WorldWideScience

Sample records for radioactive spent ion

  1. Application of ion exchange processes for the treatment of radioactive waste and management of spent ion exchangers

    International Nuclear Information System (INIS)

    2002-01-01

    This report describes the ion exchange technologies currently used and under development in nuclear industry, in particular for waste management practices, along with the experience gained in their application and with the subsequent handling, treatment and conditioning of spent ion exchange media for long term storage and/or disposal. The increased role of inorganic ion exchangers for treatment of radioactive liquid waste, both in nuclear power plant operations and in the fuel reprocessing sector, is recognised in this report. The intention of this report is to consolidate the previous publications, document recent developments and describe the state of the art in the application of ion exchange processes for the treatment of radioactive liquid waste and the management of spent ion exchange materials

  2. Oxidative degradation of low and intermediate level Radioactive organic wastes 2. Acid decomposition on spent Ion-Exchange resins

    International Nuclear Information System (INIS)

    Ghattas, N.K.; Eskander, S.B.

    1995-01-01

    The present work provides a simplified, effective and economic method for the chemical decomposition of radioactively contaminated solid organic waste, especially spent ion - exchange resins. The goal is to achieve volume reduction and to avoid technical problems encountered in processes used for similar purposes (incineration, pyrolysis). Factors efficiency and kinetics of the oxidation of the ion exchange resins in acid medium using hydrogen peroxide as oxidant, namely, duration of treatment and the acid to resin ratio were studied systematically on a laboratory scale. Moreover the percent composition of the off-gas evolved during the decomposition process was analysed. 3 figs., 5 tabs

  3. Oxidative degradation of low and intermediate level Radioactive organic wastes 2. Acid decomposition on spent Ion-Exchange resins

    Energy Technology Data Exchange (ETDEWEB)

    Ghattas, N K; Eskander, S B [Radioisotope dept., atomic energy authority, (Egypt)

    1995-10-01

    The present work provides a simplified, effective and economic method for the chemical decomposition of radioactively contaminated solid organic waste, especially spent ion - exchange resins. The goal is to achieve volume reduction and to avoid technical problems encountered in processes used for similar purposes (incineration, pyrolysis). Factors efficiency and kinetics of the oxidation of the ion exchange resins in acid medium using hydrogen peroxide as oxidant, namely, duration of treatment and the acid to resin ratio were studied systematically on a laboratory scale. Moreover the percent composition of the off-gas evolved during the decomposition process was analysed. 3 figs., 5 tabs.

  4. Application of ion beams in materials science of radioactive waste forms: focus on the performance of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garrido, Frederico [Centre de Spectrometrie Nucleaire et de Spectrometrie de Masse, CNRS-IN2P3-Universite Paris-Sud, Batiments 104-108, 91405 Orsay Campus (France)]. E-mail: garrido@csnsm.in2p3.fr; Nowicki, Lech [Andrzej Soltan Institute for Nuclear Studies, Hoza 69, 00-681 Warsaw (Poland); Thome, Lionel [Centre de Spectrometrie Nucleaire et de Spectrometrie de Masse, CNRS-IN2P3-Universite Paris-Sud, Ba-hat timents 104-108, 91405 Orsay Campus (France)

    2005-10-15

    Ion beam techniques provide unique tools for the qualification of radioactive waste forms. They address three major issues: (i) the simulation by ion irradiation of the stability of a matrix submitted to radiative environment; (ii) the doping of a material with stable or radioactive elements which simulate the species to be confined; (iii) the characterisation of a material via nuclear microanalysis techniques. Among various classes of nuclear matrices the spent nuclear fuel is widely considered as a potential candidate for the stabilisation of radioactive wastes in scenarios of long term interim storage or final geological disposal. Illustrative examples revealing the potentialities of the use of ion beams either as a pure characterisation tool - to investigate the chemical stability of the UO{sub 2} matrix under an oxygen potential - or in a combined way (e.g. irradiation/characterisation, doping/characterisation) - to explore the radiation stability and the behaviour of foreign species - are presented. Transformations (stoichiometry, depth and structure of growing hyperstoichiometric U{sub 4}O{sub 9}/U{sub 3}O{sub 7} oxides) occurring during low-temperature air oxidation of uranium dioxide single crystals are reported. Swift heavy ion irradiation of UO{sub 2} single crystals leads to a peculiar single crystal-polycrystal transformation (i.e. polygonisation of the fluorite-type structure of the material). Irradiation of UO{sub 2} at low energy shows that the damage production is directly linked to the energy deposited in nuclear elastic collisions. The lattice location of helium atoms (generated in large amount during the storage period) in interstitial octahedral positions is discussed.

  5. ''Spray'' drying unit for spent ion-exchange resins sludges and radioactive concentrates

    International Nuclear Information System (INIS)

    Raibaud, J.

    1985-01-01

    The procedure consisting in drying radwaste either in liquid form or in aqueous suspension is a very attractive solution for volume Reduction. Technicatome presents an experimental spray drying station for 50 kg/hr, using the LEAFLASH process, developed by Rhone Poulenc Recherches. This process, used at full scale in a large number of branches in industry, is applicable to the drying of various materials: bead type ion-echange resins, powered ion exchange resins, centrifuge sludges, filter sludges, evaporator bottoms [fr

  6. Spray-type drying unit for spent ion exchange resins, sludges and radioactive concentrates

    International Nuclear Information System (INIS)

    Raibaud, J.

    1986-01-01

    The process for drying radwaste in the liquid form or in aqueous suspension is a very attractive solution from the standpoint of volume reduction. Most of the existing drying facilities are not well adapted for drying the varieties of aqueous waste produced by the nuclear research centres and nuclear power plants, such as: - ion exchange resins, bead type or powdered resins, - centrifuge sludges, - settling sludges, - evaporator bottoms. Technicatome has selected the LEAFLASH process developed by Rhone Poulenc Recherches for drying the nuclear aqueous waste. This process has been well tried at full scale in a large number of industrial branches. The advantages of the process have been confirmed by the results obtained in operating a pilot facility. They include: - high flexibility in operation: - quick start-up and stoppage procedures, - adaptation to a wide spectrum of liquid waste without significant changes in the adjustment of the device. - compactness, - low power consumption, - complete drying of the waste for any initial concentration [fr

  7. Electrodialytic decontamination of spent ion exchange resins

    International Nuclear Information System (INIS)

    Nott, B.R.

    1982-01-01

    Development of a novel electrodialytic decontamination process for the selective removal of radioactive Cs from spent ion exchange resins containing large amounts of Li is described. The process involves passage of a dc electric current through a bed of the spent ion exchange resin in a specially designed electrodialytic cell. The radiocesium so removed from a volume of the spent resin is concentrated onto a much smaller volume of a Cs selective sorbent to achieve a significant radioactive waste volume reduction. Technical feasibility of the electrodialytic resin decontamination process has been demonstrated on a bench scale with a batch of simulated spent ion exchange resin and using potassium cobalt ferrocyanide as the Cs selective sorbent. A volume reduction factor between 10 and 17 has been estimated. The process appears to be economically attractive. Improvements in process economics can be expected from optimization of the process. Other possible applications of the EDRD process have been identified

  8. Incineration of spent ion exchange resin

    International Nuclear Information System (INIS)

    Hasegawa, Chiaki

    1990-01-01

    It is a pressing need to reduce radioactive waste which is generated from the maintenance and operation of a nuclear power plant. Incineration of low level combustible solid waste such as polyethylene seats, paper and others have been successfully performed since 1984 at the Shimane Nuclear Power Station. Furthermore, for extending incineration treatment to spent ion exchange resin, the incineration test was carried out in 1989. However, as the cation exchange resin contains sulfur and then incineration generates SOx gases, so the components of this facility will be in a corrosive environment. We surveyed incineration conditions to improve the corrosive environment at the exhaust gas treatment system. This paper includes these test results and improved method to incinerate spent ion exchange resin. (author)

  9. Radioactivity of spent TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P. [Reactor Department, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  10. Radioactivity of spent TRIGA fuel

    International Nuclear Information System (INIS)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-01-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive

  11. Method of processing spent ion exchange resins

    International Nuclear Information System (INIS)

    Mori, Kazuhide; Tamada, Shin; Kikuchi, Makoto; Matsuda, Masami; Aoyama, Yoshiyuki.

    1985-01-01

    Purpose: To decrease the amount of radioactive spent ion exchange resins generated from nuclear power plants, etc and process them into stable inorganic compounds through heat decomposition. Method: Spent ion exchange resins are heat-decomposed in an inert atmosphere to selectively decompose only ion exchange groups in the preceeding step while high molecular skeltons are completely heat-decomposed in an oxidizing atmosphere in the succeeding step. In this way, gaseous sulfur oxides and nitrogen oxides are generated in the preceeding step, while gaseous carbon dioxide and hydrogen requiring no discharge gas procession are generated in the succeeding step. Accordingly, the amount of discharged gases requiring procession can significantly be reduced, as well as the residues can be converted into stable inorganic compounds. Further, if transition metals are ionically adsorbed as the catalyst to the ion exchange resins, the ion exchange groups are decomposed at 130 - 300 0 C, while the high molecular skeltons are thermally decomposed at 240 - 300 0 C. Thus, the temperature for the heat decomposition can be lowered to prevent the degradation of the reactor materials. (Kawakami, Y.)

  12. Concretes characterization for spent radioactive sources

    International Nuclear Information System (INIS)

    Martinez B, J.; Monroy G, F. P.

    2013-10-01

    The present work includes the preparation and characterization of the concrete used as conditioning matrix of spent radioactive sources in the Treatment Plant of Radioactive Wastes of the Instituto Nacional de Investigaciones Nucleares (ININ). The concrete tests tubes were subjected to resistance assays to the compression, leaching, resistance to the radiation and porosity, and later on characterized by means of X rays diffraction, scanning electron microscopy and infrared spectrometry, with the purpose of evaluating if this concrete accredits the established tests by the NOM-019-Nucl-1995. The results show that the concrete use in the Treatment Plant fulfills the requirements established by the NOM-019-Nucl-1995. (author)

  13. Trapping radioactive ions

    CERN Document Server

    Kluge, Heinz-Jürgen

    2004-01-01

    Trapping devices for atomic and nuclear physics experiments with radioactive ions are becoming more and more important at accelerator facilities. While about ten years ago only one online Penning trap experiment existed, namely ISOLTRAP at ISOLDE/CERN, meanwhile almost every radioactive beam facility has installed or plans an ion trap setup. This article gives an overview on ion traps in the operation, construction or planing phase which will be used for fundamental studies with short-lived radioactive nuclides such as mass spectrometry, laser spectroscopy and nuclear decay spectroscopy. In addition, this article summarizes the use of gas cells and radiofrequency quadrupole (Paul) traps at different facilities as a versatile tool for ion beam manipulation like retardation, cooling, bunching, and cleaning.

  14. Trapping radioactive ions

    International Nuclear Information System (INIS)

    Kluge, H.-J.; Blaum, K.

    2004-01-01

    Trapping devices for atomic and nuclear physics experiments with radioactive ions are becoming more and more important at accelerator facilities. While about ten years ago only one online Penning trap experiment existed, namely ISOLTRAP at ISOLDE/CERN, meanwhile almost every radioactive beam facility has installed or plans an ion trap setup. This article gives an overview on ion traps in the operation, construction or planing phase which will be used for fundamental studies with short-lived radioactive nuclides such as mass spectrometry, laser spectroscopy and nuclear decay spectroscopy. In addition, this article summarizes the use of gas cells and radiofrequency quadrupole (Paul) traps at different facilities as a versatile tool for ion beam manipulation like retardation, cooling, bunching, and cleaning

  15. Management of spent sealed radioactive sources

    International Nuclear Information System (INIS)

    Vicente, Roberto; Sordei, Gian-Maria; Hiromoto, Goro

    2002-01-01

    The number of sealed radiation sources used in industrial, medical, and research applications in Brazil amounts to hundreds of thousands. Spent or disused sources are being collected and stored as radioactive waste in nuclear research centers, awaiting for a decision on their final disposal. However, a safe and economically feasible disposal technology is unavailable. The aim of this paper is to report the development of the concept of a repository and a treatment process that will allow the final disposal of all the spent sealed sources in a safe, dedicated, and exclusive repository. The concept of the disposal system is a deep borehole in stable geologic media, meeting the radiological performance standards and safety requirements set by international organizations. (author)

  16. Trends for minimization of radioactive waste arising from spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Polyakov, A.S.; Koltunov, V.S.; Marchenko, V.I.; Ilozhev, A.P.; Mukhin, I.V.

    2000-01-01

    Research and development of technologies for radioactive waste (RAW) minimization arising from spent nuclear fuel reprocessing are discussed. Novel reductants of Pu and Np ions, reagents of purification recycled extractant, possibility of the electrochemical methods are studied. The partitioning of high activity level waste are considered. Examples of microbiological methods decomposition of radioactive waste presented. (authors)

  17. Acceleration of radioactive ions

    International Nuclear Information System (INIS)

    Laxdal, R.E.

