, with the plant meeting throughput and decontamination performance targets, and final discharges being at or below limits of detection. The performance of the plant allows the treated waste to be discharged safely as normal industrial effluent from the AWE site. Although the project has had a challenging schedule, the project was completed on programme, to budget and with an exemplary safety record (over 280,000 hours in construction with no lost time events) largely due to a pro-active partnering approach between AWE plc and RWE NUKEM and its sub-contractors. (authors)

  18. The treatment of radioactive aqueous wastes by reverse osmosis

    International Nuclear Information System (INIS)

    Hodgson, T.D.

    Experiments were carried out to determine the rejection factors for the more important radionuclides found in aqueous wastes, to study activity deposition within reverse osmosis modules, and to obtain experience in active operation of a reverse osmosis facility. It was found that reverse osmosis is likely to be useful in aqueous radioactive waste treatment when a wide range of contaminants rather than a specific radioactive species must be removed. There appeared to be no barrier to active operation, although greater confidence in the reliability of pumps and membranes is needed. The rejection of trace quantities of radioisotopes such as Cs + or Sr ++ could be predicted from the behaviour of similar inactive ions. Activity present as polyvalent ions or colloidal aggregates is highly rejected by the membrane. Activity may be deposited onto the membrane with insoluble or scaling compounds, and is greatest on areas of the membrane shielded from the sweeping action of the liquor flow

  19. Stabilization of Savannah River National Laboratory (SRNL) Aqueous Waste by Fluidized Bed Steam Reforming (FBSR)

    International Nuclear Information System (INIS)

    Jantzen, C

    2004-01-01

    The Savannah River National Laboratory (SRNL) is a multidisciplinary laboratory operated by Westinghouse Savannah River Company (WSRC) in Aiken, South Carolina. Research and development programs have been conducted at SRNL for ∼50 years generating non-radioactive (hazardous and non-hazardous) and radioactive aqueous wastes. Typically the aqueous effluents from the R and D activities are disposed of from each laboratory module via the High Activity Drains (HAD) or the Low Activity Drains (LAD) depending on whether they are radioactive or not. The aqueous effluents are collected in holding tanks, analyzed and shipped to either H-Area (HAD waste) or the F/H Area Effluent Treatment Facility (ETF) (LAD waste) for volume reduction. Because collection, analysis, and transport of LAD and HAD waste is cumbersome and since future treatment of this waste may be curtailed as the F/H-Area evaporators and waste tanks are decommissioned, SRNL laboratory operations requested several proof of principle demonstrations of alternate technologies that would define an alternative disposal path for the aqueous wastes. Proof of principle for the disposal of SRNL HAD waste using a technology known as Fluidized Bed Steam Reforming (FBSR) is the focus of the current study. The FBSR technology can be performed either as a batch process, e.g. in each laboratory module in small furnaces with an 8'' by 8'' footprint, or in a semi-continuous Bench Scale Reformer (BSR). The proof of principle experiments described in this study cover the use of the FBSR technology at any scale (pilot or full scale). The proof of principle experiments described in this study used a non-radioactive HAD simulant

  20. Chemical precipitation processes for the treatment of aqueous radioactive waste

    International Nuclear Information System (INIS)

    1992-01-01

    Chemical precipitation by coagulation-flocculation and sedimentation has been commonly used for many years to treat liquid (aqueous) radioactive waste. This method allows the volume of waste to be substantially reduced for further treatment or conditioning and the bulk of the waste to de discharged. Chemical precipitation is usually applied in combination with other methods as part of a comprehensive waste management scheme. As with any other technology, chemical precipitation is constantly being improved to reduce cost to increase the effectiveness and safety on the entire waste management system. The purpose of this report is to review and update the information provided in Technical Reports Series No. 89, Chemical Treatment of Radioactive Wastes, published in 1968. In this report the chemical methods currently in use for the treatment of low and intermediate level aqueous radioactive wastes are described and illustrated. Comparisons are given of the advantages and limitations of the processes, and it is noted that good decontamination and volume reduction are not the only criteria according to which a particular process should be selected. Emphasis has been placed on the need to carefully characterize each waste stream, to examine fully the effect of segregation and the importance of looking at the entire operation and not just the treatment process when planning a liquid waste treatment facility. This general approach includes local requirements and possibilities, discharge authorization, management of the concentrates, ICRP recommendations and economics. It appears that chemical precipitation process and solid-liquid separation techniques will continue to be widely used in liquid radioactive waste treatment. Current research and development is showing that combining different processes in one treatment plant can provide higher decontamination factors and smaller secondary waste arisings. Some of these processes are already being incorporated into new and

  1. U.S. advanced accelerator applications program: plans to develop and test waste transmutation technologies

    International Nuclear Information System (INIS)

    Van Tuyle, G.; Bennett, D.; Arthur, E.; Cappiello, M.; Finck, P.; Hill, D.; Herczeg, J.; Goldner, F.

    2001-01-01

    The primary mission of the U.S. Advanced Accelerator Applications (AAA) Program is to establish a national nuclear technology research capability that can demonstrate accelerator-based transmutation of waste and conduct transmutation research while at the same time providing a capability for the production of tritium if required. The AAA Program was created during fiscal year 2001 from the Accelerator Transmutation of Waste (ATW) Program and the Accelerator Production of Tritium (APT) Project. This paper describes the new AAA Program, as well as its two major components: development and testing of waste transmutation technologies and construction of an integrated accelerator-driven test facility (ADTF). (author)

  2. Criticality safety analysis of accelerator transmutation waste system

    International Nuclear Information System (INIS)

    Landeyro, P.A.; Cepraga, D.G.; Orazi, A.

    1993-01-01

    The Accelerator Transmutation Waste system (ATW) is under development at the Los Alamos National Laboratory. It consists of a particle accelerator producing a proton beam having an energy of 1.5 GeV. These particles are introduced into the upper part of a molten Pb-Bi column and they produce, by a spallation reaction, a high strength neutron flux, 1.0x10 16 n/(square centimeters sec). The neutrons enter a heavy water blanket where actinides and long-lived fission products circulate in vertical tubes. The goal of this research effort is to perform an independent verification of the feasibility of actinide burning in the ATW system. The work is divided into four tasks: a) production of an actinide and long-lived fission product cross section library from JEF 2.2; b) simulation, using MCNP and KENO IV Monte Carlo codes, of the ATW configurations existing in literature; c) validation of the cross sections by comparison of Keff and reaction rate results, calculated with MCNP and KENO IV, with experimental benchmarks and intercomparison between calculations of a PWR unit cell and the computations carried out with various codes and cross section libraries (NEACRF criticality working group data); d) simulation of the ATW configuration. The two first tasks are almost complete with excellent agreement between this study's results and those of Los Alamos

  3. Calculation of the time behavior of a PWR NPP during a loss of feedwater ATWS case

    International Nuclear Information System (INIS)

    Hoeld, A.

    1988-01-01

    Event tree analyses of plant internal accidents play an important role within the safety evaluations of nuclear power reactors. The consequences after normal and abnormal operational perturbations have to be studied with respect to the safety situation of the entire plant and the possibility of additional failures in the reactor scram system be taken into account. In the analysis of anticipated transients with or without reactor scram (non-ATWS or ATWS-cases), it can, according to their initiating events, be distinguished between three important categories, namely - loss of off-site and on-site power (LOOP), - turbine-trip without opening of the bypass station, - loss of main feed water (LOFW). The last case with the additional assumption of a failure in the control rod drive will be subject of this presentation, calculating the dynamic behavior of a PWR NPP (with an end of cycle core, EOC) after such a LOFW/ATWS accident by the transient code combination ALMOD-4/UTSG-2. A short characterization of this combination will be given before consequences of such an accident and the interactions of the different plant parameters are discussed in more detail on basis of the corresponding calculation

  4. Aqueous-based thick photoresist removal for bumping applications

    Science.gov (United States)

    Moore, John C.; Brewer, Alex J.; Law, Alman; Pettit, Jared M.

    2015-03-01

    Cleaning processes account for over 25% of processing in microelectronic manufacturing [1], suggesting electronics to be one of the most chemical intensive markets in commerce. Industry roadmaps exist to reduce chemical exposure, usage, and waste [2]. Companies are encouraged to create a safer working environment, or green factory, and ultimately become certified similar to LEED in the building industry [3]. A significant step in this direction is the integration of aqueous-based photoresist (PR) strippers which eliminate regulatory risks and cut costs by over 50%. One of the largest organic solvent usages is based upon thick PR removal during bumping processes [4-6]. Using market projections and the benefits of recycling, it is estimated that over 1,000 metric tons (mt) of residuals originating from bumping processes are incinerated or sent to a landfill. Aqueous-based stripping would eliminate this disposal while also reducing the daily risks to workers and added permitting costs. Positive-tone PR dissolves in aqueous strippers while negative-tone systems are lifted-off from the substrate, bumps, pillars, and redistribution layers (RDL). While the wafers are further processed and rinsed, the lifted-off PR is pumped from the tank, collected onto a filter, and periodically back-flushed to the trash. The PR solids become a non-hazardous plastic waste while the liquids are mixed with the developer stream, neutralized, filtered, and in most cases, disposed to the sewer. Regardless of PR thickness, removal processes may be tuned to perform in <15min, performing at rates nearly 10X faster than solvents with higher bath lives. A balanced formula is safe for metals, dielectrics, and may be customized to any fab.

  5. Disposition of nuclear waste using subcritical accelerator-driven systems

    International Nuclear Information System (INIS)

    Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

    1998-01-01

    Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power

  6. Effects of aqueous environment on long-term durability of phosphate-bonded ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Jeong, S.Y.

    1996-01-01

    Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically-bonded phosphate ceramics for solidifying and stabilizing low-level mixed wastes. This technology is crucial for stabilizing waste streams that contain volatile species and off-gas secondary waste streams generated by high-temperature treatment of such wastes. Magnesium phosphate ceramic has been developed to treat mixed wastes such as ash, salts, and cement sludges. Waste forms of surrogate waste streams were fabricated by acid-base reactions between the mixtures of magnesium oxide powders and the wastes, and phosphoric acid or acid phosphate solutions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with ash waste streams spiked with hazardous and radioactive surrogates. Standard leaching tests such as ANS 16.1 and TCLP were conducted on the final waste forms. Fates of the contaminants in the final waste forms were established by electron microscopy. In addition, stability of the waste forms in aqueous environments was evaluated with long-term water-immersion tests

  7. Handling and treatment of radioactive aqueous wastes

    International Nuclear Information System (INIS)

    1992-07-01

    This report aims to provide essential guidance to developing Member States without a nuclear power programme regarding selection, design and operation of cost effective treatment processes for radioactive aqueous liquids arising as effluents from small research institutions, hospitals and industries. The restricted quantities and low activity associated with the relevant wastes will generally permit contact-handling and avoid the need for shielding requirements. The selection of liquid waste treatment involves: Characterization of arising with the possibility of segregation; Discharge requirements for decontaminated liquors, both radioactive and non-radioactive; Available technologies and costs; Conditioning of the concentrates resulting from the treatment; Storage and disposal of the conditioned concentrates. The report will serve as a technical manual providing reference material and direct step-by-step know-how to staff in radioisotope user establishments and research centres in the developing Member States without nuclear power generation. Therefore, emphasis is limited to the simpler treatment facilities, which will be included with only the robust, well-established waste management processes carefully chosen as appropriate to developing countries. 20 refs, 12 figs, 7 tabs

  8. Disposition of nuclear waste using subcritical accelerator-driven systems

    International Nuclear Information System (INIS)

    Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

    1998-01-01

    Studies have shown that the repository long-term radiological risk is from the long-lived transuranics and the fission products Tc-99 and I-129, thermal loading concerns arise mainly form the short-lived fission products Sr-90 and Cs-137. In relation to the disposition of nuclear waste, ATW is expected to accomplish the following: (1) destroy over 99.9% of the actinides; (2) destroy over 99.9% of the Tc and I; (3) separate Sr and Cs (short half-life isotopes); (4) separate uranium; (5) produce electricity. In the ATW concept, spent fuel would be shipped to a ATW site where the plutonium, other transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their only pass through the facility. This approach contrasts with the present-day reprocessing practices in Europe and Japan, during which high purity plutonium is produced and used in the fabrication of fresh mixed-oxide fuel (MOX) that is shipped off-site for use in light water reactors

  9. Risk evaluation of the alternate-3A modification to the ATWS prevention/mitigation system in a BWR-4, MARK-II power plant

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Bari, R.A.; Karol, R.; Shiu, K.

    1983-01-01

    The authors present a risk evaluation of the ATWS Alternate 3A modification proposed by NRC staff in NUREG-0460 to the ATWS prevention/mitigation system in a BWR nuclear power plant. The evaluation is done relative to three risk indices: the frequency of core damage, the expected early fatalities, and the expected latent fatalities. The ATWS prevention tree includes: the mechanical subsystem of the reactor protection system, the electrical subsystem of the reactor protection system, the recirculation pump trip and the Alternate Rod Insertion System. The mitigation tree includes: standby liquid control system, opening of the relief valves, reclosing the relief valves, failure of coolant injection, inadvertent actuation of the automatic depressurization system, inadvertent operation of high-pressure injection system and containment heat removal

  10. Fabrication of the novel hydrogel based on waste corn stalk for removal of methylene blue dye from aqueous solution

    Science.gov (United States)

    Ma, Dongzhuo; Zhu, Baodong; Cao, Bo; Wang, Jian; Zhang, Jianwei

    2017-11-01

    The novel hydrogel based on waste corn stalk was synthetized by aqueous solution polymerization technique with functional monomers in the presence of organic montmorillonite (OMMT) under ultrasonic. In this study, batch adsorption experiments were carried out to research the effect of initial dye concentration, the dosage of hydrogel, stirring speed, contact time and temperature on the adsorption of methylene blue (MB) dye. The adsorption process was best described by the pseudo-second-order kinetic model, which confirmed that it should be a chemical process. Furthermore, we ascertained the rate controlling step by establishing the intraparticle diffusion model and the liquid film diffusion model. The adsorption and synthesis mechanisms were vividly depicted in our work as well. Structural and morphological characterizations by virtue of FTIR, FESEM, and Biomicroscope supported the relationship between the adsorption performance and material's microstructure. This research is a valuable contribution for the environmental protection, which not only converts waste corn stalks into functional materials, but improves the removal of organic dye from sewage water.

  11. The feasibility study I on the blanket fuel options for the ATW/HYPER

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok; Meyer, M.K; Hayes, S.L

    2001-01-01

    The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended.

  12. The feasibility study I on the blanket fuel options for the ATW/HYPER

    International Nuclear Information System (INIS)

    Hwang, Woan; Lee, Byoung Oon; Lee, Bong Sang; Park, Won Seok; Meyer, M.K; Hayes, S.L.

    2001-01-01

    The choice of a blanket fuel cycle technology and the fuel type for HYPER/ATW are important to develop an ADS with better economics, performance and safety. Even though several fuel types have been considered as an alternative of the blanket fuels for HYPER/ATW, the metal alloy and the dispersion fuels were selected as the candidate fuels for ADS, and the technical feasibilities for both fuels are evaluated in this report. General performance characteristics, fabrication abilities, technical aspects, safety aspects, economics, and non-proliferation aspects for each fuel type are reviewed and evaluated. And some technological problems are addressed in this report, focused on the development strategy, the roadmaps, and the flexibility to meet the missions and specific designs. This study has been performed at the first stage of conceptual design. Since it is under the lack of physical properties for each fuel material, no an attempt is made to select the best fuel option, but the more better fuel options are recommended

  13. Materials compatibility and corrosion issues for accelerator transmutation of waste

    International Nuclear Information System (INIS)

    Staudhammer, K.

    1992-08-01

    The need to understand the materials issues in an accelerator transmutation of waste (ATW) system is essential. This report focuses on the spallation container material, as this material is exposed to some of the most crucial environmental conditions of simultaneous radiation and corrosion in the system. The most severe design being considered is that of liquid lead. In previous investigations of lead compatibility with materials, the chemistry of the system was derived solely from the corrosion products; however, in an ATW system, the chemistry of the lead changes not only with the derived corrosion products of the material being tested but also with the buildup of the daughter production with time. Daughter production builds up and introduces elements that may have a great effect on the corrosion activity of the liquid lead. Consequently, data on liquid lead compatibility can be regarded only as a guide and must be reevaluated when particular daughter products are added. This report is intended to be a response to specific materials issues and concerns expressed by the ATW design working group and addresses the compatibility/corrosion concerns

  14. 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)

  15. Sandia solidification process: a broad range aqueous waste solidification method

    International Nuclear Information System (INIS)

    Lynch, R.W.; Dosch, R.G.; Kenna, B.T.; Johnstone, J.K.; Nowak, E.J.

    1976-01-01

    New ion-exchange materials of the hydrous oxide type were developed for solidifying aqueous radioactive wastes. These materials have the general formula M[M'/sub x/O/sub y/H/sub z/]/sub n/, where M is an exchangeable cation of charge +n and M' may be Ti; Nb; Zr, or Ta. Affinities for polyvalent cations were found to be very high and ion-exchange capacities large (e.g., 4.0--4.5 meq/g for NaTi 2 O 5 H depending on moisture content). The effectiveness of the exchangers for solidifying high-level waste resulting from reprocessing light-water reactor fuel was demonstrated in small-scale tests. Used in conjunction with anion exchange resin, these materials reduced test solution radioactivity from approximately 0.2 Ci/ml to as low as approximately 2 nCi/ml. The residual radioactivity was almost exclusively due to 106 Ru and total α-activity was only a few pCi/ml. Alternative methods of consolidating the solidified waste were evaluated using nonradioactive simulants. Best results were obtained by pressure-sintering which yielded essentially fully dense ceramics, e.g., titanate/titania ceramics with bulk density as high as 4.7 g/cm 3 , waste oxide content as high as 1.2 g/cm 3 , and leach resistance comparable to good borosilicate glass. Based on the above results, a baseline process for solidifying high-level waste was defined and approximate economic analyses indicated costs were not prohibitive. Additional tests have demonstrated that, if desired, operating conditions could be modified to allow recovery of radiocesium (and perhaps other isotopes) during solidification of the remaining constituents of high-level waste. Preliminary tests have also shown that these materials offer promise for treating tank-stored neutralized wastes

  16. Estimates of thermal fatigue due to beam interruptions for an ALMR-type ATW

    International Nuclear Information System (INIS)

    Dunn, F. E.; Wade, D. C.

    1999-01-01

    Thermal fatigue due to beam interruptions has been investigated in a sodium cooled ATW using the Advanced Liquid Metal mod B design as a basis for the subcritical source driven reactor. A k eff of 0.975 was used for the reactor. Temperature response in the primary coolant system was calculated, using the SASSYS- 1 code, for a drop in beam current from full power to zero in 1 microsecond.. Temperature differences were used to calculate thermal stresses. Fatigue curves from the American Society of Mechanical Engineers Boiler and Pressure Vessel Code were used to determine the number of cycles various components should be designed for, based on these thermal stresses

  17. Simulation of BWR stability following an ATWS with boron injection using TRAC-BF1 with one-dimensional kinetics

    International Nuclear Information System (INIS)

    Lider, S.; Maclan, R.; Baratta, A.J.; Mahaffy, J.; Robinson, G.E.

    2004-01-01

    The scenario following an ATWS is characterized by the necessity to reduce the power in the reactor as fast as possible. The only means to insert a significant amount of negative reactivity in a BWR during an ATWS are the natural reactor negative void coefficient, and the injection of highly enriched boron through the SLCS. The ATWS management strategy suggested by BWR owner's group contemplates an initial rapid decrease in power as a result of the recirculation pump trip. This is followed by lowering of vessel water level and the injection of borated water into the lower plenum. A recent paper of Dias, et al. reports that reducing core power and lowering water level causes a reduction in boron mixing efficiency and the net effect is a longer time to shut down and an increase in Suppression Pool (SP) temperature. In the present paper, a series of analyses are made to address this issue. The preliminary results for the water level positions at TAF, TAF+1.5 m (TAF+5') and TAF+3 m (TAF+10') support the similar findings of Dias, et al. (author)

  18. Investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a BWR ATWS

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Layman, W.; Hentzen, R.D.; Gose, G.C.

    1985-01-01

    A best-estimate analysis was performed to evaluate the technique of intentionally reducing reactor coolant inventory in order to reduce power during a BWR ATWS event. The ATWS was initiated by closure of the main steam isolation valves. The analysis was performed with the RETRAN-02 computer code utilizing the one-dimensional kinetics option. The one-dimensional cross sections were developed using the SIMULATE-E and SIMTRAN-E computer codes. The MSIV closure transient provides some of the more severe conditions following a postulated failure to scram. In this transient, the only mechanism for removing energy from the vessel is through the safety relief valve system which results in a heating up of the suppression pool fluid. Consequently, the reactor power must be reduced so that the suppression pool temperature limits are not exceeded. Under the proposed emergency procedure guidelines for the ATWS event, the reactor vessel water level will be lowered to reduce system power. This analysis evaluated the dynamic response of the system as the water level was lowered to the top of active fuel evaluation. Correlating the system power and flow patterns to water level was of particular interest in the analysis. Under natural circulating conditions, the system flows, core power, and pressure responses are extremely tightly coupled and the analysis results proved to be very sensitive to the modeling of downcomer, upper plenum, and core regions

  19. Physical property parameter set for modeling ICPP aqueous wastes with ASPEN electrolyte NRTL model

    International Nuclear Information System (INIS)

    Schindler, R.E.

    1996-09-01

    The aqueous waste evaporators at the Idaho Chemical Processing Plant (ICPP) are being modeled using ASPEN software. The ASPEN software calculates chemical and vapor-liquid equilibria with activity coefficients calculated using the electrolyte Non-Random Two Liquid (NRTL) model for local excess Gibbs free energies of interactions between ions and molecules in solution. The use of the electrolyte NRTL model requires the determination of empirical parameters for the excess Gibbs free energies of the interactions between species in solution. This report covers the development of a set parameters, from literature data, for the use of the electrolyte NRTL model with the major solutes in the ICPP aqueous wastes

  20. Safety assessment of the Indonesian multipurpose reactor RSG-GAS against ATWS and hypothetical accidents

    International Nuclear Information System (INIS)

    Hastowo, H.; Nabbi, R.; Prayoto; Ismuntoyo, R.P.H.

    2004-01-01

    Investigation on ATWS and hypothetical accidents for the Indonesian Multipurpose Reactor RSG-GAS have been undertaken by computer simulation technique. Two computer codes, namely RELAP5 and PARET-ANL, were used as the main tools. The RELAP5 was utilized to perform system analysis while the PARET-ANL code was used to perform the reactor core analysis in more detail. Two different models have been applied as a basis of the simulation: Typical Working Core model (IWC-model) consisting of four regions with different radial power factors; and the hot-channel model consisting of two regions with different radial power factors. Both RELAP5 ad PARET-ANL results showed that in the occurrence of ATWS, failure on fuel element or fuel plate was limited to the region with the most highest power factor. The results also indicated that no high pressure development occurs in that region, so that mechanical damage on the fuel element or other core components due to pressure shock did not happen.(author)

  1. 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.

  2. Actinide and Xenon reactivity effects in ATW high flux systems

    International Nuclear Information System (INIS)

    Woosley, M.; Olson, K.; Henderson, D.L.

    1995-01-01

    In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

  3. Actinide and xenon reactivity effects in ATW high flux systems

    International Nuclear Information System (INIS)

    Woosley, M.; Olson, K.; Henderson, D. L.; Sailor, W. C.

    1995-01-01

    In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

  4. Actinide and Xenon reactivity effects in ATW high flux systems

    Energy Technology Data Exchange (ETDEWEB)

    Woosley, M. [Univ. of Virginia, Charlottesville, VA (United States); Olson, K.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)] [and others

    1995-10-01

    In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides.

  5. Transmutation and inventory analysis in an ATW molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Sisolak, J.E.; Truebenbach, M.T.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)

    1995-10-01

    As an extension of earlier work to determine the equilibrium state of an ATW molten salt, power producing, reactor/transmuter, the WAIT code provides a time dependent view of material inventories and reactor parameters. By considering several cases, the authors infer that devices of this type do not reach equilibrium for dozens of years, and that equilibrium design calculations are inapplicable over most of the reactor life. Fissile inventory and k{sub eff} both vary by factors of 1.5 or more between reactor startup and ultimate convergence to equilibrium.

  6. Parametric study on thermal-hydraulic response following as ATWS event

    International Nuclear Information System (INIS)

    Suh, Jeong Kwan; Bang, Young Seok; Kim, Hho Jung

    2000-01-01

    A series of sensitivity calculations for the LOFT L9-3 experiment were performed using RELAP5/MOD3 code to assess parametric effects on thermal-hydraulic response in the event of Anticipated Transient Without Scram (ATWS). The base case calculation was made by the condition which gave a good agreement for the pressure of the reactor coolant system (RCS) with the experimental data. Four parameters of PORV/spray energy loss coefficient, steam generator nodalization and moderator density coefficient (MDC) were selected during the input preparation and investigated by calculating the total discharged energy through relief valves. The energy loss coefficient of the pressurizer spray valve has a significant effect on the behavior of the RCS pressure and the change of the MDC curve within 15 % at the negative region decreased the difference of the coolant temperature between the experiment and the calculation within a range of measurement uncertainty. The finer steam generator nodalization increased the primary to secondary heat transfer rate

  7. Research program on development of advanced treatment technology for americium-containing aqueous waste in NUCEF

    Energy Technology Data Exchange (ETDEWEB)

    Mineo, Hideaki; Matsumura, Tatsuro; Tsubata, Yasuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-10-01

    A research program was prepared on the development of an advanced treatment process for the americium-containing concentrated aqueous waste in NUCEF, than allows americium recovery for the reuse and the reduction of TRU waste generation. A preliminary analysis was conducted on the separation requirements based on the components estimated for the waste. An R and D strategy was proposed from the view to reduce TRU waste generated in the processing that the highest priority is given on the control of TRU leakage such as americium into the effluent stream after americium recovery and the minimization of salt used in the separation over the decontamination of impurities from americium. The extraction chromatographic method was selected as a candidate technology for americium separation under the principle to use reagents that are functional in acidic conditions such as bidentate extractants of DHEDECMP, CMPO or diamides, considering the larger flexibilities in process modification and possible multi-component separation with compact equipment and the past achievements on the recovery of kg quantities of americium. Major R and D items extracted are screening and evaluation of extractants for americium and plutonium, optimization of separation conditions, selection of denitration method, equipment developments and development of solidification methods of discarded americium after reuse and of various kinds of separation residues. In order to cope these items, four steps of R and D program were proposed, i.e., fundamental experiment in beaker-scale on screening and evaluation of extractants, flowsheet study in bench-scale using simulated and small amount of americium aqueous waste solution to evaluate candidate process, americium recovery test in iron-shielded cell to be installed in NUCEF. It is objected to make recovery of 100g orders of americium used for research on fundamental TRU fuel properties. (J.P.N.)

  8. Evaluation of very low frequencies of ATWS and PLOHS in a loop-type FBR plant by making use of inherently safe features

    International Nuclear Information System (INIS)

    Sakata, K.; Koyama, K.; Aoi, S.; Simonelli, R.B.; Wallace, I.T.

    1987-01-01

    Frequencies of ATWS (Anticipated Transient Without Scram) and PLOHS (Protected Loss of Heat Sink) for a large loop-type FBR plant were evaluated by applying PSA methodologies. The frequencies were found to be so low that ATWS and PLOHS could be excluded from candidates of the design basis events. Furthermore, the inherently safe features introduced to the system design were verified to be very effective for reduction of the Probability of CCF (Common Cause Failure), which deteriorates reliability of both the reactor shutdown and the decay heat removal systems. (orig.)

  9. Solid waste from leather industry as adsorbent of organic dyes in aqueous-medium

    International Nuclear Information System (INIS)

    Oliveira, Luiz C.A.; Goncalves, Maraisa; Oliveira, Diana Q.L.; Guerreiro, Mario C.; Guilherme, Luiz R.G.; Dallago, Rogerio M.

    2007-01-01

    The industrial tanning of leather usually produces considerable amounts of chromium-containing solid waste and liquid effluents and raises many concerns on its environmental effect as well as on escalating landfill costs. Actually, these shortcomings are becoming increasingly a limiting factor to this industrial activity that claims for alternative methods of residue disposals. In this work, it is proposed a novel alternative destination of the solid waste, based on the removal of organic contaminants from the out coming aqueous-residue. The adsorption isotherm pattern for the wet blue leather from the Aurea tanning industry in Erechim-RS (Brazil) showed that these materials present high activity on adsorbing the reactive red textile dye as well as other compounds. The adsorbent materials were characterized by IR spectroscopy and SEM and tested for the dye adsorption (reactive textile and methylene blue dyes). The concentrations of dyes were measured by UV-vis spectrophotometry and the chromium extraction from leather waste was realized by basic hydrolysis and determined by atomic absorption. As a low cost abundant adsorbent material with high adsorption ability on removing dye methylene blue (80 mg g -1 ) and textile dye reactive red (163 mg g -1 ), the leather waste is revealed to be a interesting alternative relatively to more costly adsorbent materials

  10. Solid waste from leather industry as adsorbent of organic dyes in aqueous-medium

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Luiz C.A. [Universidade Federal de Lavras, Depto. de Quimica, Caixa Postal 37, CEP 37200.000, Lavras-MG (Brazil)]. E-mail: luizoliveira@ufla.br; Goncalves, Maraisa [Universidade Federal de Lavras, Depto. de Quimica, Caixa Postal 37, CEP 37200.000, Lavras-MG (Brazil); Oliveira, Diana Q.L. [Universidade Federal de Lavras, Depto. de Quimica, Caixa Postal 37, CEP 37200.000, Lavras-MG (Brazil); Guerreiro, Mario C. [Universidade Federal de Lavras, Depto. de Quimica, Caixa Postal 37, CEP 37200.000, Lavras-MG (Brazil); Guilherme, Luiz R.G. [Universidade Federal de Lavras, Depto. de Ciencia do solo, CEP 37200.000, Lavras-MG (Brazil); Dallago, Rogerio M. [URI-Campus Erechim, Av. 7 Setembro 1621, Centro, CEP 99700-000, Depto de Quimica, Erechim-RS (Brazil)

    2007-03-06

    The industrial tanning of leather usually produces considerable amounts of chromium-containing solid waste and liquid effluents and raises many concerns on its environmental effect as well as on escalating landfill costs. Actually, these shortcomings are becoming increasingly a limiting factor to this industrial activity that claims for alternative methods of residue disposals. In this work, it is proposed a novel alternative destination of the solid waste, based on the removal of organic contaminants from the out coming aqueous-residue. The adsorption isotherm pattern for the wet blue leather from the Aurea tanning industry in Erechim-RS (Brazil) showed that these materials present high activity on adsorbing the reactive red textile dye as well as other compounds. The adsorbent materials were characterized by IR spectroscopy and SEM and tested for the dye adsorption (reactive textile and methylene blue dyes). The concentrations of dyes were measured by UV-vis spectrophotometry and the chromium extraction from leather waste was realized by basic hydrolysis and determined by atomic absorption. As a low cost abundant adsorbent material with high adsorption ability on removing dye methylene blue (80 mg g{sup -1}) and textile dye reactive red (163 mg g{sup -1}), the leather waste is revealed to be a interesting alternative relatively to more costly adsorbent materials.

  11. The Los Alamos accelerator driven transmutation of nuclear waste (ATW) concept development of the ATW target/blanket system

    International Nuclear Information System (INIS)

    Venneri, F.; Williamson, M.A.; Ning, L.

    1997-01-01

    The studies carried out in the frame of the Accelerator Driven Transmutation Technology (ADTT) program developed at Los Alamos in order to solve the nuclear waste problem and to build a new generation of safer and non-proliferant nuclear power plants, are presented

  12. Study on Fabrication of Ni-5 at.%W Tapes for Coated Conductors from Cylinder Ingots

    DEFF Research Database (Denmark)

    Ma, L.; Suo, H. L.; Yue, Zhao

    2015-01-01

    Ni-5 at.%W (Ni5W) tapes with a strong cube texture were fabricated using the RABiTS technique and by starting from cylindrical shaped ingots. In contrast to a conventional cuboid-shaped ingot, a cylinder shaped ingot has no anisotropy along the axial direction and the resulting tape will therefore...

  13. Solidification of aqueous radioactive waste using insoluble compounds of magnesium oxide

    International Nuclear Information System (INIS)

    Carlson, J.E.

    1986-01-01

    A process is described for the treatment of radioactive waste which comprises: (a) first adding, under continuous agitation, a sufficient amount of a powdered magnesium oxide or magnesium hydroxide to an aqueous radioactive waste solution containing boric acid, the temperature of the water solution being 55-95 degrees C. to produce a magnesium borate derivative; (b) adding cement, under continuous agitation, to the magnesium borate derivative; and (c) then adding, under continuous agitation, after the cement has been dispersed, a sufficient amount of a compound selected from the group consisting of calcium oxide and calcium hydroxide to (b) to produce a gel matrix structure

  14. Solvent extraction of radionuclides from aqueous tank waste

    International Nuclear Information System (INIS)

    Moyer, B.A.; Bonnesen, P.V.; Sachleben, R.A.

    1997-01-01

    This task aims toward the development of efficient solvent-extraction processes for the removal of the fission products 99 Tc, 90 Sr, and 137 Cs from alkaline tank wastes. Processes already developed or proposed entail direct treatment of the waste solution with the solvent and subsequent stripping of the extracted contaminants from the solvent into a dilute aqueous solution. Working processes to remove Tc(and SR) separately and Cs separately have been developed; the feasibility of a combined process is under investigation. Since Tc, Sr, and Cs will be vitrified together in the high-level fraction, however, a process that could separate Tc, Sr, and Cs simultaneously, as opposed to sequentially, potentially offers the greatest impact. A figure presents a simplified diagram of a proposed solvent-extraction cycle followed by three possible treatments for the stripping solution. Some degree of recycle of the stripping solution (option a) is expected. Simple evaporation (option c) is possible prior to vitrification; this offers the greatest possible volume reduction with simple operation and no consumption of chemicals, but it is energy intensive. However, if the contaminants are concentrated (option b) by fixed-bed technology, the energy penalty of evaporation can be avoided and vitrification facilitated without any additional secondary waste being produced

  15. Considerations affecting deep-well disposal of tritium-bearing low-level aqueous waste from nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Trevorrow, L.E.; Warner, D.L.; Steindler, M.J.

    1977-03-01

    Present concepts of disposal of low-level aqueous wastes (LLAW) that contain much of the fission-product tritium from light water reactors involve dispersal to the atmosphere or to surface streams at fuel reprocessing plants. These concepts have been challenged in recent years. Deep-well injection of low-level aqueous wastes, an alternative to biospheric dispersal, is the subject of this presentation. Many factors must be considered in assessing its feasibility, including technology, costs, environmental impact, legal and regulatory constraints, and siting. Examination of these factors indicates that the technology of deep-well injection, extensively developed for other industrial wastes, would require little innovation before application to low-level aqueous wastes. Costs would be low, of the order of magnitude of 10 -4 mill/kWh. The environmental impact of normal deep-well disposal would be small, compared with dispersal to the atmosphere or to surface streams; abnormal operation would not be expected to produce catastrophic results. Geologically suitable sites are abundant in the U.S., but a well would best be co-located with the fuel-reprocessing plant where the LLAW is produced. Legal and regulatory constraints now being developed will be the most important determinants of the feasibility of applying the method

  16. Advanced methods for the treatment of organic aqueous wastes: wet air oxidation and wet peroxide oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Debellefontaine, Hubert; Chakchouk, Mehrez; Foussard, Jean Noel [Institut National des Sciences Appliquees (INSA), 31 - Toulouse (France). Dept. de Genie des Procedes Industriels; Tissot, Daniel; Striolo, Phillipe [IDE Environnement S.A., Toulouse (France)

    1994-12-31

    There is a growing concern about the problems of wastes elimination. Various oxidation techniques are suited for elimination of organic aqueous wastes, however, because of the environmental drawbacks of incineration, liquid phase oxidation should be preferred. `Wet Air Oxidation` and `Wet Peroxide Oxidation`are alternative processes which are discussed in this paper. 17 refs., 13 figs., 4 tabs.

  17. Advanced methods for the treatment of organic aqueous wastes: wet air oxidation and wet peroxide oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Debellefontaine, Hubert; Chakchouk, Mehrez; Foussard, Jean Noel [Institut National des Sciences Appliquees (INSA), 31 - Toulouse (France). Dept. de Genie des Procedes Industriels; Tissot, Daniel; Striolo, Phillipe [IDE Environnement S.A., Toulouse (France)

    1993-12-31

    There is a growing concern about the problems of wastes elimination. Various oxidation techniques are suited for elimination of organic aqueous wastes, however, because of the environmental drawbacks of incineration, liquid phase oxidation should be preferred. `Wet Air Oxidation` and `Wet Peroxide Oxidation`are alternative processes which are discussed in this paper. 17 refs., 13 figs., 4 tabs.

  18. 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.

  19. Environmental impacts of radiological consequences during the anticipated transients without scram (ATWS) events in nuclear power reactors

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    2011-01-01

    Anticipated transients without scram (ATWS), is one of the (worst case) accidents could happen if the system that provides a highly reliable means of shutting down the reactor (scram system )fails to work during a reactor event (anticipated transient).It has two general characteristics: (1) Initiation by a transient anticipated to occur one or more times in the life of reactor and ,(2) Assumed to proceed without scram.The types of events considered are those used for designing the plant .The evaluation of the radiological consequences during the assessment of the nuclear events,especially ATWS in nuclear power reactors, is very essential for environmental studies and public safety. In this paper, the root cases for nuclear events and dose calculation are presented. Scenario of accident sequences together with radiological impacts is illustrated for loss of coolant accident (LOCA) for a typical pressurized water reactor nuclear power plant. Recommendations for mitigating or preventing the release of radiation and high radioactive materials to environment are presented.

  20. Separations technology development to support accelerator-driven transmutation concepts

    International Nuclear Information System (INIS)

    Venneri, F.; Arthur, E.; Bowman, C.

    1996-01-01

    This is the final report of a one-year Laboratory-Directed Research and Development (LDRD) Project at the Los Alamos National Laboratory (LANL). This project investigated separations technology development needed for accelerator-driven transmutation technology (ADTT) concepts, particularly those associated with plutonium disposition (accelerator-based conversion, ABC) and high-level radioactive waste transmutation (accelerator transmutation of waste, ATW). Specific focus areas included separations needed for preparation of feeds to ABC and ATW systems, for example from spent reactor fuel sources, those required within an ABC/ATW system for material recycle and recovery of key long-lived radionuclides for further transmutation, and those required for reuse and cleanup of molten fluoride salts. The project also featured beginning experimental development in areas associated with a small molten-salt test loop and exploratory centrifugal separations systems

  1. Technologies for destruction of long-lived radionuclides in high-level nuclear waste: Overview and requirements

    International Nuclear Information System (INIS)

    Arthur, E.D.

    1993-01-01

    This paper, and this topical session on Nuclear Waste Minimization, Management and Remediation, focuses on two nuclear systems, and their associated technologies, that have the potential to address concerns surrounding long-lived radionuclides in high-level waste. Both systems offer technology applicable to HLW from present light-water reactors (LWR). Additionally these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes, long-term fuel supplies, and enhanced safety. The first system is the Integral Fast Reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while simultaneously burning long-lived actinide waste. IFR applications include burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, Accelerator Transmutation of Waste (ATW), is based upon an accelerator-induced intense source of thermal neutrons and is aimed at destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

  2. Green synthesis of graphene from recycled PET bottle wastes for use in the adsorption of dyes in aqueous solution.

    Science.gov (United States)

    El Essawy, Noha A; Ali, Safa M; Farag, Hassan A; Konsowa, Abdelaziz H; Elnouby, Mohamed; Hamad, Hesham A

    2017-11-01

    Polyethyleneterephthalate (PET) is an important component of post-consumer plastic waste. This study focuses on the potential of utilizing "waste-treats-waste" by synthesis of graphene using PET bottle waste as a source material. The synthesized graphene is characterized by SEM, TEM, BET, Raman, TGA, and FT-IR. The adsorption of methylene blue (MB) and acid blue 25 (AB25) by graphene is studied and parameters such as contact time, adsorbent dosage were optimized. The Response Surface Methodology (RSM) is applied to investigate the effect of three variables (dye concentration, time and temperature) and their interaction on the removal efficiency. Adsorption kinetics and isotherm are followed a pseudo-second-order model and Langmuir and Freundlich isotherm models, respectively. Thermodynamic parameters demonstrated that adsorption of dye is spontaneous and endothermic in nature. The plastic waste can be used after transformation into valuable carbon-based nanomaterials for use in the adsorption of organic contaminants from aqueous solution. Copyright © 2017 Elsevier Inc. All rights reserved.

  3. Recovery of uranium and plutonium from Redox off-standard aqueous waste streams

    Energy Technology Data Exchange (ETDEWEB)

    Holm, C.H.; Matheson, A.R.

    1949-12-31

    In the operation of countercurrent extraction columns as in the Redox process, it is possible, and probable, that from unexpected behaviour of a column, operator error, colloid formation, etc., there will result from time to time excessive losses of uranium and plutonium in the overall process. These losses will naturally accumulate in the waste streams, particularly in the aqueous waste streams. If the loss is excessively high, and such lost material can be recovered by some additional method, then if economical and within reason, the recovered materials ran be returned to a ISF column for further processing. The objective of this work has been to develop such a method to recover uranium and plutonium from such off-standard waste streams in a form whereby the uranium send plutonium can be returned to the process line and subsequently purified and separated.

  4. Aqueous nitrate waste treatment: Technology comparison, cost/benefit, and market analysis

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-01

    The purpose of this analysis is to provide information necessary for the Department of Energy (DOE) to evaluate the practical utility of the Nitrate to Ammonia and Ceramic or Glass (NAC/NAG/NAX) process, which is under development in the Oak Ridge National Laboratory. The NAC/NACx/NAX process can convert aqueous radioactive nitrate-laden waste to a glass, ceramic, or grout solid waste form. The tasks include, but are not limited to, the following: Identify current commercial technologies to meet hazardous and radiological waste disposal requirements. The technologies may be thermal or non-thermal but must be all inclusive (i.e., must convert a radionuclide-containing nitrate waste with a pH around 12 to a stable form that can be disposed at permitted facilities); evaluate and compare DOE-sponsored vitrification, grouting, and minimum additive waste stabilization projects for life-cycle costs; compare the technologies above with respect to material costs, capital equipment costs, operating costs, and operating efficiencies. For the NAC/NAG/NAX process, assume aluminum reactant is government furnished and ammonia gas may be marketed; compare the identified technologies with respect to frequency of use within DOE for environmental management applications with appropriate rationale for use; Assess the potential size of the DOE market for the NAC/NAG/NAX process; assess and off-gas issues; and compare with international technologies, including life-cycle estimates.

  5. Aqueous nitrate waste treatment: Technology comparison, cost/benefit, and market analysis

    International Nuclear Information System (INIS)

    1994-01-01

    The purpose of this analysis is to provide information necessary for the Department of Energy (DOE) to evaluate the practical utility of the Nitrate to Ammonia and Ceramic or Glass (NAC/NAG/NAX) process, which is under development in the Oak Ridge National Laboratory. The NAC/NACx/NAX process can convert aqueous radioactive nitrate-laden waste to a glass, ceramic, or grout solid waste form. The tasks include, but are not limited to, the following: Identify current commercial technologies to meet hazardous and radiological waste disposal requirements. The technologies may be thermal or non-thermal but must be all inclusive (i.e., must convert a radionuclide-containing nitrate waste with a pH around 12 to a stable form that can be disposed at permitted facilities); evaluate and compare DOE-sponsored vitrification, grouting, and minimum additive waste stabilization projects for life-cycle costs; compare the technologies above with respect to material costs, capital equipment costs, operating costs, and operating efficiencies. For the NAC/NAG/NAX process, assume aluminum reactant is government furnished and ammonia gas may be marketed; compare the identified technologies with respect to frequency of use within DOE for environmental management applications with appropriate rationale for use; Assess the potential size of the DOE market for the NAC/NAG/NAX process; assess and off-gas issues; and compare with international technologies, including life-cycle estimates

  6. Species removal from aqueous radioactive waste by deep-bed filtration.

    Science.gov (United States)

    Dobre, Tănase; Zicman, Laura Ruxandra; Pârvulescu, Oana Cristina; Neacşu, Elena; Ciobanu, Cătălin; Drăgolici, Felicia Nicoleta

    2018-05-26

    Performances of aqueous suspension treatment by deep-bed sand filtration were experimentally studied and simulated. A semiempirical deterministic model and a stochastic model were used to predict the removal of clay particles (20 μm) from diluted suspensions. Model parameters, which were fitted based on experimental data, were linked by multiple linear correlations to the process factors, i.e., sand grain size (0.5 and 0.8 mm), bed depth (0.2 and 0.4 m), clay concentration in the feed suspension (1 and 2 kg p /m 3 ), suspension superficial velocity (0.015 and 0.020 m/s), and operating temperature (25 and 45 °C). These relationships were used to predict the bed radioactivity determined by the deposition of radioactive suspended particles (>50 nm) from low and medium level aqueous radioactive waste. A deterministic model based on mass balance, kinetic, and interface equilibrium equations was developed to predict the multicomponent sorption of 60 Co, 137 Cs, 241 Am, and 3 H radionuclides (0.1-0.3 nm). A removal of 98.7% of radioactive particles was attained by filtering a radioactive wastewater volume of 10 m 3 (0.5 mm sand grain size, 0.3 m bed depth, 0.223 kg p /m 3 suspended solid concentration in the feed suspension, 0.003 m/s suspension superficial velocity, and 25 °C operating temperature). Predicted results revealed that the bed radioactivity determined by the sorption of radionuclides (0.01 kBq/kg b ) was significantly lower than the bed radioactivities caused by the deposition of radioactive particles (0.5-1.8 kBq/kg b ). Copyright © 2018 Elsevier Ltd. All rights reserved.

  7. Method for removing and decolorizing aqueous waste effluents containing dissolved or dispersed organic matter

    International Nuclear Information System (INIS)

    Case, F.N.; Ketchen, E.E.

    1975-01-01

    A method is provided for treating organic waste material dissolved or dispersed in an aqueous effluent, which comprises contacting the effluent with an inert particulate carbonaceous sorbent at an oxygen pressure up to 2000 psi, irradiating the resultant mixture with high energy radiation until a decolorized liquid is produced, and then separating the decolorized liquid

  8. 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.

  9. Integrated treatment process of hazardous and mixed wastes

    International Nuclear Information System (INIS)

    Shibuya, M.; Suzuki, K.; Fujimura, Y.; Nakashima, T.; Moriya, Y.

    1993-01-01

    An integrated waste treatment system was studied based on technologies developed for the treatment of liquid radioactive, organic, and aqueous wastes containing hazardous materials and soils contaminated with heavy metals. The system consists of submerged incineration, metal ion fixing and stabilization, and soil washing treatments. Introduction of this system allows for the simultaneous processing of toxic waste and contaminated soils. Hazardous organic wastes can be decomposed into harmless gases, and aqueous wastes can be converted into a dischargeable effluent. The contaminated soil is backfilled after the removal of toxic materials. Experimental data show that the integration system is practical for complicated toxic wastes

  10. ATWS analysis for Browns Ferry Nuclear Plant Unit 1

    International Nuclear Information System (INIS)

    Dallman, R.J.; Jouse, W.C.

    1985-01-01

    Analyses of postulated Anticipated Transients Without Scram (ATWS) were performed at the Idaho National Engineering Laboratory (INEL). The Browns Ferry Nuclear Plant Unit 1 (BFNP1) was selected as the subject of this work because of the cooperation of the Tennessee Valley Authority (TVA). The work is part of the Severe Accident Sequence Analysis (SASA) Program of the US Nuclear Regulatory Commission (NRC). A Main Steamline Isolation Valve (MSIV) closure served as the transient initiator for these analyses, which proceeded a complete failure to scram. Results from the analyses indicate that operator mitigative actions are required to prevent overpressurization of the primary containment. Uncertainties remain concerning the effectiveness of key mitigative actions. The effectiveness of level control as a power reduction procedure is limited. Power level resulting from level control only reduce the Pressure Suppression Pool (PSP) heatup rate from 6 to 4 0 F/min

  11. Aspiration tests in aqueous foam using a breathing simulator

    Energy Technology Data Exchange (ETDEWEB)

    Archuleta, M.M.