    2003-01-01

    There is an intense interest world-wide in the use of radioactive ion beams (RIBs) for experiment. In many existing or proposed facilities ions are produced or collected at source potential, ionized and re-accelerated. Within the past year three new ISOL based facilities have added dedicated post-accelerators to deliver accelerated RIBs to experiment. The paper gives an overview of RIB accelerators present and future, and explores the inherent features in the various acceleration methods with an emphasis on heavy ion linacs. The ISAC-I and ISAC-II post-accelerators are discussed as examples. Commissioning results and initial operating experience with ISAC-I will be presented

  18. Introduction on the recycling of spent and disused radioactive sources

    International Nuclear Information System (INIS)

    Zhao Mingqiang; Zang Ruihua

    2011-01-01

    It is not only a stress of environment safety, but also a waste of huge resources to send directly to store spent and disused radioactive sources. This article reviews some important aspects of management suggestions recommended by IAEA and requirements of regulations in China for disposing the spent and disused radioactive sources. The present condition and benefit of recycling spent and disused sources are analyzed. Some suggestions on carrying out recycling in China are put forward too. (authors)

  19. Pyrolysis of Spent Ion Exchange Resins

    International Nuclear Information System (INIS)

    Braehler, Georg; Slametschka, Rainer

    2012-09-01

    Ion exchangers (IEX in international language) are used to remove radionuclides from the primary coolant in all nuclear power stations with a water cooling circuit. This is done by continuously removing a volume of coolant from the primary circuit and passing it through coolers, filters and the ion exchange beds. Cation and anion exchangers, in the form of coarse-grained resin beads in pressurized-water reactors and as finely ground powdered resins in boiling water reactors, are used. The trend for new power stations is to exploit all the possibilities for avoiding the generation of contaminated liquids and then to clean, as far as possible, the solutions that are nevertheless generated using ion exchange for it to be possible to dispose of them as non-radioactive waste. This relieves the burden on evaporator facilities, or means that these can even be dispensed with entirely. Regeneration is possible in principle, but little use is made of it. As the regeneration usual in conventional technologies is not employed in nuclear power stations, it is necessary to dispose of this material as radioactive waste. On the international level, a great number of processes are offered that are intended to meet the relevant national regulations, and these will be discussed in brief with their advantages and disadvantages. The aim is then to find a process which reduces the volume, yields an inert or mineralized product, works at temperatures of no more than approximately 600 deg. C and can be run in a simple facility. Originally, the pyrolysis process was developed to treat liquid organic waste from reprocessing. A typical application is the decomposition of spent solvent (TBP, tributyl phosphate, mixed with kerosene). In this process TBP is pyrolyzed together with calcium hydroxide in a fluidized bed facility at temperatures of around 500 deg. C, the calcium hydroxide reacts with the phosphate groups directly to form calcium pyrophosphate which contains all the radioactivity

  20. Future spent nuclear fuel and radioactive waste infrastructure in Norway

    International Nuclear Information System (INIS)

    Soerlie, A.A.

    2002-01-01

    In Norway a Governmental Committee was appointed in 1991 to make an evaluation of the future steps that need to be taken in Norway to find a final solution for the spent nuclear fuel and for some other radioactive waste for which a disposal option does not exist today. The report from the Committee is now undergoing a formal hearing process. Based on the Committees recommendation and comments during the hearing the responsible Ministry will take a decision on future infrastructure in Norway for the spent nuclear fuel. This will be decisive for the future management of spent nuclear fuel and radioactive waste in Norway. (author)

  1. Report on the disposal of radioactive wastes and spent fuel elements from Baden-Wuerttemberg

    International Nuclear Information System (INIS)

    2017-04-01

    The report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg covers the following issues: legal framework for the nuclear disposal; producer of spent fuels and radioactive wastes in Baden- Report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg; low- and medium-level radioactive wastes (non heat generating radioactive wastes); spent fuels and radioactive wastes from waste processing (heat generating radioactive wastes); final disposal.

  2. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

    International Nuclear Information System (INIS)

    Huang, Chun-Ping; Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin

    2012-01-01

    Highlights: ► Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. ► Identify the properties of radioactive contaminants and performance test for water treatment materials. ► The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. ► The radioactive ions were major composed by uranium and fission products. ► Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m 3 of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as 137 Cs, 90 Sr, U, and α-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 μm filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb β-emitters, α-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously eliminate particles and adsorb ionic radionuclides from water.

  3. Spent fuel and high-level radioactive waste storage

    International Nuclear Information System (INIS)

    Trigerman, S.

    1988-06-01

    The subject of spent fuel and high-level radioactive waste storage, is bibliographically reviewed. The review shows that in the majority of the countries, spent fuels and high-level radioactive wastes are planned to be stored for tens of years. Sites for final disposal of high-level radioactive wastes have not yet been found. A first final disposal facility is expected to come into operation in the United States of America by the year 2010. Other final disposal facilities are expected to come into operation in Germany, Sweden, Switzerland and Japan by the year 2020. Meanwhile , stress is placed upon the 'dry storage' method which is carried out successfully in a number of countries (Britain and France). In the United States of America spent fuels are stored in water pools while the 'dry storage' method is still being investigated. (Author)

  4. Thermochemical treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Ojovan, M.I.; Petrov, G.A.; Dmitriev, S.A.; Trusov, B.G.; Semenov, K.N.; Klimov, V.L.

    2001-01-01

    Spent ion exchange resins (IER) is a principal type of radioactive waste constantly generated by nuclear plants of various functions. The reduction of volume of this waste and its treatment to the forms suitable for long-term disposal is an urgent problem facing the present-day atomic energetics. Nowadays the technological process THOR (Studsvik, Sweden) based on the thermodestruction of IER is the best developed and realized on the industrial scale. Unfortunately, this process requires expensive equipment and great energy consumption for the moisture to be evaporated and thereafter IER to be destroyed by heat. Meanwhile the capability of some elements (Mg, Al, Si, Ti etc.) has long been known and practical use found for active interaction with water in combustion regime. This property of the metals has been used in the development of new technology of treatment of IERs in SIA ''Radon''. Wet IER is mixed with powder metal fuel (PMF) which represents a mixture of metal powder, a quantity of burning activator and some technological additives. On initiation, the mixture of IER with PMF burns without extra energy supply to generate enough heat for the moisture to be evaporated and products of IER decomposition to be destroyed and evaporated. To burn out the products of IER evaporation the air is used. The thermodynamic simulation data and the results of experiments using a pilot plant show that radionuclides contained in IER are chemically bound in ash residue consisting of metal oxides, spinel, silicates, etc. According to the experimental data, radionuclides in amounts of 90% or more of Cs-137 and up to 95% of Sr-90 and Co-60 are fixed in the ash residue. The residue volume is several times less than the initial volume of IER. Concentrations of hazard gases in off-gases do not exceed maximum permissible ones accepted in different countries. The technological process is easy to perform, it does not require sophisticated equipment and great energy consumption which

  5. The study on radioactivity reduction of spent PWR cladding hull

    International Nuclear Information System (INIS)

    Jung, I. H.; Kim, J. H.; Park, C. J.; Jung, Y. H.; Song, K. C.; Lee, J. W.; Park, J. J.; Yang, M. S.

    2003-01-01

    Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull-from PWR spent fuel of 32,000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4-7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radioactive waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull

  6. Immobilization of radioactive waste sludge from spent fuel storage pool

    International Nuclear Information System (INIS)

    Pavlovic, R.; Plecas, I.

    1998-01-01

    In the last forty years, in FR Yugoslavia, as result of the research reactors' operation and radionuclides application in medicine, industry and agriculture, radioactive waste materials of the different categories and various levels of specific activities were generated. As a temporary solution, these radioactive waste materials are stored in the two hanger type interim storages for solid waste and some type of liquid waste packed in plastic barrels, and one of three stainless steal underground containers for other types of liquid waste. Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some aspects of immobilisation, conditioning and storage of this sludge are presented in this paper. (author

  7. Management of spent sealed radioactive sources in the European Union

    International Nuclear Information System (INIS)

    Cecille, L.; Taylor, D.

    2000-01-01

    For several years, the European Commission (EC) has been active in the field of spent sealed radioactive sources (SSRS) to improve management schemes and to prepare Euratom Directives that will impact on national legislation and regulatory schemes in European Member States (MS). The main safety issues related to the management of SSRS are described and recommendations made are presented. Additional projects are outlined. (author)

  8. Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1984-09-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy (DOE) radioactive wastes were compiled through December 31, 1983, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 37 years and characteristics of these materials are also presented, consistent with the latest DOE/Energy Information Administration (EIA) or projection of US commercial nuclear power growth and expected defense-related and private industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, airborne waste, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated, based on reported or calculated isotopic compositions. 48 figures, 107 tables

  9. Radioactive waste management decommissioning spent fuel storage. V. 3. Waste transport, handling and disposal spent fuel storage

    International Nuclear Information System (INIS)

    1985-01-01

    As part of the book entitled Radioactive waste management decommissioning spent fuel storage, vol. 3 dealts with waste transport, handling and disposal, spent fuel storage. Twelve articles are presented concerning the industrial aspects of nuclear waste management in France [fr

  10. Radioactive wastes and spent fuels management in Argentina

    International Nuclear Information System (INIS)

    Maset, Elvira R.

    2006-01-01

    CNEA was created in 1950 and since then has carried out research and development activities, production of radioisotopes, medical and industrial applications, and those activities related with the nuclear fuel cycle, including the operation of two nuclear power stations. More ever, different public and private institutions use radioactive materials in medical, industrial and research activities. These activities generate different types of radioactive waste, desuse sealed sources and spent fuel. The management of radioactive waste of all types produced in the country, as the spent nuclear fuel of power and research reactors and the used radioactive sources was always and it is at present a CNEA's responsibility. In February 2003, according to the Law No. 25.018, called 'Management of Radioactive Waste Regimen', the 'Radioactive Waste Management National Programme' was created by CNEA to fulfill the institutional functions and responsibilities established in the Law, in order to guarantee the safe management of radioactive waste according to the regulations established by the Argentine Nuclear Regulatory Agency and to the legislation in force. (author) [es

  11. Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1983-09-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled through December 31, 1982, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 40 years and characteristics of these materials are also presented, based on the latest DOE/EIA projection of US commercial nuclear power growth and expected defense-related and industrial and institutional activities. Materials considered, on a chapter-by-chapter bases, are: spent fuel, high-level waste, transuranic waste, low-level waste, active uranium mill tailings, airborne waste, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated, based on reported or calculated isotopic compositions. One chapter gives broad, summary data on the costs of spent fuel and radioactive waste management and disposal to provide an economic perspective. This chapter is not intended as a definitive guide, but it is a source of reasonable, order-of-magnitude costs and also provides references to more-detailed and scenario-specific studies. An appendix on generic flowsheets and source terms used for the projections is also included

  12. Radioactive ion exchanger fixation

    International Nuclear Information System (INIS)

    Pokonova, Yu.P.; Ivshina, O.A.; Il'ina, O.V.

    1993-01-01

    Properties of some binding agents for fixing radioactive cationites and anionites, namels cement, bitumen, carbamide and polyether resins are analyzed. It is shown that localization of ionites in carbamide resin is not very effective, the same is true of cementing process owing to considerable washing out of cesium-137 (∼ 1.1 x 10 -1 cm) and low water resistance (the samples are destructed when, storage conditions vary). Products of ionite localization in polyether feature a lower washing out (3x10 -2 - 4x10 -2 cm) and a better water resistance (water absorption rate is approximattely 1.5 x 10 -4 cm/day)/ Polyethers despite their high cost, are preferable for processing and transportation of small amounts of the ionites (up to 100 m 3 /year)

  13. Treatment of spent ion-exchange resins

    International Nuclear Information System (INIS)

    Ghattas, N.K.; Ikladious, N.E.; Eskander, S.B.

    1981-01-01

    PMMA was studied with the aim to evaluate its usefulness as an incorporation medium for the final containment of spent ion-exchange resins. The study of the effect of water content (ranging from 25 to 100%) of the incorporated resin into PMMA on the compression strength of the final solid products shows that with the increasing water content the compression strength of the final products decreases sharply. Hardness of the final products follows nearly the same trend of compression strength. Increasing gamma irradiation doses, up to 7.77x10 7 rad, PMMA shows increase in compression strength and hardness for small doses and then decreases with increasing irradiation dose due to the increase in polymerization process and the degradation of the incorporation medium

  14. Spent fuel and radioactive-waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1982-10-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled, based on the most reliable information available from government sources and the open literature, technical reports, and direct contacts. Future waste and spent fuel to be generated over the next 40 years, and characteristics of these materials are also presented, based on a present DOE/EIA projection of US commercial nuclear power growth and expected defense-related and industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, remedial action waste, active uranium mill tailings, airborne waste, and decommissioning. For each category, current and projected inventories are given through the year 2020. The land usage requirements are given for storage/disposal of low-level and transuranic wastes, and for the present inventories of inactive uranium mill tailings. For each waste category the radioactivity and thermal power are calculated. Isotopic compositions and cost data are given for each waste type and for spent fuel

  15. Ontario Hydro Research Division's program for treatment of spent ion-exchange resins

    International Nuclear Information System (INIS)

    Nott, B.R.; Dodd, D.J.R.

    1981-09-01

    A brief review of the evolution of work programmes for chemical treatment of spent ion-exchange resins in Ontario Hydro's Research Division is presented. Attention has been focussed on pre-treatment processes for the treatment of the spent resins prior to encapsulation of the products in solid matrices. Spent Resin Regeneration and Acid Stripping processes were considered in some detail. Particular attention was paid to carbon-14 on spent resins, its determination in and removal from the spent resins (with the acid stripping technique). The use of separate cation and anion resin beds instead of mixed bed resins was examined with a view to reducing the volume of resin usage and consequently the volume of waste radioactive ion-exchange resin generated. (author)

  16. Solidification of radioactive liquid wastes. A comparison of treatment options for spent resins and concentrates

    International Nuclear Information System (INIS)

    Roth, A.; Willmann, F.; Ebata, M.; Wendt, S.

    2008-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence of a final repository site, the built-up of additional volume has to be considered as very critical. Moreover, corrosive effects on cemented drums during long-term interim storage at the surface have raised doubts about the long-term stability of such waste products. In order to avoid such disadvantages solidification methods have been improved in order to get a well-defined product with a better load factor of wastes in the matrix. In a complete different approach, other technologies solidify the liquid radioactive wastes without adding of any inactive material by means of drying

  17. Pyrolysis of spent ion-exchanger resins

    International Nuclear Information System (INIS)

    Slametschka, Rainer; Braehler, Georg

    2012-01-01

    Initial tests have shown that ion exchangers (IEX) can be decomposed by pyrolysis with very good results, yielding an inert and chemically resistant product. No additives are necessary. The main constituent of the product, the pyrolysis residues or ash, is carbon. It has been discovered that the entire radioactive inventory remains in the pyrolysis residues during pyrolysis of the IEX. This is achieved by relatively low process temperatures that prevent highly volatile nuclides such as the caesium nuclides from passing into the gaseous phase. Sintered metal filters in pyrolysis plant ensure that even the radioactivity bonded to the dust remains in the pyrolysis residues. In addition to the radionuclides, the main constituents of the residue are carbon from the original polystyrene matrix and sulphur from the functional groups. The pyrolysis residues form a flowable solid material and not a melt. It is thus easy to handle and can be compacted or cemented, depending on the requirements for interim and permanent storage. Any further constituents such as inorganic filter materials or even other organic materials do not interfere with the process, they are dried, calcined or also pyrolysed. (orig.)