    1995-12-01

    Non-toxic aqueous foams are being developed by Sandia National Laboratories (SNL) for the National Institute of Justice (NIJ) for use in crowd control, cell extractions, and group disturbances in the criminal justice prison systems. The potential for aspiration of aqueous foam during its use and the resulting adverse effects associated with complete immersion in aqueous foam is of major concern to the NIJ when examining the effectiveness and safety of using this technology as a Less-Than-Lethal weapon. This preliminary study was designed to evaluate the maximum quantity of foam that might be aspirated by an individual following total immersion in an SNL-developed aqueous foam. A.T.W. Reed Breathing simulator equipped with a 622 Silverman cam was used to simulate the aspiration of an ammonium laureth sulfate aqueous foam developed by SNL and generated at expansion ratios in the range of 500:1 to 1000:1. Although the natural instinct of an individual immersed in foam is to cover their nose and mouth with a hand or cloth, thus breaking the bubbles and decreasing the potential for aspiration, this study was performed to examine a worst case scenario where mouth breathing only was examined, and no attempt was made to block foam entry into the breathing port. Two breathing rates were examined: one that simulated a sedentary individual with a mean breathing rate of 6.27 breaths/minute, and one that simulated an agitated or heavily breathing individual with a mean breathing rate of 23.7 breaths/minute. The results of this study indicate that, if breathing in aqueous foam without movement, an air pocket forms around the nose and mouth within one minute of immersion.

  12. 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.

  13. ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

    International Nuclear Information System (INIS)

    Parzer, I.; Kljenak, I.

    2005-01-01

    The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

  14. Evaluation of Proposed New LLW Disposal Activity: Disposal of Aqueous PUREX Waste Stream in the Saltstone Disposal Facility

    International Nuclear Information System (INIS)

    Cook, J.R.

    2003-01-01

    The Aqueous PUREX waste stream from Tanks 33 and 35, which have been blended in Tank 34, has been identified for possible processing through the Saltstone Processing Facility for disposal in the Saltstone Disposal Facility

  15. Magnetic Adsorption Method for the Treatment of Metal Contaminated Aqueous Waste

    International Nuclear Information System (INIS)

    Cotten, G.B.; Eldredge, H.B.; Navratil, J.D.

    1999-01-01

    There have been many recent developments in separation methods used for treating radioactive and non-radioactive metal bearing liquid wastes. These methods have included adsorption, ion exchange, solvent extraction and other chemical and physical techniques. To date very few, if any, of these processes can provide a low cost and environmentally benign solution. Recent research into the use of magnetite for wastewater treatment indicates the potential for magnetite both cost and environment drivers. A brief review of recent work in using magnetite as a sorbent is presented as well as recent work performed in our laboratory using supported magnetite in the presence of an external magnetic field. The application to groundwater and other aqueous waste streams is discussed. Recent research has focused on supporting magnetite in an economical (as compared to the magnetic polymine-epichlorohydrine resin) and inert (non-reactive, chemically or otherwise) environment that promotes both adsorption and satisfactory flow characteristics

  16. Physical inventory by use of modeling for the tritium aqueous waste recovery system

    International Nuclear Information System (INIS)

    Sienkiewicz, C.J.; Lentz, J.E.; Wiggins, D.V.

    1988-01-01

    Physical inventory requirements for the Tritium Aqueous Waste Recovery System (TAWRS) presented constraints that required unique solutions. Available analytical techniques for which sound measurement control practices existed could not be readily adapted to the system without significant modifications and expense. Based on the assumption that would accurately estimate total system inventory given a few key measurements, a model was developed for TAWRS. Tritium concentrations in two streams, the tritiated feed stream to the process and the tritiated hydrogen stream generated by the electrolysis cells, provided the key values to the model. The proposed mathematical model relates the tritium concentration throughout the system to the tritium concentration in these two streams. Testing of the system using low-level tritiated feed water was conducted to characterize tritium distribution in the system and to relate key values to total inventory. 4 refs., 2 figs.,

  17. Reliability and availability considerations in the RF systems of ATW-class accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Tallerico, P.J.; Lynch, M.T.; Lawrence, G. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    In an RF-driven, ion accelerator for waste transmutation or nuclear material production, the overall availability is perhaps the most important specification. The synchronism requirements in an ion accelerator, as contrasted to an electron accelerator, cause a failure of an RF source to have a greater consequence. These large machines also are major capital investments, so the availability determines the return on this capital. RF system design methods to insure a high availability without paying a serious cost penalty are the subject of this paper. The overall availability goal in the present designs is 75% for the entire ATW complex, and from 25 to 35% of the unavailability is allocated to the RF system, since it is one of the most complicated subsystems in the complex. The allowed down time for the RF system (including the linac and all other systems) is then only 7 to 9% of the operating time per year, or as little as 613 hours per year, for continuous operation. Since large accelerators consume large amounts of electrical power, excellent efficiency is also required with the excellent availability. The availability also influences the sizes of the RF components; smaller components may fail and yet the accelerator may still meet all specifications. Larger components are also attractive, since the cost of an RF system usually increases as the square root of the number of RF systems utilized. In some cases, there is a reliability penalty that accompanies the cost savings from using larger components. The authors discuss these factors, and present an availability model that allows one to examine these trade offs, and make rational choices in the RF and accelerator system designs.

  18. Joyo ATWS test analysis by Mimir-N2

    International Nuclear Information System (INIS)

    Yoshida, Akihiro

    2001-03-01

    The study on the passive safety test by using the Experimental Fast Reactor Joyo was performed to demonstrate the inherent safety of fast breeder reactors. An analysis code: Mimir-N2, which has been developed to analyze Joyo plant kinetics, was selected as a standard code for this study. In order to increase the reliability of the calculation, Mimir-N2 code was adjusted based on the data obtained through several plant characteristics tests carried out in Joyo. Throughout an operational data obtained in Joyo, it is supposed that the burn-up dependency observed on the power reactivity coefficient might be coming from the reactivity shift caused by a depression of a thermal expansion of fuel pellet. Based on the relationship between the measured power reactivity coefficient and the core averaged burn-up, the burn-up dependency mentioned above was estimated and introduced to Mimir-N2. As a result, calculated core and plant dynamics during the step reactivity response test, such as the response of the power range neutron monitor and the coolant temperature at the core inlet/outlet, corresponded with the measured value. Especially, it was confirmed that Mimir-N2 can simulate the perturbation caused by the thermal expansion of the core support plate. In addition, Mimir-N2 was modified to be enable to take into account for the core bowing reactivity, which is calculated by the core bowing reactivity analysis system developed for Joyo. The preliminary analysis of the plant dynamics during the ATWS events in MK-III core were carried out by using modified Mimir-N2. As a result, it was confirmed that the core bowing reactivity should not be neglected because it sometimes shows positive feedback characteristics. (author)

  19. Evolution of microstructure, texture and topography during additional annealing of cube-textured Ni–5at.%W substrate for coated conductors

    DEFF Research Database (Denmark)

    Wulff, Anders Christian; Mishin, Oleg; Grivel, Jean-Claude

    2012-01-01

    Microstructure, texture and topography have been studied in a recrystallized Ni–5at.%W substrate before and after additional annealing at 1025C for 1 h. The initial recrystallized material contained a strong cube texture and a high fraction of low angle grain boundaries. R3 boundaries were also f...

  20. Chemistry of materials relevant to aqueous reprocessing and waste management

    International Nuclear Information System (INIS)

    Srinivasan, T.G.

    2012-01-01

    Nuclear energy option will be an inevitable one with the fossil fuels depleting fast and present coal and oil based thermal power generation resulting in unwanted green house gas emission. The utilisation of the fissile resources will be more effective with closed fuel cycle option wherein the spent reactor fuel is reprocessed and the unused uranium and plutonium formed during the reactor operation is recovered and re-used. Of the aqueous and non-aqueous routes available to reprocess the spent nuclear fuels, aqueous reprocessing method of recovering the valuable uranium and plutonium by the PUREX process is in vogue for the past six decades. The process involves chopping the fuel into small lengths, leaching uranium and plutonium with concentrated nitric acid under reflux, conditioning the dissolver solution with respect to acidity and valency of U and Pu, solvent extraction with 30%TBP/n-DD to selectively extract U(VI) and Pu(IV) leaving most of the fission products into the raffinate, partitioning plutonium from uranium and reconversion of U and Pu into oxide forms after further purification. Many reagents are used to achieve near quantitative recovery of both uranium and plutonium (>99.9%) and with high decontamination factors (>10 7 ) from highly radioactive fission products. Nevertheless, the chemistry of several reagents used and the chemical processes that take place during the entire course of reprocessing and waste management operations are yet to be fully understood and gives a lot of scope for further improvements. Some examples where research requires concerted efforts are, 1) development of new extractants conforming to CHON principle, with acceptable physical properties, high stability, selectivity and resistance to third phase formation, 2) new partitioning reagents and processes which offer good efficiency and kinetics for uranium/plutonium reduction, 3) understanding the chemistry of troublesome fission products such as Tc, Ru and Zr, 4

  1. Development of cube textured Ni-5 at.%W alloy substrates for coated conductor application using a melting process

    International Nuclear Information System (INIS)

    Zhao Yue; Suo Hongli; Liu Min; Liu Danmin; Zhang Yingxiao; Zhou Meiling

    2006-01-01

    Biaxially textured Ni-5 at.%W substrates have been prepared by cold rolling, followed by three different annealing routes. In this paper, the processes of melting Ni and W metals, flat rolling, various annealing methods are described in detail. The Ni-5 at.%W tapes annealed under either high vacuum or flowing Ar (7% H 2 ) gas were characterized by X-ray pole figures, ODF, EBSD as well as AFM analysis. The texture analysis indicated that as fabricated tapes have a sharp cube texture formed after annealing at a wide temperature range of 800-1100 o C. The high quality of cube orientation on tapes was obtained after a two-step annealing (TSA), where the percentage of the cube texture component was as high as 93.5% within a misorientation angle smaller than 8 o from EBSD analysis. Furthermore, it was also observed that the number of twin boundaries in this tape decreased with respect to that of tapes annealed both in vacuum and one-step gas annealing. From AFM on 1 μm 2 areas, it was concluded that the roughness (RMS) on the tape surface reached 0.98 nm

  2. 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.

  3. Treatability study of absorbent polymer waste form for mixed waste treatment

    International Nuclear Information System (INIS)

    Herrmann, S. D.; Lehto, M. A.; Stewart, N. A.; Croft, A. D.; Kern, P. W.

    2000-01-01

    A treatability study was performed to develop and characterize an absorbent polymer waste form for application to low level (LLW) and mixed low level (MLLW) aqueous wastes at Argonne National Laboratory-West (ANL-W). In this study absorbent polymers proved effective at immobilizing aqueous liquid wastes in order to meet Land Disposal Restrictions for subsurface waste disposal. Treatment of aqueous waste with absorbent polymers provides an alternative to liquid waste solidification via high-shear mixing with clays and cements. Significant advantages of absorbent polymer use over clays and cements include ease of operations and waste volume minimization. Absorbent polymers do not require high-shear mixing as do clays and cements. Granulated absorbent polymer is poured into aqueous solutions and forms a gel which passes the paint filter test as a non-liquid. Pouring versus mixing of a solidification agent not only eliminates the need for a mixing station, but also lessens exposure to personnel and the potential for spread of contamination from treatment of radioactive wastes. Waste minimization is achieved as significantly less mass addition and volume increase is required of and results from absorbent polymer use than that of clays and cements. Operational ease and waste minimization translate into overall cost savings for LLW and MLLW treatment

  4. Uncertainty analysis of suppression pool heating during an ATWS in a BWR-5 plant

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Johnsen, G.W.; Lellouche, G.S.

    1994-03-01

    The uncertainty has been estimated of predicting the peak temperature in the suppression pool of a BWR power plant, which undergoes an NRC-postulated Anticipated Transient Without Scram (ATWS). The ATWS is initiated by recirculation-pump trips, and then leads to power and flow oscillations as they had occurred at the LaSalle-2 Power Station in March of 1988. After limit-cycle oscillations have been established, the turbines are tripped, but without MSIV closure, allowing steam discharge through the turbine bypass into the condenser. Postulated operator actions, namely to lower the reactor vessel pressure and the level elevation in the downcomer, are simulated by a robot model which accounts for operator uncertainty. All balance of plant and control systems modeling uncertainties were part of the statistical uncertainty analysis that was patterned after the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology. The analysis showed that the predicted suppression-pool peak temperature of 329.3 K (133 degrees F) has a 95-percentile uncertainty of 14.4 K (26 degrees F), and that the size of this uncertainty bracket is dominated by the experimental uncertainty of measuring Safety and Relief Valve mass flow rates under critical-flow conditions. The analysis showed also that the probability of exceeding the suppression-pool temperature limit of 352.6 K (175 degrees F) is most likely zero (it is estimated as < 5-104). The square root of the sum of the squares of all the computed peak pool temperatures is 350.7 K (171.6 degrees F)

  5. Equilibrium and Kinetic Sorption of Some Heavy Metals from Aqueous Waste Solutions Using p (AAc-HEMA)

    International Nuclear Information System (INIS)

    El-Sayed, A.H.; Khalil, F.H.; Elnesr, E.; Mansour, T.; El-Gammal, B.; El -Sabbah, M.M.B.

    2013-01-01

    Removal of heavy metals from aqueous waste solution using poly acrylic acid / 2-hydroxy ethyle methacrylate ( p-AAc/ HEMA) was investigated. Experiments were carried out as function of contact time, initial concentration, ph, particle size and temperature. Adsorption data were modeled using the pseudo-first-order, pseudo-second-order and intra-particle diffusion kinetics equations. It was shown that pseudo-second-order kinetic equation could best describe the adsorption kinetics. The results indicated that poly acrylic acid / 2-hydroxy ethyle methacrylate (p-AAc/ HEMA) is suitable as adsorbent material for adsorption of Sr 2+ , Co 2+ , Cd 2+ , Zn 2+ , Nd 3+ and Eu 3+ radio active nuclei from aqueous solutions.

  6. Use of synthetic zeolites and other inorganic sorbents for the removal of radionuclides from aqueous wastes

    International Nuclear Information System (INIS)

    Samantha, S.K.; Singh, I.J.; Jain, S.; Sathi, S.; Venkatesan, K.; Ramaswamy, M.; Theyyunni, T.K.; Siddiqui, H.R.

    1997-01-01

    Several synthetic zeolites and inorganic sorbents were tested in the laboratory for the sorption of various radionuclides present in radioactive aqueous waste streams originating from nuclear installations. The sorption of the critical radionuclides like 137 Cs, 90 Sr and 60 Co from level waste solutions was studied using the synthetic zeolites 4A, 13X and AR1 of Indian origin. Granulated forms of ammonium molybdophosphate and CaSO 4 -BaSO 4 eutectoid were tested for the sorption of cesium and strontium respectively, from acidic solutions. The removal of radiostrontium from alkaline salt-loaded intermediate level reprocessing wastes was studied using hydrous ferric oxide-activated carbon composite sorbent, hydrous titania and hydrous manganese dioxide.. The results of these investigations are expected to be of value in formulating radioactive waste treatment schemes for achieving high decontamination and volume reduction factors. (author). 12 refs, 5 figs, 18 tabs

  7. Ru decorated carbon nanotubes - a promising catalyst for reforming bio-based acetic acid in the aqueous phase

    NARCIS (Netherlands)

    de Vlieger, Dennis; Lefferts, Leonardus; Seshan, Kulathuiyer

    2014-01-01

    Catalytic reforming of biomass derived waste streams in the aqueous phase is a promising process for the production of sustainable hydrogen. Acetic acid will be a major component (up to 20 wt%) in many anticipated gasification feed streams (e.g. the aqueous fraction of pyrolysis oil). Conventional

  8. Usefulness of ANN-based model for copper removal from aqueous solutions using agro industrial waste materials

    Directory of Open Access Journals (Sweden)

    Petrović Marija S.

    2015-01-01

    Full Text Available The purpose of this study was to investigate the adsorption properties of locally available lignocelluloses biomaterials as biosorbents for the removal of copper ions from aqueous solution. Materials are generated from juice production (apricot stones and from the corn milling process (corn cob. Such solid wastes have little or no economic value and very often present a disposal problem. Using batch adsorption techniques the effects of initial Cu(II ions concentration (Ci, amount of biomass (m and volume of metal solution (V, on biosorption efficiency and capacity were studied for both materials, without any pre-treatments. The optimal parameters for both biosorbents were selected depending on a highest sorption capability of biosorbent, in removal of Cu(II. Experimental data were compared with second order polynomial regression models (SOPs and artificial neural networks (ANNs. SOPs showed acceptable coefficients of determination (0.842 - 0.997, while ANNs performed high prediction accuracy (0.980-0.986 in comparison to experimental results. [Projekat Ministarstva nauke Republike Srbije, br. TR 31003, TR 31055

  9. Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

    Energy Technology Data Exchange (ETDEWEB)

    Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States); Mayberry, J. [Science Applications International Corp., Idaho Falls, ID (United States); Frazier, G. [Univ. of Tennessee, Knoxville, TN (United States)

    1994-01-01

    This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well.

  10. Surrogate formulations for thermal treatment of low-level mixed waste, Part II: Selected mixed waste treatment project waste streams

    International Nuclear Information System (INIS)

    Bostick, W.D.; Hoffmann, D.P.; Chiang, J.M.; Hermes, W.H.; Gibson, L.V. Jr.; Richmond, A.A.; Mayberry, J.; Frazier, G.

    1994-01-01

    This report summarizes the formulation of surrogate waste packages, representing the major bulk constituent compositions for 12 waste stream classifications selected by the US DOE Mixed Waste Treatment Program. These waste groupings include: neutral aqueous wastes; aqueous halogenated organic liquids; ash; high organic content sludges; adsorbed aqueous and organic liquids; cement sludges, ashes, and solids; chloride; sulfate, and nitrate salts; organic matrix solids; heterogeneous debris; bulk combustibles; lab packs; and lead shapes. Insofar as possible, formulation of surrogate waste packages are referenced to authentic wastes in inventory within the DOE; however, the surrogate waste packages are intended to represent generic treatability group compositions. The intent is to specify a nonradiological synthetic mixture, with a minimal number of readily available components, that can be used to represent the significant challenges anticipated for treatment of the specified waste class. Performance testing and evaluation with use of a consistent series of surrogate wastes will provide a means for the initial assessment (and intercomparability) of candidate treatment technology applicability and performance. Originally the surrogate wastes were intended for use with emerging thermal treatment systems, but use may be extended to select nonthermal systems as well

  11. 47{sup th} Annual meeting on nuclear technology (AMNT 2016). Key Topics / Enhanced safety and operation excellence and decommissioning experience and Waste management solutions

    Energy Technology Data Exchange (ETDEWEB)

    Salnikova, Tatiana [AREVA GmbH, Erlangen (Germany); Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-10-15

    Summary report on the Key Topics ''Enhanced Safety and Operation Excellence'' and ''Decommissioning Experience and Waste Management Solutions'' of the 47{sup th} Annual Conference on Nuclear Technology (AMNT 2016) held in Hamburg, 10 to 12 May 2016. Other Sessions of AMNT 2016 have been and will be covered in further issues of atw.

  12. Removal of thorium(IV) from aqueous solution by biosorption onto modified powdered waste sludge. Experimental design approach

    International Nuclear Information System (INIS)

    Yunus Pamukoglu, M.; Mustafa Senyurt; Bulent Kirkan

    2017-01-01

    The biosorption of radioactive Th(IV) ions in the aqueous solutions onto the modified powdered waste sludge (MPWS) has been examined. In this context, the parameters affecting biosorption of Th(IV) from aqueous solutions has been examined by using MPWS biosorbent in Box Behnken statistical experimental design. The structure of MPWS biosorbent was characterized by using SEM and BET techniques. According to the experimental design results, MPWS and Th(IV) concentrations should be kept high to achieve the maximum efficiency in Th(IV) biosorption. On the other hand, MPWS, which is also used as a biosorbent, is an economical, effective and natural biosorbent. (author)

  13. Recovery of uranium (VI) from low level aqueous radioactive waste

    International Nuclear Information System (INIS)

    Kulshrestha, Mukul

    1996-01-01

    Investigation was undertaken to evaluate the uranium (VI) removal and recovery potential of a naturally occurring, nonviable macrofungus, Ganoderma Lucidum from the simulated low level aqueous nuclear waste. These low level waste waters discharged from nuclear mine tailings and nuclear power reactors have a typical U(VI) concentration of 10-100 mg/L. It is possible to recover this uranium economically with the advent of biosorption as a viable technology. Extensive laboratory studies have revealed Ganoderma Lucidum to be a potential biosorbent with a specific uptake of 2.75 mg/g at an equilibrium U(VI) concentration of 10 mg/L at pH 4.5. To recover the sorbed U(VI), the studies indicated 0.2N Na 2 CO 3 to be an effective elutant. The kinetics of U(VI) desorption from loaded Ganoderma Lucidum with 0.2N Na 2 CO 3 as elutant, was found to be rapid with more than 75% recovery occurring in the first five minutes, the specific metal release rate being 0.102 mg/g/min. The equilibrium data fitted to a linearised Freundlich plot and exhibited a near 100% recovery of sorbed U(VI), clearly revealing a cost-effective method of recovery of precious uranium from low level wastewater. (author). 7 refs., 3 figs., 1 tab

  14. Removal and recovery of toxic metal ions from aqueous waste sites using polymer pendant ligands

    International Nuclear Information System (INIS)

    Fish, D.

    1996-01-01

    The purpose of this project is to investigate the use of polymer pendant ligand technology to remove and recover toxic metal ions from DOE aqueous waste sites. Polymer pendant lgiands are organic ligands, anchored to crosslinked, modified divinylbenzene-polystyrene beads, that can selectively complex metal ions. The metal ion removal step usually occurs through a complexation or ion exchange phenomena, thus recovery of the metal ions and reuse of the beads is readily accomplished

  15. Purification of simulated waste water using green synthesized silver nanoparticles of Piliostigma thonningii aqueous leave extract

    Science.gov (United States)

    Shittu, K. O.; Ihebunna, O.

    2017-12-01

    Synthesis of nanoparticles from various biological systems has been reported, but among all such systems, biosynthesis of nanoparticles from plants is considered the most suitable method. The use of plant material not only makes the process eco-friendly, but also the abundance makes it more economical. The aim of this study was to biologically synthesize silver nanoparticle using Piliostigma thonningii aqueous leaf extract and applied in the purification of laboratory stimulated waste with optimization using the different conditions of silver nanoparticle production such as time, temperature, pH, concentration of silver nitrate and volume of the aqueous extract. The biosynthesized silver nanoparticles were characterized by UV-visible spectrophotometry, nanosizer, energy dispersive x-ray analysis (EDX), transmission electron microscopy (TEM) and Fourier transform infrared (FTIR) spectroscopy. The time intervals for the reaction with aqueous silver nitrate solution shows an increase in the absorbance with time and became constant giving a maximum absorbance at 415 nm at 60 min of incubation. The pH of 6.5, temperature 65 °C, 1.25 mM of silver nitrate and 5 ml of plant extract was the best condition with maximum absorbance. The results from nanosizer, UV-vis and TEM suggested the biosynthesis silver nanoparticle to be spherical ranging from 50 nm to 114 nm. The EDX confirmed the elemental synthesis of silver at 2.60 keV and FTIR suggested the capping agent to be hydroxyl (OH) group with -C=C stretching vibrations. The synthesized silver nanoparticle also shows heavy metal removal activity in laboratory simulated waste water. The safety toxicity studies show no significant difference between the orally administered silver nanoparticles treated water group and control group, while the histopathological studies show well preserved hepatic architecture for the orally administered silver nanoparticle treated waste water group when compared with the control

  16. 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.

  17. Laboratory performance testing of an extruded bitumen containing a surrogate, sodium nitrate-based, low-level aqueous waste

    International Nuclear Information System (INIS)

    Mattus, A.J.; Kaczmarsky, M.M.

    1986-01-01

    Laboratory results of a comprehensive, regulatory performance test program, utilizing an extruded bitumen and a surrogate, sodium nitrate-based waste, have been compiled at the Oak Ridge National Laboratory (ORNL). Using a 53 millimeter, Werner and Pfleiderer extruder, operated by personnel of WasteChem Corporation of Paramus, New Jersey, laboratory-scale, molded samples of type three, air blown bitumen were prepared for laboratory performance testing. A surrogate, low-level, mixed liquid waste, formulated to represent an actual on-site waste at ORNL, containing about 30 wt % sodium nitrate, in addition to eight heavy metals, cold cesium and strontium was utilized. Samples tested contained three levels of waste loading: that is, forty, fifty and sixty wt % salt. Performance test results include the ninety day ANS 16.1 leach test, with leach indices reported for all cations and anions, in addition to the EP Toxicity test, at all levels of waste loading. Additionally, test results presented also include the unconfined compressive strength and surface morphology utilizing scanning electron microscopy. Data presented include correlations between waste form loading and test results, in addition to their relationship to regulatory performance requirements

  18. Persimmon leaf bio-waste for adsorptive removal of heavy metals from aqueous solution.

    Science.gov (United States)

    Lee, Seo-Yun; Choi, Hee-Jeong

    2018-03-01

    The aim of this study was to investigate heavy metal removal using waste biomass adsorbent, persimmon leaves, in an aqueous solution. Persimmon leaves, which are biomaterials, have a large number of hydroxyl groups and are highly suitable for removal of heavy metals. Therefore, in this study, we investigated the possibility of removal of Cu, Pb, and Cd in aqueous solution by using raw persimmon leaves (RPL) and dried persimmon leaves (DPL). Removal of heavy metals by RPL and DPL showed that DPL had a 10%-15% higher removal than RPL, and the order of removal efficiency was found to be Pb > Cu > Cd. The pseudo-second order model was a better fit to the heavy metal adsorption experiments using RPL and DPL than the pseudo-first order model. The adsorption of Cu, Pb, and Cd by DPL was more suitable with the Freundlich isothermal adsorption and showed an ion exchange reaction which occurred in the uneven adsorption surface layer. The maximum adsorption capacity of Cu, Pb, and Cd was determined to be 19.42 mg/g, 22.59 mg/g, and 18.26 mg/g, respectively. The result of the adsorption experiments showed that the n value was higher than 2 regardless of the dose, indicating that the heavy metal adsorption on DPL was easy. In the thermodynamic experiment, ΔG° was a negative value, and ΔH° and ΔS° were positive values. It can be seen that the heavy metal adsorption process using DPL was spontaneous in nature and was an endothermic process. Moreover, as the temperature increased, the adsorption increased, and the affinity of heavy metal adsorption to DPL was very good. This experiment, in which heavy metals are removed using the waste biomass of persimmon leaves is an eco-friendly new bioadsorbent method because it can remove heavy metals without using chemicals while utilizing waste recycling. Copyright © 2018 Elsevier Ltd. All rights reserved.

  19. Planning and reporting of Russian transmutation research projects within ISTC. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Conde, H. [Uppsala Univ., (Sweden). Dept. of Neutron Research; Gudowski, W. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Reactor Technology; Liljenzin, J.O. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Nuclear Chemistry; Mileikovsky, C. [Pully (Switzerland)

    1997-02-01

    The International Scientific and Technical Center (ISTC) in Moscow funds research of civil interest to counteract the risk of nuclear weapon proliferation. Recently, new technical concepts, Accelerator Transmutation of Nuclear Waste (ATW), have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The Russian experts are knowledgeable and well equipped for doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been proposed to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. The present report describes the back ground, the status and near term activities of a few ISTC projects of relevance for the ATW concept, which are planned with the participation of a Swedish reference group. 4 refs.

  20. Planning and reporting of Russian transmutation research projects within ISTC. Phase 1

    International Nuclear Information System (INIS)

    Conde, H.

    1997-02-01

    The International Scientific and Technical Center (ISTC) in Moscow funds research of civil interest to counteract the risk of nuclear weapon proliferation. Recently, new technical concepts, Accelerator Transmutation of Nuclear Waste (ATW), have been proposed to incinerate and transmute long-lived radioactive nuclear waste to relax the time needed to store the waste in a geological repository. The Russian experts are knowledgeable and well equipped for doing research in the different technical fields of relevance for the transmutation concepts. Thus, a number of ISTC projects have been proposed to investigate different technical aspects of ATW with a result that a fair number of former weapon specialists have converted from military to peaceful civilian research. The present report describes the back ground, the status and near term activities of a few ISTC projects of relevance for the ATW concept, which are planned with the participation of a Swedish reference group. 4 refs

  1. Vine-shoot waste aqueous extract applied as foliar fertilizer to grapevines: Effect on amino acids and fermentative volatile content.

    Science.gov (United States)

    Sánchez-Gómez, R; Garde-Cerdán, T; Zalacain, A; Garcia, R; Cabrita, M J; Salinas, M R

    2016-04-15

    The aim of this work was to study the influence of foliar applications of different wood aqueous extracts on the amino acid content of musts and wines from Airén variety; and to study their relationship with the volatile compounds formed during alcoholic fermentation. For this purpose, the foliar treatments proposed were a vine-shoot aqueous extract applied in one and two times, and an oak extract which was only applied once. Results obtained show the potential of Airén vine-shoot waste aqueous extracts to be used as foliar fertilizer, enhancing the wine amino acid content especially when they were applied once. Similar results were observed with the aqueous oak extract. Regarding wine fermentative volatile compounds, there is a close relationship between musts and their wines amino acid content allowing us to discuss about the role of proline during the alcoholic fermentation and the generation of certain volatiles. Copyright © 2015 Elsevier Ltd. All rights reserved.

  2. Closed Fuel Cycle Waste Treatment Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collins, E. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Crum, J. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, S. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Garn, T. G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gombert, D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maio, V. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Matyas, J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nenoff, T. M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Riley, B. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sevigny, G. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thallapally, P. K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, J. H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-01

    with encapsulated nano-sized AgI crystals; Carbon-14 immobilized as a CaCO3 in a cement waste form; Krypton-85 stored as a compressed gas; An aqueous reprocessing high-level waste (HLW) raffinate waste immobilized by the vitrification process; An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel either included in the borosilicate HLW glass or immobilized in the form of a metal alloy or titanate ceramics; Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware super-compacted for disposal or purified for reuse (or disposal as low-level waste, LLW) of Zr by reactive gas separations; Electrochemical process salt HLW incorporated into a glass bonded Sodalite waste form; and Electrochemical process UDS and SS cladding hulls melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported. In addition to the above listed primary waste streams, a range of secondary process wastes are generated by aqueous reprocessing of LWR fuel, metal SFR fuel fabrication, and electrochemical reprocessing of SFR fuel. These secondary wastes have been summarized and volumes estimated by type and classification. The important waste management data gaps and research needs have been summarized for each primary waste stream and selected waste process.

  3. Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT Facility

    International Nuclear Information System (INIS)

    Grush, W.H.; Koizumi, Y.; Woerth, S.C.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data

  4. A simplified analysis of uncertainty propagation in inherently controlled ATWS events

    International Nuclear Information System (INIS)

    Wade, D.C.

    1987-01-01

    The quasi static approach can be used to provide useful insight concerning the propagation of uncertainties in the inherent response to ATWS events. At issue is how uncertainties in the reactivity coefficients and in the thermal-hydraulics and materials properties propagate to yield uncertainties in the asymptotic temperatures attained upon inherent shutdown. The basic notion to be quantified is that many of the same physical phenomena contribute to both the reactivity increase of power reduction and the reactivity decrease of core temperature rise. Since these reactivities cancel by definition, a good deal of uncertainty cancellation must also occur of necessity. For example, if the Doppler coefficient is overpredicted, too large a positive reactivity insertion is predicted upon power reduction and collapse of the ΔT across the fuel pin. However, too large a negative reactivity is also predicted upon the compensating increase in the isothermal core average temperature - which includes the fuel Doppler effect

  5. Application of biomass for the sorption of radionuclides from low level Purex aqueous wastes

    International Nuclear Information System (INIS)

    Ramanujam, A.; Gopalakrishnan, V.; Dhami, P.S.; Kannan, R.; Udupa, S.R.; Salvi, N.A.

    1997-05-01

    Microbial biomass have been found to be good biological adsorbents for radioactive nuclides such as uranium and thorium with comparatively easy desorption and recovery. Based on this, sorption studies have been carried out to assess the feasibility of using biomass Rhizopus arrhizus (RA) for the removal of radionuclides present in Purex low level waste streams. Biomass Rhizopus arrhizus (RA) appears effective for the removal of actinides and fission products from low level Purex plant waste/effluent solutions. Maximum sorption for uranium and plutonium is observed at 6-7 pH whereas for Am, Eu, Pm, Ce and Zr the sorption is maximum at pH 2 with high D values and fast kinetics in both cases. Sorption for Ru and Cs are negligible. Sorbed nuclides are recoverable by elution with 1 M HNO 3 , on once through basis. The method can be used for treating the evaporator condensates from the plant and the hold-up delay tank solution. The sodium nitrate salt concentration in the aqueous solution beyond 0.14 M seriously affects the metal uptake. The results from column experiments indicate a limited loading capacity in terms of mg of Am/U/Pu etc. per gm of RA. However, as the Purex low level effluents contain only trace level activities whose absolute ionic concentrations are much lower, the capacities observed with the present form of biomass may still be satisfactory

  6. Selective recovery of a pyridine derivative from an aqueous waste stream containing acetic acid and succinonitrile with solvent impregnated resins

    NARCIS (Netherlands)

    Bokhove, J.; Visser, T.J.; Schuur, Boelo; de Haan, A.B.

    2015-01-01

    Solvent impregnated resins (SIRs) were evaluated for the recovery of pyridine derivatives from an aqueous waste-stream containing also acetic acid and succinonitrile. For this purpose, a new solvent was developed, synthesized and impregnated in Amberlite XAD4. Sorption studies were used to determine

  7. Aqueous reprocessing - some dreams!

    International Nuclear Information System (INIS)

    Srinivasan, T.G.

    2015-01-01

    India has been pursuing a aqueous reprocessing based closed fuel cycle for both thermal and fast reactor fuels employing the PUREX process. Though the country has more than six decades of experience, the dreams or wish lists such as, a highly efficient process with textbook specifications of 99.9% recovery of U and Pu, a DF of more than 10 7 for both U and Pu from the fission products, operating with name plate capacity with high safety, low waste generation, recovery of useful fission products and minor actinides from high level waste are never ceasing and ever growing. The talk will cover safety precautions and actions to be taken in the steps listed below, to ensure a safe and successful process

  8. Technologies for destruction of long-lived radionuclides in high-level nuclear waste - overview and requirements

    International Nuclear Information System (INIS)

    Arthur, E.D.

    1993-01-01

    A major issue surrounding current nuclear power generation is the management and disposal of long-lived, high-level waste (HLW). The planned and scientifically acceptable destination for this waste is in deep underground, geologically stable, repositories. However, public concerns surrounding such disposal of long-lived nuclear wastes and other issues such as proliferation and safety negatively affect the potential role that nuclear power can play in meeting current and future national energy needs. This paper and this topical session on nuclear waste minimization, management, and remediation focus on two nuclear systems and their associated technologies that have the potential to address concerns surrounding long-lived radionuclides in HLW. Both systems offer technology applicable to HLW from current light water reactors (LWRs). In addition, these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes long-term fuel supplies, and enhanced safety. The first system is the integral fast reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while burning long-lived actinide waste. The IFR applications include the burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, accelerator transmutation of waste (ATW), is based on an accelerator-induced intense source of thermal neutrons and is aimed at the destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent-fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

  9. Mechanisms of Copper Corrosion in Aqueous Environments. A report from the Swedish National Council for Nuclear Waste's scientific workshop, on November 16, 2009

    International Nuclear Information System (INIS)

    2010-01-01

    In 2010 the Swedish Nuclear Fuel and Waste Management Company, SKB, plans to submit its license application for the final repository of spent nuclear fuel. The proposed method is the so-called KBS-3 method and implies placing the spent nuclear fuel in copper canisters, surrounded by a buffer of bentonite clay, at 500 m depth in the bedrock. The site selected by SKB to host the repository is located in the municipality of Oesthammar on the Swedish east coast. The copper canister plays a key role in the design of the repository for spent nuclear fuel in Sweden. The long-term physical and chemical stability of copper in aqueous environments is fundamental for the safety evolution of the proposed disposal concept. However, the corrosion resistance of copper has been questioned by results obtained under anoxic conditions in aqueous solution. These observations caused some head-lines in the Swedish newspapers as well as public and political concerns. Consequently, the Swedish National Council for Nuclear Waste organized a scientific workshop on the issue 'Mechanisms of Copper Corrosion in Aqueous Environments'. The purpose of the workshop was to address the fundamental understanding of the corrosion characteristics of copper regarding oxygen-free environments, and to identify what additional information is needed to assess the validity of the proposed corrosion mechanism and its implication on the containment of spent nuclear fuel in a copper canister. This seminar report is based on the presentations and discussions at the workshop. It also includes written statements by the members of the expert panel

  10. Removal of common organic solvents from aqueous waste streams via supercritical C02 extraction: a potential green approach to sustainable waste management in the pharmaceutical industry.

    Science.gov (United States)

    Leazer, Johnnie L; Gant, Sean; Houck, Anthony; Leonard, William; Welch, Christopher J

    2009-03-15

    Supercritical CO2 extraction of aqueous streams is a convenient and effective method to remove commonly used solvents of varying polarities from aqueous waste streams. The resulting aqueous layers can potentially be sewered; whereas the organic layer can be recovered for potential reuse. Supercritical fluid extraction (SFE) is a technology that is increasingly being used in commercial processes (1). Supercritical fluids are well suited for extraction of a variety of media, including solids, natural products, and liquid products. Many supercritical fluids have low critical temperatures, allowing for extractions to be done at modestly low temperatures, thus avoiding any potential thermal decomposition of the solutes under study (2). Furthermore, the CO2 solvent strength is easily tuned by adjusting the density of the supercritical fluid (The density is proportional to the pressure of the extraction process). Since many supercritical fluids are gases at ambient temperature, the extract can be concentrated by simply venting the reaction mixture to a cyclone collection vessel, using appropriate safety protocols.

  11. Alpha wastes treatment

    International Nuclear Information System (INIS)

    Thouvenot, P.

    2000-01-01

    Alter 2004, the alpha wastes issued from the Commissariat a l'Energie Atomique installations will be sent to the CEDRA plant. The aims of this installation are decontamination and wastes storage. Because of recent environmental regulations concerning ozone layer depletion, the use of CFC 113 in the decontamination unit, as previously planned, is impossible. Two alternatives processes are studied: the AVD process and an aqueous process including surfactants. Best formulations for both processes are defined issuing degreasing kinetics. It is observed that a good degreasing efficiency is linked to a good decontamination efficiency. Best results are obtained with the aqueous process. Furthermore, from the point of view of an existing waste treatment unit, the aqueous process turns out to be more suitable than the AVD process. (author)

  12. Improved Process Used to Treat Aqueous Mixed Waste Results in Cost Savings and Improved Worker Safety

    International Nuclear Information System (INIS)

    Hodge, D.S.; Preuss, D.E.; Belcher, K.J.; Rock, C.M.; Bray, W.S.; Herman, J.P.

    2006-01-01

    This paper describes an improved process implemented at Argonne National Laboratory (ANL) to treat aqueous mixed waste. This waste is comprised of radioactively-contaminated corrosive liquids with heavy metals. The Aqueous Mixed Waste Treatment System (AMWTS) system components include a reaction tank and a post-treatment holding tank with ancillary piping and pumps; and a control panel with pumping/mixing controls; tank level, temperature and pH/Oxidation Reduction Potential (ORP) indicators. The process includes a neutralization step to remove the corrosive characteristic, a chromium reduction step to reduce hexavalent chromium to trivalent chromium, and a precipitation step to convert the toxic metals into an insoluble form. Once the toxic metals have precipitated, the resultant sludge is amenable to stabilization and can be reclassified as a low-level waste if the quantity of leachable toxic metals, as determined by the TCLP, is below Universal Treatment Standards (UTS). To date, six batches in eight have passed the UTS. The AMWTS is RCRA permitted and allows for the compliant treatment of mixed waste prior to final disposal at a Department of Energy (DOE) or commercial radioactive waste disposal facility. Mixed wastes eligible for treatment include corrosive liquids (pH 12.5) containing EPA-regulated toxic metals (As, Ba, Pb, Cd, Cr, Ag, Se, Hg) at concentrations greater than the RCRA Toxicity Characteristic Leaching Procedure (TCLP) limit. The system has also been used to treat corrosive wastes with small quantities of fissionable materials. The AMWTS is a significant engineered solution with many improvements over the more labor intensive on-site treatment method being performed within a ventilation hood used previously. The previously used treatment system allowed for batch sizes of only 15-20 gallons whereas the new AMWTS allows for the treatment of batches up to 75 gallons; thereby reducing batch labor and supply costs by 40-60% and reducing analytical

  13. Simultaneous production of high-quality water and electrical power from aqueous feedstock’s and waste heat by high-pressure membrane distillation

    NARCIS (Netherlands)

    Kuipers, N.J.M.; Hanemaaijer, J.H.; Brouwer, H.; Medevoort, J. van; Jansen, A.; Altena, F.; Vleuten, P. van der; Bak, H.

    2015-01-01

    A new membrane distillation (MD) concept (MemPower) has been developed for the simultaneous production of high-quality water from various aqueous feedstocks with cogeneration of mechanical power (electricity). Driven by low-grade heat (waste, solar, geothermal, etc.) a pressurized distillate can be

  14. Standard practice for analysis of aqueous leachates from nuclear waste materials using inductively coupled plasma-atomic emission spectrometry

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice is applicable to the determination of low concentration and trace elements in aqueous leachate solutions produced by the leaching of nuclear waste materials, using inductively coupled plasma-atomic emission spectroscopy (ICP-AES). 1.2 The nuclear waste material may be a simulated (non-radioactive) solid waste form or an actual solid radioactive waste material. 1.3 The leachate may be deionized water or any natural or simulated leachate solution containing less than 1 % total dissolved solids. 1.4 This practice should be used by analysts experienced in the use of ICP-AES, the interpretation of spectral and non-spectral interferences, and procedures for their correction. 1.5 No detailed operating instructions are provided because of differences among various makes and models of suitable ICP-AES instruments. Instead, the analyst shall follow the instructions provided by the manufacturer of the particular instrument. This test method does not address comparative accuracy of different devices...

  15. Why have we stopped research on liquid centrifugal separation

    International Nuclear Information System (INIS)

    Li, N.

    1996-01-01

    Using high-temperature high-speed liquid centrifuges for lanthanides and actinides separation was originally proposed as a physical separation method in the Los Alamos ADTT/ATW concept [C. Bowman, LA-UR-92-1065 (1992)]. The authors investigated centrifugal separation in a concerted effort of experiments, theoretical analysis and numerical simulations. They discovered that owing to the ionic-composition-dependence of the sedimentation coefficients for the fission products and actinides, separation by grouping of molecular densities would not work in general in the molten salt environment. Alternatively the lanthanides and actinides could be transferred to a liquid metal carrier (e.g. bismuth) via reductive extraction and then separated by liquid centrifuges, but the material and technical challenges are severe. Meanwhile the authors have established that the reductive extraction procedure itself can be used for desired separations. Unlike conventional aqueous-based reprocessing technologies, reductive extraction separation uses only reagent (Li) that reconstitutes carrier salts (LiF-BeF 2 ) and a processing medium (Bi) that can be continuously recycled and reused, with a nearly-pure fission products waste stream. The processing units are compact and reliable, and can be built at relatively low cost while maintaining high throughput. Therefore the research effort on developing liquid centrifuges for separations in ADTT/ATW was terminated in late 1995. This paper will discuss the various aspects involved in reaching this decision

  16. Effect of power oscillations on suppression pool heating during ATWS [Anticipated Transients Without Scram] conditions

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.

    1990-01-01

    Nine selected Anticipated Transients Without Scram (ATWS) have been simulated on the BNL Engineering Plant Analyzer (EPA), to determine how power and flow oscillations, similar to those that did or could have occurred at the LaSalle-2 boiling Water Reactor (BWR), could affect the rate of Pressure Suppression Pool heating. It has been determined that the pool can reach its temperature limit of 80 degree C in 4.3 min. after Turbine Trip without Bypass, if the feedwater pumps are not tripped. The pool will not reach its limit, if Boron is injected, even when oscillations are encountered. Simultaneous turbine and recirculation pump trips, introduced under stable conditions, can lead to instability. 2 refs., 17 figs., 9 tabs

  17. 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

  18. Isotopic analysis of radioactive waste packages (an inexpensive approach)

    International Nuclear Information System (INIS)

    Padula, D.A.; Richmond, J.S.

    1983-01-01

    A computer printout of the isotopic analysis for all radioactive waste packages containing resins, or other aqueous filter media is now required at the disposal sites at Barnwell, South Carolina, and Beatty, Nevada. Richland, Washington requires an isotopic analysis for all radioactive waste packages. The NRC (Nuclear Regulatory Commission), through 10 CFR 61, will require shippers of radioactive waste to classify and label for disposal all radioactive waste forms. These forms include resins, filters, sludges, and dry active waste (trash). The waste classification is to be based upon 10 CFR 61 (Section 1-7). The isotopes upon which waste classification is to be based are tabulated. 7 references, 8 tables

  19. Application of biomass for the sorption of radionuclides from low level Purex aqueous wastes

    Energy Technology Data Exchange (ETDEWEB)

    Ramanujam, A; Gopalakrishnan, V; Dhami, P S; Kannan, R [Fuel Reprocessing Div., Bhabha Atomic Research Centre, Mumbai (India); Udupa, S R; Salvi, N A [Bio-Organic Div., Bhabha Atomic Research Centre, Mumbai (India)

    1997-05-01

    Microbial biomass have been found to be good biological adsorbents for radioactive nuclides such as uranium and thorium with comparatively easy desorption and recovery. Based on this, sorption studies have been carried out to assess the feasibility of using biomass Rhizopus arrhizus (RA) for the removal of radionuclides present in Purex low level waste streams. Biomass Rhizopus arrhizus (RA) appears effective for the removal of actinides and fission products from low level Purex plant waste/effluent solutions. Maximum sorption for uranium and plutonium is observed at 6-7 pH whereas for Am, Eu, Pm, Ce and Zr the sorption is maximum at pH 2 with high D values and fast kinetics in both cases. Sorption for Ru and Cs are negligible. Sorbed nuclides are recoverable by elution with 1 M HNO{sub 3}, on once through basis. The method can be used for treating the evaporator condensates from the plant and the hold-up delay tank solution. The sodium nitrate salt concentration in the aqueous solution beyond 0.14 M seriously affects the metal uptake. The results from column experiments indicate a limited loading capacity in terms of mg of Am/U/Pu etc. per gm of RA. However, as the Purex low level effluents contain only trace level activities whose absolute ionic concentrations are much lower, the capacities observed with the present form of biomass may still be satisfactory. 15 refs., 12 tabs.

  20. 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

  1. Regenerating using aqueous cleaners with ozone and electrolysis

    Science.gov (United States)

    McGinness, Michael P.

    1994-02-01

    A new process converts organic oil and grease contaminates in used water based cleaners into synthetic surfactants. This permits the continued use of a cleaning solution long after it would have been dumped using previously known methods. Since the organic soils are converted from contaminates to cleaning compounds the need for frequent bath dumps is totally eliminated. When cleaning solutions used in aqueous cleaning systems are exhausted and ready for disposal, they will always contain the contaminates removed from the cleaned parts and drag-in from prior cleaning steps. Even when the cleaner is biodegradable these contaminants will frequently cause the waste cleaning solution to be a hazardous waste. Chlorinated solvents are rapidly being replaced by aqueous cleaners to avoid the new ozone-depletion product-labeling-law. Many industry standard halocarbon based solvents are being completely phased out of production, and their prices have nearly tripled. Waste disposal costs and cradle-to-grave liability are also major concerns for industry today. This new process reduces the amount of water and chemicals needed to maintain the cleaning process. The cost of waste disposal is eliminated because the water and cleaning compounds are reused. Energy savings result by eliminating the need for energy currently used to produce and deliver fresh water and chemicals as well as the energy used to treat and destroy the waste from the existing cleaning processes. This process also allows the cleaning bath to be maintained at the peak performance of a new bath resulting in decreased cycle times and decreased energy consumption needed to clean the parts. This results in a more efficient and cost effective cleaning process.

  2. Waste forms based on Cs-loaded silicotitanates

    International Nuclear Information System (INIS)

    Balmer, M.L.; Bunker, B.C.

    1995-04-01

    Silicotitanate ion exchange materials are being considered for removal of radioactive Cs and Sr from tank wastes at the Hanford site. The phase evolution as a function of heat treatment temperature for several sol gel derived compositions within the Cs 2 O-SiO 2 -TiO 2 system was investigated, in order to determine if an adequate waste form can be achieved by direct thermal conversion. The Cs leach rates and Cs loss during heat treatment of select materials were measured. Some compositions which contain large amounts of Ti melt to form a glass with reasonably low aqueous leach rates. A new Cs-silicotitanate material with a structure isomorphous to pollucite was discovered. This material forms at low temperatures (700--800 C) where Cs volatility is negligible. The silicotitanate-pollucite exhibits extremely low leach rates (1.42 g/m 2 day ) at 90 C, and has been identified as a promising waste form for Cs containment

  3. Adsorption of malachite green on groundnut shell waste based powdered activated carbon

    International Nuclear Information System (INIS)

    Malik, R.; Ramteke, D.S.; Wate, S.R.

    2007-01-01

    In the present technologically fast changing situation related to waste management practices, it is desirable that disposal of plant waste should be done in a scientific manner by keeping in view economic and pollution considerations. This is only possible when the plant waste has the potential to be used as raw material for some useful product. In the present study, groundnut shell, an agricultural waste, was used for the preparation of an adsorbent by chemical activation using ZnCl 2 under optimized conditions and its comparative characterisation was conducted with commercially available powdered activated carbon (CPAC) for its physical, chemical and adsorption properties. The groundnut shell based powdered activated carbon (GSPAC) has a higher surface area, iodine and methylene blue number compared to CPAC. Both of the carbons were used for the removal of malachite green dye from aqueous solution and the effect of various operating variables, viz. adsorbent dose (0.1-1 g l -1 ), contact time (5-120 min) and adsorbate concentrations (100-200 mg l -1 ) on the removal of dye, has been studied. The experimental results indicate that at a dose of 0.5 g l -1 and initial concentration of 100 mg l -1 , GSPAC showed 94.5% removal of the dye in 30 min equilibrium time, while CPAC removed 96% of the dye in 15 min. The experimental isotherm data were analyzed using the linearized forms of Freundlich, Langmuir and BET equations to determine maximum adsorptive capacities. The equilibrium data fit well to the Freundlich isotherm, although the BET isotherm also showed higher correlation for both of the carbons. The results of comparative adsorption capacity of both carbons indicate that groundnut shell can be used as a low-cost alternative to commercial powdered activated carbon in aqueous solution for dye removal

  4. Water Mock-up for the Sodium Waste Treatment Process

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Kim, Jong Man; Kim, Byung Ho; Lee, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    It is important to safely treat the waste sodium which was produced from the sodium cooled fast reactors and the sodium facilities. About 1.3 tons of sodium waste has accumulated at KAERI from the sodium experiments which have been carried out since 1990. Also, large scaled sodium experiments are scheduled to verify the design of the sodium cooled fast reactor. As a treatment method for the waste sodium produced at the sodium facility, an investigation of the reaction procedure of the waste sodium with the sodium hydroxide aqueous has been developed. The NOAH process was developed in France for the treatment of waste sodium produced from sodium facilities and reactors. In the NOAH process, a small amount of sodium waste is continuously injected into the upper space which is formed on the free surface of the aqueous and slowly reacted with sodium hydroxide aqueous. Since the density of the sodium is lower than that of the aqueous, the injected sodium waste sometimes accumulates above the free surface of the sodium hydroxide aqueous, and its reaction rate becomes slow or suddenly increases. In the improved process, the sodium was injected into a reaction vessel filled with a sodium hydroxide aqueous through an atomizing nozzle installed on a lower level than that of the aqueous to maintain the reaction uniformly. Fig.1 shows the sodium waste process which was proposed in KAERI. The aqueous is composed of 60% sodium hydroxide, and its temperature is about 60 .deg. C. The process is an exothermic reaction. The hydrogen gas is generated, and the concentration of the sodium hydroxide increases in this process. It needs several systems for the process, i.e. a waste sodium injection, a cooling of the aqueous, hydrogen ventilation, and neutralization with nitric acid. The atomizing nozzle was designed to inject the sodium with the nitrogen gas which supplies a heat to the sodium to prevent its solidification and to uniformly mix the sodium with the aqueous. There are

  5. Generic implications of ATWS events at the Salem Nuclear Power Plant. Licensee and staff actions

    International Nuclear Information System (INIS)

    1983-08-01

    This report, Volume 2 of two volumes of NUREG-1000, describes the intermediate term actions to be taken by licensees and applicants of the US Nuclear Regulatory Commission (NRC), on the one hand, and by NRC staff, on the other, to address the generic issues raised by two anticipated transients without scram (ATWS) at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25, 1983. These actions came about as a result of the findings of NUREG-1000, Volume 1, and of reviews by the NRC Committee to Review Generic Requirements, the NRC Program Offices, and the Commission. The actions to be taken by licensees and applicants have been detailed in a letter pursuant to 10 CFR 50.54(f)

  6. RF system considerations for accelerator production of tritium and the transmutation of nuclear waste

    International Nuclear Information System (INIS)

    Tallerico, P.J.; Lynch, M.T.

    1993-01-01

    RF driven proton accelerators for the transmutation of nuclear waste (ATW) or for the production of tritium (APT) require unprecedented amounts of CW RF power at UHF frequencies. For both systems, the baseline design is for 246 MW at 700 MHz and 8,5 MW at 350 MHz. The main technical challenges are how to design and build such a large system so that it has excellent reliability, high efficiency, and reasonable capital cost. The issues associated with the selection of the RF amplifier and the sizes of the power supplies are emphasized in this paper

  7. Preliminary investigation of actinide and xenon reactivity effects in accelerator transmutation of waste high-flux systems

    International Nuclear Information System (INIS)

    Olson, K.R.; Henderson, D.L.

    1995-01-01

    The possibility of an unstable positive reactivity growth in an accelerator transmutation of waste (ATW)-type high-flux system is investigated. While it has always been clear that xenon is an important actor in the reactivity response of a system to flux changes, it has been suggested that in very high thermal flux transuranic burning systems, a positive, unstable reactivity growth could be caused by the actinides alone. Initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately. The maximum change in reactivity after a flux change caused by the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or startup. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response caused by the actinides. The capabilities and applications of both the current actinide model and the xenon model are discussed. Finally, the need for a complete dynamic model for the high-flux fluid-fueled ATW system is addressed

  8. Mechanisms of Copper Corrosion in Aqueous Environments. A report from the Swedish National Council for Nuclear Waste's scientific workshop, on November 16, 2009

    Energy Technology Data Exchange (ETDEWEB)

    2010-07-01

    In 2010 the Swedish Nuclear Fuel and Waste Management Company, SKB, plans to submit its license application for the final repository of spent nuclear fuel. The proposed method is the so-called KBS-3 method and implies placing the spent nuclear fuel in copper canisters, surrounded by a buffer of bentonite clay, at 500 m depth in the bedrock. The site selected by SKB to host the repository is located in the municipality of Oesthammar on the Swedish east coast. The copper canister plays a key role in the design of the repository for spent nuclear fuel in Sweden. The long-term physical and chemical stability of copper in aqueous environments is fundamental for the safety evolution of the proposed disposal concept. However, the corrosion resistance of copper has been questioned by results obtained under anoxic conditions in aqueous solution. These observations caused some head-lines in the Swedish newspapers as well as public and political concerns. Consequently, the Swedish National Council for Nuclear Waste organized a scientific workshop on the issue 'Mechanisms of Copper Corrosion in Aqueous Environments'. The purpose of the workshop was to address the fundamental understanding of the corrosion characteristics of copper regarding oxygen-free environments, and to identify what additional information is needed to assess the validity of the proposed corrosion mechanism and its implication on the containment of spent nuclear fuel in a copper canister. This seminar report is based on the presentations and discussions at the workshop. It also includes written statements by the members of the expert panel

  9. A review on applicability of naturally available adsorbents for the removal of hazardous dyes from aqueous waste.

    Science.gov (United States)

    Sharma, Pankaj; Kaur, Harleen; Sharma, Monika; Sahore, Vishal

    2011-12-01

    The effluent water of many industries, such as textiles, leather, paper, printing, cosmetics, etc., contains large amount of hazardous dyes. There is huge number of treatment processes as well as adsorbent which are available for the processing of this effluent water-containing dye content. The applicability of naturally available low cast and eco-friendly adsorbents, for the removal of hazardous dyes from aqueous waste by adsorption treatment, has been reviewed. In this review paper, we have provided a compiled list of low-cost, easily available, safe to handle, and easy-to-dispose-off adsorbents. These adsorbents have been classified into five different categories on the basis of their state of availability: (1) waste materials from agriculture and industry, (2) fruit waste, (3) plant waste, (4) natural inorganic materials, and (5) bioadsorbents. Some of the treated adsorbents have shown good adsorption capacities for methylene blue, congo red, crystal violet, rhodamine B, basic red, etc., but this adsorption process is highly pH dependent, and the pH of the medium plays an important role in the treatment process. Thus, in this review paper, we have made some efforts to discuss the role of pH in the treatment of wastewater.

  10. Actinide recovery using aqueous biphasic extraction: Initial developmental studies

    International Nuclear Information System (INIS)

    Chaiko, D.J.; Mensah-Biney, R.; Mertz, C.J.; Rollins, A.N.

    1992-08-01

    Aqueous biphasic extraction systems are being developed to treat radioactive wastes. The separation technique involves the selective partitioning of either solutes or colloid-size particles between two scible aqueous phases. Wet grinding of plutonium residues to an average particle size of one micron will be used to liberate the plutonium from the bulk of the particle matrix. The goal is to produce a plutonium concentrate that will integrate with existing and developing chemical recovery processes. Ideally, the process would produce a nonTRU waste stream. Coupling physical beneficiation with chemical processing will result in a substantial reduction in the volume of mixed wastes generated from dissolution recovery processes. As part of this program, we will also explore applications of aqueous biphasic extraction that include the separation and recovery of dissolved species such as metal ions and water-soluble organics. The expertise and data generated in this work will form the basis for developing more cost-effective processes for handling waste streams from environmental restoration and waste management activities within the DOE community. This report summarizes the experimental results obtained during the first year of this effort. Experimental efforts were focused on elucidating the surface and solution chemistry variables which govern partitioning behavior of plutonium and silica in aqueous biphasic extraction systems. Additional efforts were directed toward the development of wet grinding methods for producing ultrafine particles with diameters of one micron or less

  11. Actinide recovery using aqueous biphasic extraction: Initial developmental studies

    Energy Technology Data Exchange (ETDEWEB)

    Chaiko, D.J.; Mensah-Biney, R.; Mertz, C.J.; Rollins, A.N.

    1992-08-01

    Aqueous biphasic extraction systems are being developed to treat radioactive wastes. The separation technique involves the selective partitioning of either solutes or colloid-size particles between two scible aqueous phases. Wet grinding of plutonium residues to an average particle size of one micron will be used to liberate the plutonium from the bulk of the particle matrix. The goal is to produce a plutonium concentrate that will integrate with existing and developing chemical recovery processes. Ideally, the process would produce a nonTRU waste stream. Coupling physical beneficiation with chemical processing will result in a substantial reduction in the volume of mixed wastes generated from dissolution recovery processes. As part of this program, we will also explore applications of aqueous biphasic extraction that include the separation and recovery of dissolved species such as metal ions and water-soluble organics. The expertise and data generated in this work will form the basis for developing more cost-effective processes for handling waste streams from environmental restoration and waste management activities within the DOE community. This report summarizes the experimental results obtained during the first year of this effort. Experimental efforts were focused on elucidating the surface and solution chemistry variables which govern partitioning behavior of plutonium and silica in aqueous biphasic extraction systems. Additional efforts were directed toward the development of wet grinding methods for producing ultrafine particles with diameters of one micron or less.

  12. Integrated waste management - Looking beyond the solid waste horizon

    International Nuclear Information System (INIS)

    Seadon, J.K.

    2006-01-01

    Waste as a management issue has been evident for over four millennia. Disposal of waste to the biosphere has given way to thinking about, and trying to implement, an integrated waste management approach. In 1996 the United Nations Environmental Programme (UNEP) defined 'integrated waste management' as 'a framework of reference for designing and implementing new waste management systems and for analysing and optimising existing systems'. In this paper the concept of integrated waste management as defined by UNEP is considered, along with the parameters that constitute integrated waste management. The examples used are put into four categories: (1) integration within a single medium (solid, aqueous or atmospheric wastes) by considering alternative waste management options (2) multi-media integration (solid, aqueous, atmospheric and energy wastes) by considering waste management options that can be applied to more than one medium (3) tools (regulatory, economic, voluntary and informational) and (4) agents (governmental bodies (local and national), businesses and the community). This evaluation allows guidelines for enhancing success: (1) as experience increases, it is possible to deal with a greater complexity; and (2) integrated waste management requires a holistic approach, which encompasses a life cycle understanding of products and services. This in turn requires different specialisms to be involved in the instigation and analysis of an integrated waste management system. Taken together these advance the path to sustainability

  13. Long-term modeling of glass waste in portland cement- and clay-based matrices

    International Nuclear Information System (INIS)

    Stockman, H.W.; Nagy, K.L.; Morris, C.E.

    1995-12-01

    A set of ''templates'' was developed for modeling waste glass interactions with cement-based and clay-based matrices. The templates consist of a modified thermodynamic database, and input files for the EQ3/6 reaction path code, containing embedded rate models and compositions for waste glass, cement, and several pozzolanic materials. Significant modifications were made in the thermodynamic data for Th, Pb, Ra, Ba, cement phases, and aqueous silica species. It was found that the cement-containing matrices could increase glass corrosion rates by several orders of magnitude (over matrixless or clay matrix systems), but they also offered the lowest overall solubility for Pb, Ra, Th and U. Addition of pozzolans to cement decreased calculated glass corrosion rates by up to a factor of 30. It is shown that with current modeling capabilities, the ''affinity effect'' cannot be trusted to passivate glass if nuclei are available for precipitation of secondary phases that reduce silica activity

  14. Photochemical properties of Ysub(t) base in aqueous solution

    International Nuclear Information System (INIS)

    Paszyc, S.; Rafalska, M.

    1979-01-01

    Photoreactivity of Ysub(t) base (I) has been studied in aqueous solution (pH-6) saturated with oxygen. Two photoproducts (II,III), resulting from irradiation at lambda = 253.7 nm and lambda >= 290 nm were isolated and their structures determined. The quantum yield for Ysub(t) base disappearance (rho dis) is 0.002 (lambda = 313 nm). It was shown that dye- sensitised photo-oxidation of Ysub(t) base in aqueous solution occurs according to a Type I mechanism as well as with participation of singlet state oxygen. Quantum yields, fluorescence decay times and phosphorescence of Ysub(t) base have also been determined. (author)

  15. Effect of Na{sub 2}O on aqueous dissolution of nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Farooqi, Rahmat Ullah, E-mail: rufarooqi@live.com [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hrma, Pavel [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA (United States)

    2017-04-15

    Sodium oxide is present in the majority of commercial and waste glasses as a viscosity-reducing component. In some nuclear waste glasses, its source is the waste itself. As such, it can limit the waste loading because of its deleterious effect on the resistance of the glass to attack by aqueous media. The maximum tolerable content of Na{sub 2}O in glass depends on the presence and concentration of components that interact with it. To assess the acceptability limits of Na{sub 2}O in the composition region of nuclear waste glasses, we formulated 11 baseline compositions by varying the content of oxides of Si, B, Al, Ca, Zr, and Li. In each of these compositions, we varied the Na{sub 2}O fraction from 8–16 mass% to 23–30 mass%. To each of 146 glasses thus formulated, we applied the seven-day Product Consistency Test (PCT) to determine normalized B and Na releases (r{sub i}, where i ≡ B or Na). Fitting approximation functions ln(r{sub i}/gm{sup −2}) = Σb{sub ij}g{sub j} to r{sub i} data (g{sub j} is the j-th component mass fraction and b{sub ij} the corresponding component coefficient), we showed that the r{sub B} (and, consequently, the initial glass alteration rate) was proportional to the glass component mass fractions in the order Al{sub 2}O{sub 3}

  16. Valorization of aquaculture waste in removal of cadmium from aqueous solution: optimization by kinetics and ANN analysis

    Science.gov (United States)

    Aditya, Gautam; Hossain, Asif

    2018-05-01

    Cadmium is one of the most hazardous heavy metal concerning human health and aquatic pollution. The removal of cadmium through biosorption is a feasible option for restoration of the ecosystem health of the contaminated freshwater ecosystems. In compliance with this proposition and considering the efficiency of calcium carbonate as biosorbent, the shell dust of the economically important snail Bellamya bengalensis was tested for the removal of cadmium from aqueous medium. Following use of the flesh as a cheap source of protein, the shells of B. bengalensis made up of CaCO3 are discarded as aquaculture waste. The biosorption was assessed through batch sorption studies along with studies to characterize the morphology and surface structures of waste shell dust. The data on the biosorption were subjected to the artificial neural network (ANN) model for optimization of the process. The biosorption process changed as functions of pH of the solution, concentration of heavy metal, biomass of the adsorbent and time of exposure. The kinetic process was well represented by pseudo second order ( R 2 = 0.998), and Langmuir equilibrium ( R 2 = 0.995) had better fits in the equilibrium process with 30.33 mg g-1 of maximum sorption capacity. The regression equation ( R 2 = 0.948) in the ANN model supports predicted values of Cd removal satisfactorily. The normalized importance analysis in ANN predicts Cd2+ concentration, and pH has the most influence in removal than biomass dose and time. The SEM and EDX studies show clear peaks for Cd confirming the biosorption process while the FTIR study depicts the main functional groups (-OH, C-H, C=O, C=C) responsible for the biosorption process. The study indicated that the waste shell dust can be used as an efficient, low cost, environment friendly, sustainable adsorbent for the removal of cadmium from aqueous solution.

  17. Adsorption of Reactive Blue 171 from Aqueous Solution using Low Cost Activated Carbon Prepared from Agricultural Solid Waste: Albizia amara

    Directory of Open Access Journals (Sweden)

    K. Anitha

    2015-07-01

    Full Text Available The adsorption of Reactive Blue 171 (Reactive Dye from aqueous solution using activated carbon prepared from Albizia amara pod shell waste as an adsorbent have been carried out. The experimental adsorption data fitted reasonably well to Langmuir and Freundlich adsorption isotherms. Kinetic parameters as a function of Initial dye concentration have been calculated and the kinetic data were substituted in Pseudo First Order, Elovich and Pseudo Second order equations. A probable explanation is offered to account for the results of kinetic study. The thermodynamic parameter enthalpy change (∆H suggests the exothermic nature of absorption of Reactive Blue 171 onto activated Albizia amara pod shell waste carbon.

  18. A process for treatment of mixed waste containing chemical plating wastes

    International Nuclear Information System (INIS)

    Anast, K.R.; Dziewinski, J.; Lussiez, G.

    1995-01-01

    The Waste Treatment and Minimization Group at Los Alamos National Laboratory has designed and will be constructing a transportable treatment system to treat low-level radioactive mixed waste generated during plating operations. The chemical and plating waste treatment system is composed of two modules with six submodules, which can be trucked to user sites to treat a wide variety of aqueous waste solutions. The process is designed to remove the hazardous components from the waste stream, generating chemically benign, disposable liquids and solids with low level radioactivity. The chemical and plating waste treatment system is designed as a multifunctional process capable of treating several different types of wastes. At this time, the unit has been the designated treatment process for these wastes: Destruction of free cyanide and metal-cyanide complexes from spent plating solutions; destruction of ammonia in solution from spent plating solutions; reduction of Cr VI to Cr III from spent plating solutions, precipitation, solids separation, and immobilization; heavy metal precipitation from spent plating solutions, solids separation, and immobilization, and acid or base neutralization from unspecified solutions

  19. Defining a metal-based waste form for IFR pyroprocessing wastes

    International Nuclear Information System (INIS)

    McDeavitt, S.M.; Park, J.Y.; Ackerman, J.P.

    1994-01-01

    Pyrochemical electrorefining to recover actinides from metal nuclear fuel is a key element of the Integral Fast Reactor (IFR) fuel cycle. The process separates the radioactive fission products from the long-lived actinides in a molten LiCl-KCl salt, and it generates a lower waste volume with significantly less long-term toxicity as compared to spent nuclear fuel. The process waste forms include a mineral-based waste form that will contain fission products removed from an electrolyte salt and a metal-based waste form that will contain metallic fission products and the fuel cladding and process materials. Two concepts for the metal-based waste form are being investigated: (1) encapsulating the metal constituents in a Cu-Al alloy and (2) alloying the metal constituents into a uniform stainless steel-based waste form. Results are given from our recent studies of these two concepts

  20. Waste-acceptance criteria and risk-based thinking for radioactive-waste classification

    International Nuclear Information System (INIS)

    Lowenthal, M.D.

    1998-01-01

    The US system of radioactive-waste classification and its development provide a reference point for the discussion of risk-based thinking in waste classification. The official US system is described and waste-acceptance criteria for disposal sites are introduced because they constitute a form of de facto waste classification. Risk-based classification is explored and it is found that a truly risk-based system is context-dependent: risk depends not only on the waste-management activity but, for some activities such as disposal, it depends on the specific physical context. Some of the elements of the official US system incorporate risk-based thinking, but like many proposed alternative schemes, the physical context of disposal is ignored. The waste-acceptance criteria for disposal sites do account for this context dependence and could be used as a risk-based classification scheme for disposal. While different classes would be necessary for different management activities, the waste-acceptance criteria would obviate the need for the current system and could better match wastes to disposal environments saving money or improving safety or both

  1. Process auditing and performance improvement in a mixed wastewater-aqueous waste treatment plant.

    Science.gov (United States)

    Collivignarelli, Maria Cristina; Bertanza, Giorgio; Abbà, Alessandro; Damiani, Silvestro

    2018-02-01

    The wastewater treatment process is based on complex chemical, physical and biological mechanisms that are closely interconnected. The efficiency of the system (which depends on compliance with national regulations on wastewater quality) can be achieved through the use of tools such as monitoring, that is the detection of parameters that allow the continuous interpretation of the current situation, and experimental tests, which allow the measurement of real performance (of a sector, a single treatment or equipment) and comparison with the following ones. Experimental tests have a particular relevance in the case of municipal wastewater treatment plants fed with a strong industrial component and especially in the case of plants authorized to treat aqueous waste. In this paper a case study is presented where the application of management tools such as careful monitoring and experimental tests led to the technical and economic optimization of the plant: the main results obtained were the reduction of sludge production (from 4,000 t/year w.w. (wet weight) to about 2,200 t/year w.w.) and operating costs (e.g. from 600,000 €/year down to about 350,000 €/year for reagents), the increase of resource recovery and the improvement of the overall process performance.

  2. Aqueous Synthesis of Technetium-Doped Titanium Dioxide by Direct Oxidation of Titanium Powder, a Precursor for Ceramic Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Lukens, Wayne W. [Chemical; Saslow, Sarah A. [Earth

    2017-11-17

    Technetium-99 (Tc) is a problematic fission product that complicates the long-term disposal of nuclear waste due to its long half-life, high fission yield, and the environmental mobility of pertechnetate, its stable form in aerobic environments. One approach to preventing Tc contamination is through incorporation into durable waste forms based on weathering-resistant minerals such as rutile (titanium dioxide). Here, the incorporation of technetium into titanium dioxide by means of simple, aqueous chemistry is presented. X-ray absorption fine structure spectroscopy and diffuse reflectance spectroscopy indicate that Tc(IV) replaces Ti(IV) within the structure. Rather than being incorporated as isolated Tc(IV) ions, Tc is present as pairs of edge-sharing Tc(IV) octahedra similar to molecular Tc(IV) complexes such as [(H2EDTA)TcIV](u-O)2. Technetium-doped TiO2 was suspended in deionized water under aerobic conditions, and the Tc leached under these conditions was followed for 8 months. The normalized release rate of Tc (LRTc) from the TiO2 particles is low (3×10-6 g m-2 d-1), which illustrates the potential utility of TiO2 as waste form. However, the small size of the as-prepared TiO2 nanoparticles results in estimated retention of Tc for 104 years, which is only a fraction of the half-life of Tc (2×10-5 years).

  3. Optimization of strontium adsorption from aqueous solution using (mn-Zr) oxide-pan composite spheres

    International Nuclear Information System (INIS)

    Inan, S.; Altas, Y.

    2009-01-01

    The processes based on adsorption and ion exchange have a great role for the pre-concentration and separation of toxic, long lived radionuclides from liquid waste. In Nuclear waste management, the removal of long lived, radiotoxic isotopes from radioactive waste such as strontium reduces the storage problems and facilitates the disposal of the waste. Depending on the waste type, a variety of adsorbents and/or ion exchangers are used. Due to the amorphous structure of hydrous oxides and their mixtures, they don't have reproducible properties. Besides, obtained powders are very fine particles and they can cause operational problems such as pressure drop and filtration. Therefore they are not suitable for column applications. These reasons have recently expedited the study on the preparation of organic-inorganic composite adsorbent beads for industrial applications. PAN, as a stable and porous support for fine particles, provides the utilization of ion exchangers in large scale column applications. The utilization of PAN as a support material with many inorganic ion exchangers was firstly achieved by Sebesta in the beginning of 1990's. Later on, PAN based composite ion exchangers were prepared and used for the removal of radionuclides and heavy metal ions from aqueous solution and waste waters. In this study, spherical (Mn-Zr)oxide-PAN composite were prepared for separation of strontium from aqueous solution in a wide pH range. Sr 2 + adsorption of composite adsorbent was optimized by using experimental design 'Central Composite Design' model.

  4. Decontamination of alpha contaminated metallic waste by cerium IV redox process

    International Nuclear Information System (INIS)

    Shah, J.G.; Dhami, P.S.; Gandhi, P.M.; Wattal, P.K.

    2012-01-01

    Decontamination of alpha contaminated metallic waste is an important aspect in the management of waste generated during dismantling and decommissioning of nuclear facilities. Present work on cerium redox process targets decontamination of alpha contaminated metallic waste till it qualifies for the non alpha waste category for disposal in near surface disposal facility. Recovery of the alpha radio nuclides and cerium from aqueous secondary waste streams was also studied deploying solvent extraction process and established. The alpha-lean secondary waste stream has been immobilised in cement based matrix for final disposal. (author)

  5. Efficient sample preparation method based on solvent-assisted dispersive solid-phase extraction for the trace detection of butachlor in urine and waste water samples.

    Science.gov (United States)

    Aladaghlo, Zolfaghar; Fakhari, Alireza; Behbahani, Mohammad

    2016-10-01

    In this work, an efficient sample preparation method termed solvent-assisted dispersive solid-phase extraction was applied. The used sample preparation method was based on the dispersion of the sorbent (benzophenone) into the aqueous sample to maximize the interaction surface. In this approach, the dispersion of the sorbent at a very low milligram level was achieved by inserting a solution of the sorbent and disperser solvent into the aqueous sample. The cloudy solution created from the dispersion of the sorbent in the bulk aqueous sample. After pre-concentration of the butachlor, the cloudy solution was centrifuged and butachlor in the sediment phase dissolved in ethanol and determined by gas chromatography with flame ionization detection. Under the optimized conditions (solution pH = 7.0, sorbent: benzophenone, 2%, disperser solvent: ethanol, 500 μL, centrifuged at 4000 rpm for 3 min), the method detection limit for butachlor was 2, 3 and 3 μg/L for distilled water, waste water, and urine sample, respectively. Furthermore, the preconcentration factor was 198.8, 175.0, and 174.2 in distilled water, waste water, and urine sample, respectively. Solvent-assisted dispersive solid-phase extraction was successfully used for the trace monitoring of butachlor in urine and waste water samples. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  6. Purification of pectinase from mango (Mangifera indica L. cv. Chokanan) waste using an aqueous organic phase system: a potential low cost source of the enzyme.

    Science.gov (United States)

    Amid, Mehrnoush; Abdul Manap, Mohd Yazid; Mustafa, Shuhaimi

    2013-07-15

    As a novel method of purification, an aqueous organic phase system (AOPS) was employed to purify pectinase from mango waste. The effect of different parameters, such as the alcohol concentration (ethanol, 1-propanol, and 2-propanol), the salt type and concentration (ammonium sulfate, potassium phosphate and sodium citrate), the feed stock crude load, the aqueous phase pH and NaCl concentration, were investigated in the recovery of pectinase from mango peel. The partition coefficient (K), selectivity (S), purification factor (PF) and yield (Y, %) were investigated in this study as important parameters for the evaluation of enzyme recovery. The desirable partition efficiency for pectinase purification was achieved in an AOPS of 19% (w/w) ethanol and 22% (w/w) potassium phosphate in the presence of 5% (w/w) NaCl at pH 7.0. Based on the system, the purification factor of pectinase was enhanced 11.7, with a high yield of 97.1%. Copyright © 2013 Elsevier B.V. All rights reserved.

  7. Durability and degradation of HT9 based alloy waste forms with variable Ni and Cr content

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-31

    Short-term electrochemical and long-term hybrid electrochemical corrosion tests were performed on alloy waste forms in reference aqueous solutions that bound postulated repository conditions. The alloy waste forms investigated represent candidate formulations that can be produced with advanced electrochemical treatment of used nuclear fuel. The studies helped to better understand the alloy waste form durability with differing concentrations of nickel and chromium, species that can be added to alloy waste forms to potentially increase their durability and decrease radionuclide release into the environment.

  8. Acid-base behavior in hydrothermal processing of wastes. 1997 annual progress report

    International Nuclear Information System (INIS)

    1997-01-01

    'A major obstacle to the development of hydrothermal technology for treating DOE wastes has been a lack of scientific knowledge of solution chemistry, thermodynamics and transport phenomena. The progress over the last year is highlighted in the following four abstracts from manuscripts which have been submitted to journals. The authors also have made considerable progress on a spectroscopic study of the acid-base equilibria of Cr(VI). They have utilized novel spectroscopic indicators to study acid-base equilibria up to 380 C. Until now, very few systems have been studied at such high temperatures, although this information is vital for hydrothermal processing of wastes. The pH values of aqueous solutions of boric acid and KOH were measured with the optical indicator 2-naphthol at temperatures from 300 to 380 C. The equilibrium constant Kb-l for the reaction B(OH)3 + OH - = B(OH) -4 was determined from the pH measurements and correlated with a modified Born model. The titration curve for the addition of HCl to sodium borate exhibits strong acid-strong base behavior even at 350 C and 24.1 MPa. At these conditions, aqueous solutions of sodium borate buffer the pH at 9.6 t 0.25. submitted to Ind. Eng. Chem. Res. Acetic Acid and HCl Acid-base titrations for the KOH-acetic acid or NH 3 -acetic acid systems were monitored with the optical indicator 2-naphthoic acid at 350 C and 34 MPa, and those for the HCl;Cl- system with acridine at 380 C and up to 34 MPa (5,000 psia ). KOH remains a much stronger base than NH,OH at high temperature. From 298 K to the critical temperature of water, the dissociation constant for HCl decreases by 13 orders of magnitude, and thus, the basicity of Cl - becomes significant. Consequently, the addition of NaCl to HCl raises the pH. The pH titration curves may be predicted with reasonable accuracy from the relevant equilibrium constants and Pitzer''s formulation of the Debye- Htickel equation for the activity coefficients.'

  9. Metal separations using aqueous biphasic partitioning systems

    International Nuclear Information System (INIS)

    Chaiko, D.J.; Zaslavsky, B.; Rollins, A.N.; Vojta, Y.; Gartelmann, J.; Mego, W.

    1996-01-01

    Aqueous biphasic extraction (ABE) processes offer the potential for low-cost, highly selective separations. This countercurrent extraction technique involves selective partitioning of either dissolved solutes or ultrafine particulates between two immiscible aqueous phases. The extraction systems that the authors have studied are generated by combining an aqueous salt solution with an aqueous polymer solution. They have examined a wide range of applications for ABE, including the treatment of solid and liquid nuclear wastes, decontamination of soils, and processing of mineral ores. They have also conducted fundamental studies of solution microstructure using small angle neutron scattering (SANS). In this report they review the physicochemical fundamentals of aqueous biphase formation and discuss the development and scaleup of ABE processes for environmental remediation

  10. Zirconium phosphate waste forms for low-temperature stabilization of cesium-137-containing waste streams

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Tlustochowicz.

    1996-04-01

    Novel chemically bonded phosphate ceramics are being developed and fabricated for low-temperature stabilization and solidification of waste streams that are not amenable to conventional high-temperature stabilization processes because volatiles are present in the wastes. A composite of zirconium-magnesium phosphate has been developed and shown to stabilize ash waste contaminated with a radioactive surrogate of 137 Cs. Excellent retainment of cesium in the phosphate matrix system was observed in Toxicity Characteristic Leaching Procedure tests. This was attributed to the capture of cesium in the layered zirconium phosphate structure by intercalation ion-exchange reaction. But because zirconium phosphate has low strength, a novel zirconium/magnesium phosphate composite waste form system was developed. The performance of these final waste forms, as indicated by compression strength and durability in aqueous environments, satisfy the regulatory criteria. Test results indicate that zirconium-magnesium-phosphate-based final waste forms present a viable technology for treatment and solidification of cesium-contaminated wastes

  11. Waste-based alternative adsorbents for the remediation of pharmaceutical contaminated waters: Has a step forward already been taken?

    Science.gov (United States)

    Silva, Carla Patrícia; Jaria, Guilaine; Otero, Marta; Esteves, Valdemar I; Calisto, Vânia

    2018-02-01

    When adsorption is considered for water treatment, commercial activated carbon is usually the chosen adsorbent for the removal of pollutants from the aqueous phase, particularly pharmaceuticals. In order to decrease costs and save natural resources, attempts have been made to use wastes as raw materials for the production of alternative carbon adsorbents. This approach intends to increase efficiency, cost-effectiveness, and also to propose an alternative and sustainable way for the valorization/management of residues. This review aims to provide an overview on waste-based adsorbents used on pharmaceuticals' adsorption. Experimental facts related to the adsorption behaviour of each adsorbent/pharmaceutical pair and some key factors were addressed. Also, research gaps that subsist in this research area, as well as future needs, were identified. Simultaneously, this review aims to clarify the current status of the research on pharmaceuticals' adsorption by waste-based adsorbents in order to recognize if the right direction is being taken. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Fabrication of textured Ni–9.3at.%W substrate by electropulsing intermediate annealing method

    International Nuclear Information System (INIS)

    Liu, Jianan; Liu, Wei; Tang, Guoyi; Zhu, Rufei

    2014-01-01

    Highlights: •It’s the first time that EIA is used on Ni9 W substrate production. •Compared with CIA, EIA trends to sharpen the rolling texture. •Improved cube recrystallization texture is obtained by EIA. •EIA provides a highly efficient approach for Ni9 W substrate manufacture. -- Abstract: Sharp cube texture is difficult to obtain in high W content Ni–W alloy substrates used for coated conductors. In this paper, a new method called electropulsing intermediate annealing (EIA) is adopted to optimize the rolling and recrystallization texture of Ni–9.3 at.%W substrate. It is found that, compared with conventional intermediate annealing (CIA) at the same temperature, EIA trends to increase the Copper, S and Brass components, suppress the Goss component in rolling texture. Higher cube recrystallization texture is obtained at relatively low temperature by EIA in a shorter time. The effect of EIA on texture is attributed to the enhancement of recovery process resulting from the athermal effects

  13. Cross flow filtration of aqueous radioactive tank wastes

    International Nuclear Information System (INIS)

    McCabe, D.J.; Reynolds, B.A.; Todd, T.A.; Wilson, J.H.

    1997-01-01

    The Tank Focus Area (TFA) of the Department of Energy (DOE) Office of Science and Technology addresses remediation of radioactive waste currently stored in underground tanks. Baseline technologies for treatment of tank waste can be categorized into three types of solid liquid separation: (a) removal of radioactive species that have been absorbed or precipitated, (b) pretreatment, and (c) volume reduction of sludge and wash water. Solids formed from precipitation or absorption of radioactive ions require separation from the liquid phase to permit treatment of the liquid as Low Level Waste. This basic process is used for decontamination of tank waste at the Savannah River Site (SRS). Ion exchange of radioactive ions has been proposed for other tank wastes, requiring removal of insoluble solids to prevent bed fouling and downstream contamination. Additionally, volume reduction of washed sludge solids would reduce the tank space required for interim storage of High Level Wastes. The scope of this multi-site task is to evaluate the solid/liquid separations needed to permit treatment of tank wastes to accomplish these goals. Testing has emphasized cross now filtration with metal filters to pretreat tank wastes, due to tolerance of radiation and caustic

  14. Non-aqueous nanoporous gold based supercapacitors with high specific energy

    International Nuclear Information System (INIS)

    Hou, Ying; Chen, Luyang; Hirata, Akihiko; Fujita, Takeshi; Chen, Mingwei

    2016-01-01

    In this study, we report that the supercapacitor performance of polypyrrole (PPy) in non-aqueous electrolytes can be dramatically improved by highly conductive nanoporous gold which acts as both the support of active PPy and the current collector of supercapacitors. The excellent electronic conductivity, rich porous structure and large surface area of the nanoporous electrodes give rise to a high specific capacitance and low internal resistance in non-aqueous electrolytes. Combining with a wide working potential window of ~ 2 V, the non-aqueous PPy-based supercapacitors show an extraordinary energy density and power density.

  15. Annual radioactive waste tank inspection program: 1995

    International Nuclear Information System (INIS)

    McNatt, F.G. Sr.

    1996-01-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1995 to evaluate these vessels and evaluations based on data accrued by inspections performed since the tanks were constructed are the subject of this report

  16. Annual radioactive waste tank inspection program - 1992

    International Nuclear Information System (INIS)

    McNatt, F.G.

    1992-01-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1992 to evaluate these vessels and evaluations based on data accrued by inspections made since the tanks were constructed are the subject of this report

  17. Annual radioactive waste tank inspection program - 1991

    International Nuclear Information System (INIS)

    McNatt, F.G.

    1992-01-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1991 to evaluate these vessels and evaluations based on data accrued by inspections made since the tanks were constructed are the subject of this report

  18. Salts-based size-selective precipitation: toward mass precipitation of aqueous nanoparticles.

    Science.gov (United States)

    Wang, Chun-Lei; Fang, Min; Xu, Shu-Hong; Cui, Yi-Ping

    2010-01-19

    Purification is a necessary step before the application of nanocrystals (NCs), since the excess matter in nanoparticles solution usually causes a disadvantage to their subsequent coupling or assembling with other materials. In this work, a novel salts-based precipitation technique is originally developed for the precipitation and size-selective precipitation of aqueous NCs. Simply by addition of salts, NCs can be precipitated from the solution. After decantation of the supernatant solution, the precipitates can be dispersed in water again. By means of adjusting the addition amount of salt, size-selective precipitation of aqueous NCs can be achieved. Namely, the NCs with large size are precipitated preferentially, leaving small NCs in solution. Compared with the traditional nonsolvents-based precipitation technique, the current one is simpler and more rapid due to the avoidance of condensation and heating manipulations used in the traditional precipitation process. Moreover, the salts-based precipitation technique was generally available for the precipitation of aqueous nanoparticles, no matter if there were semiconductor NCs or metal nanoparticles. Simultaneously, the cost of the current method is also much lower than that of the traditional nonsolvents-based precipitation technique, making it applicable for mass purification of aqueous NCs.

  19. Immobilization of organic liquid wastes

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1985-01-01

    This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a a FFTF fuels reprocessing line. An incineration system already on-hand was determined to be too costly to operate for the 100 to 400 gallons per year organic liquid. Organic test liquids were dispersed into an aqueous phosphate liquid using an emulsifier. A total of 109 gallons of potential and radioactive aqueous immiscible organic liquid wastes from Hanford 300 Area operations were solidified with portland cement and disposed of as solid waste during a 3-month test program with in-drum mixers. Waste packing efficiencies varied from 32 to 40% and included pump oils, mineral spirits, and TBP-NPH type solvents

  20. The IUPAC aqueous and non-aqueous experimental pKa data repositories of organic acids and bases.

    Science.gov (United States)

    Slater, Anthony Michael

    2014-10-01

    Accurate and well-curated experimental pKa data of organic acids and bases in both aqueous and non-aqueous media are invaluable in many areas of chemical research, including pharmaceutical, agrochemical, specialty chemical and property prediction research. In pharmaceutical research, pKa data are relevant in ligand design, protein binding, absorption, distribution, metabolism, elimination as well as solubility and dissolution rate. The pKa data compilations of the International Union of Pure and Applied Chemistry, originally in book form, have been carefully converted into computer-readable form, with value being added in the process, in the form of ionisation assignments and tautomer enumeration. These compilations offer a broad range of chemistry in both aqueous and non-aqueous media and the experimental conditions and original reference for all pKa determinations are supplied. The statistics for these compilations are presented and the utility of the computer-readable form of these compilations is examined in comparison to other pKa compilations. Finally, information is provided about how to access these databases.

  1. Development of nuclear transmutation technology - A study on accelerator-driven transmutation of long-lived radionuclide

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; Chung, Kie Hyung; Hong, Sang Hee; Hwang, Il Soon; Park, Byung Gi; Yang, Hyung Lyeol; Kim, Duk Kyu; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

    1996-07-01

    The objective of this study is to help establish the long-range nuclear waste disposal strategy through the investigations and comparisons of various= concepts of the accelerator-driven nuclear waste transmutation reactors, which have been suggested to replace the geological waste disposal due to the technical uncertainties in the long-time scale. Nuclear data, categorized in high -and low-energy neutron cross-sections, were investigated and the structures, principles, and recent progresses of proton linac were reviews, Also the accelerator power for transmutation and the economics were referred, The comparison of the transmutation concepts concentrated on two: Japanese OMEGA program of alloy fuelled system, Minor actinide molten salt system, and Eutectic alloy system and American ATW program of aqueous system and molten salt system. From the comparative study, a state-of-art of the technology has been identified as a concept employing proton-accelerate of 800 {approx} 1600 MeV with 100 mA capacity combined with liquid lead target, molten salt blanket and on-line chemical separation using centrifuge and electrowinning technology. 34 refs., 25 tabs., 64 figs. (author)

  2. NRC Fact-Finding Task Force report on the ATWS event at Salem Nuclear Generating Station, Unit 1, on February 25, 1983

    International Nuclear Information System (INIS)

    1983-03-01

    An NRC Region I Task Force was established on March 1, 1983 to conduct fact finding and data collection with regard to the circumstances which led to an anticipated transient without scram (ATWS) event at the Public Service Electric and Gas Company's Salem Generating Station, Unit 1 on February 25, 1983. The charter of the Task Force was to determine the factual information pertinent to management and administrative controls which should have ensured proper operation of the reactor trip breakers in the solid state protection system. This report documents the findings of the Task Force along with its conclusions

  3. Liquid secondary waste. Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-01

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testing to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford’s IDF.

  4. 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.

  5. Conditioning of tritiated wastes. Part II

    International Nuclear Information System (INIS)

    Hawthorne, S.H.

    1984-01-01

    Work is continuing on the development of conditioning systems for low and intermediate level tritiated liquid and solid wastes which will prevent loss of tritium for at least 150 years. This portion of the program has concentrated on solidification and encapsulation of tritiated aqueous wastes, development of techniques, for the measurement of tritium loss in air and water, and identification and evaluation of encapsulation materials. Solidification of tritiated aqueous wastes by water extendible polyester or cements resulted in average tritium releases of approximately 1-4x10 -1 α/day with that from water extendible polyester being the lowest. The daily release rate is independent of initial tritium concentration in the waste form and can be reduced by a factor of 1000 by encapsultation of the waste within a 10 mm layer of water extendible polyester. Water extendible polyester is the preferred material for solidification and encapsulation of aqueous tritiated wastes and encapsulation of tritiated solids permitting release of only 3x10 -3 % of the original activity over 150 years. It is expected that this program which was originally scheduled for three years can now be completed in two years with complete definition of the conditioning system including the outer package

  6. Annual radioactive waste tank inspection program - 1999

    International Nuclear Information System (INIS)

    Moore, C.J.

    2000-01-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1999 to evaluate these vessels and auxiliary appurtenances along with evaluations based on data accrued by inspections performed since the tanks were constructed are the subject of this report

  7. Novel Fiber-Based Adsorbent Technology; FINAL

    International Nuclear Information System (INIS)

    Nixon, P.G.; Tsukamoto, T.; Brose, D.J.

    2001-01-01

    The overall of this Department of Energy (DOE) Phase II SBIR program was to develop a new class of highly robust fiber-based adsorbents for recovery of heavy metals from aqueous waste-streams. The fiber-based adsorbents,when commercialized,will be used for clean up metals in aqueous waste-streams emanating from DOE facilities,industry,mining,and groundwater-cleanup operations.The amount of toxic waste released by these streams is of great significance.The U.S.Environment Protection Agency (EPA) reports that in 1990 alone,4.8 billion pounds of toxic chemicals were released into the environment.Of this waste,the metals-containing waste was the second largest contributor,representing 569 million pounds. This report presents the results of the Phase II program,which successfully synthesized noval fiber-based adsorbents for the removal of Group 12 metals(i.e.mercury),Group 14 metals (lead),and Group 10 metals(platinum and palladium) from contaminated groundwater and industrial waste streams.These fiber-based adsorbents are ideally suited for the recovery of metal ions from aqueous waste streams presently not treatable due to the degrading nature of corrosive chemicals or radioactive components in the feed stream. The adsorbents developed in this program rely on chemically resistant and robust carbon fibers and fabrics as supports for metal-ion selective ligands.These adsorbents demonstrate loading capacities and selectivities for metal ions exceeding those of conventional ion-exchange resins.The adsorbents were also used to construct filter modules that demonstrate minimal fouling,minimal compaction,chemical and physical robustness,and regeneration of metal loading capacity without loss of performance

  8. Magnetic nanoparticle (Fe3O4) impregnated onto tea waste for the removal of nickel(II) from aqueous solution

    International Nuclear Information System (INIS)

    Panneerselvam, P.; Morad, Norhashimah; Tan, Kah Aik

    2011-01-01

    The removal of Ni(II) from aqueous solution by magnetic nanoparticles prepared and impregnated onto tea waste (Fe 3 O 4 -TW) from agriculture biomass was investigated. Magnetic nanoparticles (Fe 3 O 4 ) were prepared by chemical precipitation of a Fe 2+ and Fe 3+ salts from aqueous solution by ammonia solution. These magnetic nanoparticles of the adsorbent Fe 3 O 4 were characterized by surface area (BET), Scanning Electron Microscopy (SEM), Transmission Electron Microscopy (TEM) and Fourier Transform-Infrared Spectroscopy (FT-IR). The effects of various parameters, such as contact time, pH, concentration, adsorbent dosage and temperature were studied. The kinetics followed is first order in nature, and the value of rate constant was found to be 1.90 x 10 -2 min -1 at 100 mg L -1 and 303 K. Removal efficiency decreases from 99 to 87% by increasing the concentration of Ni(II) in solution from 50 to 100 mg L -1 . It was found that the adsorption of Ni(II) increases by increasing temperature from 303 to 323 K and the process is endothermic in nature. The adsorption isotherm data were fitted to Langmuir and Freundlich equation, and the Langmuir adsorption capacity, Q o , was found to be (38.3) mg g -1 . The results also revealed that nanoparticle impregnated onto tea waste from agriculture biomass, can be an attractive option for metal removal from industrial effluent.

  9. The immobilization of organic liquid wastes

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1986-01-01

    This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a FFTF fuels reprocessing line. An incineration system already on-hand was determined to be too costly to operate for the 100 to 400 gallons per year organic liquid. Organic test liquids were dispersed into an aqueous phosphate liquid using an emulsifier. A total of 109 gallons of potential and radioactive aqueous immiscible organic liquid wastes from Hanford 300 Area operations were solidified with portland cement and disposed of as solid waste during a 3 month test program with in-drum mixers. Waste packing efficiencies varied from 32 to 40% and included pump oils, mineral spirits, and TBP-NPH type solvents

  10. Backfill barriers for nuclear waste repositories in salt

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-06-01

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

  11. Containment venting as a mitigation technique for BWR MARK I plant ATWS

    International Nuclear Information System (INIS)

    Harrington, R.M.