  18. Method of solidifying radioactive ion exchange resin

    International Nuclear Information System (INIS)

    Minami, Yuji; Tomita, Toshihide

    1989-01-01

    Spent anion exchange resin formed in nuclear power plants, etc. generally catch only a portion of anions in view of the ion exchange resins capacity and most of the anions are sent while possessing activities to radioactive waste processing systems. Then, the anion exchange resins increase the specific gravity by the capture of the anions. Accordingly, anions are caused to be captured on the anion exchange resin wastes such that the specific gravity of the anion exchange resin wastes is greater than that of the thermosetting resins to be mixed. This enables satisfactory mixing with the thermosetting resins and, in addition, enables to form integral solidification products in which anion exchange resins and cation exchange resins are not locallized separately and which are homogenous and free from cracks. (T.M.)

  19. Decontamination of spent ion-exchangers contaminated with cesium radionuclides using resorcinol-formaldehyde resins.

    Science.gov (United States)

    Palamarchuk, Marina; Egorin, Andrey; Tokar, Eduard; Tutov, Mikhail; Marinin, Dmitry; Avramenko, Valentin

    2017-01-05

    The origin of the emergence of radioactive contamination not removable in the process of acid-base regeneration of ion-exchange resins used in treatment of technological media and liquid radioactive waste streams has been determined. It has been shown that a majority of cesium radionuclides not removable by regeneration are bound to inorganic deposits on the surface and inside the ion-exchange resin beads. The nature of the above inorganic inclusions has been investigated by means of the methods of electron microscopy, IR spectrometry and X-ray diffraction. The method of decontamination of spent ion-exchange resins and zeolites contaminated with cesium radionuclides employing selective resorcinol-formaldehyde resins has been suggested. Good prospects of such an approach in deep decontamination of spent ion exchangers have been demonstrated. Copyright © 2016 Elsevier B.V. All rights reserved.

  20. 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)

  1. Porous structure analysis of radioactive spent resin cementation matrix

    International Nuclear Information System (INIS)

    Zhou Yaozhong; Yun Guichun

    2004-01-01

    According to a cement product microstructure, a radioactive spent resin cementation matrix has the properties of porous matters. The distributing of the pore size and the pore microstructure stability are closely related to many crucial macro properties, including strength and permeability of the matrixes. By using a new computer-controlled Hg pressure test, a experiment methods of the matrix micro-properties was studied. By using porous structure analyses, it was found that the experimental method is useful for the future cementation research. In this test, it was also found that ASC cement matrixes of spent resin have superior microstructure to the OPC's. They have better pore size distribution, more stable structure and higher ability to hold the Hg in the matrixes than OPC's, and these properties are the important factors that make ASC cement matrixes have more stable macro-structure and lower leaching of nuclides. (authors)

  2. Regulations for the safe management of radioactive wastes and spent nuclear fuel

    International Nuclear Information System (INIS)

    Voica, Anca

    2007-01-01

    The paper presents the national, international and European regulations regarding radioactive waste management. ANDRAD is the national authority charged with nation wide coordination of safe management of spent fuel and radioactive waste including their final disposal. ANDRAD's main objectives are the following: - establishing the National Strategy concerning the safety management of radioactive waste and spent nuclear fuel; - establishing the national repositories for the final disposal of the spent nuclear fuel and radioactive waste; - developing the technical procedures and establishing norms for all stages of management of spent nuclear fuel and radioactive waste, including the disposal and the decommissioning of the nuclear and radiologic facilities

  3. Degradation of ion spent resin using the Fenton's reagent

    International Nuclear Information System (INIS)

    Araujo, Leandro Goulart de

    2013-01-01

    The most common method for spent radioactive ion exchange resin treatment is its immobilization in cement, which reduces the radionuclides release into the environment. Although this method is efficient, it increases considerably the final volume of the waste due to the low incorporation capacity. The objective of this work was to develop a degradation method of spent resins arising from the nuclear research reactor located at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), using an Advanced Oxidation Process (AOP) with Fenton's reagents. This method would allow a higher incorporation in cement. Three different resins were evaluated: cationic, anionic and a mixture of both resins. The reactions were conducted varying the catalyst concentration (25, 50, 100 and 150 mM), the volume of hydrogen peroxide (320 to 460 mL), and three different temperatures, 50, 60 and 70 deg C. Degradation of about 98% was achieved using a 50 mM catalyst solution and 330 mL of hydrogen peroxide solution. The most efficient temperature was 60 deg C. (author)

  4. Quality control of concretes for conditioning of spent radioactive sources

    International Nuclear Information System (INIS)

    Gonzalez N, M.

    2015-01-01

    The spent sealed radioactive sources are considered as a specific type of radioactive wastes and should be properly stored to ensure their integrity and prevent or limit the release of radionuclides in the geosphere. For this, these sources can be put up in concrete matrices. This research presents the evaluation and characterization of five concretes prepared with 4 brands of commercial cements: CPC Extra RS, CPC 30R Impercem of Cemex, Cruz Azul CPC 30R and CPC 30R of Apasco; three sizes of coarse aggregate (<30 mm, 29-11 mm and <10 mm) and fine aggregate (0.0797 mm) used as matrices for conditioning of spent sealed radioactive sources, in order to verify if these specific concretes accredit the standard NOM-019-Nucl-1995. After hardening for 28 days the concrete specimens were subjected to the tests: compressive strength; thermal cycles, irradiation, leaching and permeability, later to be characterized by: 1) X-ray diffraction in order to meet their crystalline phases; 2) scanning electron microscopy, to determine changes in morphology; 3) infrared spectroscopy, to determine the structural changes of concrete from its functional groups; 4) Raman spectroscopy to determine their structural changes and 5) Moessbauer spectroscopy, which determines changes in the oxidation state of iron in the concrete. According to the results and the changes presented by each concrete after applying the tests set by NOM-019-Nucl-1995, is concluded that the concrete made with cement Cemex brand (CPC 30-RS Extra), gravel of particle size 11-29 mm and sieved sand (0.0797 mm) can be used as matrices of spent sealed sources conditioning. Is remarkable a morphological and structural change of the concrete due to gamma irradiation and heat treatment. (Author)

  5. Implications of the Croatian Spent Fuel and Radioactive Waste Strategy

    International Nuclear Information System (INIS)

    Lokner, V.; Levanat, I.

    2012-01-01

    Croatian Government approved its national Radioactive waste and Spent Fuel Strategy as a part of the accession process to EU in July 2009 enabling acquisition of adequate administrative capacity by the time of accession to properly implement and enforce the relevant legislation in all areas related to nuclear safety. Strategy was formulated in line with the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. In particular, the strategy was approved to ensure that sufficient qualified staff and adequate financial services are available to support the safety of facilities for spent fuel and radioactive waste management generated by Krsko NPP during their operating lifetime and from decommissioning. Following are statements from the Croatian strategy that are relevant for Croatian position regarding capacity building for storage and disposal of LILW and SF. To be able to fulfill the obligations assumed, Croatia needs to do the following (Strategy Section 3.3): reach an agreement with Slovenia by 2013 at the latest regarding the location of objects for storing LILW; should no such agreement be reached, Croatia is to initiate preparations for assuming its half of operational LILW and for third-country export thereof, or for storing the LILW on Croatian territory, whilst also gradually assuming the part of LILW created from decommissioning; reach an agreement with Slovenia by 2018 at the latest regarding the location of a common SF storage; should no such agreement be reached, initiate preparations for assuming a third-country export of SF, or for storing half of SF on Croatian territory. This paper discusses the Strategy aims in the light of noticeable delay of 2nd revision of the Program of NPP Krsko Decommissioning and SF andLILW Disposal, the status of the planned Slovenian national repository on Vrbina site and the prospects of its use for joint Croatian/Slovenian LILW disposal - all in the context of as yet

  6. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Chun-Ping, E-mail: chunping@iner.gov.tw [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China); Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China)

    2012-09-30

    Highlights: Black-Right-Pointing-Pointer Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. Black-Right-Pointing-Pointer Identify the properties of radioactive contaminants and performance test for water treatment materials. Black-Right-Pointing-Pointer The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. Black-Right-Pointing-Pointer The radioactive ions were major composed by uranium and fission products. Black-Right-Pointing-Pointer Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m{sup 3} of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as {sup 137}Cs, {sup 90}Sr, U, and {alpha}-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 {mu}m filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb {beta}-emitters, {alpha}-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously

  7. Spent Fuel and High-Level Radioactive Waste Transportation Report

    International Nuclear Information System (INIS)

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ''comprehensive overview of the issues.'' This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list

  8. Spent fuel and high-level radioactive waste transportation report

    Energy Technology Data Exchange (ETDEWEB)

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  9. Spent fuel and high-level radioactive waste transportation report

    International Nuclear Information System (INIS)

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ''comprehensive overview of the issues.'' This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list

  10. Spent fuel and high-level radioactive waste transportation report

    International Nuclear Information System (INIS)

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ''comprehensive overview of the issues.'' This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list

  11. Management of spent high activity radioactive sources (SHARS)

    International Nuclear Information System (INIS)

    2002-09-01

    The objective of this report is to provide all people involved in the handling and management of high activity sources with sufficient information about processes that are required for the safe management of spent high activity radioactive sources (SHARS). This includes examples of spent source management that are already taking place and also a description of the range of appropriate options that are available for each stage in the management process. This report also aims to identify the important issues to be addressed in order to develop a waste management strategy as part of the integrated management strategy that takes account of international experience and the guidance and principles that have been learned from that experience. This report relates specifically to SHARS, which are spent sources that have the potential, with short exposures, to produce acute health effects if handled incorrectly. In addition, they may also incur significant economic costs in any retrieval or environmental remediation operation, following loss of or damage to such a source. The report provides guidance on the technical, administrative and economic issues associated with SHARS from the moment they cease to be in use through to disposal, including temporary storage, transport, conditioning and interim storage

  12. Method of burning ion-exchange resin contaminated with radioactivity

    International Nuclear Information System (INIS)

    Suzuki, Shigenori.

    1986-01-01

    Purpose: To process spent ion exchange resins to reduce their volume, without increasing the load on a off-gas system and in a stable state and at the same time not leaving any uncombusted portions. Method: The water slurries of the ion exchange resins contaminated with radioactive materials is dehydrated or dry combusted to reduce the water content. A binder is then added to solidify the ion exchange resin. The solidified ion exchange resins are then combusted in a furnace. This prevents the ion exchange resin from being dispersed by air and combustion gases. Furthermore, the solidified ion exchange resins in the form of small pellets burn from the surface inwards. Moreover the binder is carbonized by the combustion heat and promotes combustion to convert the ion exchange resins into a solid mass, making sure that no uncombusted portion is left. (Takahashi, M.)

  13. Biodegradation of radioactive organic liquid waste from spent fuel reprocessing

    International Nuclear Information System (INIS)

    Ferreira, Rafael Vicente de Padua

    2008-01-01

    The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab-scale hot cell, known as Celeste located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90's, and the laboratory was closed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: organic compounds characterization in the radioactive liquid waste; the occurrence of bacterial consortia from Pocos de Caldas uranium mine soil and Sao Sebastiao estuary sediments that are able to degrade organic compounds; and the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming the cementation. From the characterization analysis, TBP and ethyl acetate were chosen to be degraded. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradation level were 66% for ethyl acetate and 70% for the TBP. (author)

  14. Solidification of radioactive liquid wastes, Treatment options for spent resins and concentrates - 16405

    International Nuclear Information System (INIS)

    Roth, Andreas

    2009-01-01

    Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. Spent resin and concentrate treatment often appear as a specific task in decommissioning projects, because in the past those waste streams typically had been stored in tanks for the lifetime of the plant and needs to be retrieved, conditioned and packed prior to dismantling activities. Additionally a large amount of contaminated liquids will be generated by utilizing decontamination processes and needs to be processed further on. Such treatment options need to achieve waste products acceptable for final disposal, because due to the closure of the site no interim storage can be envisaged. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence

  15. Key convention on safe management of spent fuel and radioactive waste to enter into force

    International Nuclear Information System (INIS)

    2001-01-01

    At a ceremony at IAEA Headquarters today, Ireland deposited its instrument of ratification to an important convention on the safe management of spent fuel and radioactive waste, thereby ensuring its entry into force. The Convention will be the first international instrument to address the safety of management and storage of radioactive wastes and spent fuels in countries with and without nuclear programmes

  16. On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

    Energy Technology Data Exchange (ETDEWEB)

    Vorona, N. A.; Gavrikov, A. V., E-mail: gavrikov@ihed.ras.ru; Samokhin, A. A.; Smirnov, V. P. [Russian Academy of Sciences, Joint Institute for High Temperatures (Russian Federation); Khomyakov, Yu. S. [Innovation–Technology Center of the Project Proryv (Russian Federation)

    2015-12-15

    The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

  17. National Centre for Radioactive Ion Beams (NCRIB)

    International Nuclear Information System (INIS)

    Chintalapudi, S.N.

    1999-01-01

    A dedicated National Centre for RIB (NCRIB) proposed discussed at several forums is presented. The production of (RIB) radioactive ion beams and applications of beams leading to competitive studies in nuclear structure, nuclear reactions, condensed matter, bio-science and radioactive isotope production etc. are mentioned

  18. Program for responsible and safe disposal of spent fuel elements and radioactive wastes (National disposal program)

    International Nuclear Information System (INIS)

    2015-01-01

    The contribution covers the following topics: fundamentals of the disposal policy; amount of radioactive wastes and prognosis; disposal of radioactive wastes - spent fuel elements and wastes from waste processing, radioactive wastes with low heat production; legal framework of the nuclear waste disposal in Germany; public participation, cost and financing.

  19. Device for separating ruthenium ion from spent fuel material

    International Nuclear Information System (INIS)

    Izumida, Tatsuo; Sasahira, Akira; Ozawa, Yoshihiro; Kawamura, Fumio.

    1988-01-01

    Purpose: To separate plutonium ions efficiently and selectively from organic solvent containing tributyl phosphate used in the main step of reprocessing process. Constitution: The device comprises, as the main constituent factor, a liquid-liquid contact device for bringing not water soluble organic solvent into contact with a nitric acid solution of spent fuel substances and a liquid-liquid contact-separation device for bringing an organic solvent solution containing spent fuel substances separated with nitric acid into contact again with nitric acid. Then, a device is disposed between two liquid-liquid contact devices for staying ruthenium ions and organic solvent for a sufficient time. In this way, ruthenium ions in the organic solvent containing butyl phosphate are gradually converted into complex compounds combined with tributyl phosphate thereby enabling to separate ruthenium ions efficiently and remarkably reduce the corrosion of equipments. (Horiuchi, T.)