    1987-01-01

    Containment venting is studied as a mitigation strategy for preventing or delaying severe fuel damage following hypothetical BWR Anticipated Transient Without Scram (ATWS) accidents initiated by MSIV-closure, and compounded by failure of the Standby Liquid Control (SLC) system injection of sodium pentaborate solution and by the failure of manually initiated control rod insertion. The venting of primary containment after reaching 75 psia (0.52 MPa) is found to result in the release of the vented steam inside the reactor building, and to result in inadequate Net Positive Suction Head (NPSH) for any system pumping from the pressure suppression pool. CONTAIN code calculations show that personnel access to large portions of the reactor building would be lost soon after the initiation of venting and that the temperatures reached would be likely to result in independent equipment failures. It is concluded that containment venting would be more likely to cause or to hasten the onset of severe fuel damage than to prevent or to delay it. Two alternative strategies that do not require containment venting, but that could delay or prevent severe fuel damage, are analyzed. BWR-LTAS code results are presented for a successful mitigation strategy in which the reactor vessel is depressurized, and for one in which the reactor vessel remains at pressure

  12. Aqueous solutions/nuclear glasses interactions

    International Nuclear Information System (INIS)

    Delage, F.; Advocat, T.; Vernaz, E.; Crovisier, J.L.

    1991-01-01

    Interactions results of the borosilicate glass used in radioactive wastes confinement and aqueous solutions at various temperature and PH show that for the glass components: - the release rate evolution follows an Arrhenius law, - in acid PH, there is a selective dissolution, - in basic PH, there is a stoechiometric dissolution [fr

  13. NOCHAR Polymers: An Aqueous and Organic Liquid Solidification Process for Cadarache LOR (Liquides Organiques Radioactifs) - 13195

    International Nuclear Information System (INIS)

    Vaudey, Claire-Emilie; Renou, Sebastien; Porco, Julien; Kelley, Dennis; Cochaud, Chantal; Serrano, Roger

    2013-01-01

    To handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW) in France, two options can be considered: the incineration at CENTRACO facility and the disposal facility on ANDRA sites. The waste acceptance in these radwaste routes is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the radwaste route specifications. If the waste characteristics are incompatible with the radwaste route specifications (presence of significant quantities of chlorine, fluorine, organic component etc or/and high activity limits), it is necessary to find an alternative solution that consists of a waste pre-treatment process. In the context of the problematic Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. The first one is composed of organic liquids at 13.1 % (diphenyloxazol, mesitylene, TBP, xylene) and water at 86.9 %. The second one is composed of TBP at 8.6 % and water at 91.4 %. They contain chlorine, fluorine and sulphate and have got alpha/beta/gamma spectra with mass activities equal to some kBq.g -1 . Therefore, tritium is present and creates the second problematic waste stream. As a consequence, in order for disposal acceptance at the ANDRA site, it is necessary to pre-treat the waste. The NOCHAR polymers as an aqueous and organic liquid solidification process seem to be an adequate solution. Indeed, these polymers constitute an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing etc) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilise the liquid. Then as the

  14. 40 CFR 227.30 - High-level radioactive waste.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated waste from...

  15. Thermal-hydraulic analysis of graphite tubes for the non-aqueous system of accelerator transmutation of nuclear waste

    International Nuclear Information System (INIS)

    Potter, R.C.; Venneri, F.; Trujillo, D.A.

    1993-01-01

    Accelerator transmutation of nuclear waste offers exciting possibilities for the disposal of nuclear waste by converting it into more benign Species. The non-aqueous system discussed here contains the materials to be transmuted within a lithium-fluoride salt. The system consists of bundles of graphite tubes containing the salt Solution. The tubes are cooled as lithium flows across their exterior. These circular graphite tubes have an inner circular passage and an outer annulus. Natural convection within the tubes causes the salt to circulate. This paper deals with the thermal-hydraulics of the system; it does not consider the neutronics in detail. Heat transfer and fluid flow were modeled using a custom computer program the system behavior of an graphite tube. Different geometries were tried, while keeping the system volume the same, to determine an optimize graphite tube geometry. I considered both the parallel flow and the counterflow of the lithium coolant, and allowed limited boiling to occur to facilitate circulation. I achieved power densities as high as 200 W/cm 3 for the overall blanket

  16. Chemical Remediation of Nickel(II) Waste: A Laboratory Experiment for General Chemistry Students

    Science.gov (United States)

    Corcoran, K. Blake; Rood, Brian E.; Trogden, Bridget G.

    2011-01-01

    This project involved developing a method to remediate large quantities of aqueous waste from a general chemistry laboratory experiment. Aqueous Ni(II) waste from a general chemistry laboratory experiment was converted into solid nickel hydroxide hydrate with a substantial decrease in waste volume. The remediation method was developed for a…

  17. The German act on the reorganisation of responsibility in nuclear waste management

    International Nuclear Information System (INIS)

    Raetzke, Christian

    2017-01-01

    The author discussed the Draft on the Act in the Reorganisation of Responsibility in Nuclear Waste Management in atw 12 (2016). Now, amendments are discussed, which resulted from the legislative procedure until today's draft. Significant additions affect the authorisation for the conclusion of a public-law contract between the Federal Government and the nuclear power plant operators, the deadline for the payment of the basic amount, and the option for the operation of the interim storage facilities for a transitional period by the operators on behalf of the federal company. Since the adoption of the draft act, it has become clear that the nuclear power plant operators will pay the risk premium. This will fulfil the full logic of the new system. It has also become known, that the public law contract is now ready for signing. According to the author, the act will bring a final arrangement for financing nuclear waste disposal. However, adjustment can not be avoided in practice. The concrete implementation will be a exciting topic in many ways.

  18. Crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics for immobilization of simulated sulfate bearing high-level liquid waste

    Science.gov (United States)

    Wu, Lang; Xiao, Jizong; Wang, Xin; Teng, Yuancheng; Li, Yuxiang; Liao, Qilong

    2018-01-01

    The crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics with different content (0-30 wt %) of simulated sulfate bearing high-level liquid waste (HLLW) were evaluated. The sulfate phase segregation in vitrification process was also investigated. The results show that the glass-ceramics with 0-20 wt% of HLLW possess mainly zirconolite phase along with a small amount baddeleyite phase. The amount of perovskite crystals increases while the amount of zirconolite crystals decreases when the HLLW content increases from 20 to 30 wt%. For the samples with 20-30 wt% HLLW, yellow phase was observed during the vitrification process and it disappeared after melting at 1150 °C for 2 h. The viscosity of the sample with 16 wt% HLLW (HLLW-16) is about 27 dPa·s at 1150 °C. The addition of a certain amount (≤20 wt %) of HLLW has no significant change on the aqueous stability of glass-ceramic waste forms. After 28 days, the 90 °C PCT-type normalized leaching rates of Na, B, Si, and La of the sample HLLW-16 are 7.23 × 10-3, 1.57 × 10-3, 8.06 × 10-4, and 1.23 × 10-4 g·m-2·d-1, respectively.

  19. Characterisation of aqueous waste produced during the clandestine production of amphetamine following the Leuckart route utilising solid-phase extraction gas chromatography-mass spectrometry and capillary electrophoresis with contactless conductivity detection.

    Science.gov (United States)

    Hauser, Frank M; Hulshof, Janneke W; Rößler, Thorsten; Zimmermann, Ralf; Pütz, Michael

    2018-04-18

    Chemical waste from the clandestine production of amphetamine is of forensic and environmental importance due to its illegal nature which often leads to dumping into the environment. In this study, 27 aqueous amphetamine waste samples from controlled Leuckart reactions performed in Germany, the Netherlands, and Poland were characterised to increase knowledge about the chemical composition and physicochemical characteristics of such waste. Aqueous waste samples from different reaction steps were analysed to determine characteristic patterns which could be used for classification. Conductivity, pH, density, ionic load, and organic compounds were determined using different analytical methods. Conductivity values ranged from 1 to over 200 mS/cm, pH values from 0 to 14, and densities from 1.0 to 1.3 g/cm 3 . A capillary electrophoresis method with contactless conductivity detection (CE-C 4 D) was developed and validated to quantify chloride, sulphate, formate, ammonium, and sodium ions which were the most abundant ions in the investigated waste samples. A solid-phase extraction sample preparation was used prior to gas chromatography-mass spectrometry analysis to determine the organic compounds. Using the characterisation data of the known samples, it was possible to assign 16 seized clandestine waste samples from an amphetamine production to the corresponding synthesis step. The data also allowed us to draw conclusions about the synthesis procedure and used chemicals. The presented data and methods could support forensic investigations by showing the probative value of synthesis waste when investigating the illegal production of amphetamine. It can also act as starting point to develop new approaches to tackle the problem of clandestine waste dumping. Copyright © 2018 John Wiley & Sons, Ltd.

  20. Parallel proton transfer pathways in aqueous acid-base reactions

    NARCIS (Netherlands)

    Cox, M.J.; Bakker, H.J.

    2008-01-01

    We study the mechanism of proton transfer (PT) between the photoacid 8-hydroxy-1,3, 6-pyrenetrisulfonic acid (HPTS) and the base chloroacetate in aqueous solution. We investigate both proton and deuteron transfer reactions in solutions with base concentrations ranging from 0.25M to 4M. Using

  1. Mixed waste treatment using the ChemChar thermolytic detoxification technique

    Energy Technology Data Exchange (ETDEWEB)

    Kuchynka, D. [Mirage Systems, Sunnyvale, CA (United States)

    1995-10-01

    The diversity of mixed waste matrices contained at Department of Energy sites that require treatment preclude a single, universal treatment technology capable of handling sludges, solids, heterogeneous debris, aqueous and organic liquids and soils. This report describes the ChemChar thermolytic detoxification process. The process is a thermal, chemically reductive technology that converts the organic portion of mixed wastes to a synthesis gas, while simultaneously absorbing volatile inorganics on a carbon-based char.

  2. Oak Ridge National Laboratory Melton Valley Storage Tanks Waste Filtration Process Evaluation

    International Nuclear Information System (INIS)

    Walker, B.W.

    1998-01-01

    Cross-flow filtration is being evaluated as a pretreatment in the proposed treatment processes for aqueous high-level radioactive wastes at Oak Ridge National Laboratory (ORNL) to separate insoluble solids from aqueous waste from the Melton Valley Storage Tanks (MVST)

  3. Removal of Phenol in Aqueous Solution Using Kaolin Mineral Clay

    International Nuclear Information System (INIS)

    Sayed, M.S.

    2008-01-01

    Kaolin clay were tested for phenol removal as toxic liquid waste from aqueous waste water. Several experimental conditions such as weight and particle size of clay were investigated to study batch kinetic techniques, also the ph and concentration of the phenol solution were carried out. The stability of the Langmuir adsorption model of the equilibrium data were studied for phenol sorbent clay system. Infrared spectra, thermogravimetric and differential thermal analysis techniques were used to characterize the behavior of kaolin clay and kaolin clay saturated with phenol. The results obtained showed that kaolin clay could be used successfully as an efficient sorbent material to remove phenol from aqueous solution

  4. Treatment and minimization of heavy metal-containing wastes 1995

    International Nuclear Information System (INIS)

    Hager, J.P.; Mishra, B.; Litz, J.L.

    1995-01-01

    This symposium was held in conjunction with the 1995 Annual Meeting of the Minerals, Metals and Materials Society in Las Vegas, Nevada, February 12--16, 1995. The purpose of this meeting was to provide a forum for exchange of state-of-the-art information on treating and minimizing heavy metal-containing wastes. Papers were categorized under the following broad headings: aqueous processing; waste water treatment; thermal processing and stabilization; processing of fly ash, flue dusts, and slags; and processing of lead, mercury, and battery wastes. Individual papers have been processed separately for inclusion in the appropriate data bases

  5. Requirements for a radioactive waste data base

    International Nuclear Information System (INIS)

    Sato, Y.; Kobayashi, I.; Kikuchi, M.

    1990-01-01

    With the progress of nuclear fuel cycle in Japan, various types of radioactive waste will generate at each nuclear facility in the cycle. Therefor generated volume and stored quantity of waste will be supposed to increase. From the viewpoints of safety and public acceptance, by using mainframe computer it is necessary that the treatment of historical waste data, the estimation of generated waste volume and stored quantity and the investigation of research and development status of waste processing and disposal are carried out. This paper proposes design and development of the radioactive waste data base which is able to properly and correctly manage and grasp numerical and/or documentary information for generated radioactive waste. So the data base will be expected to use for planning the future management of radioactive waste. (author)

  6. AQUEOUS CLEANING OF PRINTED CIRCUIT BOARD STENCILS

    Science.gov (United States)

    The USEPA through NRMRL has partnered with the California Dept. of Toxic Substance Control under an ETV Pilot Project to verigy polllution prevention, recycling and waste treatment technologies. One of the projects selected for verification was the ultrasonic aqueous cleaning tec...

  7. The chemistry, waste form development, and properties of the Nitrate to Ammonia and Ceramic (NAC) process

    International Nuclear Information System (INIS)

    Mattus, A.J.; Lee, D.D.; Youngblood, E.L.; Walker, J.F. Jr.; Tiegs, T.N.

    1994-01-01

    A process for the conversion of alkaline, aqueous nitrate wastes to ammonia gas at low temperature, based upon the use of the active metal reductant aluminum, has been developed at the Oak Ridge National Laboratory (ORNL). The process is also well suited for the removal of low-level waste (LLW) radioelements and hazardous metals which report to the solid, alumina-based by-product. ne chemistry of the interaction of aluminum powders with nitrate, and other waste stream metals is presented

  8. New developments and improvements in processing of 'problematic' radioactive waste. Results of a coordinated research project 2003-2007

    International Nuclear Information System (INIS)

    2007-12-01

    This report addresses a category of wastes termed 'problematic wastes', wastes for which safe, efficient and cost effective methods for processing are not readily available. Processing options for many of these are identified and addressed. Results presented, illustrate the strategy for breaking 'problematic' waste streams down into a sequence of 'standard' issues which are amenable to solution. Decision makers and facility managers faced with problematic waste streams should be able to use this information to identify and pursue solutions to meet their needs. In this report, processing options for a total of 27 problematic waste streams that were identified and addressed by the individual laboratories participating in the Coordinated Research Project are discussed. These waste streams covered an extremely broad spectrum, ranging from simple, one component aqueous solutions originating from a research laboratory to very complex aqueous concentrates of waste resulting from reprocessing activities or reactor operation. These challenging wastes included: waste contaminated by tritium, wastes containing transuranic elements, and solid health care waste. The range of aqueous wastes included those contaminated by organic complexing agents and surfactants to pure organic waste such as contaminated oil. Correspondingly, the scale of approaches and technologies used to address these wastes is very broad. Use of this report is likely to be most effective as an initial screening tool to identify technologies best able to meet specific waste management objectives in terms of the waste generated, the technical complexity, the available economic resources, the environmental impact considerations, and the desired end product (output) of the technology. The report should assist the user to compare technologies and to reach an informed decision based on safety, technological maturity, economics, and other local needs

  9. Annual radioactive waste tank inspection program -- 1993

    International Nuclear Information System (INIS)

    McNatt, F.G. Sr.

    1994-05-01

    Aqueous radioactive wastes from Savannah River Site (SRS) separations processes are contained in large underground carbon steel tanks. Inspections made during 1993 to evaluate these vessels, and evaluations based on data accrued by inspections made since the tanks were constructed, are the subject of this report. The 1993 inspection program revealed that the condition of the Savannah River Site waste tanks had not changed significantly from that reported in the previous annual report. No new leaksites were observed. No evidence of corrosion or materials degradation was observed in the waste tanks. However, degradation was observed on covers of the concrete encasements for the out-of-service transfer lines to Tanks 1 through 8

  10. Research needs in cement-based waste forms

    International Nuclear Information System (INIS)

    McDaniel, E.W.; Spence, R.D.; Tallent, O.K.

    1990-01-01

    Cement-based waste forms are one of the most widely used waste disposal options, yet definitive knowledge of the fate of the waste species inside the waste form is lacking. A fundamental understanding of the chemistry and microstructure of the waste forms would lead to a better understanding of the mass transfer of the waste species, more confidence in predicting and extrapolating waste form performance, and design of better waste forms. Better and cheaper leach tests would lead to quicker and more cost effective screening of waste form alternatives. In addition, assessment of durability may be important to predicting waste form performance in the field. It should be noted that the research needs discussed in this report are from the perspective of investigators working in applied waste management areas, while the proposed investigations are fundamental or basic. Details as to experimental methods and tools to be used in achieving the objectives of the proposed are research beyond the scope of this paper and are better filled in by others. In broad terms, the research topics discussed are correlation of cement-based waste form physical properties to performance, waste-form fundamental chemistry and microstructure, and product performance testing

  11. Immobilization of radioactive waste through cementation using Cuban zeolitic rock as additive

    International Nuclear Information System (INIS)

    Chales Suarez, G.; Castillo Gomez, R.

    1997-01-01

    The cementation of both simulated and real low level aqueous wastes using Cuban zeolite as additive is described. Mechanical characteristics and leach testing of the cemented waste forms has been studied. The results obtained have shown that the presence of zeolite in the cemented waste for reduces considerably the leach rates of Cs and Co and moreover, mechanical characteristics (set time and compressive strength) are better when compared with direct cementation of aqueous wastes. (author). 13 refs, 8 tabs

  12. Aqueous corrosion of silicate glasses. Analogy between volcanic glasses and the French nuclear waste glass R7T7

    International Nuclear Information System (INIS)

    Goldschmidt, F.

    1991-01-01

    The behaviour of borosilicate glasses upon aqueous corrosion is controlled for long periods of time (>10,000 years) by processes which are not directly accessible by means of laboratory experiments. The analogical approach consists here to compare leaching performances between the french nuclear waste glass R7T7 and natural volcanic glasses, basaltic and rhyolitic ones. The three glasses were leached in the same conditions; open system, 90 deg C, initial pH of 9.7. Basaltic and R7T7 glasses having the same kinetic of dissolution, the basaltic glass was chosen as the best analogue. (author). refs., figs., tabs

  13. NEARSOL, Aqueous Speciation and Solubility of Actinides for Waste Disposal

    International Nuclear Information System (INIS)

    Leach, S.J.; Pryke, D.C.

    1989-01-01

    A - Description of program or function: NEARSOL models the aqueous speciation and solubility of actinides under near-field conditions for disposal using a simple thermodynamic approach. B - Method of solution: The program draws information from a thermodynamic data base consisting of solubility products and complex formation constants for all known species, and standard electrode potentials, at 25 C, corrected for ionic strength effects. By minimising the free energy of the system through a series of iterations, a precipitating solid phase is predicted which limits the solubility, and the concentration of the main aqueous species are calculated as a function of pH. Initially the program evaluates only hydroxide and carbonate species, but the effect of sulphate, phosphate and fluoride anions can also be included. The program is simple to use, requiring inputs of: 1. Actinide(s); 2. pH range; 3. Ionic strength; 4. Redox conditions; 5. Ligand concentrations. Functions are included to calculate the distribution of the protonated and un-protonated forms of carbonate and phosphate and the value of Eh as a function of pH under disposal conditions as required. The program can further evaluate the role of free calcium ions. C - Restrictions on the complexity of the problem: None

  14. Bioadsorption of a reactive dye from aqueous solution by municipal solid waste

    Directory of Open Access Journals (Sweden)

    Abdelkader Berrazoum

    2015-09-01

    Full Text Available The biosorbent was obtained from municipal solid waste (MSW of the Mostaganem city. Before use the MSW was dried in air for three days and washed several times. The sorption of yellow procion reactive dye MX-3R onto biomass from aqueous solution was investigated as function of pH, contact time and temperature. The adsorption capacity of MX-3R was 45.84 mg/g at pH 2–3 and room temperature. MX-3R adsorption decreases with increasing temperature. The Langmuir, Freundlich and Langmuir–Freundlich adsorption models were applied to describe the related isotherms. Langmuir–Freundlich equation has shown the best fitting with the experimental data. The pseudo first-order, pseudo second-order and intra-particle diffusion kinetic models were used to describe the kinetic sorption. The results clearly showed that the adsorption of MX-3R onto biosorbent followed the pseudo second-order model. The enthalpy (ΔH°, entropy (ΔS° and Gibbs free energy (ΔG° changes of adsorption were calculated. The results indicated that the adsorption of MX-3R occurs spontaneously as an exothermic process.

  15. Pitting morphologies of zirconium base alloys in aqueous and non aqueous chloride media

    International Nuclear Information System (INIS)

    Palit, G.C.; Gadiyar, H.S.

    1988-01-01

    Pitting morphology of zirconium and Zr-Cr alloys in aqueous chloride and nonaqueous methanol + 0.4 per cent HCl solution was investigated and observed to follow different modes in these two environments. While in aqueous chloride solution pitting was transgranular and randomly oriented, in methanol-chloride solution pits were observed to initiate and propagate along the grain boundaries. In aqueous chloride solution very irregular and sponge like zirconium metal was formed inside the pit while in methanol-chloride solution the pits were crystallographic in nature. Optical microscopy has revealed that pits preferentially initiate and propagate along scratch line in aqueous chloride solution, but such was not the case in nonaqueous methanol-chloride solution. The nature and the mechanism operating in the catastropic failure of these materials are investigated. (author). 10 refs., 11 figs

  16. Synthesis and functionalization of dextran-based single-chain nanoparticles in aqueous media

    OpenAIRE

    Gracia R.; Marradi M.; Cossío U.; Benito A.; Pérez-San Vicente A.; Gómez-Vallejo V.; Grande H.-J.; Llop J.; and Loinaz I.

    2017-01-01

    Water-dispersible dextran-based single-chain polymer nanoparticles (SCPNs) were prepared in aqueous media and under mild conditions. Radiolabeling of the resulting biocompatible materials allowed the study of lung deposition of aqueous aerosols after intratracheal nebulization by means of single-photon emission computed tomography (SPECT), demonstrating their potential use as imaging contrast agents.

  17. Solvent extraction of radionuclides from aqueous tank waste

    International Nuclear Information System (INIS)

    Bonnesen, P.V.; Sachleben, R.A.; Moyer, B.A.

    1996-01-01

    The purpose of this task is to develop an efficient solvent-extraction and stripping process for the removal of the fission products Tc-99, Sr-90, and Cs-137 from alkaline tank wastes, such as those stored at Hanford and Oak Ridge. As such, this task expands upon FY 1995's successful development of a solvent-extraction and stripping process for technetium separation from at sign e tank-waste solutions. This process has in fact already been extended to include the capability of removing both Tc and Sr simultaneously. In this form, the process has been given the name SRTALK and will be developed further in this program as a prelude to developing a system capable of removing Tc, Sr, and Cs together. Such a system could potentially simplify and improve fission-product removal from tank waste. In addition, it would possess the advantages already inherent in our Tc solvent-extraction process: No required feed adjustment, economical water stripping, low consumption of materials, and low waste volume

  18. Removal of acid blue 062 on aqueous solution using calcinated colemanite ore waste

    Energy Technology Data Exchange (ETDEWEB)

    Atar, Necip [Department of Chemistry, Faculty of Arts and Science, University of Dumlupinar, Kuetahya (Turkey); Olgun, Asim [Department of Chemistry, Faculty of Arts and Science, University of Dumlupinar, Kuetahya (Turkey)]. E-mail: aolgun@dumlupinar.edu.tr

    2007-07-19

    Colemanite ore waste (CW) has been employed as adsorbent for the removal of acid blue 062 anionic dye (AB 062) from aqueous solution. The adsorption of AB 062 onto CW was examined with respect to contact time, calcination temperature, particle size, pH, adsorbent dosage and temperature. The physical and chemical properties of the CW, such as particle sizes and calcinations temperature, play important roles in dye adsorption. The dye adsorption largely depends on the initial pH of the solution with maximum uptake occurring at pH 1.Three simplified kinetics models, namely, pseudo-first order, pseudo-second order, and intraparticle diffusion models were tested to investigate the adsorption mechanisms. The kinetic adsorption of AB 062 on CW follows a pseudo-second order equation. The adsorption data have been analyzed using Langmuir and Freundlich isotherms. The results indicate that the Langmuir model provides the best correlation of the experimental data. Isotherms have also been used to obtain the thermodynamic parameters such as free energy, enthalpy and entropy of the adsorption of dye onto CW.

  19. Removal of acid blue 062 on aqueous solution using calcinated colemanite ore waste

    International Nuclear Information System (INIS)

    Atar, Necip; Olgun, Asim

    2007-01-01

    Colemanite ore waste (CW) has been employed as adsorbent for the removal of acid blue 062 anionic dye (AB 062) from aqueous solution. The adsorption of AB 062 onto CW was examined with respect to contact time, calcination temperature, particle size, pH, adsorbent dosage and temperature. The physical and chemical properties of the CW, such as particle sizes and calcinations temperature, play important roles in dye adsorption. The dye adsorption largely depends on the initial pH of the solution with maximum uptake occurring at pH 1.Three simplified kinetics models, namely, pseudo-first order, pseudo-second order, and intraparticle diffusion models were tested to investigate the adsorption mechanisms. The kinetic adsorption of AB 062 on CW follows a pseudo-second order equation. The adsorption data have been analyzed using Langmuir and Freundlich isotherms. The results indicate that the Langmuir model provides the best correlation of the experimental data. Isotherms have also been used to obtain the thermodynamic parameters such as free energy, enthalpy and entropy of the adsorption of dye onto CW

  20. Liquid secondary waste: Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-31

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

  1. 46th Annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence / Sustainable reactor operation management - safe, efficient, valuable

    International Nuclear Information System (INIS)

    Fischer, Erwin

    2015-01-01

    Summary report on the following Topical Session of the 46 th Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015: - Sustainable Reactor Operation Management - Safe, Efficient, Valuable (Erwin Fischer) The other Sessions of the Key Topics - ''Outstanding Know-How and Sustainable Innovations'', - ''Enhanced Safety and Operation Excellence'' and - ''Decommissioning Experience and Waste Management Solutions'' have been covered in atw 7 (2015) and will be covered in further issues of atw.

  2. 46{sup th} Annual meeting on nuclear technology (AMNT 2015). Key topic / Enhanced safety and operation excellence / Sustainable reactor operation management - safe, efficient, valuable

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Erwin [E.ON Kernkraft GmbH, Global Unit Next Generation, Hannover (Germany)

    2015-08-15

    Summary report on the following Topical Session of the 46{sup th} Annual Conference on Nuclear Technology (AMNT 2015) held in Berlin, 5 to 7 May 2015: - Sustainable Reactor Operation Management - Safe, Efficient, Valuable (Erwin Fischer) The other Sessions of the Key Topics - ''Outstanding Know-How and Sustainable Innovations'', - ''Enhanced Safety and Operation Excellence'' and - ''Decommissioning Experience and Waste Management Solutions'' have been covered in atw 7 (2015) and will be covered in further issues of atw.

  3. Chemical separations schemes for partitioning and transmutation systems

    International Nuclear Information System (INIS)

    Laidler, J.

    2002-01-01

    In the initial phase of the U.S. Accelerator Transmutation of Waste (ATW) program, a single-tier system was foreseen in which the transuranics and long-lived fission products (specifically, 99 Tc and 129 I) recovered from spent LWR oxide fuel would be sent directly to an accelerator-driven transmuter reactor [1]. Because the quantity of fuel to be processed annually was so large (almost 1,500 tons per year), an aqueous solvent extraction process was chosen for LWR fuel processing. Without the need to separate transuranics from one another for feed to the transmuter, it became appropriate to develop an advanced aqueous separations method that became known as UREX. The UREX process employs an added reagent (acetohydroxamic acid) that suppresses the extraction of plutonium and promotes the extraction of technetium together with uranium. Technetium can then be efficiently removed from the uranium; the recovered uranium, being highly decontaminated, can be disposed of as a low-level waste or stored in an unshielded facility for future use. Plutonium and the other transuranic elements, plus the remaining fission products, are directed to the liquid waste stream. This stream is calcined, converting the transuranics and fission products to their oxides. The resulting oxide powder, now representing only about four percent of the original mass of the spent fuel, is reduced to metallic form by means of a pyrometallurgical process. Subsequently, the transuranics are separated from the fission products in another pyro-metallurgical step involving molten salt electrorefining

  4. Mercury separation from mixed wastes. Annual report

    International Nuclear Information System (INIS)

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

    1995-11-01

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

  5. Low level radioactive liquid waste decontamination by electrochemical way

    International Nuclear Information System (INIS)

    Tronche, E.

    1994-10-01

    As part of the work on decontamination treatments for low level radioactive aqueous liquid wastes, the study of an electro-chemical process has been chosen by the C.E.A. at the Cadarache research centre. The first part of this report describes the main methods used for the decontamination of aqueous solutions. Then an electro-deposition process and an electro-dissolution process are compared on the basis of the decontamination results using genuine radioactive aqueous liquid waste. For ruthenium decontamination, the former process led to very high yields (99.9 percent eliminated). But the elimination of all the other radionuclides (antimony, strontium, cesium, alpha emitters) was only favoured by the latter process (90 percent eliminated). In order to decrease the total radioactivity level of the waste to be treated, we have optimized the electro-dissolution process. That is why the chemical composition of the dissolved anode has been investigated by a mixture experimental design. The radionuclides have been adsorbed on the precipitating products. The separation of the precipitates from the aqueous liquid enabled us to remove the major part of the initial activity. On the overall process some operations have been investigated to minimize waste embedding. Finally, a pilot device (laboratory scale) has been built and tested with genuine radioactive liquid waste. (author). 77 refs., 41 tabs., 55 figs., 4 appendixes

  6. Characteristics Data Base: Programmer's guide to the High-Level Waste Data Base

    International Nuclear Information System (INIS)

    Jones, K.E.; Salmon, R.

    1990-08-01

    The High-Level Waste Data Base is a menu-driven PC data base developed as part of OCRWM's technical data base on the characteristics of potential repository wastes, which also includes spent fuel and other materials. This programmer's guide completes the documentation for the High-Level Waste Data Base, the user's guide having been published previously. 3 figs

  7. An Optical Fiber-Based Sensor Array for the Monitoring of Zinc and Copper Ions in Aqueous Environments

    Directory of Open Access Journals (Sweden)

    Steven Kopitzke

    2014-02-01

    Full Text Available Copper and zinc are elements commonly used in industrial applications as aqueous solutions. Before the solutions can be discharged into civil or native waterways, waste treatment processes must be undertaken to ensure compliance with government guidelines restricting the concentration of ions discharged in solution. While currently there are methods of analysis available to monitor these solutions, each method has disadvantages, be it high costs, inaccuracy, and/or being time-consuming. In this work, a new optical fiber-based platform capable of providing fast and accurate results when performing solution analysis for these metals is described. Fluorescent compounds that exhibit a high sensitivity and selectivity for either zinc or copper have been employed for fabricating the sensors. These sensors demonstrated sub-part-per-million detection limits, 30-second response times, and the ability to analyze samples with an average error of under 10%. The inclusion of a fluorescent compound as a reference material to compensate for fluctuations from pulsed excitation sources has further increased the reliability and accuracy of each sensor. Finally, after developing sensors capable of monitoring zinc and copper individually, these sensors are combined to form a single optical fiber sensor array capable of simultaneously monitoring concentration changes in zinc and copper in aqueous environments.

  8. Accelerator driven reactors and nuclear waste management projects in the Czech Republic

    Energy Technology Data Exchange (ETDEWEB)

    Janouch, F. [Royal Institute of Technology, Stockholm (Sweden); Mach, R. [Institute of Nuclear Physics, Rez near Prague (Czechoslovakia)

    1995-10-01

    The Czech Republic is almost the only country in the central Europe which continues with the construction of nuclear power reactors. Its small territory and dense population causes public worries concerning the disposal of the spent nuclear fuel. The Czech nuclear scientists and the power companies and the nuclear industries are therefore looking for alternative solutions. The Los Alamos ATW project had received a positive response in the Czech mass-media and even in the industrial and governmental quarters. The recent scientific symposium {open_quotes}Accelerator driven reactors and nuclear waste management{close_quotes} convened at the Liblice castle near Prague, 27-29. 6. 1994 and sponsored by the Czech Energy Company CEZ, reviewed the competencies and experimental basis in the Czech republic and made the first attempt to formulate the national approach and to establish international collaboration in this area.

  9. Studies on the Influence of Mercaptoacetic Acid (MAA) Modification of Cassava (Manihot sculenta Cranz) Waste Biomass on the Adsorption of Cu2+ and Cd2+ from Aqueous Solution

    International Nuclear Information System (INIS)

    Horsfall, M.; Spiff, A. I.; Abia, A. A.

    2004-01-01

    Cassava peelings waste, which is both a waste and pollutant, was chemically modified using mercaptoacetic acid (MAA) and used to adsorb Cu 2+ and Cd 2+ from aqueous solution over a wide range of reaction conditions at 30 .deg. C. Acid modification produced a larger surface area, which significantly enhanced the metal ion binding capacity of the biomass. An adsorption model based on the Cu 2+ /Cd 2+ adsorption differences was developed to predict the competition of the two metal ions towards binding sites for a mixed metal ion system. The phytosorption process was examined in terms of Langmuir, Freundlich and Dubinin-Radushkevich models. The models indicate that the cassava waste biomass had a greater phytosorption capacity, higher affinity and greater sorption intensity for Cu 2+ than Cd 2+ . According to the evaluation using Langmuir equation, the monolayer binding capacity obtained was 127.3 mg/g Cu 2+ and 119.6 mg/g Cd 2+ . The kinetic studies showed that the phytosorption rates could be described better by a pseudo-second order process and the rate coefficients was determined to be 2.04 x 10 -3 min -1 and 1.98 x 10 -3 min -1 for Cu 2+ and Cd 2+ respectively. The results from these studies indicated that acid treated cassava waste biomass could be an efficient sorbent for the removal of toxic and valuable metals from industrial effluents

  10. Remotely controlled reagent feed system for mixed waste treatment Tank Farm

    International Nuclear Information System (INIS)

    Dennison, D.K.; Bowers, J.S.; Reed, R.K.

    1995-02-01

    LLNL has developed and installed a large-scale. remotely controlled, reagent feed system for use at its existing aqueous low-level radioactive and mixed waste treatment facility (Tank Farm). LLNL's Tank Farm is used to treat aqueous low-level and mixed wastes prior to vacuum filtration and to remove the hazardous and radioactive components before it is discharged to the City of Livermore Water Reclamation Plant (LWRP) via the sanitary sewer in accordance with established limits. This reagent feed system was installed to improve operational safety and process efficiency by eliminating the need for manual handling of various reagents used in the aqueous waste treatment processes. This was done by installing a delivery system that is controlled either remotely or locally via a programmable logic controller (PLC). The system consists of a pumping station, four sets of piping to each of six 6,800-L (1,800-gal) treatment tanks, air-actuated discharge valves at each tank, a pH/temperature probe at each tank, and the PLC-based control and monitoring system. During operation, the reagents are slowly added to the tanks in a preprogrammed and controlled manner while the pH, temperature, and liquid level are continuously monitored by the PLC. This paper presents the purpose of this reagent feed system, provides background related to LLNL's low-level/mixed waste treatment processes, describes the major system components, outlines system operation, and discusses current status and plans

  11. Radioactive waste data base through the net: A tool to improve the development of waste management

    International Nuclear Information System (INIS)

    Sanhueza Mir, Azucena

    2003-01-01

    One of the duties in Chilean Commission for Nuclear Energy (CCHEN) is the timely reply to the International Atomic Energy Agency (IAEA) Net enable waste management data base (NEWMDB) in the waste management field. This duty is carried out by the Radioactive Waste Management Section. CCHEN has complete this data base from about one decade ago. Through the time, the data base has changed according to new available information technologies, to the point that the access using the international net is a need today. The NEWMDB objective is to exchange information and knowledge between member states related to radioactive waste management situation and to conform a world inventory of radioactive waste. The Chilean experience got from the NEWMDB first data collection cycle (1999-2000) is presented here, and recommendations to be considered for incorporation in the domestic waste management system are exposed. In so doing, the data base answer should be easy to do and totally understood by everyone whose job is waste management around the world, in the context of the glossary, criteria and conventions on this data base is supported. The composition of the NEWMDB considers a General Frame which indicates the way in which the waste management is enfaced in the country, regulations, authorities, policies, infrastructure; a Waste Classification matrix which give the equivalence between proper country waste classification and that recommended by IAEA; Waste Data which give the quantities and situation of waste in the different steps of the management such as: conditioned waste, unconditioned stored waste, etc. Finally, the Sustainable Development for radioactive waste management Indicators (SDI) for the safety and environmental radioactive waste management are estimated (Au)

  12. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  13. Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Marra, J. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-26

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

  14. Technology for commercial radioactive waste management

    International Nuclear Information System (INIS)

    1979-05-01

    Conceptual processes and facilities for treating gaseous and various transuranium (TRU) wastes produced during the past fission portion of the light water reactor fuel cycle are described in volume 2. The goal of the treatment process for TRU wastes and for long-lived radionuclides removed from the gaseous waste streams is to convert these wastes to stable products suitable for placement in geologic isolation repositories. The treatment concepts are based on available technology. They do not necessarily represent an optimum design but are representative of what could be achieved with current technology. In actual applications it is reasonable to expect that there could be some improvement over these concepts that might be reflected in either lower costs or lower environmental impacts or both. These conceptual descriptions do provide a reasonable basis for cost analysis and for development of estimates of environmental impacts. The waste treatment technologies considered here include: high-level waste solidification, packaging of fuel residue, failed equipment and noncombustible waste treatment, general trash and combustible waste treatment, degraded solvent treatment, dilute aqueous waste pretreatment, immobilization of wet and solid wastes, off-gas particle removal systems, fuel reprocessing plant dissolver off-gas treatment, process off-gas treatment, and fuel reprocessing plant atmospheric protection system

  15. Advanced oxidation treatment of high strength bilge and aqueous petroleum waste

    Energy Technology Data Exchange (ETDEWEB)

    Hulsey, R.A.; Kobylinski, E.A. [Black and Veatch, Kansas City, MO (United States); Leach, B. [EEC, Inc., Virginia Beach, VA (United States); Pearce, L. [TRITECH, Greensboro, NC (United States)

    1996-11-01

    The Craney Island Fuel Depot is the largest US Navy fuel terminal in the continental US. Services provided at this facility include fuel storage (current capacity is 1.5 million barrels), fuel reclamation (recovery of oil from oily wastewater), and physical/chemical treatment for the removal of residual oil from bilge water and from aqueous petroleum waste. Current wastewater treatment consists of storage/equalization, oil/water separation, dissolved air flotation, sand filtration, and carbon adsorption. The Navy initiated this study to comply with the State requirement that its existing physical/chemical oily wastewater treatment plant be upgraded to remove soluble organics and produce an effluent which would meet acute toxicity limits. The pilot tests conducted during the study included several variations of chemical and biological wastewater treatment processes. While biological treatment alone was capable of meeting the proposed BOD limit of 26 mg/L, the study showed that the effluent of the biological process contained a high concentration of refractory (nonbiodegradable) organics and could not consistently meet the proposed limits for COD and TOC when treating high-strength wastewater. Additional tests were conducted with advanced oxidation processes (AOPs). AOPs were evaluated for use as independent treatment processes as well as polishing processes following biological treatment. The AOP processes used for this study included combinations of ozone (O{sub 3}) ultraviolet radiation (UV), and hydrogen peroxide (H{sub 2}O{sub 2}).

  16. Measurements and models for hazardous chemical and mixed wastes. 1998 annual progress report

    International Nuclear Information System (INIS)

    Holcomb, C.; Louie, B.; Mullins, M.E.; Outcalt, S.L.; Rogers, T.N.; Watts, L.

    1998-01-01

    'Aqueous waste of various chemical compositions constitutes a significant fraction of the total waste produced by industry in the US. A large quantity of the waste generated by the US chemical process industry is waste water. In addition, the majority of the waste inventory at DoE sites previously used for nuclear weapons production is aqueous waste. Large quantities of additional aqueous waste are expected to be generated during the clean-up of those sites. In order to effectively treat, safely handle, and properly dispose of these wastes, accurate and comprehensive knowledge of basic thermophysical property information is paramount. This knowledge will lead to huge savings by aiding in the design and optimization of treatment and disposal processes. The main objectives of this project are: Develop and validate models that accurately predict the phase equilibria and thermodynamic properties of hazardous aqueous systems necessary for the safe handling and successful design of separation and treatment processes for hazardous chemical and mixed wastes. Accurately measure the phase equilibria and thermodynamic properties of a representative system (water + acetone + isopropyl alcohol + sodium nitrate) over the applicable ranges of temperature, pressure, and composition to provide the pure component, binary, ternary, and quaternary experimental data required for model development. As of May, 1998, nine months into the first year of a three year project, the authors have made significant progress in the database development, have begun testing the models, and have been performance testing the apparatus on the pure components.'

  17. Removal of Sr ions from nuclear wastes by D2EHPA+TBP based supported liquid membranes

    International Nuclear Information System (INIS)

    Chaudry, M.A.; Ahmad, I.

    2000-01-01

    Sr ions removal from nuclear wastes is of great importance. /sup 90/Sr radionuclide, due to its long half-life to disintegrate into daughter products and release of radiations, resulting from fission of uranium, produce heat and is a real problem for disposal of radioactive wastes. The separation study of Sr ions from aqueous solutions is, therefore, very important in the nuclear industry. n the present article some of the work done to develop the separation technique based on coupled transport phenomenon for Sr ions is reported. Di-2-ethyl-hexyl phosphoric acid mixed with tri-n-butyl phosphate (TBP), diluted in kerosene oil, as an organic liquid has been used as a membrane, supported in polypropylene hydrophobic films to transport Sr ions. The optimum conditions and mechanism of transport for these ions across the membrane have been described. The effect of feed complexing components i.e. tartaric acid and citric acid concentration on the flux and permeability of the Sr/sup 2+/ ions has been studied. It is shown that supported liquid membrane technique can be used as an alternate process to classical solvent extraction to remove Sr ions from nuclear industry wastes. (author)

  18. Extraction of Biomolecules Using Phosphonium-Based Ionic Liquids + K3PO4 Aqueous Biphasic Systems

    Science.gov (United States)

    Louros, Cláudia L. S.; Cláudio, Ana Filipa M.; Neves, Catarina M. S. S.; Freire, Mara G.; Marrucho, Isabel M.; Pauly, Jérôme; Coutinho, João A. P.

    2010-01-01

    Aqueous biphasic systems (ABS) provide an alternative and efficient approach for the extraction, recovery and purification of biomolecules through their partitioning between two liquid aqueous phases. In this work, the ability of hydrophilic phosphonium-based ionic liquids (ILs) to form ABS with aqueous K3PO4 solutions was evaluated for the first time. Ternary phase diagrams, and respective tie-lines and tie-lines length, formed by distinct phosphonium-based ILs, water, and K3PO4 at 298 K, were measured and are reported. The studied phosphonium-based ILs have shown to be more effective in promoting ABS compared to the imidazolium-based counterparts with similar anions. Moreover, the extractive capability of such systems was assessed for distinct biomolecules (including amino acids, food colourants and alkaloids). Densities and viscosities of both aqueous phases, at the mass fraction compositions used for the biomolecules extraction, were also determined. The evaluated IL-based ABS have been shown to be prospective extraction media, particularly for hydrophobic biomolecules, with several advantages over conventional polymer-inorganic salt ABS. PMID:20480041

  19. Methods of vitrifying waste with low melting high lithia glass compositions

    Science.gov (United States)

    Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

    2001-01-01

    The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

  20. Analysis of a high pressure ATWS [anticipated transient without scram] with very low make-up flow

    International Nuclear Information System (INIS)

    Wagner, K.C.

    1988-10-01

    A series of calculations were performed to analyze the response of General Electric Company's (GE) advanced boiling water reactor (ABWR) during an anticipated transient without scram (ATWS). This work investigated the early plant response with an assumed failure or manual inhibit of the high pressure core flooder (HPCF). Consequently, the reactor core isolation cooling (RCIC) and control rod drive (CRD) systems are the only sources of high pressure injection available to maintain core cooling. Steam leaving the reactor pressure vessel was diverted to the pressure suppression pool (PSP) via the steam line and the safety relief valves. The combination of an unscrammed core and the CRD and RCIC injection sources make this a particularly challenging transient. System energy balance calculations were performed to predict the core power and PSP heat-up rate. The amount of vessel vapor superheat and the PSP temperature were found to significantly affect the resultant core power. Consequently, detailed thermal-hydraulic calculations were performed to simulate the system response during the postulated transient. 15 refs., 15 figs., 4 tabs

  1. Stochastic simulation of pitting degradation of multi-barrier waste container in the potential repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Lee, J.H.; Atkins, J.E.; Andrews, R.W.

    1995-01-01

    A detailed stochastic waste package degradation simulation model was developed incorporating the humid-air and aqueous general and pitting corrosion models for the carbon steel corrosion-allowance outer barrier and aqueous pitting corrosion model for the Alloy 825 corrosion-resistant inner barrier. The uncertainties in the individual corrosion models were also incorporated to capture the variability in the corrosion degradation among waste packages and among pits in the same waste package. Within the scope of assumptions employed in the simulations, the corrosion modes considered, and the near-field conditions from the drift-scale thermohydrologic model, the results of the waste package performance analyses show that the current waste package design appears to meet the 'controlled design assumption' requirement of waste package performance, which is currently defined as having less than 1% of waste packages breached at 1,000 years. It was shown that, except for the waste packages that fail early, pitting corrosion of the corrosion-resistant inner barrier has a greater control on the failure of waste packages and their subsequent degradation than the outer barrier. Further improvement and substantiation of the inner barrier pitting model (currently based on an elicitation) is necessary in future waste package performance simulation model

  2. Behaviour of intermediate-level waste forms in an aqueous environment

    International Nuclear Information System (INIS)

    Amarantos, S.; DeBatist, R.; Brodersen, K.; Glasser, F.P.; Pottier, P.E.; Vejmelka, R.; Zamorani, E.

    1985-01-01

    Under Action 1 of the Second Community Programme (1980-1984), study continued of the behavoiur of low and medium activity waste matrices using 10 reference waste forms (RWFs) representative of the main waste packages produced in the Community. The aim of this paper is to outline the main results for three types of matrix: cement and derived forms, organic polymers and bitumens. The results include data on diffusion coefficients, leach rates and waste form volume changes and mass losses. They constitute a considerable advance in knowledge of confinement properties but bring to light the need for further study of radionuclide release mechanisms for the purpose of constructing long-term models of waste form behaviour in the presence of water

  3. Comparative study of aqueous and solvent methods for cleaning metals

    International Nuclear Information System (INIS)

    Briggs, J.L.; Goad, H.A.

    1976-01-01

    Studies were performed to determine the comparative effectiveness of solvent and aqueous detergent methods for cleaning various metals. The metals investigated included 304L stainless steel, beryllium, uranium-6.5 wt percent niobium alloy, and unalloyed uranium ( 238 U). The studies were initiated in response to governmental regulations restricting the use of some chlorinated solvents. Results showed that aqueous detergent cleaning was more effective than solvents, i.e. trichloroethylene and methyl chloroform, for the removal of light industrial soils. The subsequent adoption of aqueous cleaning at this plant has facilitated waste disposal, which contributed to recorded economic savings. The controlled use of aqueous detergents is environmentally acceptable and has decreased the hazards of fire and toxicity that are generally associated with solvents. 8 tables, 15 figures

  4. Effluent testing for the Oak Ridge mixed waste incinerator: Emissions test for August 27, 1990

    International Nuclear Information System (INIS)

    Bostick, W.D.; Bunch, D.H.; Gibson, L.V.; Hoffmann, D.P.; Shoemaker, J.L.

    1990-12-01

    On August 27, 1990, a special emissions test was performed at the K-1435 Toxic Substance Control Act Mixed Waste Incinerator. A sampling and analysis plan was implemented to characterize the incinerator waste streams during a 6 hour burn of actual mixed waste. The results of this characterization are summarized in the present report. Significant among the findings is the observation that less than 3% of the uranium fed to the incinerator kiln was discharged as stack emission. This value is consistent with the estimate of 4% or less derived from long-term mass balance of previous operating experience and with the value assumed in the original Environmental Impact Statement. Approximately 1.4% of the total uranium fed to the incinerator kiln appeared in the aqueous scrubber blowdown; about 85% of the total uranium in the aqueous waste was insoluble (i.e., removable by filtration). The majority of the uranium fed to the incinerator kiln appeared in the ash material, apparently associated with phosphorous as a sparingly-soluble species. Many other metals of potential regulatory concern also appeared to concentrate in the ash as sparingly-soluble species, with minimal partition to the aqueous waste. The aqueous waste was discharged to the Central Neutralization Facility where it was effectively treated by coprecipitation with iron. The treated, filtered aqueous effluent met Environmental Protection Agency interim primary drinking water standards for regulated metals

  5. Defense waste processing facility precipitate hydrolysis process

    International Nuclear Information System (INIS)

    Doherty, J.P.; Eibling, R.E.; Marek, J.C.