  20. Radioactive ions and atoms in superfluid helium

    NARCIS (Netherlands)

    Dendooven, P.G.; Purushothaman, S.; Gloos, K.; Aysto, J.; Takahashi, N.; Huang, W.; Harissopulos, S; Demetriou, P; Julin, R

    2006-01-01

    We are investigating the use of superfluid helium as a medium to handle and manipulate radioactive ions and atoms. Preliminary results on the extraction of positive ions from superfluid helium at temperatures close to 1 K are described. Increasing the electric field up to 1.2 kV/cm did not improve

  1. Cation immobilization in pyrolyzed simulated spent ion exchange resins

    International Nuclear Information System (INIS)

    Luca, Vittorio; Bianchi, Hugo L.; Manzini, Alberto C.

    2012-01-01

    Significant quantities of spent ion exchange resins that are contaminated by an assortment of radioactive elements are produced by the nuclear industry each year. The baseline technology for the conditioning of these spent resins is encapsulation in ordinary Portland cement which has various shortcomings none the least of which is the relatively low loading of resin in the cement and the poor immobilization of highly mobile elements such as cesium. The present study was conducted with cationic resin samples (Lewatit S100) loaded with Cs + , Sr 2+ , Co 2+ , Ni 2+ in roughly equimolar proportions at levels at or below 30% of the total cation exchange capacity. Low temperature thermal treatment of the resins was conducted in inert (Ar), or reducing (CH 4 ) gas atmospheres, or supercritical ethanol to convert the hydrated polymeric resin beads into carbonaceous materials that contained no water. This pyrolytic treatment resulted in at least a 50% volume reduction to give mechanically robust spherical materials. Scanning electron microscope investigations of cross-sections of the beads combined with energy dispersive analysis showed that initially all elements were uniformly distributed through the resin matrix but that at higher temperatures the distribution of Cs became inhomogeneous. Although Cs was found in the entire cross-section, a significant proportion of the Cs occurred within internal rings while a proportion migrated toward the outer surfaces to form a crustal deposit. Leaching experiments conducted in water at 25 °C showed that the divalent contaminant elements were very difficult to leach from the beads heated in inert atmospheres in the range 200–600 °C. Cumulative fractional loses of the order of 0.001 were observed for these divalent elements for temperatures below 500 °C. Regardless of the processing temperature, the cumulative fractional loss of Cs in water at 25 °C reached a plateau or steady-state within the first 24 h increasing only

  2. SE-VYZ - Decommissioning of Nuclear Installations, Radioactive Waste and Spent Fuel Management

    International Nuclear Information System (INIS)

    Anon

    2004-01-01

    In this presentations processes of radioactive waste treatment in the Bohunice Radioactive Waste Processing Center (SE-VYZ), Jaslovske Bohunice are presented. Decommissioning of the A-1 NPP is also presented. Disposal of conditioned radioactive waste in fibre concrete containers (FCC) are transported to Mochovce from Jaslovske Bohunice by the transport truck where are reposited in the National radioactive waste repository Mochovce. The Interim spent fuel storage facility (ISFSF) is included into this presentation

  3. Cement solidification of spent ion exchange resins produced by the nuclear industry

    International Nuclear Information System (INIS)

    Jaouen, C.; Vigreux, B.

    1988-01-01

    Cement solidification technology has been applied to spent ion exchange resins for many years in countries throughout the world (at reactors, research centers and spent fuel reprocessing plants). Changing specifications for storage of radioactive waste have, however, confronted the operators of such facilities with a number of problems. Problems related both to the cement solidification process (water/cement/resin interactions and chemical interactions) and to its utilization (mixing, process control, variable feed composition, etc.) have often led waste producers to prefer other, polymer-based processes, which are very expensive and virtually incompatible with water. This paper discusses research on cement solidification of ion exchange resins since 1983 and the development of application technologies adapted to nuclear service conditions and stringent finished product quality requirements

  4. Method of pyrolysis for spent ion-exchange resins

    International Nuclear Information System (INIS)

    Aoyama, Yoshiyuki; Matsuda, Masami; Kawamura, Fumio; Yusa, Hideo.

    1985-01-01

    Purpose: To prevent the generation of noxious sulfur oxide and ammonia on the pyrolysis for spent ion-exchange resins discharged from nuclear power plants. Method: In the case where the pyrolysis is made for the cationic exchange resins having sulfonic acids as the ion-exchange group, alkali metals or alkaline earth metals capable of reacting with sulfonic acid groups to form solid sulfates are previously deposited by way of ion-exchange reactions prior to the pyrolysis. In another case of the anionic exchange resins having quarternary ammonium groups as the ion-exchange groups, halogenic elements capable of reacting with the ammonium groups to form solid ammonium salts are deposited to the ion-exchange resins through ion-exchange reactions prior to the pyrolysis. As a result, the amount of the binders used can be reduced, and this method can be used in a relatively simple processing facility. (Horiuchi, T.)

  5. Treatment of spent ion exchange resins IAEA research coordination programme

    International Nuclear Information System (INIS)

    Balu, K.; Bhatia, S.C.; Wattal, P.K.; Chanana, N.

    1981-09-01

    Spent ion-exchange resins arising from steam condensate systems, reactor coolant clean-up systems and rad-waste procession, are considered as a specific solid waste management problem. This is the second report on the product characterisation with respect to thermal properties, flammability, bio-organic degradation and leaching behaviours. All these studies are based on polyester-styrene polymer as a matrix for fixation of these spent Ix-resins. Choice of this matrix was dealt with in the first report. (author)

  6. The concept of radioactive waste and spent nuclear fuel management in the Czech Republic

    International Nuclear Information System (INIS)

    Suransky, F.; Duda, V.

    2003-01-01

    The article briefly comments on the status of nuclear waste and spent nuclear fuel management in the Czech Republic in the context of the document entitled 'The Concept of Radioactive Waste and Spent Nuclear Fuel Management in the Czech Republic', which was adopted by the Czech Government in May 2002 as a national strategy in this field. (author)

  7. Integrated data management system for radioactive waste and spent fuel in Korea

    International Nuclear Information System (INIS)

    Shin, Young Ho

    2001-03-01

    An integrated data management system for the safe management of radioactive waste and spent fuel in Korea is developed to collect basic information, provide the framework for national regulation, and improve national competition and efficiency in the management of radioactive waste and spent fuel. This system can also provide public access to information such as a statistical graphs and integrated data from various waste generators to meet increased public needs and interests. So through the system, the five principles (independence, openness, clearance, efficiency and reliance) of safety regulation can be realized, and public understanding and reliance on the safety of spent fuel and radioactive waste management can be promoted by providing reliable information, it can ensure an openness within the international nuclear community and efficiently support international agreements among contracting parties by operating safe and efficient management of spent fuel and radioactive waste (IAEA joint convention on the safety of spent fuel management and on the safety of radioactive waste management), the system can compensate for the imperfections in safe regulation of radioactive waste and spent fuel management related to waste generation, storage and disposal, and make it possible to holistic control and finally re-organize the basic framework of KINS's intermediate and long term research organization and trends, regarding waste management policy is to integrate safe management and unit safe disposal. For this objectives, benchmark study was performed on similar data base system worldwide and data specification with major input/output data during the first phase of this project

  8. Radioactive heavy ion secondary beams

    International Nuclear Information System (INIS)

    Bimbot, R.

    1987-01-01

    The production of secondary radioactive beams at GANIL using the LISE spectrometer is reviewed. The experimental devices, and secondary beam characteristics are summarized. Production of neutron rich secondary beams was studied for the systems Ar40 + Be at 44 MeV/u, and 018 + Be at 45 and 65 MeV/u. Partial results were also obtained for the system Ne22 + Ta at 45 MeV/u. Experiments using secondary beams are classified into two categories: those which correspond to fast transfer of nuclei from the production target to a well shielded observation point; and those in which the radioactive beam interacts with a secondary target

  9. The PILO process: zeolites and titanates in the treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Hultgren, Aa.; Thegerstroem, C.; Forberg, S.; Westermark, T.; Faelt, L.

    1981-01-01

    Spent ion exchange resins from power reactor operation contain more than 95% of the total radioactivity of wet reactor wastes. Cementation and bituminization are the two methods applied in Sweden up to now for the immobilization of spent resins. Over the last years, however, research and development work has resulted in a proposed process (PILO), where > 99.9 % of cesium and strontium and around 90 % of other radioactive nuclides are eluted from the spent resins and sorbed in zeolites and titanates in a chromatographic process. The inorganic sorbents are dried after loading and sintered to yield long-term stable products, while the treated resins may be incinerated to give ash residues of fairly short-lived activity. The development work has included production, characterization and testing of different zeolites and titanates, bench-scale optimization of the chromatographic process using actual spent resins, heat treatment of the loaded inorganic sorbents, and resin incineration. Over-all system design studies including transport requirements, integrated process flowsheets, and cost estimates are now in progress. The aim is to have a sufficient basis during spring 1982 to decide on the merits of a PILO plant at the planned repository for low and medium level waste (SFR), to be commissioned in 1988. (Auth.)

  10. Radioactive ion beam facilities at INFN LNS

    International Nuclear Information System (INIS)

    Rifuggiato, D; Calabretta, L; Celona, L; Chines, F; Cosentino, L; Cuttone, G; Finocchiaro, P; Pappalardo, A; Re, M; Rovelli, A

    2011-01-01

    Radioactive ion beams are produced at INFN- Laboratori Nazionali del Sud (LNS) by means of the two operating accelerators, the Tandem and the Superconducting Cyclotron (CS), originally designed to accelerate stable beams. Both the ISOL (Isotope Separation On Line) and the IFF (In-Flight Fragmentation) methods are exploited to produce RIBs in two different ways at different energies: in the first case, the Cyclotron is the primary accelerator and the Tandem accelerates the secondary beams, while in the second case radioactive fragments are produced by the Cyclotron beam in a thin target with energies comparable to the primary beam energy. The ISOL facility is named EXCYT (Exotics at the Cyclotron and Tandem) and was commissioned in 2006, when the first radioactive beam ( 8 Li) has been produced. The IFF installation is named FRIBs (in Flight Radioactive Ion Beams), and it has started to produce radioactive beams in 2001, placing a thin target in the extraction beam line of the Cyclotron. The development of both facilities to produce and accelerate radioactive ion beams at LNS, is briefly described, with some details on the future prospects that are presently under consideration or realization.

  11. Treatment of spent ion-exchange resins in shaft-type reactor with fuel-plasma source of heating

    International Nuclear Information System (INIS)

    Dmitriev, S. A.; Knyazev, I. A.; Lifanov, F. A.; Polkanov, M. A.; Shvetsov, S. Yu; Savkin, A. E.

    1999-01-01

    The method of high-temperature conditioning the spent radioactive ion-exchange resins in combination with other combustible and incombustible radioactive waste in the plasma shaft furnace with obtaining a crystalline glass-like matrix as a final product has been developed. The method was tested on the pilot plant consisted of ceramic plasma melter, steel water-cooled shaft furnace, lined by fire-resistant concrete, and system of gas purification. The capacity of the furnace was within the limits from 10 to 15 kg/h. The volume and mass reduction factors of treated waste were 36 and 7.6, accordingly. The content of gaseous products of thermal decomposition of a waste at an output of the shaft furnace and properties of obtained slag compound were determined. Based on test results the proposals on creation of compact plasma plant for treatment of mixed radioactive waste including spent ion exchange resins were developed. (author). 3 refs., 2 tabs., 2 figs

  12. Experimental studies with radioactive ion beams

    International Nuclear Information System (INIS)

    Sastry, D.L.; Sree Krishna Murty, G.; Chandrasekhar Rao, M.V.S.

    1991-01-01

    The sources of information presented are essentially taken from the papers reported at several international seminars and those appeared in the Journal of Nuclear Instruments and Methods in Physics Research. Production and usage of radioactive ion beams (RIB) in research have received the attention of scientists all over the world during the past six years. The first radioactive ion beams ( 19 Ne) were produced at Bevalac for the purpose of medical research using a primary beam of energy 800 MeV/a.m.u. (author). 19 refs., 2 figs., 3 tabs

  13. Recovery Of Electrodic Powder From Spent Lithium Ion Batteries (LIBs

    Directory of Open Access Journals (Sweden)

    Shin S.M.

    2015-06-01

    Full Text Available This study was focused on recycling process newly proposed to recover electrodic powder enriched in cobalt (Co and lithium (Li from spent lithium ion battery. In addition, this new process was designed to prevent explosion of batteries during thermal treatment under inert atmosphere. Spent lithium ion batteries (LIBs were heated over the range of 300°C to 600°C for 2 hours and each component was completely separated inside reactor after experiment. Electrodic powder was successfully recovered from bulk components containing several pieces of metals through sieving operation. The electrodic powder obtained was examined by X-ray diffraction (XRD, energy dispersive X-ray spectroscopy (EDS, and atomic absorption spectroscopy (AA and furthermore image of the powder was taken by scanning electron microscopy (SEM. It was finally found that cobalt and lithium were mainly recovered to about 49 wt.% and 4 wt.% in electrodic powder, respectively.

  14. Incineration of spent ion exchange resins in a triphasic mixture at Belgoprocess

    International Nuclear Information System (INIS)

    Deckers, J.; Luycx, P.

    2003-01-01

    Up to 1998, spent ion exchange resins have been fed to the incinerator in combination with various other solid combustible wastes at Belgoprocess. However, thanks to sustained efforts to reduce radioactive waste production in all nuclear facilities in Belgium, the annual production of solid combustible waste is now much too small to allow this practice to be continued. Since the incinerator at Belgoprocess is not capable of treating spent ion exchange resins as such, it was decided to adopt the use of foam as a carrier to feed the resins to the incinerator. The mixture is a pseudohomogeneous charged foam, ensuring easy handling and allowing incineration in the existing furance, while a number of additives may be included, such as oil to increase the calorific value of the mixture and accelerate combustion. The first incineration campaign of spent ion exchange resins in a triphasic foam mixture, in conjunction with other liquid and solid combustible wastes, will be started in January 2000. The foam, comprising 70% by weight of resins, 29% by weight of water and 1% by weight of surfactant will be pulverized in the incinerator through an injection lance, at a feed rate of 40 to 100 kg/h. The incinerator and associated off-gas treatment system can be operated at standard conditions. Belgoprocess is the subsidiary of the Belgian national agency for the management of radioactive waste, known by its Dutch and French acronyms, NIRAS and ONDRAF respectively. The company ensures the treatment, conditioning and interim storage of nearly all radioactive waste produced in Belgium. (orig.)

  15. A study on the safety of spent fuel management. Radioactive source term modelling

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Kwan Sik; Lee, Hoo Keun; Park, Keun Il; Hwoang, Jung Ki; Chung, Choong Hwan [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1992-02-01

    The types and probabilities of events which may occur during the process of reception, transfer and storage of spent fuels in an away-from-reactor (AFR) spent fuel storage facility were analyzed in order to calculate the amount of radioactive material released to operation area and atmosphere, and the basic model for predicting the radioactive source-term under normal and abnormal operations were developed. Also, oxidation and dissolution of U0{sub 2} pellet was investigated to estimate the amount of radioactive materials released from spent fuel and the release characteristics of radionuclides from defected spent fuel rods was analyzed. Basic information using FIRAC code to analyze the ventilation system during fire accident was prepared and FIRIN was detached from FIRAC modified to simulate the compartment fire by personal computer. (Author).

  16. What are Spent Nuclear Fuel and High-Level Radioactive Waste?

    International Nuclear Information System (INIS)

    2002-01-01

    Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository

  17. Preparation of the National Radioactive Waste and Spent Fuel Management Programme in Slovenia

    International Nuclear Information System (INIS)

    Kralj, M.; Zeleznik, N.; Mele, I.; Veselic, M.

    2006-01-01

    The first separate National Radioactive Waste and Spent Fuel Management Programme (National Programme) was prepared in Slovenia in 2005, as a separate part of the National Environmental Action Programme that was adopted in June 2005. In the previous National Environmental Action Programme from the year 1999, the radioactive waste and spent fuel management was mentioned only briefly in the paragraph on radiation and nuclear safety with two main objectives: to provide an effective management of radioactive waste, and to keep the environmental ionising radiation under control. The new National Programme from 2005 includes all topics being relevant for the management of the radioactive waste and spent fuel, from the legislation and identification of different waste streams in Slovenia, to the management of radioactive waste and spent fuel, the decommissioning of nuclear facilities and management of (TE)NORM. It deals also with the relevant actors in the radioactive waste management, communication and information activities, and the financial aspects of the radioactive waste and spent fuel management. The National Programme was already adopted by the Slovenian Government in October 2005 and will go to Parliament proceedings. The Technical bases for the National Programme was prepared by ARAO and presented to the government in the beginning of 2005. The frames for this document were taken from relevant strategic documents: the Programme of decommissioning the nuclear power plant Krsko and the radioactive waste and spent fuel management, prepared in 2004 by Slovenian and Croatian experts (ARAO and APO), the Proposal of LILW Management Strategy (1999), the Strategy of Spent Fuel Management (1996), and the Resolution on the National Energy Programme (2004). ARAO made a detailed study on the amount and types of radioactive waste produced in Slovenia and future arising with emphasis on the minimization on radioactive waste production. It considered all producers of LILW and

  18. The dispersion of alpha and beta radioactivity to the environmental from spent fuel testing in RMI

    International Nuclear Information System (INIS)

    Yuwono, I.; Pudjadi, E.

    1996-01-01

    The destructive testing of 2 spent fuels in RMI and radioactivity air release monitoring to the environmental have been done. The monitoring equipment used alpha-beta particulate monitor, Berthold LB 150 D type. The calculations using the Gaussian plume model and distributions factor showed there were no radiological effect of alpha and beta radioactivity dispersion and contribution to the environmental. The maximum average construction of alpha and beta radioactivity are 0.002% and 0.05%. (author)

  19. National Centre for Radioactive Ion Beams (NCRIB)

    International Nuclear Information System (INIS)

    Chintalapudi, S.N.

    1999-01-01

    Radioactive Ion (nuclear) Beams have become prolific recently. Nuclear physics and associated subjects have staged a comeback to almost the beginning with the advent of RIB. A dedicated National Centre for RIB (NCRIB) proposed, discussed at several forums and under serious consideration is described

  20. Characterization of primary coolant purification system samples for assay of spent ion exchanger radionuclide inventor

    International Nuclear Information System (INIS)

    Sajin Prasad, S.; Pant, Amar; Sharma, Ranjit; Pal, Sanjit

    2018-01-01

    The primary coolant system water of a research reactor contains various fission and activation products and the water is circulated continuously through ion exchange resin cartridges, to reduce the radioactive ionic impurity present in it. The coolant purification system comprises of an ion exchange cooler, two micro filters, and a battery of six ion exchanger beds, associated valves, piping and instrumentation (Heavy water System Operating manual, 2014). The spent cartridge is finally disposed off as active solid waste which contains predominantly long lived fission and activation products. The heavy water coolant is also used to cool the structural assemblies after passing through primary heat exchanger and a metallic strainer, which accumulates the fission and activation products. When there is a reduction of coolant flow through these strainers, they are removed for cleaning and decontamination. This paper describes the characterization of ion exchange resin samples and liquid effluent generated during ultra sonic decontamination of strainer. The results obtained can be used as a methodology for the assay of the spent ion exchanger cartridges radionuclide inventory, during its disposal

  1. Method of separating radioactive nuclides from ion exchange resins

    International Nuclear Information System (INIS)

    Suzuki, Kazunori; Saikoku, Masami; Taneta, Daisuke; Yagi, Takuro.

    1987-01-01

    Purpose: To enable to safely process radioactive nuclides from spent ion exchange resins by using existent processing facilities. Method: Ion exchange resins in aqueous medium are at first placed to the ultrasonic wave irradiation site and put into such a state where clads and resins are easily separatable from each other by weakening the bonding force between them. Since the clads are magnetic material such as Fe 3 O 4 or NiFe 2 O 4 , the clads can be collected in the direction of the magnetic force by exerting the magnetic field simultaneously. The collected clads are transported by means of the aqueous medium to a collecting tank by removing the effect of magnetic field, for example, by interrupting the current supply to the electromagnet. Finally, they were subjected to stabilization and fixation into inorganic hardening agent such as cement hardener. Thus, processions can be made safely by using existent facilities. (Takahashi, M.)

  2. Experiments with SIRA - the radioactive ion separator

    International Nuclear Information System (INIS)

    Angelique, J.C.; Orr, N.A.

    1998-01-01

    There are two main techniques to obtain radioactive ion beams. One, consisting in the fragmentation of projectile in a thin target followed by a separation carried out with LISE or SISSI type spectrometers or by an alpha spectrometer is used currently at GANIL. The second one, the ISOL (Isotope Separator One-Line) is presently under study on the SIRa benchmark, as part of the SPIRaL (Source de Production d'Ions Radioactifs en Ligne). A high energy light ion beam is stopped by a thick target to produce radioactive nuclei by various reactions in the target. The target, usually of carbon, is heated at around 1800 deg. C in order to accelerate the migration of the atoms produced at the target surface. These atoms are then diffused by a transfer tube up to plasma region where they are ionized and then accelerated. As projectiles the GANIL project makes use of a large variety of heavy ions. A table containing the radioactive ion beam characteristics (charge state and lifetime), the primary beams, the yields and the expected intensities to be obtained with SPIRaL is presented. Also, data concerning the production rates of rare gases obtained during 1993 to 1994 are given

  3. Quantum-CEP processing spent ion exchange resins from nuclear power stations

    International Nuclear Information System (INIS)

    Kaczmarsky, Myron

    1997-01-01

    Quantum-CEP (Q-CEP) is an innovative and proprietary technology developed by Molten Metal Technology, Inc, which can process radioactive and mixed waste streams to decontaminate and recover resources of commercial value while achieving significant volume reduction and radionuclide stabilization. A Q-CEP facility, located in Oak Ridge, Tennessee, processes low-level radioactive spent ion exchange resins (IER) from U.S commercial nuclear power plants. The first campaign processing low level radioactive spent IER was successfully completed in December 1996. Other milestones, since December, include operating parallel Trains A and B simultaneously and processing 25,000 lb. dry resin (50,000 lb. wet resin) or six equivalent High Integrity Containers (HICs) in one batch campaign, in March; and processing 50,000 lb. dry resin or 12 equivalent HICs in one batch campaign in May. This paper presents results from the March campaign (97-008) in which 25,000 lb. of dry spent IER from five nuclear power stations were processed. This campaign has been selected since it is representative of campaigns completed during the first five months of operation. Key highlights for this campaign include processing six HICs in batch campaign while achieving a volume reduction of 24: 1. Key performance targets for the facility are to process an average of six HICs per campaign batch and achieve a volume reduction of 30: 1. The average batch size and other performance parameters have steadily improved during the initial operating period with radioactive resin. The progress was dramatically demonstrated by the May campaign during which 12 HICs were processed - achieving a volume reduction estimated to exceed 50: 1. The campaigns in March and May demonstrate that the facility's design and technology are capable of achieving and even exceeding the facility's key target performance parameters

  4. Radioactive Ions for Surface Characterization

    CERN Multimedia

    2002-01-01

    The collaboration has completed a set of pilot experiments with the aim to develop techniques for using radioactive nuclei in surface physics. The first result was a method for thermal deposition of isolated atoms (Cd, In, Rb) on clean metallic surfaces. \\\\ \\\\ Then the diffusion history of deposited Cd and In atoms on two model surfaces, Mo(110) and Pd(111), was followed through the electric field gradients (efg) acting at the probe nuclei as measured with the Perturbed Angular Correlation technique. For Mo(110) a rather simple history of the adatoms was inferred from the experiments: Atoms initially landing at terrace sites diffuse from there to ledges and then to kinks, defects always present at real surfaces. The next stage is desorption from the surface. For Pd a scenario that goes still further was found. Following the kink stage the adatoms get incorporated into ledges and finally into the top surface layer. For all these five sites the efg's could be measured.\\\\ \\\\ In preparation for a further series o...

  5. BEARS: Radioactive ion beams at LBNL

    International Nuclear Information System (INIS)

    Powell, J.; Guo, F.Q.; Haustein, P.E.

    1998-01-01

    BEARS (Berkeley Experiments with Accelerated Radioactive Species) is an initiative to develop a radioactive ion-beam capability at Lawrence Berkeley National Laboratory. The aim is to produce isotopes at an existing medical cyclotron and to accelerate them at the 88 inch Cyclotron. To overcome the 300-meter physical separation of these two accelerators, a carrier-gas transport system will be used. At the terminus of the capillary, the carrier gas will be separated and the isotopes will be injected into the 88 inch Cyclotron's Electron Cyclotron Resonance (ECR) ion source. The first radioactive beams to be developed will include 20-min 11 C and 70-sec 14 O, produced by (p,n) and (p,α) reactions on low-Z targets. A test program is currently being conducted at the 88 inch Cyclotron to develop the parts of the BEARS system. Preliminary results of these tests lead to projections of initial 11 C beams of up to 2.5 x 10 7 ions/sec and 14 O beams of 3 x 10 5 ions/sec

  6. Microwave process employed to study the immobilization feasibility of spent ion exchange resins in polymeric matrices

    Energy Technology Data Exchange (ETDEWEB)

    Caratin, Reinaldo L.; Araujo, Sumair G. de; Landini, Liliane; Neves, Sabrina C.; Lugao, Ademar B. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)], E-mail: rcaratin@ipen.br, E-mail: sgaraujo@ipen.br, E-mail: llandini@ipen.br, E-mail: scneves@ipen.br, E-mail: ablugao@ipen.br

    2007-07-01

    Nuclear activities generate radioactive wastes in several physical states, radioactive levels and kinds of radioactive emission. Hence, a lot of techniques have been developed and optimized to do the immobilization of these materials, according to local and international regulations to protect human being and environment. Another great concern is the indiscriminate disposal of used polymeric materials (such as plastic and rubber) or production leftovers in landfills, which remain for many years before they are naturally decomposed. In this work, it was studied the possibility of carrying out the immobilization of spent ion exchange resins (contaminated with ionising radiation), by using polymeric matrices of bitumen and rubber (as solidification materials for the storage of low level radioactive waste). The samples were mixed at different percentages and were heated in a microwave device (2,450 MHz) at IPEN/CNEN-SP, varying the irradiation time and power. The objective of the immobilization is converting the wastes into forms that are leach resistant and physically and chemically stable for disposal. Characterizations of these materials have been performed according to ABNT-NBR standards. The results indicated the previous idea of the necessary minimum temperature to keep the matrix for future embedding of radioactive waste, in solid state. (author)

  7. Radioactive wastes. Commune convention about the safety of spent fuel management and about the radioactive waste management

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    This common convention do not give detailed safety standards but general obligations whom objective is the development of a safety culture in the world. It concerns the spent fuels (valuable and valued by the reprocessing) and radioactive wastes (matter without any later use). (N.C.)

  8. Formulation study on immobilization of spent ion exchange resins in polymer cement

    International Nuclear Information System (INIS)

    Xia Lili; Lin Meiqiong; Bao Liangjin; Fan Xianhua

    2006-01-01

    The aim of this study is to develop a formulation of cement-solidified spent radioactive ion exchange resin form. The solidified form consists of a sort of composite cement, epoxide resin emulsion, and spent ion exchange resins. The composite cement is made up of quick-setting sulphoaluminate cement, silica powder, zeolite, and fly ash in the proportion 1:0.05:0.10:0.05. Sixteen combinations of composite cement, epoxide resin emulsion and mixed anion-cation exchange resins are selected according to a three-factors-four-levels normal design table with the compression strength as the evaluation criterion. The resulted formulation is as follows: the mass ratio of polymer emulsion to composite cement is 0.55:1, the loading of mixed anion-cation exchange resins is 0.3, and the anionic-to-cationic exchange resins ratio is 2:1. The polymer cement solidified forms were tested after 28 d curing for Cs + and Sr 2+ leaching rates, pH and conductivity of the leaching water, and radiation-resistant property in addition to their compressive strength. The measurement results indicate that the performance of thus prepared solidified forms can meet the requirements of the National Standard GB14569.1-93 for near earth's surface disposal of low radioactive waste. (authors)

  9. Integrated data management system for radioactive waste and spent fuel in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yong Taek [Korea Power Engineering Co., Inc., Yongin (Korea, Republic of)

    2002-05-15

    An integrated data management system for the safe management of radioactive waste and spent fuel in Korea is developed to collect basic information, provide the framework for national regulation and improve national competition and efficiency in the management of radioactive waste and spent fuel. This system can also provide public access to information such as a statistical graphs and integrated data from various waste generators to meet increased public needs and interests. Through the system, the five principles(independence, openness, clearance, efficiency and reliance) of safety regulation can be realized and public understanding and reliance on the safety of spent fuel and radioactive waste management can be promoted. By providing reliable information and openness within the international nuclear community can be ensured and efficient support of international agreements among contracting parties can be ensured. By operating safe and efficient management of spent fuel and radioactive waste (IAEA joint convention on the safety of spent fuel management and on the safety of radioactive waste management), the system can compensate for the imperfections in safe regulation of radioactive waste and spent fuel management related to waste generation, storage and disposal, and make it possible for holistic control and reorganization of the basic framework of KINS's intermediate and long term research organization and trends, regarding waste management policy so as to integrate safe management and unit safe disposal. To meet this objectives, design of the database system structure and the study of input/output data validation and verification methodology was performed during the second phase of this project.