    1986-03-01

    Sodium tetraphenylborate and sodium titanate are used to assist in the concentration of soluble radionuclide in the Savannah River Plant's high-level waste. In the Defense Waste Processing Facility, concentrated tetraphenylborate/sodium titanate slurry containing cesium-137, strontium-90 and traces of plutonium from the waste tank farm is hydrolyzed in the Salt Processing Cell forming organic and aqueous phases. The two phases are then separated and the organic phase is decontaminated for incineration outside the DWPF building. The aqueous phase, containing the radionuclides and less than 10% of the original organic, is blended with the insoluble radionuclides in the high-level waste sludge and is fed to the glass melter for vitrification into borosilicate glass. During the Savannah River Laboratory's development of this process, copper (II) was found to act as a catalyst during the hydrolysis reactions, which improved the organic removal and simplified the design of the reactor

  6. Leaching behavior of phosphate-bonded ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Jeong, S.Y.; Dorf, M.

    1996-04-01

    Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics for solidifying and stabilizing low-level mixed wastes. This technology is crucial for stabilizing waste streams that contain volatile species and off-gas secondary waste streams generated by high-temperature treatment of such wastes. We have developed a magnesium phosphate ceramic to treat mixed wastes such as ash, salts, and cement sludges. Waste forms of surrogate waste streams were fabricated by acid-base reactions between the mixtures of magnesium oxide powders and the wastes, and phosphoric acid or acid phosphate solutions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with ash waste streams spiked with hazardous and radioactive surrogates. Standard leaching tests such as ANS 16.1 and TCLP were conducted on the final waste forms. Fates of the contaminants in the final waste forms were established by electron microscopy. In addition, stability of the waste forms in aqueous environments was evaluated with long-term water-immersion tests

  7. The disposal of radioactive solvent waste

    International Nuclear Information System (INIS)

    Dean, B.; Baker, W.T.

    1976-01-01

    As the use of radioisotope techniques increases, laboratories are faced with the problem of disposing of considerable quantities of organic solvent and aqueous liquid wastes. Incineration or collection by a waste contractor both raise problems. Since most of the radiochemicals are preferentially water soluble, an apparatus for washing the radiochemicals out into water and discharging into the normal drainage system in a high diluted form is described. Despite the disadvantages (low efficiency, high water usuage, loss of solvent in presence of surface active agents, precipitation of phosphors from dioxan based liquids) it is felt that the method has some merit if a suitably improved apparatus can be designed at reasonable cost. (U.K.)

  8. The phase transport and reactions of γ-irradiated aqueous-ionic liquids

    International Nuclear Information System (INIS)

    Howett, S.; Joseph, J.; Noel, J.J.; Wren, J.C.

    2010-01-01

    A novel technology based on the transfer of chemical species across water/ionic liquid interfaces via specific complexation reactions is currently being considered for the separation and sequestration of metal ion contaminants from radioactive waste effluents in the nuclear fuel cycle. An ideal solvent for these applications should have a high intrinsic selectivity for a targeted metal or group of metals (e.g., trans-Pu actinides, lanthanides, or other fission products), an efficient switching mechanism (between complexation and decomplexation), and a high immiscibility with aqueous solutions. These characteristics must be maintained in the chemical, radiation, and mass transport environments present during the separation process. Ionic liquids (ILs) have an almost negligible vapour pressure and high thermal stability. Their ability to dissolve a wide range of substrate molecules and potential to be highly resilient in radiation fields make ILs particularly promising media. The separation efficiency of the biphasic system will depend on many parameters, including the aqueous oxidation state of the targeted metal ion, and the thermodynamics and kinetics of interfacial transport and metal-ligand complex formation at the water/IL interface or in the IL phase. The most uncertain and unstudied area for these applications is the effect of ionizing radiation on the stability and separation efficiency of the biphasic system. The present study investigates the effect of γ-radiation on gas/IL and water/IL interfacial stability and mass transfer with trihexyltetradecylphosphonium bis(trifluoromethyl-sulfonyl)imide, a phosphonium-based IL. The IL, in contact with either gas or water, was irradiated at a dose rate of 6.4 kGy·h -1 . Gas-phase samples were analyzed by gas chromatography-mass spectrometry (GC-MS) and the changes in the IL and aqueous phases were monitored by conductivity measurements and Raman spectroscopy. In this paper we discuss these observations and their

  9. NABTIT-a computer program for non-aqueous acid-base titration.

    Science.gov (United States)

    Budevsky, O; Zikolova, T; Tencheva, J

    1988-11-01

    A program NABTIT written in BASIC has been developed for the treatment of data (ml/mV) obtained from potentiometric acid-base titrations in non-aqueous solvents. No preliminary information on equilibrium constants is required for the input. The treatment of the data is based on known equations and uses least-squares procedures. The essence of the method is to determine the equivalence volume (V(e)) accurately, and to use the data acquired by adding titrant after V(e) for the pH*-calibration of the non-aqueous potentiometric cell. As a by-product or the calculations, the pK* value of the substance titrated is also obtained, and in some cases the autoprotolysis constant of the medium (pK*(s)). Good agreement between experiment and theory was found in the treatment of data obtained for water and methanol-water mixtures.

  10. 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.

  11. Process for disposal of aqueous solutions containing radioactive isotopes

    Science.gov (United States)

    Colombo, Peter; Neilson, Jr., Robert M.; Becker, Walter W.

    1979-01-01

    A process for disposing of radioactive aqueous waste solutions whereby the waste solution is utilized as the water of hydration to hydrate densified powdered portland cement in a leakproof container; said waste solution being dispersed without mechanical inter-mixing in situ in said bulk cement, thereafter the hydrated cement body is impregnated with a mixture of a monomer and polymerization catalyst to form polymer throughout the cement body. The entire process being carried out while maintaining the temperature of the components during the process at a temperature below 99.degree. C. The container containing the solid polymer-impregnated body is thereafter stored at a radioactive waste storage dump such as an underground storage dump.

  12. Process for disposal of aqueous solutions containing radioactive isotopes

    International Nuclear Information System (INIS)

    Colombo, P.; Neilson, R.M. Jr.; Becker, W.W.

    1979-01-01

    A process for disposing of radioactive aqueous waste solutions whereby the waste solution is utilized as the water of hydration to hydrate densified powdered portland cement in a leakproof container; said waste solution being dispersed without mechanical inter-mixing in situ in said bulk cement, thereafter the hydrated cement body is impregnated with a mixture of a monomer and polymerization catalyst to form polymer throughout the cement body. The entire process being carried out while maintaining the temperature of the components during the process at a temperature below 99 0 C. The container containing the solid polymer-impregnated body is thereafter stored at a radioactive waste storage dump such as an underground storage dump

  13. Removal of Hexavalent Chromium from Aqueous Solutions using ...

    African Journals Online (AJOL)

    The hexavalent chromium exists in aquatic media as water soluble complex anions and persist. These are concentrated in industrial waste water especially from the tannery industries and release of effluents from industries adversely affects the environment. The removal of heavy metals from aqueous solutions is carried ...

  14. Glass and nuclear wastes

    International Nuclear Information System (INIS)

    Sombret, C.

    1982-10-01

    Glass shows interesting technical and economical properties for long term storage of solidified radioactive wastes by vitrification or embedding. Glass composition, vitrification processes, stability under irradiation, thermal stability and aqueous corrosion are studied [fr

  15. GIS-based tools to identify tradeoffs between waste management and remediation strategies from radiological dispersal device incidents

    Energy Technology Data Exchange (ETDEWEB)

    Lemieux, P.; Wood, J.; Snyder, E. [U.S. Environmental Protection Agency, Research Triangle Park, NC (United States); Boe, T. [Oak Ridge Inst. for Science and Education, Research Triangle Park, NC (United States); Schulthiesz, D.; Peake, T.; Ierardi, M. [U.S. Environmental Protection Agency, Washington, DC (United States); Hayes, C.; Rodgers, M. [Eastern Research Group, Inc., Morrisville, NC (United States)

    2011-07-01

    Management of waste and debris from the detonation of a Radiological Dispersal Device (RDD) will likely comprise a significant portion of the overall remediation effort and possibly contribute to a significant portion of the overall remediation costs. As part of the recent National Level Exercise, Liberty RadEx, that occurred in Philadelphia in April 2010, a methodology was developed by EPA to generate a first-order estimate of a waste inventory for the hypothetical RDD from the exercise scenario. Determination of waste characteristics and whether the generated waste is construction and demolition (C&D) debris, municipal solid waste (MSW), hazardous waste, mixed waste, or low level radioactive waste (LLRW), and characterization of the wastewater that is generated from the incident or subsequent cleanup activities will all influence the cleanup costs and timelines. Decontamination techniques, whether they involve chemical treatment, abrasive removal, or aqueous washing, will also influence the waste generated and associated cleanup costs and timelines. This paper describes the ongoing effort to develop a tool to support RDD planning and response activities by assessing waste quantities and characteristics as a function of potential mitigation strategies and targeted cleanup levels. (author)

  16. GIS-based tools to identify tradeoffs between waste management and remediation strategies from radiological dispersal device incidents

    International Nuclear Information System (INIS)

    Lemieux, P.; Wood, J.; Snyder, E.; Boe, T.; Schulthiesz, D.; Peake, T.; Ierardi, M.; Hayes, C.; Rodgers, M.

    2011-01-01

    Management of waste and debris from the detonation of a Radiological Dispersal Device (RDD) will likely comprise a significant portion of the overall remediation effort and possibly contribute to a significant portion of the overall remediation costs. As part of the recent National Level Exercise, Liberty RadEx, that occurred in Philadelphia in April 2010, a methodology was developed by EPA to generate a first-order estimate of a waste inventory for the hypothetical RDD from the exercise scenario. Determination of waste characteristics and whether the generated waste is construction and demolition (C&D) debris, municipal solid waste (MSW), hazardous waste, mixed waste, or low level radioactive waste (LLRW), and characterization of the wastewater that is generated from the incident or subsequent cleanup activities will all influence the cleanup costs and timelines. Decontamination techniques, whether they involve chemical treatment, abrasive removal, or aqueous washing, will also influence the waste generated and associated cleanup costs and timelines. This paper describes the ongoing effort to develop a tool to support RDD planning and response activities by assessing waste quantities and characteristics as a function of potential mitigation strategies and targeted cleanup levels. (author)

  17. Recovery of technetium from nuclear fuel wastes

    International Nuclear Information System (INIS)

    Carlin, W.W.

    1975-01-01

    Technetium is removed from aqueous, acidic waste solutions. The acidic waste solution is mixed with a flocculant, e.g., an alkaline earth metal hydroxide or oxide, to precipitate certain fission products. Technetium remains in solution and in the resulting supernatant alkaline aqueous phase. The supernatant alkaline aqueous phase is made acidic and electrolyzed in an electrolytic cell under controlled cathodic potential conditions to deposit technetium on the cathode. Elemental technetium is removed from the cathode. Technetium is separated from other plated fission product metals by extraction from an alkaline solution with an organic extractant, such as pyridine, having affinity for technetium. Technetium is separated from the organic extractant by steam distillation and the resulting aqueous phase treated with ammoniacal reagent to precipitate technetium as ammonium pertechnetate. The precipitate may be acidified to form an aqueous acidic solution of fission product metal values and the solution electrolyzed in an electrolytic cell under controlled cathodic potential conditions and at a potential sufficiently negative to plate out from the solution those fission product metals desired. The metal deposit is stripped from the cathode and stored until its radioactivity has diminished. (U.S.)

  18. Biosorptive removal of cobalt (II) ions from aqueous solution by ...

    African Journals Online (AJOL)

    hope&shola

    2010-11-29

    Nov 29, 2010 ... chemical reaction, decay, adsorption and biodegradation. The presence of ... sted coffee (Dakiky et al., 2002), waste tea (Ahluwalia and Goyal, 2005) ..... green removal from aqueous solution by citric acid modified rice straw.

  19. Solidification of TRU wastes in a ceramic matrix

    International Nuclear Information System (INIS)

    Loida, A.; Schubert, G.

    1991-01-01

    Aluminumsilicate based ceramic materials have been evaluated as an alternative waste form for the incorporation of TRU wastes. These waste forms are free of water and - cannot generate hydrogen radiolyticly, - they show good compatibility between the compounds of the waste and the matrix, - they are resistent against aqueous solutions, heat and radiation. R and D-work has been performed to demonstrate the suitability of this waste form for the immobilization of TRU-wastes. Four kinds of original TRU-waste streams and a mixture of all of them have been immobilized by ceramization, using glove box and remote operation technique as well. Clay minerals, (kaolinite, bentonite) and reactive corundum were selected as ceramic raw materials (KAB 78) in an appropriate ratio yielding 78 wt% Al 2 O 3 and 22 wt%SiO 2 . The main process steps are (i) pretreatment of the liquid waste (concentration, denitration, neutralization, solid- liquid separation), (ii) mixing with ceramic raw materials and forming, (iii) heat treatment with T max. of 1300 0 C for 15 minutes. The waste load of the ceramic matrix has been increased gradually from 20 to 50, in some cases to 60 wt.%

  20. Determination of particle size distribution of salt crystals in aqueous slurries

    International Nuclear Information System (INIS)

    Miller, A.G.

    1977-10-01

    A method for determining particle size distribution of water-soluble crystals in aqueous slurries is described. The salt slurries, containing sodium salts of predominantly nitrate, but also nitrite, sulfate, phosphate, aluminates, carbonate, and hydroxide, occur in radioactive, concentrated chemical waste from the reprocessing of nuclear fuel elements. The method involves separating the crystals from the aqueous phase, drying them, and then dispersing the crystals in a nonaqueous medium based on nitroethane. Ultrasonic treatment is important in dispersing the sample into its fundamental crystals. The dispersed crystals are sieved into appropriate size ranges for counting with a HIAC brand particle counter. A preponderance of very fine particles in a slurry was found to increase the difficulty of effecting complete dispersion of the crystals because of the tendency to retain traces of aqueous mother liquor. Traces of moisture produce agglomerates of crystals, the extent of agglomeration being dependent on the amount of moisture present. The procedure is applicable to particles within the 2 to 600 μm size range of the HIAC particle counter. The procedure provides an effective means for measuring particle size distribution of crystals in aqueous salt slurries even when most crystals are less than 10 μm in size. 19 figures

  1. Potential benefits of waste transmutation to the U.S. high-level waste respository

    Energy Technology Data Exchange (ETDEWEB)

    Michaels, G.E. [Oak Ridge National Laboratory, TN (United States)

    1995-10-01

    This paper reexamines the potential benefits of waste transmutation to the proposed U.S. geologic repository at the Yucca Mountain site based on recent progress in the performance assessment for the Yucca Mountain base case of spent fuel emplacement. It is observed that actinides are assumed to have higher solubility than in previous studies and that Np and other actinides now dominate the projected aqueous releases from a Yucca Mountain repository. Actinides are also indentified as the dominant source of decay heat in the repository, and the effect of decay heat in perturbing the hydrology, geochemistry, and thermal characteristics of Yucca Mountain are reviewed. It is concluded that the potential for thermally-driven, buoyant, gas-phase flow at Yucca Mountain introduces data and modeling requirements that will increase the costs of licensing the site and may cause the site to be unattractive for geologic disposal of wastes. A transmutation-enabled cold repository is proposed that might allow licensing of a repository to be based upon currently observable characteristics of the Yucca Mountain site.

  2. Introducing a European Partnership. First issue of 'European Nuclear Features'. A joint publication of atw, Nuclear Espana, Revue Generale Nucleare (2004)

    International Nuclear Information System (INIS)

    2004-01-01

    'European Nuclear Features' is a joint publication of the three specialized technical journals, Nuclear Espana (Spain), Revue Generale Nucleaire (France), and atw - International Journal for Nuclear Power (Germany), planned for six issues annually. ENF is to further greatly the international European exchange of information and news about energy and nuclear power. News items, comments, and scientific and technical contributions will cover important aspects of the field. The first issue of ENF contains contributions about these topics, among others: - European Nuclear Society and Foratom: Strengthening the Nuclear Network. - Report: EPR - the European Pressurized Water Reactor. - Finland: Starting Construction of the Fifth Nuclear Power Plant. - Czech Republic: Nuclear Power Report for 2003/2004. - The Decommissioning Project of the Bohunice-1 and -2 Units. - FRM-II: TUM Research Neutron Source Generates Its First Neutrons. (orig.)

  3. Cytotoxicity testing of aqueous extract of bitter leaf (Vernonia ...

    African Journals Online (AJOL)

    Cytotoxicity testing of aqueous extract of bitter leaf (Vernonia amygdalina Del) and sniper. 1000EC (2,3 ... man and animals.1 It is estimated that 80% of the popula- ..... evaluation of waste, surface and ground water quality using the Allium test ...

  4. Use of fixation techniques in processing radioactive wastes from nuclear power plants in Czechoslovakia

    International Nuclear Information System (INIS)

    Seliga, M.

    1977-01-01

    The current state of radioactive waste disposal from the Bohunice nuclear power plant is described. The method of vacuum cementation was chosen for solidifying liquid radioactive wastes. This method makes it possible to obtain a product whose properties, namely strength, leachability, and radiation stability allow for the production of blocks without packing material. Also solved was the fixation of liquid radioactive waste using bituminization based on mixing liquid radioactive waste with aqueous bitumen emulsion in a film evaporator in which the mixture of liquid radioactive wastes and bitumen emulsion evaporate producing solid bitumen. The parameters are given of the cementation and bituminization lines which are designed for use in nuclear power plants with WWER type reactors. (J.B.)

  5. 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

  6. Integrated Data Base: Status and waste projections

    International Nuclear Information System (INIS)

    Klein, J.A.

    1990-01-01

    The Integrated Data Base (IDB) is the official US Department of Energy (DOE) data base for spent fuel and radioactive waste inventories and projections. DOE low-level waste (LLW) is just one of the many waste types that are documented with the IDB. Summary-level tables and figures are presented illustrating historical and projected volume changes of DOE LLW. This information is readily available through the annual IDB publication. Other presentation formats are also available to the DOE community through a request to the IDB Program. 4 refs., 6 figs., 5 tabs

  7. Effluent testing for the Oak Ridge Mixed Waste Incinerator: Emissions test for August 27, 1990

    International Nuclear Information System (INIS)

    Bostick, W.D.; Bunch, D.H.; Gibson, L.V.; Hoffmann, D.P.; Shoemaker, J.L.

    1991-01-01

    On August 27, 1990, a special emissions test was performed at the K-1435 Toxic Substance Control Act Mixed Waste Incinerator. A sampling and analysis plan was implemented to characterize the incinerator waste streams during a 6 hour burn of actual mixed waste. The results of this characterization are summarized in the present report. Significant among the findings is the observation that less than 3% of the uranium fed to the incinerator kiln was discharged as stack emission. This value is consistent with the estimate of 4% or less derived from long-term mass balance of previous operating experience and with the value assumed in the original Environmental Impact Statement. Approximately 1.4% of the total uranium fed to the incinerator kiln appeared in the aqueous scrubber blowdown; about 85% of the total uranium in the aqueous waste was insoluble (i.e., removable by filtration). The majority of the uranium fed to the incinerator kiln appeared in the ash material, apparently associated with phosphorous as a sparingly-soluble species. Many other metals of potential regulatory concern also appeared to concentrate in the ash as sparingly-soluble species, with minimal partition to the aqueous waste. The aqueous waste was discharged to the Central Neutralization Facility where it was effectively treated by coprecipitation with iron. The treated, filtered aqueous effluent met Environmental Protection Agency interim primary drinking water standards for regulated metals. 4 refs., 2 figs., 10 tabs

  8. SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J. [Savannah River National Laboratory, Aiken, SC (United States); Marra, J. [Savannah River National Laboratory, Aiken, SC (United States)

    2014-10-02

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

  9. SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization

    International Nuclear Information System (INIS)

    Amoroso, J.; Marra, J.

    2014-01-01

    A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear fuel. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing

  10. Transuranic contaminated waste form characterization and data base

    International Nuclear Information System (INIS)

    McArthur, W.C.; Kniazewycz, B.G.

    1980-07-01

    This report outlines the sources, quantities, characteristics and treatment of transuranic wastes in the United States. This document serves as part of the data base necessary to complete preparation and initiate implementation of transuranic wastes, waste forms, waste container and packaging standards and criteria suitable for inclusion in the present NRC waste management program. No attempt is made to evaluate or analyze the suitability of one technology over another. Indeed, by the nature of this report, there is little critical evaluation or analysis of technologies because such analysis is only appropriate when evaluating a particular application or transuranic waste streams. Due to fiscal restriction, the data base is developed from a myriad of technical sources and does not necessarily contain operating experience and the current status of all technologies. Such an effort was beyond the scope of this report

  11. Transuranic contaminated waste form characterization and data base

    Energy Technology Data Exchange (ETDEWEB)

    McArthur, W.C.; Kniazewycz, B.G.

    1980-07-01

    This report outlines the sources, quantities, characteristics and treatment of transuranic wastes in the United States. This document serves as part of the data base necessary to complete preparation and initiate implementation of transuranic wastes, waste forms, waste container and packaging standards and criteria suitable for inclusion in the present NRC waste management program. No attempt is made to evaluate or analyze the suitability of one technology over another. Indeed, by the nature of this report, there is little critical evaluation or analysis of technologies because such analysis is only appropriate when evaluating a particular application or transuranic waste streams. Due to fiscal restriction, the data base is developed from a myriad of technical sources and does not necessarily contain operating experience and the current status of all technologies. Such an effort was beyond the scope of this report.

  12. Hanford Waste Physical and Rheological Properties: Data and Gaps

    Energy Technology Data Exchange (ETDEWEB)

    Wells, Beric E.; Kurath, Dean E.; Mahoney, Lenna A.; Onishi, Yasuo; Huckaby, James L.; Cooley, Scott K.; Burns, Carolyn A.; Buck, Edgar C.; Tingey, Joel M.; Daniel, Richard C.; Anderson, K. K.

    2011-08-01

    The Hanford Site in Washington State manages 177 underground storage tanks containing approximately 250,000 m3 of waste generated during past defense reprocessing and waste management operations. These tanks contain a mixture of sludge, saltcake and supernatant liquids. The insoluble sludge fraction of the waste consists of metal oxides and hydroxides and contains the bulk of many radionuclides such as the transuranic components and 90Sr. The saltcake, generated by extensive evaporation of aqueous solutions, consists primarily of dried sodium salts. The supernates consist of concentrated (5-15 M) aqueous solutions of sodium and potassium salts. The 177 storage tanks include 149 single-shell tanks (SSTs) and 28 double -hell tanks (DSTs). Ultimately the wastes need to be retrieved from the tanks for treatment and disposal. The SSTs contain minimal amounts of liquid wastes, and the Tank Operations Contractor is continuing a program of moving solid wastes from SSTs to interim storage in the DSTs. The Hanford DST system provides the staging location for waste feed delivery to the Department of Energy (DOE) Office of River Protection’s (ORP) Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP is being designed and constructed to pretreat and then vitrify a large portion of the wastes in Hanford’s 177 underground waste storage tanks.

  13. Biological conversion of the aqueous wastes from hydrothermal liquefaction of algae and pine wood by Rhodococci

    Energy Technology Data Exchange (ETDEWEB)

    He, Yucai; Li, Xiaolu; Xue, Xiaoyun; Swita, Marie S.; Schmidt, Andrew J.; Yang, Bin

    2017-01-01

    In this study, R. opacus PD630, R. jostii RHA1, R. jostii RHA1 VanA-, and their co-culture were employed to convert hydrothermal liquefaction aqueous waste (HTLAW) into lipids. After 11 days, the COD reduction of algal-HTLAW reached 93.4% and 92.7% by R. jostii RHA1 and its mutant VanA-, respectively. Woody-HTLAW promoted lipid accumulation of 0.43 g lipid/g cell dry weight in R. opacus PD630 cells. Additionally, the total number of chemicals in HTLAW decreased by over 1/3 after 7 days of coculture, and 0.10 g/L and 0.46 g/L lipids were incrementally accumulated in the cellular mass during the fermentation of wood- and algal-HTLAW, respectively. The GC-MS data supported that different metabolism pathways were followed when these Rhodococci strains degraded algae- and woody-HTLAW. These results indicated promising potential of bioconversion of under-utilized carbon and toxic compounds in HTLAW into useful products by selected Rhodococci.

  14. Testing and modelling the performance of inorganic exchangers for radionuclide removal from aqueous nuclear waste

    International Nuclear Information System (INIS)

    Harjula, R.; Lehto, J.; Paajanen, A.; Saarinen, L.

    1997-01-01

    Three different inorganic sorbents/ion exchangers have been tested in this work. Granular hexacyanoferrate-based ion exchanger was developed for Cs removal from radioactive liquid waste at NPPs. It was tested for Cs removal from waste solutions containing different complexing agents and detergents. Radiation stability and thermal stability test has shown, that this sorbent can be used for treatment of medium-active waste treatment. Active carbon materials were tested for Co removal from liquid waste effluents at NPPs. It was found that 60 Co cannot be removed from the evaporator concentrates with reasonable efficiency and a combined process with up-stream precipitation step is needed for better Co separation efficiency. Granular modified titanium oxide was tested for 90 Sr removal from the waste effluents and showed very high efficiency. A mathematical model was developed to analyze ion exchange performance in feeds of different chemical and radiochemical compositions. (author). 9 refs, 7 figs, 3 tabs

  15. Pyro-processes and the wastes

    International Nuclear Information System (INIS)

    Kurata, Masaki; Tokiwai, Moriyasu; Inoue, Tadashi; Nishimura, Tomohiro

    2000-01-01

    Reprocessing using pyrometallurgical processes is generally considered to have economical benefits comparing with conventional aqueous processes because of the combination of simpler process and equipments, less criticality, and more compact facilities. On the other hand, the pyrometallurgical processes must generate peculiar wastes and R and D on those wastes is slightly inferior, as compared with the main processes. In this paper, process flows of major pyrometallurgical processes are firstly summarized and, then, the present R and D condition on the wastes are shown. (author)

  16. Bioprocessing of a stored mixed liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Wolfram, J.H.; Rogers, R.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Finney, R. [Mound Applied Technologies, Miamisburg, OH (United States)] [and others

    1995-12-31

    This paper describes the development and results of a demonstration for a continuous bioprocess for mixed waste treatment. A key element of the process is an unique microbial strain which tolerates high levels of aromatic solvents and surfactants. This microorganism is the biocatalysis of the continuous flow system designed for the processing of stored liquid scintillation wastes. During the past year a process demonstration has been conducted on commercial formulation of liquid scintillation cocktails (LSC). Based on data obtained from this demonstration, the Ohio EPA granted the Mound Applied Technologies Lab a treatability permit allowing the limited processing of actual mixed waste. Since August 1994, the system has been successfully processing stored, {open_quotes}hot{close_quotes} LSC waste. The initial LSC waste fed into the system contained 11% pseudocumene and detectable quantities of plutonium. Another treated waste stream contained pseudocumene and tritium. Data from this initial work shows that the hazardous organic solvent, and pseudocumene have been removed due to processing, leaving the aqueous low level radioactive waste. Results to date have shown that living cells are not affected by the dissolved plutonium and that 95% of the plutonium was sorbed to the biomass. This paper discusses the bioprocess, rates of processing, effluent, and the implications of bioprocessing for mixed waste management.

  17. Adsorptive removal of hydrophobic organic compounds by carbonaceous adsorbents: a comparative study of waste-polymer-based, coal-based activated carbon, and carbon nanotubes.

    Science.gov (United States)

    Lian, Fei; Chang, Chun; Du, Yang; Zhu, Lingyan; Xing, Baoshan; Liu, Chang

    2012-01-01

    Adsorption of the hydrophobic organic compounds (HOCs) trichloroethylene (TCE), 1,3-dichlorobenzene (DCB), 1,3-dinitrobenzene (DNB) and gamma-hexachlorocyclohexane (HCH) on five different carbonaceous materials was compared. The adsorbents included three polymer-based activated carbons, one coal-based activated carbon (F400) and multiwalled carbon nanotubes (MWNT). The polymer-based activated carbons were prepared using KOH activation from waste polymers: polyvinyl chloride (PVC), polyethyleneterephthalate (PET) and tire rubber (TR). Compared with F400 and MWNT, activated carbons derived from PVC and PET exhibited fast adsorption kinetics and high adsorption capacity toward the HOCs, attributed to their extremely large hydrophobic surface area (2700 m2/g) and highly mesoporous structures. Adsorption of small-sized TCE was stronger on the tire-rubber-based carbon and F400 resulting from the pore-filling effect. In contrast, due to the molecular sieving effect, their adsorption on HCH was lower. MWNT exhibited the lowest adsorption capacity toward HOCs because of its low surface area and characteristic of aggregating in aqueous solution.

  18. Parametric and kinetic study of adsorptive removal of dyes from aqueous solutions using an agriculture waste

    Science.gov (United States)

    Bencheikh, imane; el hajjaji, souad; abourouh, imane; Kitane, Said; Dahchour, Abdelmalek; El M'Rabet, Mohammadine

    2017-04-01

    Wastewater treatment is the subject of several studies through decades. Interest is continuously oriented to provide cheaper and efficient methods of treatment. Several methods of treatment exit including coagulation flocculation, filtration, precipitation, ozonation, ion exchange, reverse osmosis, advanced oxidation process. The use of these methods proved limited because of their high investment and operational cost. Adsorption can be an efficient low-cost process to remove pollutants from wastewater. This method of treatment calls for an solid adsorbent which constitutes the purification tool. Agricultural wastes have been widely exploited in this case .As we know the agricultural wastes are an important source of water pollution once discharged into the aquatic environment (river, sea ...). The valorization of such wastes and their use allows the prevention of this problem with an economic and environment benefits. In this context our study aimed testing the wastewater treatment capacity by adsorption onto holocellulose resulting from the valorization of an agriculture waste. In this study, methylene blue (MB) and methyl orange (MO) are selected as models pollutants for evaluating the holocellulose adsorbent capacity. The kinetics of adsorption is performed using UV-visible spectroscopy. In order to study the effect of the main parameters for the adsorption process and their mutual interaction, a full factorial design (type nk) has been used.23 full factorial design analysis was performed to screen the parameters affecting dye removal efficiency. Using the experimental results, a linear mathematical model representing the influence of the different parameters and their interactions was obtained. The parametric study showed that efficiency of the adsorption system (Dyes/ Holocellulose) is mainly linked to pH variation. The best yields were observed for MB at pH=10 and for MO at pH=2.The kinetic data was analyzed using different models , namely , the pseudo

  19. Removal of mercury from aqueous solutions using activated carbon prepared from agricultural by-product/waste.

    Science.gov (United States)

    Rao, M Madhava; Reddy, D H K Kumar; Venkateswarlu, Padala; Seshaiah, K

    2009-01-01

    Removal of mercury from aqueous solutions using activated carbon prepared from Ceiba pentandra hulls, Phaseolus aureus hulls and Cicer arietinum waste was investigated. The influence of various parameters such as effect of pH, contact time, initial metal ion concentration and adsorbent dose for the removal of mercury was studied using a batch process. The experiments demonstrated that the adsorption process corresponds to the pseudo-second-order-kinetic models and the equilibrium adsorption data fit the Freundlich isotherm model well. The prepared adsorbents ACCPH, ACPAH and ACCAW had removal capacities of 25.88 mg/g, 23.66 mg/g and 22.88 mg/g, respectively, at an initial Hg(II) concentration of 40 mg/L. The order of Hg(II) removal capacities of these three adsorbents was ACCPH>ACPAH>ACCAW. The adsorption behavior of the activated carbon is explained on the basis of its chemical nature. The feasibility of regeneration of spent activated carbon adsorbents for recovery of Hg(II) and reuse of the adsorbent was determined using HCl solution.

  20. Combustion of animal or vegetable based liquid waste products

    International Nuclear Information System (INIS)

    Wikman, Karin; Berg, Magnus

    2002-04-01

    In this project experiences from combustion of animal and vegetable based liquid waste products have been compiled. Legal aspects have also been taken into consideration and the potential for this type of fuel on the Swedish energy market has been evaluated. Today the supply of animal and vegetable based liquid waste products for energy production in Sweden is limited. The total production of animal based liquid fat is about 10,000 tonnes annually. The animal based liquid waste products origin mainly from the manufacturing of meat and bone meal. Since meat and bone meal has been banned from use in animal feeds it is possible that the amount of animal based liquid fat will decrease. The vegetable based liquid waste products that are produced in the processing of vegetable fats are today used mainly for internal energy production. This result in limited availability on the commercial market. The potential for import of animal and vegetable based liquid waste products is estimated to be relatively large since the production of this type of waste products is larger in many other countries compared to Sweden. Vegetable oils that are used as food or raw material in industries could also be imported for combustion, but this is not reasonable today since the energy prices are relatively low. Restrictions allow import of SRM exclusively from Denmark. This is today the only limit for increased imports of animal based liquid fat. The restrictions for handle and combustion of animal and vegetable based liquid waste products are partly unclear since this is covered in several regulations that are not easy to interpret. The new directive for combustion of waste (2000/76/EG) is valid for animal based waste products but not for cadaver or vegetable based waste products from provisions industries. This study has shown that more than 27,400 tonnes of animal based liquid waste products and about 6,000 tonnes of vegetable based liquid waste products were used for combustion in Sweden

  1. Extraction of long-lived radionuclides from caustic Hanford tank waste supernatants

    International Nuclear Information System (INIS)

    Chaiko, D.J.; Mertz, C.J.; Vojta, Y.

    1995-07-01

    A series of polymer-based extraction systems, based on the use of polyethylene glycols (PEGs) or polypropylene glycols (PPGs), was demonstrated to be capable of selective extraction and recovery of long-lived radionuclides, such as 99 Tc and 129 I, from Hanford SY-101 tank waste, neutralized current acid waste, and single-shell tank waste simulants. During the extraction process, anionic species like TcO 4 - and I - are selectively transferred to the less dense PEG-rich aqueous phase. The partition coefficients for a wide range of inorganic cations and anions, such as sodium, potassium, aluminum, nitrate, nitrite, and carbonate, are all less than one. The partition coefficients for pertechnetate ranged from 12 to 50, depending on the choice of waste simulant and temperature. The partition coefficient for iodide was about 5, while that of iodate was about 0.25. Irradiation of the PEG phase with gamma-ray doses up to 20 Mrad had no detectable effect on the partition coefficients. The most selective extraction systems examined were those based on PPGs, which exhibited separation factors in excess of 3000 between TcO 4 - and NO 3 - /NO 2- . An advantage of the PPG-based system is minimization of secondary waste production. These studies also highlighted the need for exercising great care in extrapolating the partitioning behavior with tank waste simulants to actual tank waste

  2. Relative contributions of natural and waste-derived organics to the subsurface transport of radionuclides

    International Nuclear Information System (INIS)

    Toste, A.P.; Myers, R.B.

    1985-06-01

    Our laboratory is studying the role of organic compounds in the subsurface transport of radionuclides at shallow-land burial sites of low-level nuclear waste, including a commercial site at Maxey Flats, Kentucky, and an aqueous waste disposal site. At the Maxey Flats site, several radionuclides, notably Pu and 60 Co, appear to exist as anionic, organic complexes. Waste-derived organics, particularly chelating agents such as EDTA, HEDTA and associated degradation products (e.g., ED3A), are abundant in aqueous waste leachates and appear to account for the complexation. EDTA, and probably other waste-derived chelating agents as well, are chelated to the Pu and 60 Co in the leachates, potentially mobilizing these radionuclides. In contrast, at the low-level aqueous waste disposal site, naturally-occurring organics, ranging from low molecular weight (MW) acids to high MW humic acids, account for the bulk of the groundwater's organic content. Certain radionuclides, notably 60 Co, 103 Ru and 125 Sb, are mobile as anionic complexes. These radionuclides are clearly associated with higher MW organics, presumably humic and fulvic acids with nominal MW's > 1000. It is clear, therefore, that naturally-occurring organics may play an important role in radionuclide transport, particularly at nuclear waste burial sites containing little in the way of waste-derived organics

  3. Radiolytic gas generation from cement-based waste hosts for DOE low-level radioactive wastes

    International Nuclear Information System (INIS)

    Dole, L.R.; Friedman, H.A.

    1986-01-01

    Using cement-based immobilization binders with simulated radioactive waste containing sulfate, nitrate, nitrite, phosphate, and fluoride anions, the gamma- and alpha-radiolytic gas generation factors (G/sub t/, molecules/100 eV) and gas compositions were measured on specimens of cured grouts. These tests studied the effects of; (1) waste composition; (2) the sample surface-to-volume ratio; (3) the waste slurry particle size; and (4) the water content of the waste host formula. The radiolysis test vessels were designed to minimize the ''dead'' volume and to simulate the configuration of waste packages

  4. Solvent extraction of radionuclides from aqueous tank waste

    International Nuclear Information System (INIS)

    Bonnesen, P.; Sachleben, R.; Moyer, B.

    1996-01-01

    The purpose of this task is to develop an efficient solvent-extraction and stripping process to remove the fission products 99 Tc, 90 Sr, and 137 Cs from alkaline tank waste, such as those stored at Hanford and Oak Ridge. As such, this task expands on FY 1995's successful development of a solvent-extraction and stripping process for technetium separation from alkaline tank-waste solutions. This process now includes the capability of removing both technetium and strontium simultaneously. In this form, the process has been named SRTALK and will be developed further in this program as a prelude to developing a system capable of removing technetium, strontium, and cesium

  5. RADDA - Comparison of results of three ATWS/ATWC scenarios simulated with the help of POLCA-T and S3K/RELAP5

    International Nuclear Information System (INIS)

    Peltonen, J.

    2008-03-01

    The effects of ATWS and ATWC-events with control rods failing to enter the core has been evaluated in this project. To understand the uncertainties in using modern 3D-calculation methods two different codes were used in the project. The outputs from the two code packages were compared. Within the project the used code were first evaluated against a real event, pancake core at Forsmark 3. The results give important knowledge of the core responses for such events and on how to use different code to perform such calculations. The NKS report is only one minor part of the total project. The project was sponsored by TVO, Forsmark, OKG, Ringhals, SKI besides the NKS-funding. The results could be used for PSA-studies and for deterministically safety analysis. (au)

  6. Apparatus for fixing radioactive waste

    International Nuclear Information System (INIS)

    Murphy, J.D.; Pirro, J. Jr.; Lawrence, M.; Wisla, S.F.

    1975-01-01

    Fixing radioactive waste is disclosed in which the waste is collected as a slurry in aqueous media in a metering tank located within the nuclear facilities. Collection of waste is continued from time to time until a sufficient quantity of material to make up a full shipment to a burial ground has been collected. The slurry is then cast in shipping containers for shipment to a burial ground or the like by metering through a mixer into which fixing materials are simultaneously metered at a rate to yield the desired proportions of materials. (U.S.)

  7. Activity-Based Approach for Teaching Aqueous Solubility, Energy, and Entropy

    Science.gov (United States)

    Eisen, Laura; Marano, Nadia; Glazier, Samantha

    2014-01-01

    We describe an activity-based approach for teaching aqueous solubility to introductory chemistry students that provides a more balanced presentation of the roles of energy and entropy in dissolution than is found in most general chemistry textbooks. In the first few activities, students observe that polar substances dissolve in water, whereas…

  8. Extraction and recovery of mercury and lead from aqueous waste streams using redox-active layered metal chalcogenides. Annual progress report, September 15, 1996 - September 14, 1997

    International Nuclear Information System (INIS)

    Dorhout, P.K.; Strauss, S.H.

    1997-01-01

    'The authors have begun to examine the extraction and recovery of heavy elements from aqueous waste streams using redox-active metal chalcogenides. They have been able to prepare extractants from known chalcogenide starting materials, studied the efficacy of the extractants for selective removal of soft metal ions from aqueous phases, studied the deactivation of extractants and the concomitant recovery of soft metal ions from the extractants, and characterized all of the solids and solutions thus far in the study. The study was proposed as two parallel tasks: Part 1 and Part 2 emphasize the study and development of known metal chalcogenide extractants and the synthesis and development of new metal chalcogenide extractants, respectively. The two tasks were divided into sub-sections that study the extractants and their chemistry as detailed below: Preparation and reactivity of metal chalcogenide host solids Extraction of target waste (guest) ions from simulated waste streams Examination of the guest-host solids recovery of the guest metal and reuse of extractant Each section of the two tasks was divided into focused subsections that detail the specific problems and solutions to those problems that were proposed. The extent to which those tasks have been accomplished and the continued efforts of the team are described in detail below. (b) Progress and Results. The DOE-supported research has proceeded largely as proposed and has been productive in its first 12 months. Two full-paper manuscripts were submitted and are currently under peer review. A third paper is in preparation and will be submitted shortly. In addition, 5 submitted or invited presentations have been made.'

  9. 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

  10. Actinide removal from aqueous solution with activated magnetite

    International Nuclear Information System (INIS)

    Kochen, R.L.; Thomas, R.L.

    1987-01-01

    An actinide aqueous waste treatment process using activated magnetite has been developed at Rocky Flats. The use and effectiveness of various magnetites in lowering actinide concentrations in aqueous solution are described. Experiments indicate that magnetite particle size and pretreatment (activation of the magnetite surface with hydroxyl ions greatly influence the effective use of magnetite as an actinide adsorbent. With respect to actinide removal, Ba(OH) 2 -activated magnetite was more effective over a broader pH range than was NaOH-activated magnetite. About 50% less Ba(OH) 2 -activated magnetite was required to lower plutonium concentration from 10 -4 to 10 -8 g/l. 7 refs., 8 tabs

  11. 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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  13. Advanced pyrochemical technologies for minimizing nuclear waste

    International Nuclear Information System (INIS)

    Bronson, M.C.; Dodson, K.E.; Riley, D.C.

    1994-01-01

    The Department of Energy (DOE) is seeking to reduce the size of the current nuclear weapons complex and consequently minimize operating costs. To meet this DOE objective, the national laboratories have been asked to develop advanced technologies that take uranium and plutonium, from retired weapons and prepare it for new weapons, long-term storage, and/or final disposition. Current pyrochemical processes generate residue salts and ceramic wastes that require aqueous processing to remove and recover the actinides. However, the aqueous treatment of these residues generates an estimated 100 liters of acidic transuranic (TRU) waste per kilogram of plutonium in the residue. Lawrence Livermore National Laboratory (LLNL) is developing pyrochemical techniques to eliminate, minimize, or more efficiently treat these residue streams. This paper will present technologies being developed at LLNL on advanced materials for actinide containment, reactors that minimize residues, and pyrochemical processes that remove actinides from waste salts

  14. Cement-Based Materials for Nuclear Waste Storage

    CERN Document Server

    Cau-di-Coumes, Céline; Frizon, Fabien; Lorente, Sylvie

    2013-01-01

    As the re-emergence of nuclear power as an acceptable energy source on an international basis continues, the need for safe and reliable ways to dispose of radioactive waste becomes ever more critical. The ultimate goal for designing a predisposal waste-management system depends on producing waste containers suitable for storage, transportation and permanent disposal. Cement-Based Materials for Nuclear-Waste Storage provides a roadmap for the use of cementation as an applied technique for the treatment of low- and intermediate-level radioactive wastes.Coverage includes, but is not limited to, a comparison of cementation with other solidification techniques, advantages of calcium-silicate cements over other materials and a discussion of the long-term suitability and safety of waste packages as well as cement barriers. This book also: Discusses the formulation and production of cement waste forms for storing radioactive material Assesses the potential of emerging binders to improve the conditioning of problemati...

  15. Adsorptive removal of hydrophobic organic compounds by carbonaceous adsorbents: A comparative study of waste-polymer-based,coal-based activated carbon, and carbon nanotubes

    Institute of Scientific and Technical Information of China (English)

    Fei Lian; Chun Chang; Yang Du; Lingyan Zhu; Baoshan Xing; Chang Liu

    2012-01-01

    Adsorption of the hydrophobic organic compounds (HOCs) trichloroethylene (TCE),1,3-dichlorobenzene (DCB),1,3-dinitrobenzene (DNB) and γ-hexachlorocyclohexane (HCH) on five different carbonaceous materials was compared.The adsorbents included three polymer-based activated carbons,one coal-based activated carbon (F400) and multiwalled carbon nanotubes (MWNT).The polymerbased activated carbons were prepared using KOH activation from waste polymers:polyvinyl chloride (PVC),polyethyleneterephthalate (PET) and tire rubber (TR).Compared with F400 and MWNT,activated carbons derived from PVC and PET exhibited fast adsorption kinetics and high adsorption capacity toward the HOCs,attributed to their extremely large hydrophobic surface area (2700 m2/g) and highly mesoporous structures.Adsorption of small-sized TCE was stronger on the tire-rubber-based carbon and F400 resulting from the pore-filling effect.In contrast,due to the molecular sieving effect,their adsorption on HCH was lower.MWNT exhibited the lowest adsorption capacity toward HOCs because of its low surface area and characteristic of aggregating in aqueous solution.

  16. Equilibrium modeling of removal of drimarine yello HG-3GL dye from aqueous solutions by low cost agricultural waste

    International Nuclear Information System (INIS)

    Bhatti, S.N.H.N.; Sadaf, S.; Sadaf, S.; Farrukh, Z.; Noreen, S.

    2014-01-01

    Pollution control is one of the leading issues of society today. The present study was designed to remove the Drimarine Yellow HF-3GL dye from aqueous solutions through biosorption. Sugarcane bagasse was used as biosorbent in native, acetic acid treated and immobilized form. Batch study was conducted to optimize different system variables like pH of solution, medium temperature, biosorbent concentration, initial dye concentration and contact time. Maximum dye removal was observed at pH 2, biosorbent dose of 0.05 g/50 mL and 40 degree C temperature. The equilibrium was achieved in 45-90 min. Different kinetic and equilibrium models were applied to the experimental results. The biosorption kinetic data was found to follow the pseudo second order kinetic model. Freundlich adsorption isotherm model showed a better fitness to the equilibrium data. The value of Gibbs free energy revealed that biosorption of Drimarine Yellow HF-3GL dye by native and pretreated sugarcane bagasse was a spontaneous process. Presence of salt and heavy metal ions in aqueous solution enhanced the biosorption capacity while presence of surfactants decreased the biosorption potential of biosorbent. Dye was desorbed by 1M NaOH solution. Fixed bed column study of Drimarine Yellow HF-3GL was carried out to optimize different parameters like bed height, flow rate and initial dye concentration. It was observed that biosorption capacity increases with increase in initial dye concentration and bed height but decreases with the increase in flow rate. The data of column study was explained very well by BDST model. FT-IR analysis confirmed the involvement of various functional groups, mainly hydroxyl, carboxyl and amine groups. The results proved that sugarcane bagasse waste biomass can be used as a favorable biosorbent for the removal of dyes from aqueous solutions. (author)

  17. Recovery of metals from simulant spent lithium-ion battery as organophosphonate coordination polymers in aqueous media

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Emilie; Andre, Marie-Laure; Navarro Amador, Ricardo [ICSM, Institut de Chimie Séparative de Marcoule, UMR 5257, CEA/CNRS/ENSCM/UM, Bât 426, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Hyvrard, François; Borrini, Julien [SARPI VEOLIA, Direction Technique et Innovations, Zone portuaire de Limay-Porcheville, 427 route du Hazay, 78520 Limay (France); Carboni, Michaël, E-mail: michael.carboni@cea.fr [ICSM, Institut de Chimie Séparative de Marcoule, UMR 5257, CEA/CNRS/ENSCM/UM, Bât 426, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Meyer, Daniel [ICSM, Institut de Chimie Séparative de Marcoule, UMR 5257, CEA/CNRS/ENSCM/UM, Bât 426, BP 17171, 30207 Bagnols-sur-Cèze cedex (France)

    2016-11-05

    Highlights: • Original waste disposal strategies for battery. • Precipitation of metals as coordination polymers. • Organo-phosphonate coordination polymers. • Selective extraction of manganese or co-precipitation of manganese/cobalt. • The recycling process give a promising application on any waste solution. - Abstract: An innovative approach is proposed for the recycling of metals from a simulant lithium-ion battery (LIBs) waste aqueous solution. Phosphonate organic linkers are introduced as precipitating agents to selectively react with the metals to form coordination polymers from an aqueous solution containing Ni, Mn and Co in a hydrothermal process. The supernatant is analyzed by ICP-AES to quantify the efficiency and the selectivity of the precipitation and the materials are characterized by Scanning Electron Microscopy (SEM), Powder X-Ray Diffraction (PXRD), Thermogravimetric Analyses (TGA) and nitrogen gas sorption (BET). Conditions have been achieved to selectively precipitate Manganese or Manganese/Cobalt from this solution with a high efficiency. This work describes a novel method to obtain potentially valuable coordination polymers from a waste metal solution that can be generalized on any waste solution.

  18. A highly reversible anthraquinone-based anolyte for alkaline aqueous redox flow batteries

    Science.gov (United States)

    Cao, Jianyu; Tao, Meng; Chen, Hongping; Xu, Juan; Chen, Zhidong

    2018-05-01

    The development of electroactive organic materials for use in aqueous redox flow battery (RFB) electrolytes is highly attractive because of their structural flexibility, low cost and sustainability. Here, we report on a highly reversible anthraquinone-based anolyte (1,8-dihydroxyanthraquinone, 1,8-DHAQ) for alkaline aqueous RFB applications. Electrochemical measurements reveal the substituent position of hydroxyl groups for DHAQ isomers has a significant impact on the redox potential, electrochemical reversibility and water-solubility. 1,8-DHAQ shows the highest redox reversibility and rapidest mass diffusion among five isomeric DHAQs. The alkaline aqueous RFB using 1,8-DHAQ as the anolyte and potassium ferrocyanide as the catholyte yields open-circuit voltage approaching 1.1 V and current efficiency and capacity retention exceeding 99.3% and 99.88% per cycle, respectively. This aqueous RFB produces a maximum power density of 152 mW cm-2 at 100% SOC and 45 °C. Choline hydroxide was used as a hydrotropic agent to enhance the water-solubility of 1,8-DHAQ. 1,8-DHAQ has a maximum solubility of 3 M in 1 M KOH with 4 M choline hydroxide.

  19. Secondary Waste Cementitious Waste Form Data Package for the Integrated Disposal Facility Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-16

    A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at the Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.

  20. Method of decomposing treatment for radioactive organic phosphate wastes

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  1. Frozen soil barriers for hazardous waste confinement

    International Nuclear Information System (INIS)

    Dash, J.G.; Leger, R.; Fu, H.Y.

    1997-01-01

    Laboratory and full field measurements have demonstrated the effectiveness of artificial ground freezing for the containment of subsurface hazardous and radioactive wastes. Bench tests and a field demonstration have shown that cryogenic barriers are impenetrable to aqueous and non aqueous liquids. As a result of the successful tests the US Department of Energy has designated frozen ground barriers as one of its top ten remediation technologies

  2. Silk Fibroin Aqueous-Based Adhesives Inspired by Mussel Adhesive Proteins.

    Science.gov (United States)

    Burke, Kelly A; Roberts, Dane C; Kaplan, David L

    2016-01-11

    Silk fibroin from the domesticated silkworm Bombyx mori is a naturally occurring biopolymer with charged hydrophilic terminal regions that end-cap a hydrophobic core consisting of repeating sequences of glycine, alanine, and serine residues. Taking inspiration from mussels that produce proteins rich in L-3,4-dihydroxyphenylalanine (DOPA) to adhere to a variety of organic and inorganic surfaces, the silk fibroin was functionalized with catechol groups. Silk fibroin was selected for its high molecular weight, tunable mechanical and degradation properties, aqueous processability, and wide availability. The synthesis of catechol-functionalized silk fibroin polymers containing varying amounts of hydrophilic polyethylene glycol (PEG, 5000 g/mol) side chains was carried out to balance silk hydrophobicity with PEG hydrophilicity. The efficiency of the catechol functionalization reaction did not vary with PEG conjugation over the range studied, although tuning the amount of PEG conjugated was essential for aqueous solubility. Adhesive bonding and cell compatibility of the resulting materials were investigated, where it was found that incorporating as little as 6 wt % PEG prior to catechol functionalization resulted in complete aqueous solubility of the catechol conjugates and increased adhesive strength compared with silk lacking catechol functionalization. Furthermore, PEG-silk fibroin conjugates maintained their ability to form β-sheet secondary structures, which can be exploited to reduce swelling. Human mesenchymal stem cells (hMSCs) proliferated on the silks, regardless of PEG and catechol conjugation. These materials represent a protein-based approach to catechol-based adhesives, which we envision may find applicability as biodegradable adhesives and sealants.

  3. Heterogeneous catalysis contribution for the denitration of aqueous nuclear radioactive waste with formic acid

    International Nuclear Information System (INIS)

    Guenais, S.

    2001-01-01

    The chemical denitration aims to reduce the nitric acid concentration in nuclear fuel reprocessing aqueous wastes by adding formic acid as a reducing agent. The denitration reaction exhibits an induction period, which duration is related to the time needed by the key intermediate of the reaction, i.e. nitrous acid, to reach a threshold concentration in the reaction medium. The addition of a Pt/SiO 2 catalyst in the reaction mixture suppresses the induction period of the chemical denitration. The aim of the present work is to identify the role of Pt/SiO 2 in the catalytic denitration mechanism. The experimental work is based on the comparison of catalytic tests performed with various catalysts, in order to identify the intrinsic characteristics of Pt/SiO 2 that might influence its activity for the reaction. Catalytic denitration results show that Pt/SiO 2 acts only by speeding up the nitrous acid generation in the solution until its concentration reaches the threshold level of 0,01 mol L -1 in the experimental conditions. Catalysts activity is evaluated by quantifying the nitrous acid generated on the platinum surface during the induction period of the homogeneous denitration reaction. The large platinum aggregates reactivity is greater than the one of nano-sized particles. The study of the key step of the catalytic denitration reaction, the catalytic generation of nitrous acid, clarifies the role of Pt/SiO 2 . The homogeneous denitration is initiated thanks to a redox cycle on the catalyst surface: an initial oxidation of Pt 0 by nitric acid, which is reduced into nitrous acid, followed by the reduction of the passivated 'Pt ox ' by formic acid. Furthermore, a platinum reduction by formic acid prior to the catalytic test prevents any platinum leaching from the catalyst into the nitric solution, being all the more significant as platinum dispersion is high. (author)

  4. A reaction-based fluorescent sensor for detection of cyanide in aqueous media

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shan-Teng; Sie, Yi-Wun [Department of Chemistry, National Changhua University of Education, Changhua 50058, Taiwan (China); Wan, Chin-Feng [School of Applied Chemistry, Chung Shan Medical University, Taichung City 40201, Taiwan (China); Wu, An-Tai, E-mail: antai@cc.ncue.edu.tw [Department of Chemistry, National Changhua University of Education, Changhua 50058, Taiwan (China)

    2016-05-15

    A simple boronic acid derivative was utilized as a reaction-based receptor for CN{sup −} in aqueous solution. The receptor showed a selective and sensitive response to CN{sup −} over other various anions via nucleophilic addition of CN{sup −} to the imine moiety of the boronic-based receptor.

  5. Radioactive waste management

    International Nuclear Information System (INIS)

    Blomek, D.

    1980-01-01

    The prospects of nuclear power development in the USA up to 2000 and the problems of the fuel cycle high-level radioactive waste processing and storage are considered. The problems of liquid and solidified radioactive waste transportation and their disposal in salt deposits and other geologic formations are discussed. It is pointed out that the main part of the high-level radioactive wastes are produced at spent fuel reprocessing plants in the form of complex aqueous mixtures. These mixtures contain the decay products of about 35 isotopes which are the nuclear fuel fission products, about 18 actinides and their daughter products as well as corrosion products of fuel cans and structural materials and chemical reagents added in the process of fuel reprocessing. The high-level radioactive waste management includes the liquid waste cooling which is necessary for the short and middle living isotope decay, separation of some most dangerous components from the waste mixture, waste solidification, their storage and disposal. The conclusion is drawn that the seccessful solution of the high-level radioactive waste management problem will permit to solve the problem of the fuel cycle radioactive waste management as a whole. The salt deposits, shales and clays are the most suitable for radioactive waste disposal [ru

  6. Photovoltaic's silica-rich waste sludge as supplementary cementitious material

    NARCIS (Netherlands)

    Quercia, G.; Van der Putten, J.J.G.; Brouwers, H.J.H.

    2013-01-01

    Waste sludge, a solid recovered from wastewater of photovoltaic-industries, composes of agglomerates of nano-particles like SiO2 and CaCO3. This sludge deflocculates in aqueous solutions into nano-particles smaller than 1000 nm. Thus, this sludge is potentially hazardous waste when is improperly

  7. General-base catalysed hydrolysis and nucleophilic substitution of activated amides in aqueous solutions

    NARCIS (Netherlands)

    Buurma, NJ; Blandamer, MJ; Engberts, JBFN; Buurma, Niklaas J.

    The reactivity of 1-benzoyl-3-phenyl-1,2,4-triazole (1a) was studied in the presence of a range of weak bases in aqueous solution. A change in mechanism is observed from general-base catalysed hydrolysis to nucleophilic substitution and general-base catalysed nucleophilic substitution. A slight

  8. Scavenging of priority organic pollutants from aqueous waste using granular activated carbon

    Energy Technology Data Exchange (ETDEWEB)

    Singh, S.; Yenkie, M.K.N. [Central Fuel Research Institute, Nagpur (India)

    2006-04-15

    Many organic compounds present in industrial and domestic wastewaters are carcinogenic in nature. Removal of these organic compounds from wastewater has become a great challenge to wastewater treatment technologies, as many of them are non-biodegradable in nature. Adsorption on granular activated carbon (GAC) has emerged an efficient and economically viable technology for removal of final traces of a broad spectrum of toxic organic compounds from domestic and industrial wastewater. In the present investigation adsorption of some priority organic pollutants, namely phenol, o-cresol, p-nitrophenol, m-methoxyphenol, benzoic acid and salicylic acid on granular activated carbon, was studied in a batch system at laboratory scale. Experiments were carried out to determine adsorption isotherms and kinetics for adsorbate when present in aqueous solutions as single, bi- and tri-solute systems. The commercially available bituminous coal based granular activated carbon Filtrasorb 300 (F-300) was used as adsorbent. The results indicate that p-nitrophenol is most strongly adsorbed as compared to other phenols studied. Aqueous phase solubility of the adsorbate plays a deciding role in multi-component systems as more hydrophobic p-nitrophenol adsorbs to a greater extent than less hydrophobic phenol, o-cresol and m-methoxyphenol. The preferential adsorption of strongly adsorbable solute over a weakly adsorbable one has been observed, as the solutes are competing for the available surface area of the adsorbent for adsorption.

  9. Extraction of long-lived radionuclides from caustic Hanford tank waste supernatants

    Energy Technology Data Exchange (ETDEWEB)

    Chaiko, D.J.; Mertz, C.J.; Vojta, Y. [and others

    1995-07-01

    A series of polymer-based extraction systems, based on the use of polyethylene glycols (PEGs) or polypropylene glycols (PPGs), was demonstrated to be capable of selective extraction and recovery of long-lived radionuclides, such as {sup 99}Tc and {sup 129}I, from Hanford SY-101 tank waste, neutralized current acid waste, and single-shell tank waste simulants. During the extraction process, anionic species like TcO{sub 4}{sup {minus}} and I{sup {minus}} are selectively transferred to the less dense PEG-rich aqueous phase. The partition coefficients for a wide range of inorganic cations and anions, such as sodium, potassium, aluminum, nitrate, nitrite, and carbonate, are all less than one. The partition coefficients for pertechnetate ranged from 12 to 50, depending on the choice of waste simulant and temperature. The partition coefficient for iodide was about 5, while that of iodate was about 0.25. Irradiation of the PEG phase with gamma-ray doses up to 20 Mrad had no detectable effect on the partition coefficients. The most selective extraction systems examined were those based on PPGs, which exhibited separation factors in excess of 3000 between TcO{sub 4}{sup {minus}} and NO{sub 3}{sup {minus}}/NO{sub 2}{sub {minus}}. An advantage of the PPG-based system is minimization of secondary waste production. These studies also highlighted the need for exercising great care in extrapolating the partitioning behavior with tank waste simulants to actual tank waste.

  10. Retention of barium and europium radionuclides from aqueous solutions on ash-based sorbents by application of radiochemical techniques.

    Science.gov (United States)

    Noli, Fotini; Kapnisti, Maria; Buema, Gabriela; Harja, Maria

    2016-10-01

    New materials were synthesized for application in sorption of radionuclides from aqueous solutions. The elaboration was performed by conversion of power plant ash using the hydrothermal method under optimum experimental conditions. Sodalite, Na-Y, and analcime were formed from ash precursor during the treatment, exhibiting thermal stability as revealed by the characterization by X-ray diffraction (XRD) and thermogravimetric differential thermal analysis (TG-DTA). The Brunauer-Emmett-Teller (BET) surface area and pore volume were determined and they presented higher values than plant ash. The ability of the new products to retain Ba and Eu radionuclides was studied in aqueous solutions using (133)Ba and (152)Eu as tracers and γ-ray spectroscopy under batch experiments. The experimental data were modeled by the Langmuir and Freundlich equations, whereas sorption kinetics measurements were performed at 293, 308, and 323K and thermodynamic parameters were calculated. The release of the sorbed ions into the environment was also tested by leaching experiments. The results of these tests indicated that the synthesized materials are very efficient in removing the aforementioned metals from aqueous solutions and can be considered as potential low-cost sorbents in nuclear waste management. Copyright © 2016 Elsevier Ltd. All rights reserved.

  11. Speciation analysis on Eu(3) in aqueous solution using laser-induced breakdown spectroscopy

    International Nuclear Information System (INIS)

    Hotokezaka, H.; Tanaka, S.; Nagasaki, S.

    2001-01-01

    Investigation of the chemical behaviour of lanthanides and actinides in the geosphere is important for the safety assessment of high-level radioactive waste disposal. However, determination of speciation for lanthanides and actinides is difficult, because it is too hard to distinguish between metal ion and colloidal metal in aqueous solution. Laser-induced breakdown spectroscopy (LIBS) can detect both ions and microparticles of metals in aqueous solution, especially, high sensitive to microparticles. In this study, we analysed Eu(III) ion and Eu 2 O 3 particle in aqueous solution by LIBS, and measured the hydrolysis behaviour of Eu(III) in aqueous solution. Furthermore, we tried to detect the plasma emission of Eu(III) ions sorbed on TiO 2 particles, and also tried to observe the adsorption behaviour of Eu(III) ions onto TiO 2 particles in aqueous solution. (authors)

  12. POTENTIALS OF AGRICULTURAL WASTE AND GRASSES IN ...

    African Journals Online (AJOL)

    Shima

    Potentials of some agricultural waste and grasses were investigated. ... to education, printing, publishing and ... technical form, paper is an aqueous deposit ..... Period of. Soaking. Overnight. Overnight. Overnight. Overnight. Overnight.

  13. Mixed Waste Management Facility

    International Nuclear Information System (INIS)

    Brummond, W.; Celeste, J.; Steenhoven, J.

    1993-08-01

    The DOE has developed a National Mixed Waste Strategic Plan which calls for the construction of 2 to 9 mixed waste treatment centers in the Complex in the near future. LLNL is working to establish an integrated mixed waste technology development and demonstration system facility, the Mixed Waste Management Facility (MWMF), to support the DOE National Mixed Waste Strategic Plan. The MWMF will develop, demonstrate, test, and evaluate incinerator-alternatives which will comply with regulations governing the treatment and disposal of organic mixed wastes. LLNL will provide the DOE with engineering data for design and operation of new technologies which can be implemented in their mixed waste treatment centers. MWMF will operate under real production plant conditions and process samples of real LLNL mixed waste. In addition to the destruction of organic mixed wastes, the development and demonstration will include waste feed preparation, material transport systems, aqueous treatment, off-gas treatment, and final forms, thus making it an integrated ''cradle to grave'' demonstration. Technologies from offsite as well as LLNL's will be tested and evaluated when they are ready for a pilot scale demonstration, according to the needs of the DOE

  14. Treatment of aqueous wastes contaminated with Congo Red dye by electrochemical oxidation and ozonation processes

    International Nuclear Information System (INIS)

    Faouzi Elahmadi, Mohammed; Bensalah, Nasr; Gadri, Abdellatif

    2009-01-01

    Synthetic aqueous wastes polluted with Congo Red (CR) have been treated by two advanced oxidation processes: electrochemical oxidation on boron doped diamond anodes (BDD-EO) and ozonation under alkaline conditions. For same concentrations, galvanostatic electrolyses have led to total COD and TOC removals but ozonation process can reach only 85% and 81% of COD and TOC removals, respectively. UV-vis qualitative analyses have shown different behaviors of CR molecules towards ozonation and electrochemical oxidation. Rapid discoloration has been observed during ozonation, whereas color persistence till the end of galvanostatic electrolyses has been seen during BDD-EO process. It seems that the oxidation mechanisms involved in the two processes are different: simultaneous destruction of azoic groups is suggested during ozonation process but consecutive destruction of these groups is proposed during BDD-EO. However, energetic study has evidenced that BDD-EO appears more efficient and more economic than ozonation in terms of TOC removals. These results have been explained by the fact that during BDD-EO, other strong oxidants electrogenerated from the electrolyte oxidation such as persulfates and direct-oxidation of CR and its byproducts on BDD anodes complement the hydroxyl radicals mediated oxidation to accomplish the total mineralization of organics.

  15. Hazardous Waste Minimization Assessment: Fort Campbell, Kentucky

    Science.gov (United States)

    1991-03-01

    gal/h -- $8,250 (solvents: chlorinated and $8,600 fluorinated ) 114 Table 39 Aqueous Waste Volume Reduction Equipment Suppliers* Supplier Model Capacity...heavy chloride/hydrochloric acid metal solutions (chromium), nitric acid (zinc, magnesium) Printing (Ink) pigments, dyes, varnish , titanium oxide, iron...lacquers, epoxy. aLkyds. acrylics) :inshing Varnish . shellac, lacquer 13001 Waste flammable liquid. NOS Flammable liquid UN1993 Preserving Creosote

  16. Studies on the recovery of 233U from phosphate containing aqueous waste using DBDECMP as extractant

    International Nuclear Information System (INIS)

    Sagar, V.B.; Oak, M.S.; Pawar, S.M.; Sivaramakrishnan, C.K.; Patil, S.K.

    1990-01-01

    A method for the recovery and purification of 233 U from phosphate containing analytical waste is developed. Extraction studies with Di-butyl N,N-diethylcarbamoylmethylphosphonate (DBDECMP) in xylene were carried out to explore the feasibility of separation and purification of 233 U from such wastes. Based on the data obtained, optimum conditions for the recovery of 233 U are suggested. (author) 11 refs.; 1 fig.; 3 tabs

  17. Physicochemical changes of cements by ground water corrosion in radioactive waste storage

    International Nuclear Information System (INIS)

    Contreras R, A.; Badillo A, V. E.; Robles P, E. F.; Nava E, N.

    2009-10-01

    Knowing that the behavior of cementations materials based on known hydraulic cement binder is determined essentially by the physical and chemical transformation of cement paste (water + cement) that is, the present study is essentially about the cement paste evolution in contact with aqueous solutions since one of principal risks in systems security are the ground and surface waters, which contribute to alteration of various barriers and represent the main route of radionuclides transport. In this research, cements were hydrated with different relations cement-aqueous solution to different times. The pastes were analyzed by different solid observation techniques XRD and Moessbauer with the purpose of identify phases that form when are in contact with aqueous solutions of similar composition to ground water. The results show a definitive influence of chemical nature of aqueous solution as it encourages the formation of new phases like hydrated calcium silicates, which are the main phases responsible of radionuclides retention in a radioactive waste storage. (Author)

  18. Kinetics and thermodynamics of aqueous Cu(II) adsorption on heat ...

    African Journals Online (AJOL)

    This study investigated the kinetics and thermodynamics of copper(II) removal from aqueous solutions using spent bleaching earth (SBE). The spent bleaching earth, a waste material from edible oil processing industries, was reactivated by heat treatment at 370 oC after residual oil extraction in excess methyl-ethyl ketone.

  19. Waste reduction at the Savannah River Site

    International Nuclear Information System (INIS)

    Stevens, W.E.; Lee, R.A.; Reynolds, R.W.

    1990-01-01

    The Savannah River Site (SRS) is a key installation for the production and research of nuclear materials for national defense and peace time applications and has been operating a full nuclear fuel cycle since the early 1950s. Wastes generated include high level radioactive, transuranic, low level radioactive, hazardous, mixed, sanitary, and aqueous wastes. Much progress has been made during the last several years to reduce these wastes including management systems, characterization, and technology programs. The reduction of wastes generated and the proper handling of the wastes have always been a part of the Site's operation. This paper summarizes the current status and future plans with respect to waste reduction to waste reduction and reviews some specific examples of successful activities

  20. Waste Package and Material Testing for the Proposed Yucca Mountain High Level Waste Repository

    International Nuclear Information System (INIS)

    Doering, Thomas; Pasupathi, V.

    2002-01-01

    Over the repository lifetime, the waste package containment barriers will perform various functions that will change with time. During the operational period, the barriers will function as vessels for handling, emplacement, and waste retrieval (if necessary). During the years following repository closure, the containment barriers will be relied upon to provide substantially complete containment, through 10,000 years and beyond. Following the substantially complete containment phase, the barriers and the waste package internal structures help minimize release of radionuclides by aqueous- and gaseous-phase transport. These requirements have lead to a defense-in-depth design philosophy. A multi-barrier design will result in a lower breach rate distributed over a longer period of time, thereby ensuring the regulatory requirements are met. The design of the Engineered Barrier System (EBS) has evolved. The initial waste package design was a thin walled package, 3/8 inch of stainless steel 304, that had very limited capacity, (3 PWR and 4 BWR assemblies) and performance characteristics, 300 to 1,000 years. This design required over 35,000 waste packages compared to today's design of just over 10,000 waste packages. The waste package designs are now based on a defense-in-depth/multi-barrier philosophy and have a capacity similar to the standard storage and rail transported spent nuclear fuel casks. Concurrent with the development of the design of the waste packages, a comprehensive waste package materials testing program has been undertaken to support the selection of containment barrier materials and to develop predictive models for the long-term behavior of these materials under expected repository conditions. The testing program includes both long-term and short-term tests and the results from these tests combination with the data published in the open literature are being used to develop models for predicting performance of the waste packages

  1. Kinetic and thermodynamic aspects of Cu(II) and Cr(III) removal from aqueous solutions using rose waste biomass

    International Nuclear Information System (INIS)

    Iftikhar, Abdur Rauf; Bhatti, Haq Nawaz; Hanif, Muhammad Asif; Nadeem, Razyia

    2009-01-01

    Distillation waste of rose petals was used to remove Cu(II) and Cr(III) from aqueous solutions. The results demonstrated the dependency of metal sorption on pH, sorbent dose, sorbent size, initial bulk concentration, time and temperature. A dosage of 1 g/L of rose waste biomass was found to be effective for maximum uptake of Cu(II) and Cr(III). Optimum sorption temperature and pH for Cu(II) and Cr(III) were 303 ± 1 K and 5, respectively. The Freundlich regression model and pseudo-second-order kinetic model were resulted in high correlation coefficients and described well the sorption of Cu(II) and Cr(III) on rose waste biomass. At equilibrium q max (mg/g) of Cu(II) and Cr(III) was 55.79 and 67.34, respectively. The free energy change (ΔG o ) for Cu(II) and Cr(III) sorption process was found to be -0.829 kJ/mol and -1.85 kJ/mol, respectively, which indicates the spontaneous nature of sorption process. Other thermodynamic parameters such as entropy change (ΔS o ), enthalpy (ΔH o )and activation energy (ΔE) were found to be 0.604 J mol -1 K -1 , -186.95 kJ/mol and 68.53 kJ/mol, respectively for Cu(II) and 0.397 J mol -1 K -1 , -119.79 kJ/mol and 114.45 kJ/mol, respectively for Cr(III). The main novelty of this work was the determination of shortest possible sorption time for Cu(II) and Cr(III) in comparison to earlier studies. Almost over 98% of Cu(II) and Cr(III) were removed in only first 20 min at an initial concentration of 100 mg/L

  2. The German act on the reorganisation of responsibility in nuclear waste management; Des Gesetz zur Neuordnung der Verantwortung in der kerntechnischen Entsorgung

    Energy Technology Data Exchange (ETDEWEB)

    Raetzke, Christian [CONLAR Consulting on Nuclear Law and Regulation, Leipzig (Germany)

    2017-04-15

    The author discussed the Draft on the Act in the Reorganisation of Responsibility in Nuclear Waste Management in atw 12 (2016). Now, amendments are discussed, which resulted from the legislative procedure until today's draft. Significant additions affect the authorisation for the conclusion of a public-law contract between the Federal Government and the nuclear power plant operators, the deadline for the payment of the basic amount, and the option for the operation of the interim storage facilities for a transitional period by the operators on behalf of the federal company. Since the adoption of the draft act, it has become clear that the nuclear power plant operators will pay the risk premium. This will fulfil the full logic of the new system. It has also become known, that the public law contract is now ready for signing. According to the author, the act will bring a final arrangement for financing nuclear waste disposal. However, adjustment can not be avoided in practice. The concrete implementation will be a exciting topic in many ways.

  3. Study of the europium behavior in aqueous media

    International Nuclear Information System (INIS)

    Fernandez R, E.; Jimenez R, M.; Solache R, M.; Martinez M, V.

    1999-01-01

    Europium as waste can produce a pollution problem in water that is in contact with it, what would has a heavy environmental impacts, because of the possibilities of diffusion of these wastes from their place of confinement or storage until the geo and biosphere. The solution of such problem requires of a lot of knowledge over the behavior of several chemical elements such as europium in aqueous solutions. In this work it was used a low ion force (0.02 M). The data set will allow extrapolate the hydrolytic behavior of europium in too much minors ion force media, such as the ground waters, including in ion force zero

  4. Determination of performance criteria for high-level solidified nuclear waste from the commercial nuclear fuel cycle: a probabilistic safety analysis

    International Nuclear Information System (INIS)

    Heckman, R.A.

    1978-01-01

    To minimize the radiological risk from the operation of a waste management system for the safe disposal of high-level waste, performance characteristics of the solidified waste form must be specified. The minimum waste form characteristics that must be specified are the radionuclide volatilization fraction, airborne particulate dispersion fraction, and the aqueous dissolution characteristics. The results indicate that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. The actual values of expected risk are sensitive to modeling assumptions and data base uncertainties. The transportation step appears to be the most limiting in determining the required performance characteristics

  5. The potential of coconut fibers in raw and treated forms to remove 241Am from aqueous solutions

    International Nuclear Information System (INIS)

    Fonseca, Heverton C.O.; Jesus, Nella N.M. de; Nobre, Vanessa B.; Potiens Junior, Ademar; Sakata, Solange K.

    2013-01-01

    In the Radioactive Waste Management (GRR) at the Nuclear and Energy Research Institute (IPEN/CNEN-SP) vegetal biomass has been studied as adsorbent to remove radioisotopes from radioactive liquid wastes as part of the radioactive waste treatment program. In this work coconuts fiber was evaluate as biosorbent to remove 241 Am from aqueous solutions and many different parameters were studied such as particle size (35 and 80 mesh) and contact time (between 5 and 60 minutes). In order to evaluate if the treated biomass could remove more 241 Am the experiments were also performed using raw biomass and treated with H 2 O 2 in basic conditions. When the experiment was carried out using raw coconuts fiber with 80 mesh, neutral conditions after 5 minutes of contact time 99% of the radionuclide was removed from the aqueous solution. This result shows the potential of this biomass to remove 241 Am from radioactive liquid wastes. (author)

  6. 1990 waste tank inspection program

    International Nuclear Information System (INIS)

    McNatt, F.G.

    1990-01-01

    Aqueous radioactive wastes from Savannah River Site separations processes are contained in large underground carbon steel tanks. Tank conditions are evaluated by inspection using periscopes, still photography, and video systems for visual imagery. Inspections made in 1990 are the subject of this report

  7. Lixiviation of plutonium contaminated solid wastes by aqueous solution of electro-generated reducing agents

    International Nuclear Information System (INIS)

    Agarande, Michelle

    1991-01-01

    This study concerns the development of the new concept for the decontamination of plutonium bearing solid wastes, based on the lixiviation of the wastes using electro-generated reducing agents. First, a comparative study of the kinetics of the dissolution of pure PuO 2 (prepared by calcination of Pu (IV) oxalate at 450 C) in sulfuric acid media, with different reducing agents, was realized. Qualitatively these reagents can be sorted in three groups: 1 / fast kinetics for Cr(II), V(II) and U(III); 2 / slow kinetics for Ti(III); 3 / very slow kinetics for V(III) and U(VI). In order to contribute to the design of an electrochemical reactor for the generation of the reducing agents usable for the lixiviation of plutonium bearing solid wastes, the study of the diffusion coefficients of both oxidized and reduced forms of different redox couples, at different temperatures, was undertaken. The results of this study also permits, from the knowledge of the diffusional activation energy of the ions, to conclude that the dissolution of pure plutonium dioxide under the action of these reducing agents is not diffusion limited. The feasibility of the plutonium decontamination treatment of synthetic or real solid wastes was then studied at laboratory scale using electro-generated V(II), which is with Cr(II) among the best reagents. The efficiency of the treatment was good, (80 pc Pu solubilisation yield), especially in the case of cellulosic or miscellaneous organic wastes. (author) [fr

  8. Immobilization of aqueous radioactive cesium wastes by conversion to aluminosilicate minerals

    International Nuclear Information System (INIS)

    Barney, G.S.

    1975-05-01

    Radioactive cesium (primarily 137 Cs) is a major toxic constituent of liquid wastes from nuclear fuel processing plants. Because of the long half-life, highly penetrating radiation, and mobility of 137 Cs, it is necessary to convert wastes containing this radioisotope into a solid form which will prevent movement to the biosphere during long-term storage. A method for converting cesium wastes to solid, highly insoluble, thermally stable aluminosilicate minerals is described. Aluminum silicate clays (bentonite, kaolin, or pyrophyllite) or hydrous aluminosilicate gels are reacted with basic waste solutions to form pollucite, cesium zeolite (Cs-D), Cs-F, cancrinite, or nepheline. Cesium is trapped in the aluminosilicate crystal lattice of the mineral and is permanently immobilized. The identity of the mineral product is dependent on the waste composition and the SiO 2 /Al 2 O 3 ratio of the clay or gel. The stoichiometry and kinetics of mineral formation reactions are described. The products are evaluated with respect to leachability, thermal stability, and crystal morphology. (U.S.)

  9. DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS

    International Nuclear Information System (INIS)

    Jantzen, C

    2006-01-01

    Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied

  10. Organic non-aqueous cation-based redox flow batteries

    Science.gov (United States)

    Zhang, Lu; Huang, Jinhua; Burrell, Anthony

    2018-05-08

    The present invention provides a non-aqueous redox flow battery comprising a negative electrode immersed in a non-aqueous liquid negative electrolyte, a positive electrode immersed in a non-aqueous liquid positive electrolyte, and a cation-permeable separator (e.g., a porous membrane, film, sheet, or panel) between the negative electrolyte from the positive electrolyte. During charging and discharging, the electrolytes are circulated over their respective electrodes. The electrolytes each comprise an electrolyte salt (e.g., a lithium or sodium salt), a transition-metal free redox reactant, and optionally an electrochemically stable organic solvent. Each redox reactant is selected from an organic compound comprising a conjugated unsaturated moiety, a boron cluster compound, and a combination thereof. The organic redox reactant of the positive electrolyte comprises a tetrafluorohydroquinone ether compound or a tetrafluorocatechol ether compound.

  11. Photovoltaic's silica-rich waste sludge as supplementary cementitious materials (SCM)

    NARCIS (Netherlands)

    Quercia Bianchi, G.; van der Putten, J.J.G.; Brouwers, H.J.H.; Uzoegbo, H.C.; Schmidt, W.

    2013-01-01

    Waste sludge, a solid recovered from wastewater of photovoltaic-industries, composes of agglomerates of nano-particles like SiO2 and CaCO3. This sludge deflocculates in aqueous solutions into nano-particles smaller than 1000 nm. Thus, this sludge is potentially hazardous waste when is improperly

  12. Biosorption of nickel (II) ions from aqueous solutions by tapioca peel ...

    African Journals Online (AJOL)

    Tapioca peel, waste from native tapioca starch industry in Thailand, was used for the biosorption of nickel from aqueous solution. The experimental parameter focuses on the influence of contact time, solution pH, initial concentration and temperature using batch experiments. The results indicated that the biosorption ...

  13. Method of solidifying radioactive wastes

    International Nuclear Information System (INIS)

    Tomita, Toshihide; Minami, Yuji; Matsuura, Hiroyuki; Kageyama, Hisashi; Kobori, Junzo.

    1984-01-01

    Purpose: To perform the curing sufficiently even when copper hydroxide that interferes the curing reaction is contained in radioactive wastes. Method: Solidification of radioactive wastes containing copper hydroxide using thermoset resins is carried out under the presence of an alkaline material. The thermoset resin used herein is an polyester resin comprising unsaturated polyester and a polymerizable monomer. The alkaline substance usable herein can include powder or an aqueous solution of hydroxides or oxides of sodium, magnesium, calcium or the like. (Yoshino, Y.)

  14. Waste to Want: Polymer nanocomposites using nanoclays extracted from Oil based drilling mud waste

    International Nuclear Information System (INIS)

    Adegbotolu, Urenna V; Njuguna, James; Pollard, Pat; Yates, Kyari

    2014-01-01

    Due to the European Union (EU) waste frame work directive (WFD), legislations have been endorsed in EU member states such as UK for the Recycling of wastes with a vision to prevent and reduce landfilling of waste. Spent oil based drilling mud (drilling fluid) is a waste from the Oil and Gas industry with great potentials for recycling after appropriate clean-up and treatment processes. This research is the novel application of nanoclays extracted from spent oil based drilling mud (drilling fluid) clean-up as nanofiller in the manufacture of nanocomposite materials. Research and initial experiments have been undertaken which investigate the suitability of Polyamide 6 (PA6) as potential polymer of interest. SEM and EDAX were used to ascertain morphological and elemental characteristics of the nanofiller. ICPOES has been used to ascertain the metal concentration of the untreated nanofiller to be treated (by oil and heavy metal extraction) before the production of nanocomposite materials. The challenges faced and future works are also discussed

  15. USARCENT AOR Contingency Base Waste Stream Analysis: An Analysis of Solid Waste Streams at Five Bases in the U. S. Army Central (USARCENT) Area of Responsibility

    Science.gov (United States)

    2013-03-31

    and Plastics Waste in As Bench Scale Combustor. University of Technology, Malaysia . http://eprints.utm.my/2854/1/75186.pdf. ASTM – ASTM...prevalent types of solid waste are food (19.1% by average sample weight), wood (18.9%), and plastics (16.0%) based on analysis of bases in...within the interval shown. Food and wood wastes are the largest components of the average waste stream (both at ~19% by weight), followed by plastic

  16. Carbon microspheres as ball bearings in aqueous-based lubrication.

    Science.gov (United States)

    St Dennis, J E; Jin, Kejia; John, Vijay T; Pesika, Noshir S

    2011-07-01

    We present an exploratory study on a suspension of uniform carbon microspheres as a new class of aqueous-based lubricants. The surfactant-functionalized carbon microspheres (∼0.1 wt %) employ a rolling mechanism similar to ball bearings to provide low friction coefficients (μ ≈ 0.03) and minimize surface wear in shear experiments between various surfaces, even at high loads and high contact pressures. The size range, high monodispersity, and large yield stress of the C(μsphere), as well as the minimal environmental impact, are all desirable characteristics for the use of a C(μsphere)-SDS suspension as an alternative to oil-based lubricants in compatible devices and machinery.

  17. Organic non-aqueous cation-based redox flow batteries

    Science.gov (United States)

    Jansen, Andrew N.; Vaughey, John T.; Chen, Zonghai; Zhang, Lu; Brushett, Fikile R.

    2016-03-29

    The present invention provides a non-aqueous redox flow battery comprising a negative electrode immersed in a non-aqueous liquid negative electrolyte, a positive electrode immersed in a non-aqueous liquid positive electrolyte, and a cation-permeable separator (e.g., a porous membrane, film, sheet, or panel) between the negative electrolyte from the positive electrolyte. During charging and discharging, the electrolytes are circulated over their respective electrodes. The electrolytes each comprise an electrolyte salt (e.g., a lithium or sodium salt), a transition-metal free redox reactant, and optionally an electrochemically stable organic solvent. Each redox reactant is selected from an organic compound comprising a conjugated unsaturated moiety, a boron cluster compound, and a combination thereof. The organic redox reactant of the positive electrolyte is selected to have a higher redox potential than the redox reactant of the negative electrolyte.

  18. The potential of coconut fibers in raw and treated forms to remove {sup 241}Am from aqueous solutions

    Energy Technology Data Exchange (ETDEWEB)

    Fonseca, Heverton C.O.; Jesus, Nella N.M. de; Nobre, Vanessa B.; Potiens Junior, Ademar; Sakata, Solange K., E-mail: sksakata@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    In the Radioactive Waste Management (GRR) at the Nuclear and Energy Research Institute (IPEN/CNEN-SP) vegetal biomass has been studied as adsorbent to remove radioisotopes from radioactive liquid wastes as part of the radioactive waste treatment program. In this work coconuts fiber was evaluate as biosorbent to remove {sup 241}Am from aqueous solutions and many different parameters were studied such as particle size (35 and 80 mesh) and contact time (between 5 and 60 minutes). In order to evaluate if the treated biomass could remove more {sup 241}Am the experiments were also performed using raw biomass and treated with H{sub 2}O{sub 2} in basic conditions. When the experiment was carried out using raw coconuts fiber with 80 mesh, neutral conditions after 5 minutes of contact time 99% of the radionuclide was removed from the aqueous solution. This result shows the potential of this biomass to remove {sup 241}Am from radioactive liquid wastes. (author)

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

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  20. Bench scale experiments for the remediation of Hanford Waste Treatment Plant low activity waste melter off-gas condensate

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Poirier, Michael [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)

    2017-08-11

    The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

  1. Photovoltaic's silica-rich waste sludge as supplementary cementitious materials (SCM)

    NARCIS (Netherlands)

    Quercia Bianchi, G.; van der Putten, J.J.G.; Husken, G.; Brouwers, H.J.H.

    2013-01-01

    Waste sludge, a solid recovered from wastewater of photovoltaic-industries, composes of agglomerates of nano-particles like SiO2 and CaCO3. This sludge deflocculates in aqueous solutions into nano-particles smaller than 1 µm. Thus, this sludge constitutes a potentially hazardous waste when it is

  2. Handling and storage of high-level liquid wastes from reprocessing of spent fuel

    International Nuclear Information System (INIS)

    Finsterwalder, L.

    1982-01-01

    The high level liquid wastes arise from the reprocessing of irradiated nuclear fuels, which are dissolved in aqueous acid solution, and the plutonium and unburned uranium removed in the chemical separation plant. The remaining solution, containing more than 99% of the dissolved fission products, together with impurities from cladding materials, corrosion products, traces of unseparated plutonium and uranium and most of the transuranic elements, constitutes the high-level waste. At present, these liquid wastes are usually concentrated by evaporation and stored as an aqueous nitric acid solution in high-integrity stainless-steel tanks. There is now world-wide agreement that, for the long term, these liquid wastes should be converted to solid form and much work is in progress to develop techniques for the solidification of these wastes. This paper considers the design requirements for such facilities and the experience gained during nearly 30 years of operation. (orig./RW)

  3. Separation of transuranium elements and fission products from medium activity aqueous liquid wastes

    International Nuclear Information System (INIS)

    Gompper, K.; Kunze, S.; Eden, G.; Loesch, G.; Zemski, C.

    1986-01-01

    In the course of work performed between January 1981 and June 1985 on the separation of TRU elements and fission products three liquid alpha containing waste streams were treated: - medium level waste solutions, - waste solutions from the acid digestion of burnable alpha containing solid residues, - waste solutions from mixed oxide fuel element fabrication. The method of separation was initially developed and optimized with simulating substances. Subesequently it was tested with real waste solutions

  4. Valorization of Waste Lipids through Hydrothermal Catalytic Conversion to Liquid Hydrocarbon Fuels with in Situ Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dongwook; Vardon, Derek R.; Murali, Dheeptha; Sharma, Brajendra K.; Strathmann, Timothy J.

    2016-03-07

    We demonstrate hydrothermal (300 degrees C, 10 MPa) catalytic conversion of real waste lipids (e.g., waste vegetable oil, sewer trap grease) to liquid hydrocarbon fuels without net need for external chemical inputs (e.g., H2 gas, methanol). A supported bimetallic catalyst (Pt-Re/C; 5 wt % of each metal) previously shown to catalyze both aqueous phase reforming of glycerol (a triacylglyceride lipid hydrolysis coproduct) to H2 gas and conversion of oleic and stearic acid, model unsaturated and saturated fatty acids, to linear alkanes was applied to process real waste lipid feedstocks in water. For reactions conducted with an initially inert headspace gas (N2), waste vegetable oil (WVO) was fully converted into linear hydrocarbons (C15-C17) and other hydrolyzed byproducts within 4.5 h, and H2 gas production was observed. Addition of H2 to the initial reactor headspace accelerated conversion, but net H2 production was still observed, in agreement with results obtained for aqueous mixtures containing model fatty acids and glycerol. Conversion to liquid hydrocarbons with net H2 production was also observed for a range of other waste lipid feedstocks (animal fat residuals, sewer trap grease, dry distiller's grain oil, coffee oil residual). These findings demonstrate potential for valorization of waste lipids through conversion to hydrocarbons that are more compatible with current petroleum-based liquid fuels than the biodiesel and biogas products of conventional waste lipid processing technologies.

  5. Solidification of radioactive aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Aikawa, Hideaki; Kato, Kiyoshi; Wadachi, Yoshiki

    1970-09-07

    A process for solidifying a radioactive waste solution is provided, using as a solidifying agent a mixture of calcined gypsum and burnt vermiculite. The quantity ratio of the mixture is preferred to be 1:1 by volume. The quantity of impregnation is 1/2 of the volume of the total quantity of the solidifying agent. In embodiments, 10 liters of plutonium waste solution was mixed with a mixture of 1:1 calcined gypsum and burnt vermiculite contained in a 20-liter cylindrical steel container lined with asphalt. The plutonium waste solution from the laboratory was neutralized with a caustic soda aqueous solution to prevent explosion due to the nitration of organic compounds. The neutralization is not always necessary. A market available dental gypsum was calcined at 400 to 500/sup 0/C and a vermiculite from Illinois was burnt at 1,100/sup 0/C to prepare the agents. The time required for the impregnation with 10 liters of plutonium solution was four minutes. After impregnation, the temperature rose to 40/sup 0/C within 30 minutes to one hour. Next, it was cooled to room temperature by standing for 3-4 hours. Solidification time was about 1 hour. The Japan Atomic Energy Research Insitute had treated and disposed about 1,000 tons of plutonium waste by this process as of August 19, 1970.

  6. Process for removing sulfate anions from waste water

    Science.gov (United States)

    Nilsen, David N.; Galvan, Gloria J.; Hundley, Gary L.; Wright, John B.

    1997-01-01

    A liquid emulsion membrane process for removing sulfate anions from waste water is disclosed. The liquid emulsion membrane process includes the steps of: (a) providing a liquid emulsion formed from an aqueous strip solution and an organic phase that contains an extractant capable of removing sulfate anions from waste water; (b) dispersing the liquid emulsion in globule form into a quantity of waste water containing sulfate anions to allow the organic phase in each globule of the emulsion to extract and absorb sulfate anions from the waste water and (c) separating the emulsion including its organic phase and absorbed sulfate anions from the waste water to provide waste water containing substantially no sulfate anions.

  7. Mobile fission and activation products in nuclear waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Umeki, H; Evans, N; Czervinski, K; Bruggeman, Ch; Poineau, F; Breynaert, A; Reiler, P; Pablo, J de; Pipon, Y; Molnar, M; Nishimura, T; Kienzler, B; Van Iseghem, P; Crovisier, J L; Wieland, E; Mace, N; Pablo, J de; Spahiu, K; Cui, D; Lida, Y; Charlet, L; Liu, X; Sato, H; Goutelard, F; Savoye, S; Glaus, M; Poinssot, C; Seby, F; Sato, H; Tournassat, Ch; Montavon, G; Rotenberg, B; Spahiu, K; Smith, G; Marivoet, J; Landais, P; Bruno, J; Johnson, H; Umeki, L; Geckeis, H; Giffaut, E; Grambow, B; Dierckx, A

    2007-07-01

    This document gathers 33 oral presentations that were made at this workshop dedicated to the mobility of some radionuclides in nuclear waste disposal. The workshop was organized into 6 sessions: 1) performance assessment, 2) speciation/interaction in aqueous media, 3) radioactive wastes, 4) redox processes at interfaces, 5) diffusion processes, and 6) retention processes.

  8. Mobile fission and activation products in nuclear waste disposal

    International Nuclear Information System (INIS)

    Umeki, H.; Evans, N.; Czervinski, K.; Bruggeman, Ch.; Poineau, F.; Breynaert, A.; Reiler, P.; Pablo, J. de; Pipon, Y.; Molnar, M.; Nishimura, T.; Kienzler, B.; Van Iseghem, P.; Crovisier, J.L.; Wieland, E.; Mace, N.; Pablo, J. de; Spahiu, K.; Cui, D.; Lida, Y.; Charlet, L.; Liu, X.; Sato, H.; Goutelard, F.; Savoye, S.; Glaus, M.; Poinssot, C.; Seby, F.; Sato, H.; Tournassat, Ch.; Montavon, G.; Rotenberg, B.; Spahiu, K.; Smith, G.; Marivoet, J.; Landais, P.; Bruno, J.; Johnson, H.; Umeki, L.; Geckeis, H.; Giffaut, E.; Grambow, B.; Dierckx, A.

    2007-01-01

    This document gathers 33 oral presentations that were made at this workshop dedicated to the mobility of some radionuclides in nuclear waste disposal. The workshop was organized into 6 sessions: 1) performance assessment, 2) speciation/interaction in aqueous media, 3) radioactive wastes, 4) redox processes at interfaces, 5) diffusion processes, and 6) retention processes

  9. Backfill barriers for nuclear waste repositories in salt

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-06-01

    Backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier. Bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved wires, or the sorption properties of radionuclide species.

  10. Feasibility assessment of copper-base waste package container materials in a tuff repository

    International Nuclear Information System (INIS)

    Acton, C.F.; McCright, R.D.

    1986-01-01

    This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized

  11. Nuclear energy and politics in Russian ATWS conditions

    International Nuclear Information System (INIS)

    Gagarinski, A.

    1998-01-01

    Relations between politics and nuclear power in the countries of sustainable development has been many times discussed during the short history of nuclear energy, and regularly arising new events, even very important (in Sweden, USA, etc.), just add to the formed understanding of the problem. Russia for 10 years lives in conditions of a transition period, which seems similar to ATWS-type accidents at nuclear power plants. In these conditions the effect of politics on nuclear power and vice versa are seen very clearly, and, more important, change swiftly, which may present interest for the countries with smoother public processes. The role of political processes in nuclear power is obvious and may be reduced to three main factors: change of political system and transition to market economy have placed nuclear power, though still within state sector, in an absolutely new economic condition, which determine its today's situation as 'Survival'; new possibilities of political influence and opposition to nuclear power (mainly struggle against construction of new nuclear fuel cycle objects) on a levels of authority (local, regional, federal); impact of the USSR collapse on the situation in Russian nuclear power was due sooner to temporary weakening of control and regulatory structures, than to the fact, that some fuel cycle elements have found themselves abroad (the factor of uranium resources' loss is unimportant at present). Nuclear safety was chosen to be the subject of Moscow 1996 Summit, initiated with the purpose of Russia coming closer to G7. The Summit has confirmed the thesis on the possibility of nuclear power o play an important role in the world energy demand in accordance with sustainable development goals. successful activities of Russia-USA Commission for economic and technological cooperation, known as 'Gore-Chernomyrdin' Commission, is to a large extent determined by positive nuclear decisions. Eastern direction of Russian nuclear export (Iran, China

  12. Learning the ABCs: Activity based costing in waste operations

    International Nuclear Information System (INIS)

    Zocher, Marc A.