  10. Condensed matter physics with radioactive ion beams

    International Nuclear Information System (INIS)

    Haas, H.

    1996-01-01

    An overview of the present uses of radioactive ion beams from ISOLDE for condensed matter research is presented. As simple examples of such work, tracer studies of diffusion processes with radioisotopes and blocking/channeling measurements of emitted particles for lattice location are discussed. Especially the application of nuclear hyperfine interaction techniques such as PAC or Moessbauer spectroscopy has become a powerful tool to study local electronic and structural properties at impurities. Recently, interesting information on impurity properties in semiconductors has been obtained using all these methods. The extreme sensitivity of nuclear techniques makes them also well suited for investigations of surfaces, interfaces, and biomolecules. Some ideas for future uses of high energy radioactive ion beams beyond the scope of the present projects are outlined: the study of diffusion in highly immiscible systems by deep implantation, nuclear polarization with the tilted-foil technique, and transmutation doping of wide-bandgap semiconductors. (orig.)

  11. The present state and future prospects of the radioactive waste and spent fuel management in Lithuania

    International Nuclear Information System (INIS)

    Gylys, J.

    2001-01-01

    The Ignalina nuclear power plant (NPP) is the main source of the spent fuel and the radioactive waste (RW) in Lithuania. Now Lithuania is fully responsible for the management and disposal of its RW and spent fuel. The present scheme of spent fuel, solid, and liquid waste treatment is incomplete. The Department of Thermal and Nuclear Energy at Kaunas University of Technology proposed the new idea - cellular foam apparatus for the concentration of the liquid waste at Ignalina NPP. Some data and main results of the investigation of such apparatus are presented here. (author)

  12. Design of systems for handling radioactive ion exchange resin beads

    International Nuclear Information System (INIS)

    Shapiro, S.A.; Story, G.L.

    1979-01-01

    The flow of slurries in pipes is a complex phenomenon. There are little slurry data available on which to base the design of systems for radioactive ion exchange resin beads and, as a result, the designs vary markedly in operating plants. With several plants on-line, the opportunity now exists to evaluate the designs of systems handling high activity spent resin beads. Results of testing at Robbins and Meyers Pump Division to quantify the behavior of resin bead slurries are presented. These tests evaluated the following slurry parameters; resin slurry velocity, pressure drop, bead degradation, and slurry concentration effects. A discussion of the general characteristics of resin bead slurries is presented along with a correlation to enable the designer to establish the proper flowrate for a given slurry composition and flow regime as a function of line size. Guidelines to follow in designing a resin handling system are presented

  13. Implementation of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

    International Nuclear Information System (INIS)

    Stewart, L.; Tonkay, D.

    2004-01-01

    This paper discusses the implementation of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The Joint Convention: establishes a commitment with respect to safe management of spent nuclear fuel and radioactive waste; requires the Parties to ''take appropriate steps'' to ensure the safety of their spent fuel and waste management activities, but does not delineate standards the Parties must meet; and seeks to attain, through its Contracting Parties, a higher level of safety with respect to management of their spent nuclear fuel, disused sealed sources, and radioactive waste

  14. Detection systems for radioactive ion beams

    International Nuclear Information System (INIS)

    Savajols, H.

    2002-01-01

    Two main methods are used to produce radioactive ion beams: -) the ISOL method (isotope separation on-line) in which the stable beam interacts with a thick target, the reaction products diffuse outside the target and are transferred to a source where they are ionized, a mass separator and a post-accelerator drive the selected radioactive ions to the right energy; -) the in-flight fragmentation method in which the stable beam interacts with a thin target, the reaction products are emitted from the target with a restricted angular distribution and a velocity close to that of the incident beam, the experimenter has to take advantage from the reaction kinetics to get the right particle beam. Characteristic time is far longer with the ISOL method but the beam intensity is much better because of the use of a post-accelerator. In both cases, the beam intensity is lower by several orders of magnitude than in the case of a stable beam. This article presents all the constraints imposed by radioactive beams to the detection systems of the reaction products and gives new technical solutions according to the type of nuclear reaction studied. (A.C.)

  15. Disintegration and dissolution of spent radioactive cationic exchange resins using Fenton-like oxidation process

    Energy Technology Data Exchange (ETDEWEB)

    Wan, Zhong; Xu, Lejin [Collaborative Innovation Center for Advanced Nuclear Energy Technology, INET, Tsinghua University, Beijing 100084 (China); Wang, Jianlong, E-mail: wangjl@tsinghua.edu.cn [Collaborative Innovation Center for Advanced Nuclear Energy Technology, INET, Tsinghua University, Beijing 100084 (China); Beijing Key Laboratory of Radioactive Wastes Treatment, Tsinghua University, Beijing 100084 (China)

    2015-09-15

    Highlights: • The spent radioactive resins could be oxidized by Fenton-like process. • The influencing factors on resin oxidation were evaluated. • Chemical oxygen demand (COD) reduction rate was more than 99%. • SEM and Raman spectrum were used to analyze the resins morphological change. - Abstract: The treatment and disposal of the spent radioactive resins is essential for the sustainable development of the nuclear industry. In this paper, the disintegration and dissolution of spent cationic resins were studied by Fenton-like process. The influencing factors on resin dissolution, such as pH, temperature, type and concentration of catalysts were evaluated. The results showed that the spent resins could be effectively dissolved at pH < 1, [Fe{sup 2+}] = 0.2 M and T = 97 ± 2 °C. Chemical oxygen demand (COD) reduction rate was more than 99%. The scanning electron microscopy and the Raman spectrum were used to observe the morphological changes of the spent resins during the dissolution process. Fenton-like oxidation is an efficient method for the volume reduction and stabilization of the spent resins before further immobilization.

  16. Disintegration and dissolution of spent radioactive cationic exchange resins using Fenton-like oxidation process

    International Nuclear Information System (INIS)

    Wan, Zhong; Xu, Lejin; Wang, Jianlong

    2015-01-01

    Highlights: • The spent radioactive resins could be oxidized by Fenton-like process. • The influencing factors on resin oxidation were evaluated. • Chemical oxygen demand (COD) reduction rate was more than 99%. • SEM and Raman spectrum were used to analyze the resins morphological change. - Abstract: The treatment and disposal of the spent radioactive resins is essential for the sustainable development of the nuclear industry. In this paper, the disintegration and dissolution of spent cationic resins were studied by Fenton-like process. The influencing factors on resin dissolution, such as pH, temperature, type and concentration of catalysts were evaluated. The results showed that the spent resins could be effectively dissolved at pH < 1, [Fe 2+ ] = 0.2 M and T = 97 ± 2 °C. Chemical oxygen demand (COD) reduction rate was more than 99%. The scanning electron microscopy and the Raman spectrum were used to observe the morphological changes of the spent resins during the dissolution process. Fenton-like oxidation is an efficient method for the volume reduction and stabilization of the spent resins before further immobilization

  17. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions

  18. Integrated Data Base for 1989: Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1989-11-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1988. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, commercial reactor and fuel cycle facility decommissioning waste, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous, highly radioactive materials that may require geologic disposal. 45 figs., 119 tabs

  19. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  20. Integrated data base for 1990: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1990-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1989. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 22 refs., 48 figs., 109 tabs

  1. Integrated Data Base for 1991: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1991-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1990. These data are based on the most reliable information available form government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated generally through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered are spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 160 refs., 61 figs., 142 tabs

  2. High spin studies with radioactive ion beams

    International Nuclear Information System (INIS)

    Garrett, J.D.

    1992-01-01

    The variety of new research possibilities afforded by the culmination of the two frontier areas of nuclear structure: high spin and studies far from nuclear stability (utilizing intense radioactive ion beams) are discussed. Topics presented include: new regions of exotic nuclear shape (e.g. superdeformation, hyperdeformation, and reflection-asymmetric shapes); the population of and consequences of populating exotic nuclear configurations; and complete spectroscopy (i.e. the overlap of state of the art low-and high-spin studies in the same nucleus)

  3. High spin studies with radioactive ion beams

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, J D [Oak Ridge National Lab., TN (United States)

    1992-08-01

    The variety of new research possibilities afforded by the culmination of the two frontier areas of nuclear structure: high spin and studies far from nuclear stability (utilizing intense radioactive ion beams) are discussed. Topics presented include: new regions of exotic nuclear shape (e.g. superdeformation, hyperdeformation, and reflection-asymmetric shapes); the population of and consequences of populating exotic nuclear configurations; and, complete spectroscopy (i.e. the overlap of state of the art low- and high-spin studies in the same nucleus). (author). 47 refs., 8 figs.

  4. Proceedings of national seminar on physics with radioactive ion beams

    International Nuclear Information System (INIS)

    Chintalapudi, S.N.; Shyam, R.

    1991-01-01

    This volume containing the proceedings of the national seminar on physics with radioactive ion beams gives a broad overview of the developments taking place in the area of nuclear physics and accelerator physics with special emphasis on the utilization of radioactive ion beams for various studies. Topics covered include studies on nuclear structure and nuclear astrophysics and the wide ranging applications of radioactive ion beams in these and other areas of nuclear sciences. Papers relevant to INIS are indexed separately

  5. Immobilisation Of Spent Ion Exchange Resins Using Portland Cement Blending With Organic Material

    International Nuclear Information System (INIS)

    Zalina Laili; Mohd Abdul Wahab; Nur Azna Mahmud

    2014-01-01

    Immobilisation of spent ion exchange resins (spent resins) using Portland cement blending with organic material for example bio char was investigated. The performance of cement-bio char matrix for immobilisation of spent ion exchange resins was evaluated based on their compression strength and leachability under different experimental conditions. The results showed that the amount of bio char and spent resins loading effect the compressive strength of the waste form. Several factors affecting the leaching behaviour of immobilised spent resins in cement-bio char matrix. (author)

  6. Integrated data base for 1988: Spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1988-09-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1987. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected defense-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reportd for miscellaneous, highly radioactive materials that may require geologic disposal. 89 refs., 46 figs., 104 tabs

  7. Operational programs for national radioactive waste and spent fuel management programme in Slovenia

    International Nuclear Information System (INIS)

    Zeleznik, Nadja; Kralj, Metka; Mele, Irena

    2007-01-01

    The first separate National Radioactive Waste and Spent Fuel Management Programme (National Programme) was prepared in Slovenia in 2005 as a supplementary part of the National Environmental Action Programme and was adopted in February 2006 by the Slovenian Parliament. The new National Programme includes all topics being relevant for the management of the radioactive waste and spent fuel which are produced in Slovenia, from the legislation and identification of different waste streams, to the management of radioactive waste and spent fuel, the decommissioning of nuclear facilities and management of (TE)NORM in the near future from 2006 up to the 2015. The National Programme identified the existing and possible future problems and proposed the technical solutions and action plans for two distinctive periods: 2006-2009 and 2010- 2015. According to the requirement of Act on Protection against Ionising Radiation and Nuclear Safety the national Agency for Radwaste Management (ARAO) prepared the operational programmes for the four year period with technical details on implementation of the National programme. ARAO gained the detailed plans of different involved holders and proposed 9 operational programmes with aims, measures, individual organizations in charge, expenses and resources for each of the programmes. The Operational programmes were already reviewed by the Ministry of Environment and Physical Planning and are under acceptance. The orientation of the radioactive waste management according to the National Programme and operational activities within additional limitations based on the strategical decisions of Slovenian Government is presented in the paper. (authors)

  8. Standard format and content for a license application to store spent fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    1989-09-01

    Subpart B, ''License Application, Form, and Contents,'' of 10 CFR Part 72, ''Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste,'' specifies the information to be covered in an application for a license to store spent fuel in an independent spent fuel storage installation (ISFSI) or to store spent fuel and high-level radioactive waste in a monitored retrievable storage facility (MRS). However, Part 72 does not specify the format to be followed in the license application. This regulatory guide suggests a format acceptable to the NRC staff for submitting the information specified in Part 72 for license application to store spent fuel in an ISFSI or to store spent fuel and high-level radioactive waste in an MRS

  9. A strategy for upgrading management of spent fuel and radioactive waste at the Ignalina nuclear power plant

    International Nuclear Information System (INIS)

    Balan, V.; Penkov, V.; Bergman, C.; Gustafsson, B.

    2001-01-01

    The waste management strategy of the former Soviet Union, based on its Norms and Rules, was implemented at Ignalina nuclear power Plant (INPP). This means in brief that the spent nuclear fuel (SNF) should be reprocessed in the Soviet Union and the management of radioactive waste should be done in connection with the eventual decommissioning of the reactors, The major facilities for management of radioactive waste were evaporation of liquid waste and subsequent bitumenisation of the sludge, treatment of liquid with ion-exchange techniques and subsequent storage of the resins in tanks. Solid waste was sorted according to its activity content and stored on site. Following the independence of Lithuania in 1990, Lithuania is successively replacing the former Norms and Rules with Lithuanian laws and regulations. Lithuania has signed both the Convention of Nuclear Safety and the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The paper will give an overview of the work done, in progress and planned, primarily from an operational point of view. It will give examples of practical problems which have to be overcome and how very tough prioritisation has to be made because of lack of resources. Many of the problems are associated with the fact that the waste management strategy successively has to be changed at the same time as the generation of SNF and radioactive waste is continued. A lot has already been achieved, especially regarding the SNF. However, the main work is still to be done until the INPP can declare that its waste is properly managed in full compliance with a national waste management strategy which eventually will end with the disposal in licensed repositories. (author)

  10. Treatment of spent ion-exchange resins for storage and disposal

    International Nuclear Information System (INIS)

    1985-01-01

    This report describes the experience gained by different countries on storage of spent ion exchange resins, immobilization of them into various matrices and the development of new methods in decomposition and solidification of spent resins. The report contains all the results of the Coordinated Research Programme together with additional data available from countries not participating in this programme. A review of practical industrial experience in treating spent ion exchange resins is given in the annex

  11. Leaching of 60 Co and 137 Cs from spent ion exchange resins in ...

    Indian Academy of Sciences (India)

    Cement; radioactive waste; composite; waste management. Abstract. The leaching rate of 60Co and 137Cs from the spent cation exchange resins in cement–bentonite matrix has been studied. The solidification matrix was a standard Portland cement mixed with 290–350 (kg/m3) spent cation exchange resins, with or ...