    1992-01-01

    The United States Department of Energy (DOE) is facing a challenging new national role based on current world events, changing public perception and awareness, and a legacy of wastes generated in the past. Clearly, the DOE must put mechanisms in place to comply with environmental rules, regulations, and good management practices so that public health risk is minimized while programmatic costs are controlled. DOE has begun this process and has developed a Five-Year Plan to describe the activities necessary to comply with both cleanup, or environmental restoration, and waste management of existing waste streams. The focus of this paper is how to best manage the treatment, storage, disposal, and transportation of waste throughout the DOE weapons complex by using Activity Based Costing (ABC) to both plan and control expenditures in DOE Waste Management (WM). The basics of ABC, along with an example, will be detailed. (author)

  13. Process and technological wastes compaction through a fluidized bed incineration process

    International Nuclear Information System (INIS)

    Guiroy, J.J.

    1993-01-01

    The various fluidized bed systems (dense or circulating) are reviewed and the advantages of the circulation fluidized bed are highlighted (excellent combustion performance, clean combustion, large operating range, poly-functionality with regards to waste type, ...). Applications to contaminated graphite (with the problem of ash management) and to plant process wastes (ion exchangers, technological wastes, aqueous effluents); study of the neutralization and chlorine emission

  14. Decontamination liquid waste processing method

    International Nuclear Information System (INIS)

    Enda, Masami; Hosaka, Katsumi.

    1992-01-01

    Liquid wastes after electrolytic reduction are caused to flow through an anionic exchange membrane in a diffusion dialysis step, and liquid wastes and dialyzed water are passed in a countercurrent manner. Since acids in the liquid wastes transfer on the side of the dialyzed water due to the difference of concentration between the liquid wastes and the dialyzed water, acids can be easily recovered from the liquid wastes. If the acid-removed liquid wastes are put to electrodeposition in an electrodepositing step, the electrodepositing reactions between radioactive materials such as Co ion, Mn ion and leached metals such as Fe ions and Cr ions are caused preferentially to hydrogen generation reaction on a metal deposition cathode. Accordingly, metal ions can be easily separated from the liquid wastes. Since the separated liquid wastes are an aqueous solution in which cerium ions as a decontaminant and an acid at low concentration are dissolved, the concentration thereof is controlled by mixing them to acid recovering water after the diffusion dialysis and they can be reused as the decontaminant. (T.M.)

  15. Synergistic effects of irradiation of waste water

    International Nuclear Information System (INIS)

    Woodbridge, D.D.; Cooper, P.C.; Vandenburg, A.J.; Musselman, H.D.; Lowe, H.N.; Florida Inst. of Tech., Melbourne; Army Facilities Engineering Support Agency, Fort Belvoir, Va.

    1975-01-01

    Theoretical considerations of the use of high level radiation in the treatment of waste water have failed to consider the effects of the hydrated electron and the potential of possible synergistic effects of combining chlorine, oxygen, and irradiation. An extensive testing program at the University Center for Pollution Research of Florida Institute of Technology over the past four years has shown that irradiation of waste water samples immersed in an aqueous environment provide bacterial kill and reduction in organic pollution far greater than that obtained from theoretical considerations of G values and earlier experiments where the waste samples were not immersed in an aqueous environment. These testing programs have investigated the synergistic effects of combining oxygen and irradiation. Each of these combined treatments resulted in an increased bacterial kill factor. Tests on Staphylococcus aureus bacteria and fecal streptococcus bacteria indicate that the synergistic effects observed for fecal coliform bacteria also apply to the pathogenic bacteria. A statistical analysis of the data obtained shows the interrelationships between the various effects on the bacteria. A definite shielding factor due to the turbidity of the waste water has been shown to exist. Synergistic effects have been shown to significantly offset the shielding effects. Optimization of these synergistic effects can greatly increase the effectiveness of irradiation in the treatment of waste water. (orig.) [de

  16. Aqueous Angiography: Real-Time and Physiologic Aqueous Humor Outflow Imaging.

    Directory of Open Access Journals (Sweden)

    Sindhu Saraswathy

    Full Text Available Trabecular meshwork (TM bypass surgeries attempt to enhance aqueous humor outflow (AHO to lower intraocular pressure (IOP. While TM bypass results are promising, inconsistent success is seen. One hypothesis for this variability rests upon segmental (non-360 degrees uniform AHO. We describe aqueous angiography as a real-time and physiologic AHO imaging technique in model eyes as a way to simulate live AHO imaging.Pig (n = 46 and human (n = 6 enucleated eyes were obtained, orientated based upon inferior oblique insertion, and pre-perfused with balanced salt solution via a Lewicky AC maintainer through a 1mm side-port. Fluorescein (2.5% was introduced intracamerally at 10 or 30 mm Hg. With an angiographer, infrared and fluorescent (486 nm images were acquired. Image processing allowed for collection of pixel information based on intensity or location for statistical analyses. Concurrent OCT was performed, and fixable fluorescent dextrans were introduced into the eye for histological analysis of angiographically active areas.Aqueous angiography yielded high quality images with segmental patterns (p<0.0001; Kruskal-Wallis test. No single quadrant was consistently identified as the primary quadrant of angiographic signal (p = 0.06-0.86; Kruskal-Wallis test. Regions of high proximal signal did not necessarily correlate with regions of high distal signal. Angiographically positive but not negative areas demonstrated intrascleral lumens on OCT images. Aqueous angiography with fluorescent dextrans led to their trapping in AHO pathways.Aqueous angiography is a real-time and physiologic AHO imaging technique in model eyes.

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

    International Nuclear Information System (INIS)

    Langton, C.A.

    1984-01-01

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

  18. Evaluation of aqueous Na+/Cs+ separation by electrodialysis

    International Nuclear Information System (INIS)

    Buehler, M.F.; Lawrence, W.E.; Norton, J.D.

    1993-12-01

    In support of the Hanford Site cleanup, electrodialysis is being investigated as a method to separate aqueous sodium (Na + ) and cesium (Cs + ) ions. The approach has many advantages over existing separation technologies; in particular, electrodialysis creates little secondary waste while producing usable acid and base streams. The fundamentals of electrodialysis are presented in this report to provide a foundation for interpreting experimental data. A flat-plate laboratory-scale apparatus was used to determine the feasibility of separating Na + /Cs + mixtures by electrodialysis. The results showed that Cs + is preferentially separated over Na + by a factor of 2 to 3 using a Nafion reg-sign 417 cationic membrane. The separation is relatively insensitive to solution ionic strength and flow-rate variations. The current efficiency of the separation ranges from 0.60 to 0.65 depending on the applied voltage. The laboratory-scale system was characterized by dimensional analysis, which demonstrated that the process could be scaled up to a size attractive for the volume of waste at the Hanford Site. Preliminary experiments on a bench-scale system were also conducted. The initial results showed that the current-voltage response of the laboratory- and the bench-scale unit is identical

  19. EQ3/6 geochemical modeling task plan for Nevada Nuclear Waste Storage Investigations (NNWSI)

    Energy Technology Data Exchange (ETDEWEB)

    Isherwood, D.; Wolery, T.

    1984-04-10

    This task plan outlines work needed to upgrade the EQ3/6 geochemical code and expand the supporting data bases to allow the Nevada Nuclear Waste Storage Investigations (NNWSI) to model chemical processes important to the storage of nuclear waste in a tuff repository in the unsaturated zone. The plan covers the fiscal years 1984 to 1988. The scope of work includes the development of sub-models in the EQ3/6 code package for studying the effects of sorption, precipitation kinetics, redox disequilibrium, and radiolysis on radionuclide speciation and solubility. The work also includes a glass/water interactions model and a geochemical flow model which will allow us to study waste form leaching and reactions involving the waste package. A special emphasis is placed on verification of new capabilities as they are developed and code documentation to meet NRC requirements. Data base expansion includes the addition of elements and associated aqueous species and solid phases that are specific to nuclear waste (e.g., actinides and fission products) and the upgrading and documentation of the thermodynamic data for other species of interest.

  20. Anticipated transients without scram for light water reactors. Appendices. Staff report

    International Nuclear Information System (INIS)

    1978-04-01

    Information is presented concerning scram failure probability, rod drive failure data, ATWS rule and ATWS requirements, treatment of steam generator tube failures in ATWS evaluation, radiological consequences assessments, ATWS study to include parameter variations and equipment reliability in probabilistic accident analysis, PWR MTC for ATWS, safety valve flows, ATWS contribution to risk, fuel integrity, value-impact analysis, and analytical methods

  1. Supported liquid inorganic membranes for nuclear waste separation

    Science.gov (United States)

    Bhave, Ramesh R; DeBusk, Melanie M; DelCul, Guillermo D; Delmau, Laetitia H; Narula, Chaitanya K

    2015-04-07

    A system and method for the extraction of americium from radioactive waste solutions. The method includes the transfer of highly oxidized americium from an acidic aqueous feed solution through an immobilized liquid membrane to an organic receiving solvent, for example tributyl phosphate. The immobilized liquid membrane includes porous support and separating layers loaded with tributyl phosphate. The extracted solution is subsequently stripped of americium and recycled at the immobilized liquid membrane as neat tributyl phosphate for the continuous extraction of americium. The sequestered americium can be used as a nuclear fuel, a nuclear fuel component or a radiation source, and the remaining constituent elements in the aqueous feed solution can be stored in glassified waste forms substantially free of americium.

  2. Advances in technologies for the treatment of low and intermediate level radioactive liquid wastes

    International Nuclear Information System (INIS)

    1994-01-01

    In recent years the authorized maximum limits for radioactive discharges into the environment have been reduced considerably, and this, together with the requirement to minimize the volume of waste for storage or disposal and to declassify some wastes from intermediate to low level or to non-radioactive wastes, has initiated studies of ways in which improvements can be made to existing decontamination processes and also to the development of new processes. This work has led to the use of more specific precipitants and to the establishment of ion exchange treatment and evaporation techniques. Additionally, the use of combinations of some existing processes or of an existing process with a new technique such as membrane filtration is becoming current practice. New biotechnological, solvent extraction and electrochemical methods are being examined and have been proven at laboratory scale to be useful for radioactive liquid waste treatment. In this report an attempt has been made to review the current research and development of mature and advanced technologies for the treatment of low and intermediate level radioactive liquid wastes, both aqueous and non-aqueous. Non-aqueous radioactive liquid wastes or organic liquid wastes typically consist of oils, reprocessing solvents, scintillation liquids and organic cleaning products. A brief state of the art of existing processes and their application is followed by the review of advances in technologies, covering chemical, physical and biological processes. 213 refs, 33 figs, 3 tabs

  3. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    Energy Technology Data Exchange (ETDEWEB)

    Van Hecke, K.; Goethals, P.

    2006-07-15

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  4. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    International Nuclear Information System (INIS)

    Van Hecke, K.; Goethals, P.

    2006-01-01

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  5. Anticipated transients without scram for light water reactors

    International Nuclear Information System (INIS)

    1978-12-01

    In the first two volumes of this report, Anticipated Transients without Scram for Light Water Reactors NUREG-0460, dated April 1978, the NRC staff reviewed the information on this subject that had been developed in the past and evaluated the susceptibility of current nuclear plants to ATWS events using fault tree/event tree analysis techniques. Based on that evaluation, the staff concluded that some corrective measures were required to reduce the risk of severe consequences arising from possible ATWS events. Since the issuance of NUREG-0460, new safety and cost information has become available on ATWS. Also, new insights have been developed on the general subject of quantitative risk assessment. The purpose of this supplement to NUREG-0460 is to summarize the important additions to the information base and to propose a course of action from among a variety of alternatives for resolving the ATWS concern

  6. An overview of Russian experience and capabilities for development of ATW/ABC systems

    Science.gov (United States)

    Kazaritsky, Vladimir D.

    1995-09-01

    Several Russian institutes are expected to undertake a feasibility study of nuclear power systems based on proton accelerators. The examined systems are intended for conversion of surplus Pu and transmutation of long-lived radioactive waste. This research motivated by the demilitarization agreements and criticism of traditional nuclear power is focused on environmental protection.

  7. An overview of Russian experience and capabilities for development of ATW/ABC systems

    Energy Technology Data Exchange (ETDEWEB)

    Kazaritsky, V.D. [Institute for Theoretical and Experimental Physics, Moscow (Russian Federation)

    1995-10-01

    Several Russian institutes are expected to undertake a feasibility study of nuclear power systems based on proton accelerators. The examined systems are intended for conversion of surplus Pu and transmutation of long-lived radioactive waste. This research motivated by the demilitarisation agreements and criticism of traditional nuclear power is focused on environmental protection.

  8. Extraction of peptide tagged cutinase in detergent-based aqueous two-phase systems

    NARCIS (Netherlands)

    Rodenbrock, A.; Selber, K.; Egmond, M.R.; Kula, M.-R.

    2010-01-01

    Detergent-based aqueous two-phase systems have the advantage to require only one auxiliary chemical to induce phase separation above the cloud point. In a systematic study the efficiency of tryptophan-rich peptide tags was investigated to enhance the partitioning of an enzyme to the detergent-rich

  9. Kinetics of adsorption of dyes from aqueous solution using activated carbon prepared from waste apricot

    International Nuclear Information System (INIS)

    Onal, Yunus

    2006-01-01

    Adsorbent (WA11Zn5) has been prepared from waste apricot by chemical activation with ZnCl 2 . Pore properties of the activated carbon such as BET surface area, pore volume, pore size distribution, and pore diameter were characterized by N 2 adsorption and DFT plus software. Adsorption of three dyes, namely, Methylene Blue (MB), Malachite Green (MG), Crystal Violet (CV), onto activated carbon in aqueous solution was studied in a batch system with respect to contact time, temperature. The kinetics of adsorption of MB, MG and CV have been discussed using six kinetic models, i.e., the pseudo-first-order model, the pseudo-second-order model, the Elovich equation, the intraparticle diffusion model, the Bangham equation, the modified Freundlich equation. Kinetic parameters and correlation coefficients were determined. It was shown that the second-order kinetic equation could describe the adsorption kinetics for three dyes. The dyes uptake process was found to be controlled by external mass transfer at earlier stages (before 5 min) and by intraparticle diffusion at later stages (after 5 min). Thermodynamic parameters, such as ΔG, ΔH and ΔS, have been calculated by using the thermodynamic equilibrium coefficient obtained at different temperatures and concentrations. The thermodynamics of dyes-WA11Zn5 system indicates endothermic process

  10. Waste Management Using Request-Based Virtual Organizations

    Science.gov (United States)

    Katriou, Stamatia Ann; Fragidis, Garyfallos; Ignatiadis, Ioannis; Tolias, Evangelos; Koumpis, Adamantios

    Waste management is on top of the political agenda globally as a high priority environmental issue, with billions spent on it each year. This paper proposes an approach for the disposal, transportation, recycling and reuse of waste. This approach incorporates the notion of Request Based Virtual Organizations (RBVOs) using a Service Oriented Architecture (SOA) and an ontology that serves the definition of waste management requirements. The populated ontology is utilized by a Multi-Agent System which performs negotiations and forms RBVOs. The proposed approach could be used by governments and companies searching for a means to perform such activities in an effective and efficient manner.

  11. Immobilization of fission products in phosphate ceramic waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Singh, D.; Wagh, A. [Argonne National Lab., IL (United States)

    1997-10-01

    Chemically bonded phosphate ceramics (CBPCs) have several advantages that make them ideal candidates for containing radioactive and hazardous wastes. In general, phosphates have high solid-solution capacities for incorporating radionuclides, as evidenced by several phosphates (e.g., monazites and apatites) that are natural analogs of radioactive and rare-earth elements. The phosphates have high radiation stability, are refractory, and will not degrade in the presence of internal heating by fission products. Dense and hard CBPCs can be fabricated inexpensively and at low temperature by acid-base reactions between an inorganic oxide/hydroxide powder and either phosphoric acid or an acid-phosphate solution. The resulting phosphates are extremely insoluble in aqueous media and have excellent long-term durability. CBPCs offer the dual stabilization mechanisms of chemical fixation and physical encapsulation, resulting in superior waste forms. The goal of this task is develop and demonstrate the feasibility of CBPCs for S/S of wastes containing fission products. The focus of this work is to develop a low-temperature CBPC immobilization system for eluted {sup 99}Tc wastes from sorption processes.

  12. Candidate Low-Temperature Glass Waste Forms for Technetium-99 Recovered from Hanford Effluent Management Facility Evaporator Concentrate

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Mei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tang, Ming [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rim, Jung Ho [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chamberlin, Rebecca M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-24

    Alternative treatment and disposition options may exist for technetium-99 (99Tc) in secondary liquid waste from the Hanford Direct-Feed Low-Activity Waste (DFLAW) process. One approach includes development of an alternate glass waste form that is suitable for on-site disposition of technetium, including salts and other species recovered by ion exchange or precipitation from the EMF evaporator concentrate. By recovering the Tc content from the stream, and not recycling the treated concentrate, the DFLAW process can potentially be operated in a more efficient manner that lowers the cost to the Department of Energy. This report provides a survey of candidate glass formulations and glass-making processes that can potentially incorporate technetium at temperatures <700 °C to avoid volatilization. Three candidate technetium feed streams are considered: (1) dilute sodium pertechnetate loaded on a non-elutable ion exchange resin; (2) dilute sodium-bearing aqueous eluent from ion exchange recovery of pertechnetate, or (3) technetium(IV) oxide precipitate containing Sn and Cr solids in an aqueous slurry. From the technical literature, promising candidate glasses are identified based on their processing temperatures and chemical durability data. The suitability and technical risk of three low-temperature glass processing routes (vitrification, encapsulation by sintering into a glass composite material, and sol-gel chemical condensation) for the three waste streams was assessed, based on available low-temperature glass data. For a subset of candidate glasses, their long-term thermodynamic behavior with exposure to water and oxygen was modeled using Geochemist’s Workbench, with and without addition of reducing stannous ion. For further evaluation and development, encapsulation of precipitated TcO2/Sn/Cr in a glass composite material based on lead-free sealing glasses is recommended as a high priority. Vitrification of pertechnetate in aqueous anion exchange eluent solution

  13. Transuranic waste form characterization and data base. Executive summary

    International Nuclear Information System (INIS)

    1980-01-01

    The Transuranic Waste Form Characterization and Data Base (Volume 1) provides a wide range of information from which a comprehensive data base can be established and from which standards and criteria can be developed for the present NRC waste management program. Supplementary information on each of the areas discussed in Volume 1 is presented in Appendices A through K (Volumes 2 and 3). The structure of the study (Volume 1) is outlined and appendices of Volumes 2 and 3 correlate with each main section of the report. The Executive Summary reviews the sources, quantities, characteristics and treatment of transuranic wastes in the United States. Due to the variety of potential treatment processes for transuranic wastes, the end products for long-term storage may have corresponding variations in quantities and characteristics

  14. 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

  15. Nitrate release from waste rock dumps in the Elk Valley, British Columbia, Canada.

    Science.gov (United States)

    Mahmood, Fazilatun N; Barbour, S Lee; Kennedy, C; Hendry, M Jim

    2017-12-15

    The origin, distribution and leaching of nitrate (NO 3 - ) from coal waste rock dumps in the Elk Valley, British Columbia, Canada were defined using chemical and NO 3 - isotope analyses (δ 15 N- and δ 18 O-NO 3 - ) of solids samples of pre- and post-blast waste rock and from thick (up to 180m) unsaturated waste rock dump profiles constructed between 1982 and 2012 as well as water samples collected from a rock drain located at the base of one dump and effluent from humidity cell (HC) and leach pad (LP) tests on waste rock. δ 15 N- and δ 18 O-NO 3 - values and NO 3 - concentrations of waste rock and rock drain waters confirmed the source of NO 3 - in the waste rock to be explosives and that limited to no denitrification occurs in the dump. The average mass of N released during blasting was estimated to be about 3-6% of the N in the explosives. NO 3 - concentrations in the fresh-blast waste rock and recently placed waste rock used for the HC and LP experiments were highly variable, ranging from below detection to 241mg/kg. The mean and median concentrations of these samples ranged from 10-30mg/kg. In this range of concentrations, the initial aqueous concentration of fresh-blasted waste rock could range from approximately 200-600mg NO 3 - -N/L. Flushing of NO 3 - from the HCs, LPs and a deep field profile was simulated using a scale dependent leaching efficiency (f) where f ranged from 5-15% for HCs, to 35-80% for the LPs, to 80-90% for the field profile. Our findings show aqueous phase NO 3 - from blasting residuals is present at highly variable initial concentrations in waste rock and the majority of this NO 3 - (>75%) should be flushed by recharging water during displacement of the first stored water volume. Copyright © 2017 Elsevier B.V. All rights reserved.

  16. Transuranic contaminated waste container characterization and data base. Revision I

    International Nuclear Information System (INIS)

    Kniazewycz, B.G.

    1980-05-01

    The Nuclear Regulatory Commission (NRC) is developing regulations governing the management, handling and disposal of transuranium (TRU) radioisotope contaminated wastes as part of the NRC's overall waste management program. In the development of such regulations, numerous subtasks have been identified which require completion before meaningful regulations can be proposed, their impact evaluated and the regulations implemented. This report was prepared to assist in the development of the technical data base necessary to support rule-making actions dealing with TRU-contaminated wastes. An earlier report presented the waste sources, characteristics and inventory of both Department of Energy (DOE) generated and commercially generated TRU waste. In this report a wide variety of waste sources as well as a large TRU inventory were identified. The purpose of this report is to identify the different packaging systems used and proposed for TRU waste and to document their characteristics. This document then serves as part of the data base necessary to complete preparation and initiate implementation of TRU waste container and packaging standards and criteria suitable for inclusion in the present TRU waste management program. It is the purpose of this report to serve as a working document which will be used as appropriate in the TRU Waste Management Program. This report, and those following, will be compatible not only in format, but also in reference material and direction

  17. Removal of bulk contaminants from radioactive waste water at Bruce A using a clay based flocculent system

    International Nuclear Information System (INIS)

    Davloor, R.; Harper, B.

    2011-01-01

    Bruce Power's Bruce Nuclear Generating Station 'A', located on Lake Huron, has a treatment system that processes all aqueous radioactive waste water originating from the station. This Active Liquid Waste Treatment System (ALWTS) consists of collection tanks for the collection of radioactive waste water, a Pre-Treatment System (PTS) for the removal of bulk contaminants and suspended solids, a Reverse Osmosis System (ROS) to remove dissolved solids, an Evaporation and Solidification System (ESS) to concentrate and immobilize solids contained in concentrated waste streams from the ROS, and discharge tanks for the dispersal of the treated water. The ALWTS has been in continuous service since 1999 and is used to treat approximately 100,000 litres of Active Liquid Waste (ALW) each day. With the exception of tritium, it discharges waste water containing near zero concentrations of radioactive and conventional contaminants to the lake. The original design of the Bruce A ALWTS used a Backwashable Filtration System (BFS) to provide solids free water to the ROS, as measured by the Silt Density Index (SDI). During commissioning, the BFS was not successful in backwashing the solids from the filter elements. For approximately one year, a temporary solution was implemented using a Disposable Filtration System (DFS). A cationic polymer was added upstream of the DFS to agglomerate the solids. The system proved to be highly unreliable. It was difficult to agglomerate solids in the waste stream containing high amounts of detergent. As a result, DFS consumption was high and very costly. The SDI specification for the RO membrane was not always met, resulting in a quick decline of performance of the first stage ROS membranes in the treatment process. In addition, the excess cationic polymer in the RO feed caused the membranes to become fouled. In-house station staff, together with personnel from Colloid Environmental Technologies (CETCO) Company, worked to develop and

  18. Removal of bulk contaminants from radioactive waste water at Bruce A using a clay based flocculent system

    Energy Technology Data Exchange (ETDEWEB)

    Davloor, R.; Harper, B. [Bruce Power, Tiverton, ON (Canada)

    2011-07-01

    Bruce Power's Bruce Nuclear Generating Station 'A', located on Lake Huron, has a treatment system that processes all aqueous radioactive waste water originating from the station. This Active Liquid Waste Treatment System (ALWTS) consists of collection tanks for the collection of radioactive waste water, a Pre-Treatment System (PTS) for the removal of bulk contaminants and suspended solids, a Reverse Osmosis System (ROS) to remove dissolved solids, an Evaporation and Solidification System (ESS) to concentrate and immobilize solids contained in concentrated waste streams from the ROS, and discharge tanks for the dispersal of the treated water. The ALWTS has been in continuous service since 1999 and is used to treat approximately 100,000 litres of Active Liquid Waste (ALW) each day. With the exception of tritium, it discharges waste water containing near zero concentrations of radioactive and conventional contaminants to the lake. The original design of the Bruce A ALWTS used a Backwashable Filtration System (BFS) to provide solids free water to the ROS, as measured by the Silt Density Index (SDI). During commissioning, the BFS was not successful in backwashing the solids from the filter elements. For approximately one year, a temporary solution was implemented using a Disposable Filtration System (DFS). A cationic polymer was added upstream of the DFS to agglomerate the solids. The system proved to be highly unreliable. It was difficult to agglomerate solids in the waste stream containing high amounts of detergent. As a result, DFS consumption was high and very costly. The SDI specification for the RO membrane was not always met, resulting in a quick decline of performance of the first stage ROS membranes in the treatment process. In addition, the excess cationic polymer in the RO feed caused the membranes to become fouled. In-house station staff, together with personnel from Colloid Environmental Technologies (CETCO) Company, worked to develop and

  19. Aqueous biphasic systems involving alkylsulfate-based ionic liquids

    Energy Technology Data Exchange (ETDEWEB)

    Deive, Francisco J. [Instituto de Tecnologia Quimica e Biologica, UNL, Av. Republica, Apartado 127, 2780-901 Oeiras (Portugal); Department of Chemical Engineering, University of Vigo, P.O. Box 36310, Vigo (Spain); Rodriguez, Ana [Department of Chemical Engineering, University of Vigo, P.O. Box 36310, Vigo (Spain); Marrucho, Isabel M., E-mail: imarrucho@itqb.unl.pt [Instituto de Tecnologia Quimica e Biologica, UNL, Av. Republica, Apartado 127, 2780-901 Oeiras (Portugal); Rebelo, Luis P.N. [Instituto de Tecnologia Quimica e Biologica, UNL, Av. Republica, Apartado 127, 2780-901 Oeiras (Portugal)

    2011-11-15

    Highlights: > K{sub 3}PO{sub 4}, K{sub 2}CO{sub 3}, Na{sub 2}CO{sub 3}, and (NH{sub 4}){sub 2}SO{sub 4} act as phase promoter in aqueous solutions of ILs. > Remarkable influence of alkyl-chain length on solubility curves of alkylsulfate-based ILs. > Merchuck correlation was used for describing these systems. > {Delta}S{sub hyd} and Hofmeister series were used to discuss the different salting out effects. - Abstract: The specific effects of K{sub 3}PO{sub 4}, K{sub 2}CO{sub 3}, Na{sub 2}CO{sub 3}, and (NH{sub 4}){sub 2}SO{sub 4}, as high charge-density inorganic salts and thus inducers of the formation of aqueous biphasic systems (ABS) containing several ethyl-methylimidazolium alkylsulfate ionic liquids, C{sub 2}MIM C{sub n}SO{sub 4} (n = 2, 4, 6, or 8), have been assessed at T = 298.15 K. The results are analyzed in the light of the Hofmeister series. The influence of different alkyl chain lengths in the anion, together with the ability of the selected inorganic salts to induce the formation of ABS, is discussed. Phase diagrams have been determined through turbidimetry, including tie lines assignments from mass phase ratios according to the lever - arm rule. The Merchuck equation was satisfactorily used to correlate the solubility curve.

  20. Curcumin based optical sensing of fluoride in organo-aqueous media using irradiation technique

    Science.gov (United States)

    Venkataraj, Roopa; Radhakrishnan, P.; Kailasnath, M.

    2017-06-01

    The present work describes the degradation of natural dye Curcumin in organic-aqueous media upon irradiation by a multi-wavelength source of light like mercury lamp. The presence of anions in the solution leads to degradation of Curcumin and this degradation is especially enhanced in the case of fluoride ion. The degradation of Curcumin is investigated by studying the change in its absorption and fluorescence characteristics in organoaqueous solution upon irradiation. A broad detection range of fluoride ranging from 2.3×10-6-2.22×10-3 M points to the potential of the method of visible light irradiation enabling aqueous based sensing of fluoride using Curcumin.

  1. Generation projection of solid and liquid radioactive wastes and spent radioactive sources in Mexico

    International Nuclear Information System (INIS)

    Garcia A, E.; Hernandez F, I. Y.; Fernandez R, E.; Monroy G, F.; Lizcano C, D.

    2014-10-01

    This work is focused to project the volumes of radioactive aqueous liquid wastes and spent radioactive sources that will be generated in our country in next 15 years, solids compaction and radioactive organic liquids in 10 years starting from the 2014; with the purpose of knowing the technological needs that will be required for their administration. The methodology involves six aspects to develop: the definition of general objectives, to specify the temporary horizon of projection, data collection, selection of the prospecting model and the model application. This approach was applied to the inventory of aqueous liquid wastes, as well as radioactive compaction organic and solids generated in Mexico by non energy applications from the 2001 to 2014, and of the year 1997 at 2014 for spent sources. The applied projection models were: Double exponential smoothing associating the tendency, Simple Smoothing and Lineal Regression. For this study was elected the first forecast model and its application suggests that: the volume of the compaction solid wastes, aqueous liquids and spent radioactive sources will increase respectively in 152%, 49.8% and 55.7%, while the radioactive organic liquid wastes will diminish in 13.15%. (Author)

  2. Comparative Cost of Colour Removal from Textile Effluents Using Agriculture Wastes

    International Nuclear Information System (INIS)

    Afifi, T.H.; Aboul Fetouh, M.S.; Nassar, F.A.; Riyad, Y.M.

    1999-01-01

    In recent years, investigations have been oriented towards practical use of low cost materials in the treatment of wastewater polluted by dyestuffs. The use of bagasse pith and maize cob as agricultural wastes for the colour removal of dyestuffs, namely, Direct Orange 34, Direct Red 23, Reactive Violet 2 and Reactive Blue 19 from aqueous solution at different concentrations has been investigated. The adsorption capacity for each dye- adsorbent system has been determined. The relative costs of dye removal were reported based on adsorption capacity only. The aim of the present work is to assess the feasibility of two low-cost agriculture-wastes materials to adsorb both direct and reactive dyestuffs on economic basis

  3. Radiolysis of nucleosides in aqueous solutions: base liberation by the base attack mechanism

    International Nuclear Information System (INIS)

    Fujita, S.

    1984-01-01

    On the radiolysis of uridine and some other nucleosides in aqueous solution, a pH-dependent liberation of uracil or the corresponding base was found. e - sub(aq) and HOsup(anion radicals) 2 gave no freed bases, although many oxidizing radicals, including OH, Clsup(anion radicals) 2 , Brsup(anion radicals) 2 , (CNS)sup(anion radicals) 2 and SOsup(anion radicals) 4 , did cause the release of unaltered bases, depending on the pH of the solutions. The base yields were generally high at pH >= 11, with the exception of SOsup(anion radicals) 4 , which gave a rather high yield of uracil (from uridine) even in the pH region of - , present at high pH as the dissociated form of OH, may act partly as an oxidizing radical. A plausible mechanism of 3 1 -radical formation is discussed. (author)

  4. LEACHING BOUNDARY IN CEMENT-BASED WASTE FORMS

    Science.gov (United States)

    Cement-based fixation systems are among the most commonly employed stabilization/solidification techniques. These cement haste mixtures, however, are vulnerable to ardic leaching solutions. Leaching of cement-based waste forms in acetic acid solutions with different acidic streng...

  5. Characterization of radionuclude behavior in low-level waste sites

    International Nuclear Information System (INIS)

    Toste, A.P.; Kirby, L.J.; Robertson, D.E.; Abel, K.H.; Perkins, R.W.

    1982-10-01

    Our laboratory is investigating the subsurface migration of radionuclides in groundwater at the Maxey Flats, Kentucky, shallow land-burial site and at a low-level aqueous waste disposal facility. At Maxey Flats, radionuclide and tracer data indicate groundwater communication between a waste trench and an adjacent experimental study area. Areal distributions of radionuclides in surface soil confirm that contamination at Maxey Flats has been largely contained on site. Of the radionuclides detected in the surface soil, only 3 H and 60 Co concentrations appear to be derived from waste. Plutonium exists in the anoxic subsurface waters at Maxey Flats as a reduced, anionic complex; some of the plutonium appears to be complexed with EDTA, whereas organic acids seem to be associated with 137 Cs and 90 Sr. At the aqueous waste disposal site, 3 H and mainly anionic species of certain radionuclides, including 60 Co, 106 Ru, 99 Tc, 131 I, and traces of 238 239 240 Pu, appear to migrate from a trench through soil adjacent to the trench. Radionuclides in the particulate and cationic forms appear to be efficiently retained by the soil. In general, observations indicate that the physicochemical form of the radionuclides mediates their subsurface migration in groundwater at both waste disposal sites

  6. Thermoplastic encapsulation of waste surrogates by high-shear mixing

    International Nuclear Information System (INIS)

    Lageraaen, P.R.; Kalb, P.D.; Patel, B.R.

    1995-12-01

    Brookhaven National Laboratory (BNL) has developed a robust, extrusion-based polyethylene encapsulation process applicable to a wide range of solid and aqueous low-level radioactive, hazardous and mixed wastes. However, due to the broad range of physical and chemical properties of waste materials, pretreatment of these wastes is often required to make them amenable to processing with polyethylene. As part of the scope of work identified in FY95 open-quotes Removal and Encapsulation of Heavy Metals from Ground Water,close quotes EPA SERDP No. 387, that specifies a review of potential thermoplastic processing techniques, and in order to investigate possible pretreatment alternatives, BNL conducted a vendor test of the Draiswerke Gelimat (thermokinetic) mixer on April 25, 1995 at their test facility in Mahwah, NJ. The Gelimat is a batch operated, high-shear, high-intensity fluxing mixer that is often used for mixing various materials and specifically in the plastics industry for compounding additives such as stabilizers and/or colorants with polymers

  7. Disposal and reclamation of southwestern coal and uranium wastes

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1979-01-01

    The types of solid wastes and effluents produced by the southwestern coal and uranium mining and milling industries are considered, and the current methods for the disposal and reclamation of these materials discussed. The major means of disposing of the solid wastes from both industries is by land fill or in some instances ponding. Sludges or aqueous wastes are normally discharged into settling and evaporative ponds. Basic reclamation measures for nearly all coal and uranium waste disposal sites include solids stabilization, compacting, grading, soil preparation, and revegetation. Impermeable liners and caps are beginning to be applied to disposal sites for some of the more harmful coal and uranium waste materials

  8. Chemical species of plutonium in Hanford radioactive tank waste

    International Nuclear Information System (INIS)

    Barney, G.S.

    1997-01-01

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

  9. 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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  11. Method of solidifying radioactive wastes

    International Nuclear Information System (INIS)

    Maeda, Masahiko; Kira, Satoshi; Watanabe, Naotoshi; Nagaoka, Takeshi; Akane, Junta.

    1982-01-01

    Purpose: To obtain solidification products of radioactive wastes having sufficient monoaxial compression strength and excellent in water durability upon ocean disposal of the wastes. Method: Solidification products having sufficient strength and filled with a great amount of radioactive wastes are obtained by filling and solidifying 100 parts by weight of chlorinated polyethylene resin and 100 - 500 parts by weight of particular or powderous spent ion exchange resin as radioactive wastes. The chlorinated polyethylene resin preferably used herein is prepared by chlorinating powderous or particulate polyethylene resin in an aqueous suspending medium or by chlorinating polyethylene resin dissolved in an organic solvent capable of dissolving the polyethylene resin, and it is crystalline or non-crystalline chlorinated polyethylene resin comprising 20 - 50% by weight of chlorine, non-crystalline resin with 25 - 40% by weight of chlorine being particularly preferred. (Horiuchi, T.)

  12. Liquid fuels from food waste: An alternative process to co-digestion

    Science.gov (United States)

    Sim, Yoke-Leng; Ch'ng, Boon-Juok; Mok, Yau-Cheng; Goh, Sok-Yee; Hilaire, Dickens Saint; Pinnock, Travis; Adams, Shemlyn; Cassis, Islande; Ibrahim, Zainab; Johnson, Camille; Johnson, Chantel; Khatim, Fatima; McCormack, Andrece; Okotiuero, Mary; Owens, Charity; Place, Meoak; Remy, Cristine; Strothers, Joel; Waithe, Shannon; Blaszczak-Boxe, Christopher; Pratt, Lawrence M.

    2017-04-01

    Waste from uneaten, spoiled, or otherwise unusable food is an untapped source of material for biofuels. A process is described to recover the oil from mixed food waste, together with a solid residue. This process includes grinding the food waste to an aqueous slurry, skimming off the oil, a combined steam treatment of the remaining solids concurrent with extrusion through a porous cylinder to release the remaining oil, a second oil skimming step, and centrifuging the solids to obtain a moist solid cake for fermentation. The water, together with any resulting oil from the centrifuging step, is recycled back to the grinding step, and the cycle is repeated. The efficiency of oil extraction increases with the oil content of the waste, and greater than 90% of the oil was collected from waste containing at least 3% oil based on the wet mass. Fermentation was performed on the solid cake to obtain ethanol, and the dried solid fermentation residue was a nearly odorless material with potential uses of biochar, gasification, or compost production. This technology has the potential to enable large producers of food waste to comply with new laws which require this material to be diverted from landfills.

  13. Method of disposing radioactive wastes

    International Nuclear Information System (INIS)

    Isozaki, Kei.

    1983-01-01

    Purpose : To enable safety ocean disposal of radioactive wastes by decreasing the leaching rate of radioactive nucleides, improving the quick-curing nature and increasing the durability. Method : A mixture comprising 2 - 20 parts by weight of alkali metal hydroxide and 100 parts by weight of finely powdered aqueous slags from a blast furnace is added to radioactive wastes to solidify them. In the case of medium or low level radioactive wastes, the solidification agent is added by 200 parts by weight to 100 parts by weight of the wastes and, in the case of high level wastes, the solidification agent is added in such an amount that the wastes occupy about 20% by weight in the total of the wastes and the solidification agent. Sodium hydroxide used as the alkali metal hydroxide is partially replaced with sodium carbonate, a water-reducing agent such as lignin sulfonate is added to improve the fluidity and suppress the leaching rate and the wastes are solidified in a drum can. In this way, corrosions of the vessel can be suppressed by the alkaline nature and the compression strength, heat stability and the like of the product also become excellent. (Sekiya, K.)

  14. Pyrochemical separation of radioactive components from inert materials in ICPP high-level calcined waste

    International Nuclear Information System (INIS)

    Del Debbio, J.A.; Nelson, L.O.; Todd, T.A.

    1995-05-01

    Since 1963, calcination of aqueous wastes from reprocessing of DOE-owned spent nuclear fuels has resulted in the accumulation of approximately 3800 m 3 of high-level waste (HLW) at the Idaho Chemical Processing Plant (ICPP). The waste is in the form of a granular solid called calcine and is stored on site in stainless steel bins which are encased in concrete. Due to the leachability of 137 Cs and 90 Sr and possibly other radioactive components, the calcine is not suitable for final disposal. Hence, a process to immobilize calcine in glass is being developed. Since radioactive components represent less than 1 wt % of the calcine, separation of actinides and fission products from inert components is being considered to reduce the volume of HLW requiring final disposal. Current estimates indicate that compared to direct vitrification, a volume reduction factor of 10 could result in significant cost savings. Aqueous processes, which involve calcine dissolution in nitric acid followed by separation of actinide and fission products by solvent extraction and ion exchange methods, are being developed. Pyrochemical separation methods, which generate small volumes of aqueous wastes and do not require calcine dissolution, have been evaluated as alternatives to aqueous processes. This report describes three proposed pyrochemical flowsheets and presents the results of experimental studies conducted to evaluate their feasibility. The information presented is a consolidation of three reports, which should be consulted for experimental details

  15. Final Report: Fiscal Year 1997 demonstration of omnivorous non-thermal mixed waste treatment: Direct chemical oxidation of organic solids and liquids using peroxydisulfate

    International Nuclear Information System (INIS)

    Cooper, J.F.; Ballazs G.B.

    1998-01-01

    Direct Chemical Oxidation (DCO) is a non-thermal, ambient pressure, aqueous-based technology for the oxidative destruction of the organic components of hazardous or mixed waste streams. The process has been developed for applications in waste treatment, chemical demilitarization and decontamination at LLNL since 1992. The process uses solutions of the peroxydisulfate ion (typically sodium or ammonium salts) to completely mineralize the organics to carbon dioxide and water. The expended oxidant may be electrolytically regenerated to minimize secondary waste. The paper briefly describes: free radical and secondary oxidant formation; electrochemical regeneration; offgas stream; and throughput

  16. Final Report: Fiscal Year 1997 demonstration of omnivorous non-thermal mixed waste treatment: Direct chemical oxidation of organic solids and liquids using peroxydisulfate

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, J.F.

    1998-01-01

    Direct Chemical Oxidation (DCO) is a non-thermal, ambient pressure, aqueous-based technology for the oxidative destruction of the organic components of hazardous or mixed waste streams. The process has been developed for applications in waste treatment, chemical demilitarization and decontamination at LLNL since 1992. The process uses solutions of the peroxydisulfate ion (typically sodium or ammonium salts) to completely mineralize the organics to carbon dioxide and water. The expended oxidant may be electrolytically regenerated to minimize secondary waste. The paper briefly describes: free radical and secondary oxidant formation; electrochemical regeneration; offgas stream; and throughput.

  17. Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model

    Energy Technology Data Exchange (ETDEWEB)

    Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

    2013-02-01

    This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

  18. ANSTO's waste forms for the 31. century

    International Nuclear Information System (INIS)

    Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

    2004-01-01

    ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and 99 Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

  19. ANSTO's waste forms for the 31. century

    Energy Technology Data Exchange (ETDEWEB)

    Vance, E R; Begg, B D; Day, R A; Moricca, S; Perera, D S; Stewart, M W. A.; Carter, M L; McGlinn, P J; Smith, K L; Walls, P A; Robina, M La

    2004-07-01

    ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

  20. Salicylimine-Based Colorimetric and Fluorescent Chemosensor for Selective Detection of Cyanide in Aqueous Buffer

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Jin Young; Hwang, In Hong; Kim, Hyun; Song, Eun Joo; Kim, Kyung Beom; Kim, Cheal [Seoul National Univ., Seoul (Korea, Republic of)

    2013-07-15

    A simple colorimetric and fluorescent anion sensor 1 based on salicylimine showed a high selectivity and sensitivity for detection of cyanide in aqueous solution. The receptor 1 showed high selectivity toward CN{sup -} ions in a 1:1 stoichiometric manner, which induces a fast color change from colorless to orange and a dramatic enhancement in fluorescence intensity selectively for cyanide anions over other anions. Such selectivity resulted from the nucleophilic addition of CN{sup -} to the carbon atom of an electron-deficient imine group. The sensitivity of the fluorescence-based assay (0.06 μM) is below the 1.9 μM suggested by the World Health Organization (WHO) as the maximum allowable cyanide concentration in drinking water, capable of being a practical system for the monitoring of CN. concentrations in aqueous samples.

  1. Method of decomposing radioactive organic solvent wastes

    International Nuclear Information System (INIS)

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

    1986-01-01

    Purpose: To decompose radioactive organic solvent wastes or radioactive hydrocarbon solvents separated therefrom into organic materials under moderate conditions, as well as greatly decrease the amount of secondary wastes generated. Method: Radioactive organic solvent wastes comprising an organic phosphoric acid ester ingredient and a hydrocarbon ingredient as a diluent therefor, or radioactive hydrocarbon solvents separated therefrom are oxidatively decomposed by hydrogen peroxide in an aqueous phosphoric acid solution of phosphoric acid metal salts finally into organic materials to perform decomposing treatment for the radioactive organic solvent wastes. The decomposing reaction is carried out under relatively moderate conditions and cause less burden to facilities or the likes. Further, since the decomposed liquid after the treatment can be reused for the decomposing reaction as a catalyst solution secondary wastes can significantly be decreased. (Yoshihara, H.)

  2. 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

  3. Tomatoes in oil recovery. [Plant waste additives improve yield

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    The waste from processing tomato, squash and pepper stalks found unexpected use in recovery of oil. Even a negligible amount thereof in an aqueous solution pumped into an oil-bearing formation turned out to be sufficient to increase the yield. Substances of plant origin, which improve dramatically the oil-flushing properties of water, not only increase the recovery of oil, but reduce the volume of fluid to be pumped into the stratum. The staff of the Institute of Deep Oil and Gas Deposits of the Azerbaijan Academy of Sciences, who proved the technological and economical advantages of using the waste from plant processing, transmitted their findings to the oil workers of Baku. The scientists have concluded that there is a good raw material base in this republic for utilizing this method on oil-bearing formations.

  4. Liquid Radioactive Wastes Treatment: A Review

    Directory of Open Access Journals (Sweden)

    Yung-Tse Hung

    2011-05-01

    Full Text Available Radioactive wastes are generated during nuclear fuel cycle operation, production and application of radioisotope in medicine, industry, research, and agriculture, and as a byproduct of natural resource exploitation, which includes mining and processing of ores, combustion of fossil fuels, or production of natural gas and oil. To ensure the protection of human health and the environment from the hazard of these wastes, a planned integrated radioactive waste management practice should be applied. This work is directed to review recent published researches that are concerned with testing and application of different treatment options as a part of the integrated radioactive waste management practice. The main aim from this work is to highlight the scientific community interest in important problems that affect different treatment processes. This review is divided into the following sections: advances in conventional treatment of aqueous radioactive wastes, advances in conventional treatment of organic liquid wastes, and emerged technological options.

  5. Removal of Methylene Blue from aqueous solution using spent bleaching earth

    Science.gov (United States)

    Saputra, E.; Saputra, R.; Nugraha, M. W.; Irianty, R. S.; Utama, P. S.

    2018-04-01

    The waste from industrial textile waste is one of the environmental problems, it is required effective and efficient processing. In this study spent bleaching earth was used as absorbent. It was found that the absorbent was effective to remove methylene blue from aqueous solution with removal efficiency 99.97 % in 120 min. Several parameters such as pH, amount of absorbent loading, stirring speed are found as key factor influencing removal of methylene blue. The mechanism of adsorption was also studied, and it was found that Langmuir isotherm fitted to data of experiment with adsorption capacity 0.5 mg/g.

  6. Removal of actinide elements from liquid scintillation cocktail wastes using liquid-liquid extraction and demulsification techniques

    International Nuclear Information System (INIS)

    Foltz, K.; Landsberger, S.; Srinivasan, B.; Vandegrift, G.F.

    1994-01-01

    For many years liquid scintillation cocktail (LSC) wastes have been generated and stored at Argonne National Laboratory (ANL). These wastes are stored in thousands of 10--20 m scintillation vials, many of which contain elements with Z > 88. Because storage space is limited, disposal of this waste is pressing. These wastes could be commercially incinerated if the radionuclides with Z>88 are reduced to sufficiently low levels. However, there is currently no deminimus level for these radionuclides, and separation techniques are still being tested. The University of Illinois is conducting experiments to separate radionuclides with Z > 88 from simulated LSC wastes by using liquid-liquid extraction (LLX) and demulsification techniques. The actinide elements are removed from the LSC by extraction into an aqueous phase after the cocktail has been demulsified. The aqueous and organic phases are separated and the organic phase, now free from radionuclides with Z > 88, can be sent to a commercial incineration facility. The aqueous phase may be treated and disposed of using existing techniques. The LLX separation techniques used solutions of sodium oxalate, aluminum nitrate, and tetrasodium EDTA at varying concentrations. These extractants were mixed with the simulated waste in a 1:1 volume ratio. Using 1.0M Na 4 EDTA salt solutions, decontamination ratios as high as 230 were achieved

  7. 77 FR 36447 - Hazardous Waste Management System; Identification and Listing of Hazardous Waste

    Science.gov (United States)

    2012-06-19

    ... underflow water is an aqueous solution which seeps through the treatment zone (soils) of the North Landfarm... Trichloroethylene ND 5.00E-01 2,4,6-Trichlorophenol....... ND 2.00E+00 Vinyl chloride ND 1.56E-01 Zinc 6.05E-02 3... groundwater contamination resulting from disposal of the petitioned waste in a surface impoundment, and that a...