  12. Synthesis of Lithium Fluoride from Spent Lithium Ion Batteries

    Directory of Open Access Journals (Sweden)

    Daniela S. Suarez

    2017-05-01

    Full Text Available Lithium (Li is considered a strategic element whose use has significantly expanded. Its current high demand is due to its use in lithium ion batteries for portable electronic devices, whose manufacture and market are extensively growing every day. These days there is a great concern about the final disposal of these batteries. Therefore, the possibility of developing new methodologies to recycle their components is of great importance, both commercially and environmentally. This paper presents results regarding important operational variables for the dissolution of the lithium and cobalt mixed-oxide (LiCoO2 cathodes from spent lithium ion batteries (LIBs with hydrofluoric acid. The recovery and synthesis of Co and Li compounds were also investigated. The dissolution parameters studied were: temperature, reaction time, solid-liquid ratio, stirring speed, and concentration of HF. The investigated recovery parameters included: pH, temperature, and time with and without stirring. The final precipitation of lithium fluoride was also examined. The results indicate that an increase in the HF concentration, temperature, and reaction time favors the leaching reaction of the LiCoO2. Dissolutions were close to 60%, at 75 °C and 120 min with a HF concentration of 25% (v/v. The recovery of Co and Li were 98% and 80%, respectively, with purities higher than 94%. Co and Li compounds, such as Co3O4 and LiF, were synthesized. Furthermore, it was possible to almost completely eliminate the F− ions as CaF2.

  13. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Klein, J.A.; Storch, S.N.; Ashline, R.C.

    1994-03-01

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal

  14. Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1992-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1991. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal

  15. Spent nuclear fuel and high level radioactive waste transportation. White paper

    International Nuclear Information System (INIS)

    1985-06-01

    The High-Level Radioactive Waste Committee of the Western Interstate Energy Board has been involved in a year-long cooperative project with the US Department of Energy (DOE) to develop an information base on the transportation of spent nuclear fuel and high-level radioactive waste (HLW) so that western states can be constructive and informed participants in the repository program under the Nuclear Waste Policy Act (NWPA). The historical safety record of transportation of HLW and spent fuel is excellent; no release of these radioactive materials has ever occurred during transportation. Projected shipments under the NWPA will, however, greatly exceed current shipments in the US. For example, over the past five years, 119 metric tons of civilian spent fuel have been shipped in this country, while shipments to the first and second repository are each expected to peak at 3000 metric tons per year. The Committee believes that the successful development and operation of a national HLW/spent fuel transportation system can best be accomplished through an open process based on the common sense approach of taking all reasonable measures to minimize public risk and performing whatever actions are reasonably required to promote public acceptance. Therefore, the Committee recommends that the Department of Energy further the goals of the NWPA by developing a Comprehensive Transportation Plan which adopts a systematic, comprehensive, and integrated approach to resolving all spent fuel and HLW transportation issues in a timely manner. The suggested scope of such a plan is discussed in this White paper. Many of the suggested elements of such a plan are similar to those being developed by the Department of energy for inclusion in the Department's Transportation Institutional Plan

  16. Magneto-plasma separating technologies and their possible application for conversion spent fuel and radioactive waste

    International Nuclear Information System (INIS)

    Kovtun, Yu.V.; Skyibenko, Je.Yi.; Yuferov, V.B.

    2007-01-01

    A problem of spent fuel (SF) and radioactive waste (RAW) processing is considered in the views of using magneto-plasma technologies. Basing on this analysis, the block-diagram of RAW processing by the technology using a magneto-plasma separator is offered. The paper describes the device for material element separation, where the main physical mechanism of plasma formation and heating are collective processes involved by the plasma-beam interaction. The dimensions of a pilot-separating device are determined

  17. Radioactive waste management and spent nuclear fuel storing. Options and priorities

    International Nuclear Information System (INIS)

    Popescu, Ion

    2001-01-01

    As a member of the states' club using nuclear energy for peaceful applications, Romania approaches all the activities implied by natural uranium nuclear fuel cycle, beginning with uranium mining and ending with electric energy generation. Since, in all steps of the nuclear fuel cycle radioactive wastes are resulting, in order to protect the environment and the life, the correct and competent radioactive waste management is compulsory. Such a management implies: a. Separating the radioisotopes in all the effluences released into environment; b. Treating separately the radioisotopes to be each properly stored; c. Conditioning waste within resistant matrices ensuring long term isolation of the radioactive waste destined to final disposal; d. Building radioactive waste repositories with characteristics of isolation guaranteed for long periods of time. To comply with the provisions of the International Convention concerning the safety of the spent nuclear fuel and radioactive waste management, signed on 5 September 1997, Romania launched its program 'Management of Radioactive Wastes and Dry Storing of Spent Nuclear Fuel' having the following objectives: 1. Establishing the technology package for treating and conditioning the low and medium active waste from Cernavoda NPP to prepare them for final disposal; 2. Geophysical and geochemical investigations of the site chosen for the low and medium active final disposal (DFDSMA); 3. Evaluating the impact on environment and population of the DFDSMA; 4. Providing data necessary in the dry intermediate storing of spent nuclear fuel and the continuous and automated surveillance; 5. Establishing multiple barriers for spent nuclear fuel final disposal in order to establish the repository in granitic rocks and salt massives; 6. Designing and testing containers for final disposal of spent nuclear fuel guaranteeing the isolation over at least 500 years; 7. Computational programs for evaluation of radionuclide leakage in environment in

  18. Development of the Holifield Radioactive Ion Beam Facility

    International Nuclear Information System (INIS)

    Tatum, B.A.

    1997-01-01

    The Holifield Radioactive Ion Beam Facility (HRIBF) construction project has been completed and the first radioactive ion beam has been successfully accelerated. The project, which began in 1992, has involved numerous facility modifications. The Oak Ridge Isochronous Cyclotron has been converted from an energy booster for heavy ion beams to a light ion accelerator with internal ion source. A target-ion source and mass analysis system have been commissioned as key components of the facility's radioactive ion beam injector to the 25MV tandem electrostatic accelerator. Beam transport lines have been completed, and new diagnostics for very low intensity beams have been developed. Work continues on a unified control system. Development of research quality radioactive beams for the nuclear structure and nuclear astrophysics communities continues. This paper details facility development to date

  19. Micro-organisms and divers exposure to radioactivity in spent fuel pools at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Muniz de A, D. [Underwater Construction Corporation, Latin America, Fortaleza, Ceara (Brazil); Silva, R. [Universidade Federal do Rio de Janeiro, Instituto de Biofisica Carlos Chagas Filho, 21941-902 Rio de Janeiro (Brazil); Gomes N, C. A., E-mail: dmuniz@uccdive.com [Universidade Federal do Rio de Janeiro, Instituto de Biologia, Environmental Engineering Program, 21941-902 Rio de Janeiro (Brazil)

    2017-09-15

    Many nuclear power plants (NPPs) around the world are in the process of extending their lifespan from 40 to 60 years of operation. The NPP; Angra 1 (Brazil) has performed a thorough evaluation of its Life Extension Engineering project. In this context, the spent fuel pool (SFP) was one of the areas studied in order to demonstrate the plants integrity for a life extension. Micro-organisms growing on the liner of the fuel transfer channel (Ftc) and SFP can form a film of bacteria, which is highly resistant to radiation. This paper aims to compare the micro-organisms found in NPP Angra 1 with those reported to other NPPs and also relates their occurrence with the radiation levels at the sites. It also compares divers exposure to radioactivity during underwater activities in the SFP. Fourteen samples were collected on the surface of the liners of the Ftc, the SFP and the drains within the fuel building (FB) of Angra 1. For the identification of the micro-organisms, a metagenomics analysis was performed by random sequencing (Shotgun) and the use of Ion Torrent PGM Sequence r. Twelve micro-organisms phyla were identified; Acido-bacteria, Actino-bacteria, Bacteroidetes, Chlamydiae, Chlorobi, Chloroflexi, Cyano-bacteria, Deinococcus-Thermus, Firmicutes, Planctomycetes, Proteo-bacteria, and Verrucomicrobia as well as organisms not classified. In the SFP of Angra 1, the bacteria survived the exposure to a radiation of 0.416 Gy/h (high radiation). Deinococcus-thermus, bacteria identified in Angra 1, has resisted an exposure to 30,000 Gy/h in another plant. (Author)

  20. Micro-organisms and divers exposure to radioactivity in spent fuel pools at nuclear power plants

    International Nuclear Information System (INIS)

    Muniz de A, D.; Silva, R.; Gomes N, C. A.

    2017-09-01

    Many nuclear power plants (NPPs) around the world are in the process of extending their lifespan from 40 to 60 years of operation. The NPP; Angra 1 (Brazil) has performed a thorough evaluation of its Life Extension Engineering project. In this context, the spent fuel pool (SFP) was one of the areas studied in order to demonstrate the plants integrity for a life extension. Micro-organisms growing on the liner of the fuel transfer channel (Ftc) and SFP can form a film of bacteria, which is highly resistant to radiation. This paper aims to compare the micro-organisms found in NPP Angra 1 with those reported to other NPPs and also relates their occurrence with the radiation levels at the sites. It also compares divers exposure to radioactivity during underwater activities in the SFP. Fourteen samples were collected on the surface of the liners of the Ftc, the SFP and the drains within the fuel building (FB) of Angra 1. For the identification of the micro-organisms, a metagenomics analysis was performed by random sequencing (Shotgun) and the use of Ion Torrent PGM Sequence r. Twelve micro-organisms phyla were identified; Acido-bacteria, Actino-bacteria, Bacteroidetes, Chlamydiae, Chlorobi, Chloroflexi, Cyano-bacteria, Deinococcus-Thermus, Firmicutes, Planctomycetes, Proteo-bacteria, and Verrucomicrobia as well as organisms not classified. In the SFP of Angra 1, the bacteria survived the exposure to a radiation of 0.416 Gy/h (high radiation). Deinococcus-thermus, bacteria identified in Angra 1, has resisted an exposure to 30,000 Gy/h in another plant. (Author)

  1. Transport and extraction of radioactive ions stopped in superfluid helium

    NARCIS (Netherlands)

    Huang, WX; Dendooven, P; Gloos, K; Takahashi, N; Arutyunov, K; Pekola, JP; Aysto, J

    A new approach to convert a high energy beam to a low energy one, which is essential for the next generation radioactive ion beam facilities, has been proposed and tested at Jyvaskyla, Finland. An open Ra-223 alpha-decay-recoil source has been used to produce radioactive ions in superfluid helium.

  2. Spent sealed radioactive sources conditioning technology for the disposal at the national repository Baita-Bihor

    International Nuclear Information System (INIS)

    Bujoreanu, D.; Popescu, I.V.

    2006-01-01

    A spent sealed radioactive source(SRS) is a high integrity capsule which contains a small amount of concentrated radionuclide with an activity which ranges from a few MBq up to levels of hundreds TBq. Presently, there are now many spent and unusable SRS in Romania, which have been used a long time in various industrial applications (smoke detectors, weld testing etc.). Considering the activity of the Radioactive Waste Treatment Plant (STDR) at the Institute for Nuclear Research Pitesti regarding radioactive source collecting from various economic agents, several radioactive sources are held in the intermediate storage deposit facility on the institute platform awaiting conditioning for the final disposal. This paper presents the conditioning technology for this sources, which has as ultimate purpose to completion of a product which matches the waste acceptance requirements imposed by the National Authority Control of Nuclear Activities, CNCAN, for the disposal site DNDR Baita - Bihor. The technology used for obtaining the final product allows two options for the immobilization of the sources in the 218 L steel drum and these are: Sources placed in the original packages and which can not be dismantled will be isolated by encapsulation in 10 litters metal capsules and then conditioned in 218 l steel drum, with a concrete biological shielding; Sources removed from the initial package are isolated in stainless steel capsules, which are to be conditioned in the same 218 L steel drum. The final product obtained as a result of the concrete conditioning operations of the spent SRS in 218 L steel drum is the steel drum - concrete - low radioactive waste assembly which presents itself as a concrete block which includes one or more capsules containing SRS. (author)

  3. 77 FR 58416 - Comparative Environmental Evaluation of Alternatives for Handling Low-Level Radioactive Waste...

    Science.gov (United States)

    2012-09-20

    ... for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins From Commercial Nuclear Power... Radioactive Waste Spent Ion Exchange Resins from Commercial Nuclear Power Reactors. DATES: Please submit... Evaluation of Alternatives for Handling Low-Level Radioactive Waste Spent Ion Exchange Resins from Commercial...