  8. Design of a static mixer reactor for copper recovery from waste streams

    NARCIS (Netherlands)

    Van Wageningen, W.F.C.

    2005-01-01

    The main goal of the project was the development of a plug flow reactor for the reduction of heavy metals (Cu2+) from industrial waste streams. Potential application of the reduction process inside The Netherlands lies in the IC and galvanic industry, where small waste streams containing aqueous

  9. Modeling and optimization by particle swarm embedded neural network for adsorption of zinc (II) by palm kernel shell based activated carbon from aqueous environment.

    Science.gov (United States)

    Karri, Rama Rao; Sahu, J N

    2018-01-15

    Zn (II) is one the common pollutant among heavy metals found in industrial effluents. Removal of pollutant from industrial effluents can be accomplished by various techniques, out of which adsorption was found to be an efficient method. Applications of adsorption limits itself due to high cost of adsorbent. In this regard, a low cost adsorbent produced from palm oil kernel shell based agricultural waste is examined for its efficiency to remove Zn (II) from waste water and aqueous solution. The influence of independent process variables like initial concentration, pH, residence time, activated carbon (AC) dosage and process temperature on the removal of Zn (II) by palm kernel shell based AC from batch adsorption process are studied systematically. Based on the design of experimental matrix, 50 experimental runs are performed with each process variable in the experimental range. The optimal values of process variables to achieve maximum removal efficiency is studied using response surface methodology (RSM) and artificial neural network (ANN) approaches. A quadratic model, which consists of first order and second order degree regressive model is developed using the analysis of variance and RSM - CCD framework. The particle swarm optimization which is a meta-heuristic optimization is embedded on the ANN architecture to optimize the search space of neural network. The optimized trained neural network well depicts the testing data and validation data with R 2 equal to 0.9106 and 0.9279 respectively. The outcomes indicates that the superiority of ANN-PSO based model predictions over the quadratic model predictions provided by RSM. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. MIDAS/PK code development using point kinetics model

    International Nuclear Information System (INIS)

    Song, Y. M.; Park, S. H.

    1999-01-01

    In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

  11. Radioactive wastes: sources, treatment, and disposal

    International Nuclear Information System (INIS)

    Wymer, R.G.; Blomeke, J.O.

    1975-01-01

    Sources, treatment, and disposal of radioactive wastes are analyzed in an attempt to place a consideration of the problem of permanent disposal at the level of established or easily attainable technology. In addition to citing the natural radioactivity present in the biosphere, the radioactive waste generated at each phase of the fuel cycle (mills, fabrication plants, reactors, reprocessing plants) is evaluated. The three treatment processes discussed are preliminary storage to permit decay of the short-lived radioisotopes, solidification of aqueous wastes, and partitioning the long-lived α emitters for separate and long-term storage. Dispersion of radioactive gases to the atmosphere is already being done, and storage in geologically stable structures such as salt mines is under active study. The transmutation of high-level wastes appears feasible in principle, but exceedingly difficult to develop

  12. Radioactive waste processing container

    International Nuclear Information System (INIS)

    Ishizaki, Kanjiro; Koyanagi, Naoaki; Sakamoto, Hiroyuki; Uchida, Ikuo.

    1992-01-01

    A radioactive waste processing container used for processing radioactive wastes into solidification products suitable to disposal such as underground burying or ocean discarding is constituted by using cements. As the cements, calcium sulfoaluminate clinker mainly comprising calcium sulfoaluminate compound; 3CaO 3Al 2 O 3 CaSO 4 , Portland cement and aqueous blast furnace slug is used for instance. Calciumhydroxide formed from the Portland cement is consumed for hydration of the calcium sulfoaluminate clinker. According, calcium hydroxide is substantially eliminated in the cement constituent layer of the container. With such a constitution, damages such as crackings and peelings are less caused, to improve durability and safety. (I.N.)

  13. Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier

    Energy Technology Data Exchange (ETDEWEB)

    T. Wolery

    2005-02-22

    This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks.

  14. Environment on the Surfaces of the Drip Shield and Waste Package Outer Barrier

    International Nuclear Information System (INIS)

    T. Wolery

    2005-01-01

    This report provides supporting analysis of the conditions at which an aqueous solution can exist on the drip shield or waste package surfaces, including theoretical underpinning for the evolution of concentrated brines that could form by deliquescence or evaporation, and evaluation of the effects of acid-gas generation on brine composition. This analysis does not directly feed the total system performance assessment for the license application (TSPA-LA), but supports modeling and abstraction of the in-drift chemical environment (BSC 2004 [DIRS 169863]; BSC 2004 [DIRS 169860]). It also provides analyses that may support screening of features, events, and processes, and input for response to regulatory inquiries. This report emphasizes conditions of low relative humidity (RH) that, depending on temperature and chemical conditions, may be dry or may be associated with an aqueous phase containing concentrated electrolytes. Concentrated solutions at low RH may evolve by evaporative concentration of water that seeps into emplacement drifts, or by deliquescence of dust on the waste package or drip shield surfaces. The minimum RH for occurrence of aqueous conditions is calculated for various chemical systems based on current understanding of site geochemistry and equilibrium thermodynamics. The analysis makes use of known characteristics of Yucca Mountain waters and dust from existing tunnels, laboratory data, and relevant information from the technical literature and handbooks

  15. Modelling aqueous corrosion of nuclear waste phosphate glass

    Energy Technology Data Exchange (ETDEWEB)

    Poluektov, Pavel P.; Schmidt, Olga V.; Kascheev, Vladimir A. [Bochvar All-Russian Scientific Research Institute for Inorganic Materials (VNIINM), Moscow (Russian Federation); Ojovan, Michael I., E-mail: m.ojovan@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Mappin Street, Sheffield, S1 3JD (United Kingdom)

    2017-02-15

    A model is presented on nuclear sodium alumina phosphate (NAP) glass aqueous corrosion accounting for dissolution of radioactive glass and formation of corrosion products surface layer on the glass contacting ground water of a disposal environment. Modelling is used to process available experimental data demonstrating the generic inhibiting role of corrosion products on the NAP glass surface. - Highlights: • The radionuclides yield is determined by the transport from the glass through the surface corrosion layer. • Formation of the surface layer is due to the dissolution of the glass network and the formation of insoluble compounds. • The model proposed accounts for glass dissolution, formation of corrosion layer, specie diffusion and chemical reactions. • Analytical solutions are found for corrosion layer growth rate and glass components component leaching rates.

  16. Effect of miscibility and soil water content in movement of mixed waste

    International Nuclear Information System (INIS)

    Park, W.J.

    1989-01-01

    Since commercial low-level waste sites will not accept mixed low level wastes for disposal any longer, safer disposal of these wastes as well as hazardous waste becomes the growing concern. The objective of this study were to estimate the effect of some characteristics of organic material, such as solubility, density and volatility, on the movement in soil under various moisture contents. Attempts were made to fit the measured data to theoretical models for the movement of aqueous and airborne components through the vadose zone. Four different C-14 labeled organic materials, Methyl Alcohol, Toluene, Formic Acid, and Bromobenzene, differing in density, solubility, and volatility, were injected into test columns packed with a mixture of sands having known particles sizes and porosity. The method employed to make calibrated unsaturated conditions proved to be adequate for four different designated moisture contents, permitting sampling of both airborne and aqueous components at the same time. Significant solubility and density effects were found for the different organic materials associated with movement through water channels or air-filled pores, which became available at various unsaturated conditions. To analyze this mobility mechanism as a function of inherent properties of organic materials, a couple of mathematical equations were presented to describe both airborne release and aqueous migration and their wider applicability was discussed

  17. Aqueous recovery of plutonium from pyrochemical processing residues

    International Nuclear Information System (INIS)

    Gray, L.W.; Gray, J.H.

    1984-01-01

    Pyrochemical processes provide rapid methods to reclaim plutonium from scrap residues. Frequently, however, these processes yield an impure plutonium product and waste residues that are contaminated with actinides and are therefore nondiscardable. The Savannah River Laboratory and Plant and the Rocky Flats Plant are jointly developing new processes using both pyrochemistry and aqueous chemistry to generate pure product and discardable waste. An example of residue being treated is that from the molten salt extraction (MSE), a mixture of NaCl, KCl, MgCl 2 , PuCl 3 , AmCl 3 , PuO 2 , and Pu 0 . This mixture is scrubbed with molten aluminum containing a small amount of magnesium to produce a nonhomogeneous Al-Pu-Am-Mg alloy. This process, which rejects most of the NaCl-KCl-MgCl 2 salts, results in a product easily dissolved in 6M HNO 3 -0.1M HF. Any residual chloride in the product is removed by precipitation with Hg(I) followed by centrifuging. Plutonium and americium are then separated by the standard Purex process. The americium, initially diverted to the solvent extraction waste stream, can either be recovered or sent to waste

  18. Boosting TAG Accumulation with Improved Biodiesel Production from Novel Oleaginous Microalgae Scenedesmus sp. IITRIND2 Utilizing Waste Sugarcane Bagasse Aqueous Extract (SBAE).

    Science.gov (United States)

    Arora, Neha; Patel, Alok; Pruthi, Parul A; Pruthi, Vikas

    2016-09-01

    This investigation utilized sugarcane bagasse aqueous extract (SBAE), a nontoxic, cost-effective medium to boost triacylglycerol (TAG) accumulation in novel fresh water microalgal isolate Scenedesmus sp. IITRIND2. Maximum lipid productivity of 112 ± 5.2 mg/L/day was recorded in microalgae grown in SBAE compared to modified BBM (26 ± 3 %). Carotenoid to chlorophyll ratio was 12.5 ± 2 % higher than in photoautotrophic control, indicating an increase in photosystem II activity, thereby increasing growth rate. Fatty acid methyl ester (FAME) profile revealed presence of C14:0 (2.29 %), C16:0 (15.99 %), C16:2 (4.05 %), C18:0 (3.41 %), C18:1 (41.55 %), C18:2 (12.41), and C20:0 (1.21 %) as the major fatty acids. Cetane number (64.03), cold filter plugging property (-1.05 °C), and oxidative stability (12.03 h) indicated quality biodiesel abiding by ASTM D6751 and EN 14214 fuel standards. Results consolidate the candidature of novel freshwater microalgal isolate Scenedesmus sp. IITRIND2 cultivated in SBAE, aqueous extract made from copious, agricultural waste sugarcane bagasse to increase the lipid productivity, and could further be utilized for cost-effective biodiesel production.

  19. A new pyrazoline-based fluorescent sensor for Al3+ in aqueous solution.

    Science.gov (United States)

    Hu, Shengli; Song, Jingjing; Wu, Gongying; Cheng, Cuixia; Gao, Qing

    2015-02-05

    A new pyrazoline-based fluorescent sensor was synthesized and the structure was confirmed by single crystal X-ray diffraction. The sensor responds to Al(3+) with high selectivity among a series of cations in aqueous methanol. This sensor forms a 1:1 complex with Al(3+) and displays fluorescent quenching. Copyright © 2014 Elsevier B.V. All rights reserved.

  20. Accelerator-driven thermal fission systems may provide energy supply advantages

    International Nuclear Information System (INIS)

    Linford, R.K.

    1992-01-01

    This presentation discusses the energy supply advantages of using accelerator-driven thermal fission systems. Energy supply issues as related to cost, fuel supply stability, environmental impact, and safety are reviewed. It is concluded that the Los Alamos Accelerator Transmutation of Waste (ATW) concept, discussed here, has the following advantages: improved safety in the form of low inventory and subcriticality; reduced high-level radioactive waste management timescales for both fission products and actinides; and a very long-term fuel supply requiring no enrichment

  1. Synergistic effects of irradiation of waste-water

    International Nuclear Information System (INIS)

    Woodbridge, D.D.

    1975-01-01

    Water is an absolute necessity for all forms of animal and plant life. As man's requirements for water increase, the need for better methods of purification also increase. Technology has been slow to develop new methods of water treatment for the direct utilization of waste-water. Many new construction projects are at a standstill because waste-water treatment methods have not been developed to handle adequately the ever-increasing flow of sewage. Theoretical considerations of the use of high-level radiation in the treatment of waste-water have failed to consider the effects of the hydrated electron, and the potential of the possible synergistic effects of combining chlorine, oxygen and irradiation. An extensive testing programme at the University Center for Pollution Research of the Florida Institute of Technology over the past four years has shown that irradiation of waste-water samples immersed in an aqueous environment provide bacterial kill and reduction in organic pollution far greater than that obtained from theoretical considerations of G values and earlier experiments where the waste samples were not immersed in an aqueous environment. These testing programmes have investigated the synergistic effects of combining oxygen and irradiation. Each of these combined treatments resulted in an increased bacterial kill factor. Tests on Staphylococcus aureus bacteria and faecal streptococcus bacteria indicate that the synergistic effects observed for faecal coliform bacteria also apply to the pathogenic bacteria. A statistical analysis of the data obtained shows the relationships between the various effects on the bacteria. A definite shielding factor from the turbidity of the waste-water has been shown to exist. Synergistic effects have been shown to offset significantly the shielding effects. Optimization of these synergistic effects can greatly increase the effectiveness of irradiation in the treatment of waste-water. (author)

  2. Preparation of Nanosilver and Nanogold Based on Dog Rose Aqueous Extract

    OpenAIRE

    Pulit, Jolanta; Banach, Marcin

    2014-01-01

    This paper describes a process of obtaining nanosilver and nanogold based on chemical reduction using substances contained in the aqueous extract of dog rose (Rosa canina). The resulting products were subjected to spectrophotometric analysis (UV-Vis), and testing of the nanoparticles’ size and suspension stability was carried out by measuring the electrokinetic potential, ζ, via dynamic light scattering (DLS). Solid samples were imaged by scanning electron microscopy (SEM). The obtained data ...

  3. Offshore disposal of oil-based drilling fluid waste

    International Nuclear Information System (INIS)

    Malachosky, E.; Shannon, B.E.; Jackson, J.E.

    1991-01-01

    Offshore drilling operations in the Gulf of Mexico may use oil-based drilling fluids to mitigate drilling problems. The result is the generation of a significant quantity of oily cuttings and mud. The transportation of this waste for onshore disposal is a concern from a standpoint of both personnel safety and potential environmental impact. A process for preparing a slurry of this waste and the subsequent disposal of the slurry through annular pumping has been put into use by ARCO Oil and Gas Company. The disposal technique has been approved by the Minerals Management Service (MMS). The slurried waste is displaced down a casing annulus into a permeable zone at a depth below the surface casing setting depth. The annular disposal includes all cuttings and waste oil mud generated during drilling with oil-based fluids. This disposal technique negates the need for cuttings storage on the platform, transportation to shore, and the environmental effects of onshore surface disposal. The paper describes the environmental and safety concerns with onshore disposal, the benefits of annular disposal, and the equipment and process used for the preparation and pumping of the slurry

  4. Thermodynamic study of the adsorption of chromium ions from aqueous solution on waste corn cobs material

    Directory of Open Access Journals (Sweden)

    Rafael A. Fonseca-Correa

    2014-12-01

    Full Text Available The paper shows the results of a study obtaining activated carbon from corn cobs and determining its use as an adsorbent for the removal of Cr3+ from aqueous solutions. The finely ground precursor was subjected to pyrolysis at 600 and 900 °C in a nitrogen atmosphere and chemical activation with H2O2 and HNO3. The effects of pyrolysis conditions and activation method on the physicochemical properties of the materials obtained were tested. The samples were characterised chemically and texturally. Were obtained microporous activated carbons of well-developed surface area varying from 337 to 1213 m2/g and exhibited differences acid-base character of the surface. The results obtained shows that a suitable good option of the activation procedure for corncobs permits the production of economic adsorbents with high sorption capacity for Cr3+ from aqueous solutions. A detailed study of immersion calorimetry was performed with carbons prepared from corn cobs to establish possible relationships with these materials between the enthalpies of immersion and textural and chemical parameters.

  5. Secondary Waste Form Down Selection Data Package – Ceramicrete

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Westsik, Joseph H.

    2011-08-31

    binder is formed through an acid-base reaction between calcined magnesium oxide (MgO; a base) and potassium hydrogen phosphate (KH{sub 2}PO{sub 4}; an acid) in aqueous solution. The reaction product sets at room temperature to form a highly crystalline material. During the reaction, the hazardous and radioactive contaminants also react with KH{sub 2}PO{sub 4} to form highly insoluble phosphates. In this data package, physical property and waste acceptance data for Ceramicrete waste forms fabricated with wastes having compositions that were similar to those expected for secondary waste effluents, as well as secondary waste effluent simulants from the Hanford Tank Waste Treatment and Immobilization Plant were reviewed. With the exception of one secondary waste form formulation (25FA+25 W+1B.A. fabricated with the mixed simulant did not meet the compressive strength requirement), all the Ceramicrete waste forms that were reviewed met or exceeded Integrated Disposal Facility waste acceptance criteria.

  6. Conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    Beamer, N.V.; Bourns, W.T.; Buckley, L.P.; Speranzini, R.A.

    1981-12-01

    A Waste Treatment Centre (WTC) is being constructed at the Chalk River Nuclear Laboratories to develop and demonstrate processes for converting reactor wastes to a form suitable for disposal. The WTC contains a starved air incinerator for reducing the volume of combustible solid wastes, a reverse osmosis section for reducing the volume of liquid wastes and an immobilization section for incorporating the conditioned wastes in bitumen. The incinerator is commissioned on inactive waste: approximately 16.5 Mg of waste packaged in polyethylene bags has been incinerated in 17 burns. Average weight and volume reductions of 8.4:1 and 32:1, respectively, have been achieved. Construction of the reverse osmosis section of WTC is complete and inactive commissioning will begin in 1982 January. The reverse osmosis section was designed to process 30,000 m 3 /a of dilute radioactive waste. The incinerator ash and concentrated aqueous waste will be immobiblized in bitumen using a horizontal mixer and wiped-film evaporator. Results obtained during inactive commissioning of the incinerator are described along with recent results of laboratory programs directed at demonstrating the reverse osmosis and bituminization processes

  7. Management of hazardous wastes in the laboratories of the Instituto Tecnologico de Costa Rica (phase III)

    International Nuclear Information System (INIS)

    Salas Jimenez, Juan Carlos; Quesada Carvajal, Hilda; Harada, Katsuhiro

    2009-01-01

    A scaling at pilot plant level was performanced for the treatment of wastes are stored in significant quantities at the Instituto Tecnologico de Costa Rica (ITCR). These wastes are aqueous of heavy metals from laboratories and of the nitriding process slag. Dr. Katsuhiro Harada, Japanese aid worker, suggested a treatment methodology that was tested and adapted to the characteristics of hazardous wastes generated in the ITCR. In addition, an operating procedure was suggested to centralize the treatment of waste produced in different labs but they have similar chemical characteristics; therefore can be treated with the same chemical method. For these cases it is easier and cheaper to concentrate the treatment in one place, and in the case of extremely hazardous waste, whose treatment and disposal are somewhat complicated to implement, it is advisable to establish a specialized laboratory with trained personnel for management. A hazardous waste laboratory equipped with a reactor, sludge filter and laboratory equipment for analysis. The methods tested in the pilot plant for the treatment of aqueous wastes of heavy metals and cyanide slag were effective. (author) [es

  8. Investigation of variable compositions on the removal of technetium from Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M. 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); Pareizs, John M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-29

    The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the offgas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

  9. Chemical modeling of irreversible reactions in nuclear waste-water-rock systems

    International Nuclear Information System (INIS)

    Wolery, T.J.

    1981-02-01

    Chemical models of aqueous geochemical systems are usually built on the concept of thermodynamic equilibrium. Though many elementary reactions in a geochemical system may be close to equilibrium, others may not be. Chemical models of aqueous fluids should take into account that many aqueous redox reactions are among the latter. The behavior of redox reactions may critically affect migration of certain radionuclides, especially the actinides. In addition, the progress of reaction in geochemical systems requires thermodynamic driving forces associated with elementary reactions not at equilibrium, which are termed irreversible reactions. Both static chemical models of fluids and dynamic models of reacting systems have been applied to a wide spectrum of problems in water-rock interactions. Potential applications in nuclear waste disposal range from problems in geochemical aspects of site evaluation to those of waste-water-rock interactions. However, much further work in the laboratory and the field will be required to develop and verify such applications of chemical modeling

  10. Disposal method of radioactive wastes

    International Nuclear Information System (INIS)

    Uetake, Naoto; Fukazawa, Tetsuo.

    1986-01-01

    Purpose: To improve the safety of underground disposal of radioactive wastes for a long period of time by surrounding the periphery of the radioactive wastes with materials that can inhibit the migration of radioactive nuclides and are physically and chemically stable. Method: Hardening products prepared from a water-hardenable calcium silicate compound and an aqueous solution of alkali silicate have compression strength as comparable with that of concretes, high water tightness and adsorbing property to radioactive isotopes such as cobalt similar to that of concretes and they also show adsorption to cesium which is not adsorbed to concretes. Further, the kneaded slurry thereof is excellent in the workability and can be poured even into narrow gaps. Accordingly, by alternately charging granular radioactive wastes and this slurry before hardening into the ground, the radioactive wastes can be put to underground disposal stably with simple procedures. (Kamimura, M.)

  11. A risk-based decision-aiding tool for waste disposal

    International Nuclear Information System (INIS)

    Weiner, R.F.; Reiser, A.S.; Elcock, C.G.; Nevins, S.

    1997-01-01

    N-CART (the National Spent Nuclear Fuel Program Cost Analysis and Risk Tool) is being developed to aid in low-risk, cost-effective, timely management of radioactive waste and spent nuclear fuel, and can therefore be used in management of mixed waste. N-CART provides evaluation of multiple alternatives and presents the consequences of proposed waste management activities in a clear and concise format. N-CART's decision-aiding analyses include comparisons and sensitivity analyses of multiple alternatives and allows the user to perform quick turn-around open-quotes what ifclose quotes studies to investigate various scenarios. Uncertainties in data (such as cost and schedule of various activities) are represented as distributions. N-CART centralizes documentation of the bases of program alternatives and program decisions, thereby supporting responses to stakeholders concerns. The initial N-CART design considers regulatory requirements, costs, and schedules for alternative courses of action. The final design will include risks (public health, occupational, economic, scheduling), economic benefits, and the impacts of secondary waste generation. An optimization tool is being incorporated that allows the user to specify the relative importance of cost, time risks, and other bases for decisions. The N-CART prototype can be used to compare the costs and schedules of disposal alternatives for mixed low-level radioactive waste (MLLW) and greater-than-Class-C (GTCC) waste, as well as spent nuclear fuel (SNF) and related scrap material

  12. Radioactive wastes: the challenge of volumes reduction

    International Nuclear Information System (INIS)

    Lepetit, V.

    2005-01-01

    The reduction of radioactive waste volumes is a priority for the French atomic energy commission (CEA) and for the Areva group. This article gives a rapid overview of the equipments and processes used to separate the valorizable materials from the ultimate wastes: pulsed separation columns and evaporators for the liquid-liquid extraction, compactification of spent fuel hulls, remote handling systems, recoverable colloid for surface decontamination, decontaminating foam, hydrothermal oxidation of organic and aqueous effluents, cold crucible vitrification etc. (J.S.)

  13. DEVELOPMENT OF PROTOTYPE TITANATE ION EXCHANGE LOADED MEMBRANES FOR STRONTIUM, CESIUM AND ACTINIDE DECONTAMINATION FROM AQUEOUS MEDIA

    Energy Technology Data Exchange (ETDEWEB)

    Oji, L; Keisha Martin, K; David Hobbs, D

    2008-05-30

    We have successfully incorporated high surface area particles of titanate ion exchange materials (monosodium titanate and crystalline silicotitanate) with acceptable particle size distribution into porous and inert support membrane fibrils consisting of polytetrafluoroethylene (Teflon{reg_sign}), polyethylene and cellulose materials. The resulting membrane sheets, under laboratory conditions, were used to evaluate the removal of surrogate radioactive materials for cesium-137 and strontium-90 from high caustic nuclear waste simulants. These membrane supports met the nominal requirement for nonchemical interaction with the embedded ion exchange materials and were porous enough to allow sufficient liquid flow. Some of this 47-mm size stamped out prototype titanium impregnated ion exchange membrane discs was found to remove more than 96% of dissolved cesium-133 and strontium-88 from a caustic nuclear waste salt simulants. Since in traditional ion exchange based column technology monosodium titanate (MST) is known to have great affinity for the sorbing of other actinides like plutonium, neptunium and even uranium, we expect that the MST-based membranes developed here, although not directly evaluated for uptake of these three actinides because of costs associated with working with actinides which do not have 'true' experimental surrogates, would also show significant affinity for these actinides in aqueous media. It was also observed that crystalline silicotitanate impregnated polytetrafluoroethylene or polyethylene membranes became less selective and sorbed both cesium and strontium from the caustic aqueous salt simulants.

  14. Removal of Pb(II) ions from aqueous solution by a waste mud from copper mine industry: equilibrium, kinetic and thermodynamic study.

    Science.gov (United States)

    Ozdes, Duygu; Gundogdu, Ali; Kemer, Baris; Duran, Celal; Senturk, Hasan Basri; Soylak, Mustafa

    2009-07-30

    The objective of this study was to assess the adsorption potential of a waste mud (WM) for the removal of lead (Pb(II)) ions from aqueous solutions. The WM was activated with NaOH in order to increase its adsorption capacity. Adsorption studies were conducted in a batch system as a function of solution pH, contact time, initial Pb(II) concentration, activated-waste mud (a-WM) concentration, temperature, etc. Optimum pH was specified as 4.0. The adsorption kinetic studies indicated that the overall adsorption process was best described by pseudo-second-order kinetics. The equilibrium adsorption capacity of a-WM was obtained by using Langmuir and Freundlich isotherm models and both models fitted well. Adsorption capacity for Pb(II) was found to be 24.4 mg g(-1) for 10 g L(-1) of a-WM concentration. Thermodynamic parameters including the Gibbs free energy (Delta G degrees), enthalpy (Delta H degrees), and entropy (DeltaS degrees) indicated that the adsorption of Pb(II) ions on the a-WM was feasible, spontaneous and endothermic, at temperature range of 0-40 degrees C. Desorption studies were carried out successfully with diluted HCl solutions. The results indicate that a-WM can be used as an effective and no-cost adsorbent for the treatment of industrial wastewaters contaminated with Pb(II) ions.

  15. Removal of Pb(II) ions from aqueous solution by a waste mud from copper mine industry: Equilibrium, kinetic and thermodynamic study

    International Nuclear Information System (INIS)

    Ozdes, Duygu; Gundogdu, Ali; Kemer, Baris; Duran, Celal; Senturk, Hasan Basri; Soylak, Mustafa

    2009-01-01

    The objective of this study was to assess the adsorption potential of a waste mud (WM) for the removal of lead (Pb(II)) ions from aqueous solutions. The WM was activated with NaOH in order to increase its adsorption capacity. Adsorption studies were conducted in a batch system as a function of solution pH, contact time, initial Pb(II) concentration, activated-waste mud (a-WM) concentration, temperature, etc. Optimum pH was specified as 4.0. The adsorption kinetic studies indicated that the overall adsorption process was best described by pseudo-second-order kinetics. The equilibrium adsorption capacity of a-WM was obtained by using Langmuir and Freundlich isotherm models and both models fitted well. Adsorption capacity for Pb(II) was found to be 24.4 mg g -1 for 10 g L -1 of a-WM concentration. Thermodynamic parameters including the Gibbs free energy (ΔG o ), enthalpy (ΔH o ), and entropy (ΔS o ) indicated that the adsorption of Pb(II) ions on the a-WM was feasible, spontaneous and endothermic, at temperature range of 0-40 o C. Desorption studies were carried out successfully with diluted HCl solutions. The results indicate that a-WM can be used as an effective and no-cost adsorbent for the treatment of industrial wastewaters contaminated with Pb(II) ions.

  16. Removal of Pb(II) ions from aqueous solution by a waste mud from copper mine industry: Equilibrium, kinetic and thermodynamic study

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-30

    The objective of this study was to assess the adsorption potential of a waste mud (WM) for the removal of lead (Pb(II)) ions from aqueous solutions. The WM was activated with NaOH in order to increase its adsorption capacity. Adsorption studies were conducted in a batch system as a function of solution pH, contact time, initial Pb(II) concentration, activated-waste mud (a-WM) concentration, temperature, etc. Optimum pH was specified as 4.0. The adsorption kinetic studies indicated that the overall adsorption process was best described by pseudo-second-order kinetics. The equilibrium adsorption capacity of a-WM was obtained by using Langmuir and Freundlich isotherm models and both models fitted well. Adsorption capacity for Pb(II) was found to be 24.4 mg g{sup -1} for 10 g L{sup -1} of a-WM concentration. Thermodynamic parameters including the Gibbs free energy ({Delta}G{sup o}), enthalpy ({Delta}H{sup o}), and entropy ({Delta}S{sup o}) indicated that the adsorption of Pb(II) ions on the a-WM was feasible, spontaneous and endothermic, at temperature range of 0-40 {sup o}C. Desorption studies were carried out successfully with diluted HCl solutions. The results indicate that a-WM can be used as an effective and no-cost adsorbent for the treatment of industrial wastewaters contaminated with Pb(II) ions.

  17. Electrochemical characterization of Zr-based thin film metallic glass in hydrochloric aqueous solution

    International Nuclear Information System (INIS)

    Chuang, Ching-Yen; Liao, Yi-Chia; Lee, Jyh-Wei; Li, Chia-Lin; Chu, Jinn P.; Duh, Jenq-Gong

    2013-01-01

    Recently thin film metallic glass represents a class of promising engineering materials for structural applications. In this work, the Zr-based thin film metallic glass (TFMG) was fabricated on the Si and AISI 420 substrates using a Zr–Cu–Ni–Al alloy and pure Zr metal targets by a pulsed DC magnetron sputtering system. The chemical compositions, crystalline structures, microstructures and corrosion behavior in hydrochloric (HCl) aqueous solutions of Zr-based TFMGs were investigated. The results showed that the surface morphologies of Zr-based TFMG were very smooth. A compact and dense structure without columnar structure was observed. The amorphous structure of Zr-based TFMG was characterized by the X-ray diffractometer and transmission electron microscopy analyses. After the potentiodynamic polarization test, the better corrosion resistance was achieved for the Zr-based TFMG coated AISI 420 in 1 mM HCl aqueous solution. Based on the surface morphologies and chemical analysis results of the corroded surfaces, the pitting, crevice corrosion and filiform corrosion were found. The corrosion mechanisms of the Zr-based TFMG were discussed in this work. - Highlights: ► Zr-based thin film metallic glass with amorphous structure. ► Better corrosion resistance of Zr-based thin film metallic glass observed. ► Pitting, crevice and filiform corrosion reactions revealed. ► The Cu-rich corrosion products found in the pit. ► Nanowire and flaky corrosion products formed adjacent to the filiform corrosion path

  18. 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

  19. Sustainable production of valuable compound 3-succinoyl-pyridine by genetically engineering Pseudomonas putida using the tobacco waste.

    Science.gov (United States)

    Wang, Weiwei; Xu, Ping; Tang, Hongzhi

    2015-11-17

    Treatment of solid and liquid tobacco wastes with high nicotine content remains a longstanding challenge. Here, we explored an environmentally friendly approach to replace tobacco waste disposal with resource recovery by genetically engineering Pseudomonas putida. The biosynthesis of 3-succinoyl-pyridine (SP), a precursor in the production of hypotensive agents, from the tobacco waste was developed using whole cells of the engineered Pseudomonas strain, S16dspm. Under optimal conditions in fed-batch biotransformation, the final concentrations of product SP reached 9.8 g/L and 8.9 g/L from aqueous nicotine solution and crude suspension of the tobacco waste, respectively. In addition, the crystal compound SP produced from aqueous nicotine of the tobacco waste in batch biotransformation was of high purity and its isolation yield on nicotine was 54.2%. This study shows a promising route for processing environmental wastes as raw materials in order to produce valuable compounds.

  20. Removal of Radioactive Pollutants by Liquid Emulsion Membrane From Liquid Waste

    International Nuclear Information System (INIS)

    Yossef, Y.A.A.

    2013-01-01

    Radioactive liquid waste should be safely managed because it is potentially hazardous to human health and the environment. Several methods were used for treatment of liquid waste, such as liquid emulsion membrane (LEM). In this work, liquid emulsion membrane using Tri-butyl phosphate (TBP) plus Bis (2-ethylhexyl) phosphate (HDEHP) as mobile carriers, hydrochloric acid (HCl) as stripping agents and an emulsifying agent (span 80) was used for the extraction of uranium ions from radioactive liquid waste. Various parameters influencing the permeation of uranium ions through the membrane have been optimized to separate uranium ions from radioactive liquid waste such as: the effects of membrane material, carrier concentration, operating conditions, etc. were examined; moreover, the transport mechanism of this uranium was also studied. The internal mass transfer in the water/oil (W/O) emulsion drop, the external mass transfer around the drop, the rates of formation, and the decomposition of the complex at the external aqueous-organic interface were considered. The results show that, the liquid emulsion membrane which consists of (25% by volume HDEHP, 0.005 M + 75% by volume TBP, 0.01 M) as extractant (carrier), span 80, 4% (v/v) (sorbitan monooleate) as surfactant agent, hydrochloric acid (HCl), (1.0 M) as stripping agent. From the results, the maximum extraction percent of uranium ions (nearly about of 100%) occurred at the operating conditions: stirring speed =500 rpm, the ratio between LEM and feed phase (liquid waste) = 20 ml: 100 ml, the ratio between organic phase (membrane phase) to internal aqueous phase (stripping phase) = 1.0 and the ph value of the external aqueous phase equal to 5.0.

  1. On the ATW-concepts: ITP approach and opportunities

    Science.gov (United States)

    Simonenko, V. A.; Grebyonkin, K. F.

    1995-09-01

    It is discussed the interest of Russian Federal Nuclear Center-Institute of Technical Physics at Chelyabinsk-70 in the research of Accelerator Driven Technologies applications for radioactive waste transmutation, cumulated actinides burning, energy production. The ITP background and opportunities for this research are presented. It is shown the ITP possibilities for testing and experimental development of Accelerator Driven Technologies.

  2. On the ATW-concepts: ITP approach and opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Simonenko, V.A.; Gregyonkin, K.F. [Institute of Technical Physics, Chelyabinsk (Russian Federation)

    1995-10-01

    It is discussed the interest of Russian Federal Nuclear Center - Institute of Technical Physics at Chelyabinsk-70 in the research of Accelerator Driven Technologies applications for radioactive waste transmutation, cumulated actinides burning, energy production. The ITP background and opportunities for this research are presented. It is shown the ITP possibilities for testing and experimental development of Accelerator Driven Technologies.

  3. Laboratory optimization tests of technetium decontamination of Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

    Energy Technology Data Exchange (ETDEWEB)

    Taylor-Pashow, Kathryn M.L. [Savannah River Site (SRS), Aiken, SC (United States); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-11-01

    The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable simplified operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

  4. 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)

  5. 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)

  6. Waste minimization for land-based drilling operations

    International Nuclear Information System (INIS)

    Thurber, N.E.

    1992-01-01

    This paper discusses engineering variables that should be addressed to minimize waste-toxicity and generation while drilling land-based wells. Proper balance of these variables provides both operational and environmental benefits

  7. Scanning electron microscopic study of hazardous waste flakes of polyethylene terephthalate (PET) by aminolysis and ammonolysis

    Energy Technology Data Exchange (ETDEWEB)

    Mittal, Alok, E-mail: aljymittal@yahoo.co.in [Department of Chemistry, Maulana Azad National Institute of Technology (A Deemed University), Bhopal 462051 (India); Soni, R.K.; Dutt, Krishna; Singh, Swati [Department of Chemistry, Ch. Charan Singh University, Meerut 250004 (India)

    2010-06-15

    Polyethylene terephthalate (PET) waste flakes were degraded with aqueous methylamine and aqueous ammonia, respectively at room temperature in the presence and absence of quaternary ammonium salt as a catalyst for different periods of time. The aminolysed and ammonolysed PET samples were investigated for the surface morphology with the help of scanning electron micrograph (SEM). It shows that the semi-crystalline PET waste samples reduce to monodisperse rods before fully degradation to the end products. The presence of the catalyst provides site for degradation of PET waste and enhances the rate of degradation. The SEM shows early developments of fissures in comparison to the one in absence of quaternary ammonium salt used as catalyst.

  8. Scanning electron microscopic study of hazardous waste flakes of polyethylene terephthalate (PET) by aminolysis and ammonolysis

    International Nuclear Information System (INIS)

    Mittal, Alok; Soni, R.K.; Dutt, Krishna; Singh, Swati

    2010-01-01

    Polyethylene terephthalate (PET) waste flakes were degraded with aqueous methylamine and aqueous ammonia, respectively at room temperature in the presence and absence of quaternary ammonium salt as a catalyst for different periods of time. The aminolysed and ammonolysed PET samples were investigated for the surface morphology with the help of scanning electron micrograph (SEM). It shows that the semi-crystalline PET waste samples reduce to monodisperse rods before fully degradation to the end products. The presence of the catalyst provides site for degradation of PET waste and enhances the rate of degradation. The SEM shows early developments of fissures in comparison to the one in absence of quaternary ammonium salt used as catalyst.

  9. Standard test method for static leaching of monolithic waste forms for disposal of radioactive waste

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method provides a measure of the chemical durability of a simulated or radioactive monolithic waste form, such as a glass, ceramic, cement (grout), or cermet, in a test solution at temperatures <100°C under low specimen surface- area-to-leachant volume (S/V) ratio conditions. 1.2 This test method can be used to characterize the dissolution or leaching behaviors of various simulated or radioactive waste forms in various leachants under the specific conditions of the test based on analysis of the test solution. Data from this test are used to calculate normalized elemental mass loss values from specimens exposed to aqueous solutions at temperatures <100°C. 1.3 The test is conducted under static conditions in a constant solution volume and at a constant temperature. The reactivity of the test specimen is determined from the amounts of components released and accumulated in the solution over the test duration. A wide range of test conditions can be used to study material behavior, includin...

  10. Chemical characterization of SRP waste tank sludges and supernates

    International Nuclear Information System (INIS)

    Gray, L.W.; Donnan, M.Y.; Okamoto, B.Y.

    1979-08-01

    Most high-level liquid wastes at the Savannah River Plant (SRP) are byproducts from plutonium and enriched uranium recovery processes. The high-level liquid wastes generated by these separations processes are stored in large, underground, carbon-steel tanks. The liquid wastes consist of: supernate (an aqueous solution containing sodium, nitrate, nitrite, hydroxyl, and aluminate ions), sludge (a gelatinous material containing insoluble components of the waste, such as ferric and aluminum hydroxides, and mercuric and manganese oxides), and salt cake (crystals, such as sodium nitrate, formed by evaporation of water from supernate). Analyses of SRP wastes by laser-Raman spectrometry, atomic absorption spectrometry, spark-source mass spectrometry, neutron activation analysis, colorimetry, ion chromatography, and various other wet-chemical and radiochemical methods are discussed. These analyses are useful in studies of waste tank corrosion and of forms for long-term waste storage

  11. Waste Minimization Study on Pyrochemical Reprocessing Processes

    International Nuclear Information System (INIS)

    Boussier, H.; Conocar, O.; Lacquement, J.

    2006-01-01

    Ideally a new pyro-process should not generate more waste, and should be at least as safe and cost effective as the hydrometallurgical processes currently implemented at industrial scale. This paper describes the thought process, the methodology and some results obtained by process integration studies to devise potential pyro-processes and to assess their capability of achieving this challenging objective. As example the assessment of a process based on salt/metal reductive extraction, designed for the reprocessing of Generation IV carbide spent fuels, is developed. Salt/metal reductive extraction uses the capability of some metals, aluminum in this case, to selectively reduce actinide fluorides previously dissolved in a fluoride salt bath. The reduced actinides enter the metal phase from which they are subsequently recovered; the fission products remain in the salt phase. In fact, the process is not so simple, as it requires upstream and downstream subsidiary steps. All these process steps generate secondary waste flows representing sources of actinide leakage and/or FP discharge. In aqueous processes the main solvent (nitric acid solution) has a low boiling point and evaporate easily or can be removed by distillation, thereby leaving limited flow containing the dissolved substance behind to be incorporated in a confinement matrix. From the point of view of waste generation, one main handicap of molten salt processes, is that the saline phase (fluoride in our case) used as solvent is of same nature than the solutes (radionuclides fluorides) and has a quite high boiling point. So it is not so easy, than it is with aqueous solutions, to separate solvent and solutes in order to confine only radioactive material and limit the final waste flows. Starting from the initial block diagram devised two years ago, the paper shows how process integration studies were able to propose process fittings which lead to a reduction of the waste variety and flows leading at an 'ideal

  12. Final product analysis in the e-beam and gamma radiolysis of aqueous solutions of metoprolol tartrate

    Energy Technology Data Exchange (ETDEWEB)

    Slegers, Catherine [Universite Catholique de Louvain, Unite d' Analyse Chimique et Physico-chimique des Medicaments, CHAM 72.30, Avenue E. Mounier, 72, B-1200 Brussels (Belgium)]. E-mail: catherine.slegers@cham.ucl.ac.be; Tilquin, Bernard [Universite Catholique de Louvain, Unite d' Analyse Chimique et Physico-chimique des Medicaments, CHAM 72.30, Avenue E. Mounier, 72, B-1200 Brussels (Belgium)

    2006-09-15

    The radiostability of metoprolol tartrate aqueous solutions and the influence of the absorbed dose (0-50 kGy), dose rate (e-beam (EB) vs. gamma ({gamma})) and radioprotectors (pharmaceutical excipients) are investigated by HPLC-UV analyses and through computer simulations. The use of radioprotecting excipients is more promising than an increase in the dose rate to lower the degradation of metoprolol tartrate aqueous solutions for applications such as radiosterilization. The decontamination of metoprolol tartrate from waste waters by EB processing appears highly feasible.

  13. Revised Arrangements for the Management of Solid and Non-Aqueous Radioactive Waste - 12452

    Energy Technology Data Exchange (ETDEWEB)

    Fullbrook, Michael; Walker, Johann; Macnab, Alec [Atomic Weapons Establishment, Aldermaston (United Kingdom)

    2012-07-01

    In 2010, Atomic Weapons Establishment (AWE) identified a requirement to implement revised management arrangements for the generation, storage and disposal of radioactive waste. A thorough review of the current arrangements/processes was undertaken which included both legal compliance requirements and the identification of business improvement opportunities. On completion of this review a suitable project team was established and in 2011 an integrated Radioactive Waste Management process was implemented throughout the business. Initial results have shown measurable improvements within Radioactive Waste management compliance, operator understanding and increased business efficiency. Through the development and implementation of the revised working arrangements AWE has been able to continue to demonstrate both legal compliance to its regulators along with business efficiency and effectiveness improvements. Simple to follow process maps have improved employees understanding of Radioactive Waste management requirements, provided them with easily accessible information and ensured the business operates in a single coherent manner. The implementation of a modern electronic data management system has ensured all waste related information is easily retrievable and appropriately maintained. The additional functions that have been built into the system have reduced the potential for human error and increased the overall efficiency of the Waste Management department through the use of the automated report generation functionality. (authors)

  14. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO2 fuel reprocessing waste

    International Nuclear Information System (INIS)

    Tait, J.C.

    1993-05-01

    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of 129 I, 85 Kr and 14 C. (author). 104 refs., 9 tabs., 5 figs

  15. Process for the encapsulation of radioactive wastes

    International Nuclear Information System (INIS)

    Pordes, O.; Plows, J.P.; Hill, M.L.

    1980-01-01

    Radioactive waste material, particularly radioactive ion exchange resin in the wet condition, is encapsulated in a polyurethane by dispersing the waste in an aqueous emulsion of an organic polyol, a polyisocyanate and an hydraulic cement and allowing the emulsion to set to form a monolithic block. If desired the emulsion may also contain additional filler e.g. sand or aggregate to increase the density of the final product. Preferred polyurethanes are those made from a polyester polyol and an organic diisocyanate, particularly hexamethylene diisocyanate. (author)

  16. Waste printing paper as analogous adsorbents for heavy metals in ...

    African Journals Online (AJOL)

    user

    heavy metals uptake from aqueous solutions but the recovery efficacy as economic and environmental ... system. 1 . Wastes containing metals are directly or indirectly discharge into the environment ... According to World health Organization. 5.

  17. 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

  18. Utilization of Aluminum Waste with Hydrogen and Heat Generation

    Science.gov (United States)

    Buryakovskaya, O. A.; Meshkov, E. A.; Vlaskin, M. S.; Shkolnokov, E. I.; Zhuk, A. Z.

    2017-10-01

    A concept of energy generation via hydrogen and heat production from aluminum containing wastes is proposed. The hydrogen obtained by oxidation reaction between aluminum waste and aqueous solutions can be supplied to fuel cells and/or infrared heaters for electricity or heat generation in the region of waste recycling. The heat released during the reaction also can be effectively used. The proposed method of aluminum waste recycling may represent a promising and cost-effective solution in cases when waste transportation to recycling plants involves significant financial losses (e.g. remote areas). Experiments with mechanically dispersed aluminum cans demonstrated that the reaction rate in alkaline solution is high enough for practical use of the oxidation process. In theexperiments aluminum oxidation proceeds without any additional aluminum activation.

  19. Hydrogen Generation from Sugars via Aqueous-Phase Reforming

    International Nuclear Information System (INIS)

    Randy D Cortright

    2006-01-01

    Virent Energy Systems, Inc. is commercializing the Aqueous Phase Reforming (APR) process that allows the generation of hydrogen-rich gas streams from biomass-derived compounds such as glycerol, sugars, and sugar alcohols. The APR process is a unique method that generates hydrogen from aqueous solutions of these oxygenated compounds in a single step reactor process compared to the three or more reaction steps required for hydrogen generation via conventional processes that utilize non-renewable fossil fuels. The key breakthrough of the APR process is that the reforming of these aqueous solutions is done in the liquid phase. The patented APR process occurs at temperatures (150 C to 270 C) where the water-gas shift reaction is favorable, making it possible to generate hydrogen with low amounts of CO in a single chemical reactor. Furthermore, the APR process occurs at pressures (typically 15 to 50 bar) where the hydrogen-rich effluent can be effectively purified using either membrane technology or pressure swing adsorption technology. The utilization of biomass-based compounds allows the APR process to be a carbon neutral method to generate hydrogen. In the near term, the feed-stock of interest is waste glycerol that is being generated in large quantities as a byproduct in the production of bio-diesel. Virent has developed the APR system for on-demand generation of hydrogen-rich fuel gas from either glycerol or sorbitol (the sugar alcohol formed by hydrogenation of glucose) to fuel a stationary internal combustion engine driven generator (10 kW). Under a USDOE funded project, Virent is currently developing the APR process to generate high yields of hydrogen from corn-derived glucose. This project objective is to achieve the DOE 2010 cost target for distributed production from renewable liquid fuels of 3.60 dollars/gge (gasoline gallon equivalent) delivered. (authors)

  20. Microbiology and radioactive waste disposal

    International Nuclear Information System (INIS)

    Colasanti, R.; Coutts, D.; Pugh, S.Y.R.; Rosevear, A.

    1990-03-01

    The present Nirex Safety Assessment Research Programme on microbiology is based on experimental as well as theoretical work. It has concentrated on the study of how mixed, natural populations of microbes might survive and grow on the organic component of Low Level Radioactive Wastes (LLW) and PCM (Plutonium Contaminated Waste) in a cementitious waste repository. The present studies indicate that both carbon dioxide and methane will be produced by microbial action within the repository. Carbon dioxide will dissolve and react with the concrete to a limited extent so methane will be the principal component of the produced gas. The concentration of hydrogen, derived from corrosion, will be depressed by microbial action and that this will further elevate methane levels. Actual rates of production will be lower than that in a domestic landfill due to the more extreme pH. Microbial action will clearly affect the aqueous phase chemistry where organic material is present in the waste. The cellulosic fraction is the main determinant of cell growth and the appearance of soluble organics. The structure of the mathematical model which has been developed, predicts the general features which are intuitively expected in a developing microbial population. It illustrates that intermediate compounds will build up in the waste until growth of the next organism needed for sequential degradation is initiated. The soluble compounds in the pore water and the mixture of microbes present in the waste will vary with time and sustain biological activity over a prolonged period. Present estimates suggest that most microbial action in the repository will be complete after 400 years. There is scope for the model to deal with environmental factors such as temperature and pH and to introduce other energy sources such as hydrogen. (author)