  4. Graphite Recycling from Spent Lithium-Ion Batteries.

    Science.gov (United States)

    Rothermel, Sergej; Evertz, Marco; Kasnatscheew, Johannes; Qi, Xin; Grützke, Martin; Winter, Martin; Nowak, Sascha

    2016-12-20

    The present work reports on challenges in utilization of spent lithium-ion batteries (LIBs)-an increasingly important aspect associated with a significantly rising demand for electric vehicles (EVs). In this context, the feasibility of anode recycling in combination with three different electrolyte extraction concepts is investigated. The first method is based on a thermal treatment of graphite without electrolyte recovery. The second method additionally utilizes a subcritical carbon-dioxide (subcritical CO 2 )-assisted electrolyte extraction prior to thermal treatment. And the final investigated approach uses supercritical carbon dioxide (scCO 2 ) as extractant, subsequently followed by the thermal treatment. It is demonstrated that the best performance of recycled graphite anodes can be achieved when electrolyte extraction is performed using subcritical CO 2 . Comparative studies reveal that, in the best case, the electrochemical performance of recycled graphite exceeds the benchmark consisting of a newly synthesized graphite anode. As essential efforts towards electrolyte extraction and cathode recycling have been made in the past, the electrochemical behavior of recycled graphite, demonstrating the best performance, is investigated in combination with a recycled LiNi 1/3 Co 1/3 Mn 1/3 O 2 cathode. © 2016 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  5. Integrated data base report - 1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1997-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions

  6. Acid leaching of mixed spent Li-ion batteries

    Directory of Open Access Journals (Sweden)

    A.A. Nayl

    2017-05-01

    Full Text Available Acid leaching for different types of mixed spent Li-ion mobile batteries is carried out after alkali decomposition using NH4OH followed by H2SO4 + H2O2 leaching. In the alkali decomposition step, the effects of reaction time, NH4OH concentration, liquid/solid mass ratio and reaction temperature on the decomposition process are investigated to remove Al, Cu, Mn, Ni, Co, and Li. After alkaline treatment, the alkali paste is treated to leach the remaining metals using H2SO4 + H2O2. The significant effects of reaction time, acid concentration, H2O2 concentration, liquid/solid mass ratios and reaction temperature on the leaching rate are studied. More than 97% of Al, Mn, Ni, Co, and Li and about 65% Cu are leached in two stages. Kinetic analysis shows that, the data fit with chemical reaction control mechanism and the activation energies for the investigated metals using the Arrhenius equation ranged from 30.1 to 41.4 kJ/mol. Recovered metals are precipitated from the leaching liquor at varying pH values using NaOH solution and Na2CO3. Firstly, Mn is precipitated as MnCO3 at pH = 7.5. Secondly, at pH = 9.0, nickel is precipitated as NiCO3. Thirdly, as the pH of the leaching liquor reaches 11–12, Co(OH2 is precipitated and the remaining Li is readily precipitated as Li2CO3 using a saturated Na2CO3 solution. Based on the experimental data, a flow sheet is developed and tested for the recovery process.

  7. Integrated data base for 1986: spent fuel and radioactive waste inventories, projections, and characteristics. Revision 2

    International Nuclear Information System (INIS)

    1986-09-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US Department of Energy (DOE) radioactive wastes through December 31, 1985, based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. Current projections of future waste and spent fuel to be generated through the year 2020 and characteristics of these materials are also presented. The information forecasted is consistent with the expected defense-related and private industrial and institutional activities and the latest DOE/Energy Information Administration (EIA) projections of US commercial nuclear power growth. The materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, remedial action waste, and decommissioning waste. For each category, current and projected inventories are given through the year 2020, and the radioactivity and thermal power are calculated based on reported or calculated isotopic compositions

  8. Recent radioactive ion beam program at RIKEN and related topics

    Indian Academy of Sciences (India)

    Keywords. RIKEN; radioactive ion beams; magic numbers. PACS No. 21.10.-k. 1. Introduction. In RIKEN, there are several heavy ion accelerators. Main accelerator is the RIKEN ring cyclotron (RRC) with K = 540, that has been operated from 1986. The RRC has two injectors; one is heavy ion linear accelerator that has been ...

  9. Study of plasma off-gas treatment from spent ion exchange resin pyrolysis.

    Science.gov (United States)

    Castro, Hernán Ariel; Luca, Vittorio; Bianchi, Hugo Luis

    2017-03-23

    Polystyrene divinylbenzene-based ion exchange resins are employed extensively within nuclear power plants (NPPs) and research reactors for purification and chemical control of the cooling water system. To maintain the highest possible water quality, the resins are regularly replaced as they become contaminated with a range of isotopes derived from compromised fuel elements as well as corrosion and activation products including 14 C, 60 Co, 90 Sr, 129 I, and 137 Cs. Such spent resins constitute a major proportion (in volume terms) of the solid radioactive waste generated by the nuclear industry. Several treatment and conditioning techniques have been developed with a view toward reducing the spent resin volume and generating a stable waste product suitable for long-term storage and disposal. Between them, pyrolysis emerges as an attractive option. Previous work of our group suggests that the pyrolysis treatment of the resins at low temperatures between 300 and 350 °C resulted in a stable waste product with a significant volume reduction (>50%) and characteristics suitable for long-term storage and/or disposal. However, another important issue to take into account is the complexity of the off-gas generated during the process and the different technical alternatives for its conditioning. Ongoing work addresses the characterization of the ion exchange resin treatment's off-gas. Additionally, the application of plasma technology for the treatment of the off-gas current was studied as an alternative to more conventional processes utilizing oil- or gas-fired post-combustion chambers operating at temperatures in excess of 1000 °C. A laboratory-scale flow reactor, using inductively coupled plasma, operating under sub-atmospheric conditions was developed. Fundamental experiments using model compounds have been performed, demonstrating a high destruction and removal ratio (>99.99%) for different reaction media, at low reactor temperatures and moderate power consumption

  10. Accelerator complex for a radioactive ion beam facility at ATLAS

    International Nuclear Information System (INIS)

    Nolen, J.A.

    1995-01-01

    Since the superconducting heavy ion linac ATLAS is an ideal post-accelerator for radioactive beams, plans are being developed for expansion of the facility with the addition of a driver accelerator, a production target/ion source combination, and a low q/m pre-accelerator for radioactive ions. A working group including staff from the ANL Physics Division and current ATLAS users are preparing a radioactive beam facility proposal. The present paper reviews the specifications of the accelerators required for the facility

  11. Report on the disposal of radioactive wastes and spent fuel elements from Baden-Wuerttemberg; Bericht ueber die Entsorgung von radioaktiven Abfaellen und abgebrannten Brennelementen aus Baden-Wuerttemberg

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2017-04-15

    The report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg covers the following issues: legal framework for the nuclear disposal; producer of spent fuels and radioactive wastes in Baden- Report on the disposal of radioactive wastes and spent fuel elements from Baden- Wuerttemberg; low- and medium-level radioactive wastes (non heat generating radioactive wastes); spent fuels and radioactive wastes from waste processing (heat generating radioactive wastes); final disposal.

  12. Wet oxidative destruction of spent ion-exchange resins using hydrogen peroxide

    International Nuclear Information System (INIS)

    Srinivas, C.; Ramaswamy, M.; Theyyunni, T.K.

    1994-01-01

    Spent organic ion exchange resins are generated in large quantities during the operation of nuclear facilities. Wet oxidation as a mode of treatment of these gel-type ion exchange resins was investigated using H 2 O 2 as oxidant in the presence of CuSO 4 as catalyst. Experiments using commercial samples were conducted at 95-100 deg C under reflux conditions at atmospheric pressure. It was found that the reaction of cation resin with H 2 O 2 was instantaneous whereas with anion resin, there was a lag time. For efficient utilization of the oxidant, low rate of addition of H 2 O 2 , 0.01M concentration of CuSO 4 and neutral pH in mixed resin reactions, were found to be useful. Foaming was noted during reactions involving anion resins. This could be controlled by silicone-based agents. The residual solution left after resin oxidation is aqueous in nature and is expected to contain all the radioactivity originally present in the resin. Preliminary experiments showed that it could be efficiently trapped using available inorganic sorbents. Wet oxidation system offers a simple method of converting organic waste into environmentally acceptable inorganic products. (author). 8 refs., 6 figs., 2 tabs

  13. Spent fuel and radioactive waste inventories and projections as of December 31, 1980

    International Nuclear Information System (INIS)

    1981-09-01

    Current inventories and characteristics of commercial spent fuels and both commercial and US Department of Energy radioactive wastes were compiled, based on the most reliable information available from Government sources and the open literature, technical reports, and direct contacts. Future waste generation rates and characteristics of these materials to be accumulated over the remainder of this century are also presented, based on a present DOE/EIA projection of US commercial nuclear power growth and expected defense-related and industrial and institutional activities. Materials considered, on a chapter-by-chapter basis, are: spent fuel, high-level wastes, transuranic waste, low-level waste, remedial action waste, active uranium mill tailings, and airborne waste. For each category, current and projected inventories are given through the year 2000. The land usage requirements are given for storage/disposal of low-level and transuranic wastes, and for the present inventories of inactive uranium mill tailings

  14. Data processing in the integrated data base for spent fuel and radioactive waste

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Morrison, G.W.; Notz, K.J.

    1984-01-01

    The Integrated Data Base (IDB) Program at Oak Ridge National Laboratory (ORNL) produces for the U.S. Department of Energy (DOE) the official spent fuel and radioactive waste inventories and projections for the United States through the year 2020. Inventory data are collected and checked for consistency, projection data are calculated based on specified assumptions, and both are converted to a standard format. Spent fuel and waste radionclides are decayed as a function of time. The resulting information constitutes the core data files called the Past/Present/Future (P/P/F) data base. A data file management system, SAS /sup R/, is used to retrieve the data and create several types of output: an annual report, an electronic summary data file designed for IBM-PC /sup R/ -compatible computers, and special-request reports

  15. Spent fuel and radioactive waste: an integrated data base of inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Notz, K.J.; Forsberg, C.W.; Mastal, E.F.

    1984-01-01

    The Integrated Data Base (IDB) Program provides official US Department of Energy (DOE) data on spent fuel and radioactive waste inventories, projections, and characteristics. This information is provided through the cooperative efforts of the IDB Program and DOE lead offices, lead sites, major programs, and generator sites. The program is entering its fifth year, and major accomplishments are summarized in three broad areas: (1) the annual inventory report, including ORIGEN2 applications and a Quality Assurance (QA) plan; (2) the summary data file and direct user access; and (3) data processing methodology and support to other programs. Plans for future work in these areas are outlined briefly, including increased utilization of personal computers. Some examples of spent fuel data are given in terms of projected quantities for two growth scenarios, burnup and age profile of the existing inventory, and the approximate specific thermal power relative to high-level waste (HLW) from various sources. 4 refs., 2 figs., 3 tabs

  16. Intermediate storage of radioactive waste and spent nuclear fuel at the Kola Peninsula

    International Nuclear Information System (INIS)

    Bohmer, N.

    1999-01-01

    The problem of nuclear waste and disused nuclear submarines are a product of the arms race and the Cold War. Russia still continues to build new nuclear submarines, but there are very few provisions being made to properly store old nuclear submarines, and develop sufficient storage facilities for spent nuclear fuel and other radioactive waste. A solution to this problem is proposed: to construct a new regional interim storage facilities at Kola for the spent nuclear fuel instead of transporting it to Mayak, the existing reprocessing plant. This storage should have the capacity to handle the fuel in the existing storage and the fuel still on board of retired nuclear submarines. Its lifetime should be 50 years. later it would be possible to make a decision on the future of this fuel

  17. A radioactive ion beam facility using photofission

    CERN Document Server

    Diamond, W T

    1999-01-01

    Use of a high-power electron linac as the driver accelerator for a Radioactive Ion Beam (RIB) facility is proposed. An electron beam of 30 MeV and 100 kW can produce nearly 5x10 sup 1 sup 3 fissions/s from an optimized sup 2 sup 3 sup 5 U target and about 60% of this from a natural uranium target. An electron beam can be readily transmitted through a thin window at the exit of the accelerator vacuum system and transported a short distance through air to a water-cooled Bremsstrahlung-production target. The Bremsstrahlung radiation can, in turn, be transported through air to the isotope-production target. This separates the accelerator vacuum system, the Bremsstrahlung target and the isotope-production target, reducing remote handling problems. The electron beam can be scanned over a large target area to reduce the power density on both the Bremsstrahlung and isotope-production targets. These features address one of the most pressing technological challenges of a high-power RIB facility, namely the production o...

  18. The production of accelerated radioactive ion beams

    International Nuclear Information System (INIS)

    Olsen, D.K.

    1993-01-01

    During the last few years, substantial work has been done and interest developed in the scientific opportunities available with accelerated radioactive ion beams (RIBs) for nuclear physics, astrophysics, and applied research. This interest has led to the construction, development, and proposed development of both first- and second-generation RIB facilities in Asia, North America, and Europe; international conferences on RIBs at Berkeley and Louvain-la-Neuve; and many workshops on specific aspects of RIB production and science. This paper provides a discussion of both the projectile fragmentation, PF, and isotope separator on-line, ISOL, approach to RIB production with particular emphasis on the latter approach, which employs a postaccelerator and is most suitable for nuclear structure physics. The existing, under construction, and proposed facilities worldwide are discussed. The paper draws heavily from the CERN ISOLDE work, the North American IsoSpin Laboratory (ISL) study, and the operating first-generation RIB facility at Louvain-la-Neuve, and the first-generation RIB project currently being constructed at ORNL

  19. Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    1996-12-01

    The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens

  20. Spent fuel transportation on highways: the radioactive dose to the traffic

    International Nuclear Information System (INIS)

    Yadigaroglu, G.

    1975-01-01

    The radioactive exposure of the traffic moving on the same highway as spent fuel shipments has been neglected in the past. Methods developed for calculating peak exposures, the number of individuals receiving a dose in excess of a certain limiting value, and the cumulative population doses for the occupants of the vehicles under a variety of highway and accident conditions allow comparisons to the corresponding stationary-population doses. Consideration of both routine direct-radiation exposures and accidental releases indicates that the traffic doses can be of equal or greater importance than the stationary-population doses

  1. An analysis of the technical status of high level radioactive waste and spent fuel management systems

    Science.gov (United States)

    English, T.; Miller, C.; Bullard, E.; Campbell, R.; Chockie, A.; Divita, E.; Douthitt, C.; Edelson, E.; Lees, L.

    1977-01-01

    The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

  2. Problems on radioactive waste and spent nuclear fuel management in the European Arctic Region of Russia

    International Nuclear Information System (INIS)

    Krukov, Evgeny B.

    1999-01-01

    In the Arkhangelsk and Murmansk regions of Russia, radioactive wastes and spent nuclear fuel from the Northern Fleet and Mineconomiki, the technological repairing plant Atomflot and the Kola nuclear power plant and other activities is accumulating steadily and there is no adequate waste management system in the region. There is an action plan to remedy the situation, but it has been delayed because of insufficient funds. This presentation lists the volumes of liquid and solid radioactive wastes from these sources in 1996 and the expected volumes in 2020. It also lists the specific problems of the present waste management and main proposals of the action plan. In addition to federal funds, a number of projects are financed through international co-operation

  3. United States Program on Spent Nuclear Fuel and High-Level Radioactive Waste Management

    International Nuclear Information System (INIS)

    Stewart, L.

    2004-01-01

    The President signed the Congressional Joint Resolution on July 23, 2002, that designated the Yucca Mountain site for a proposed geologic repository to dispose of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The United States (U.S.) Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently focusing its efforts on submitting a license application to the U.S. Nuclear Regulatory Commission (NRC) in December 2004 for construction of the proposed repository. The legislative framework underpinning the U.S. repository program is the basis for its continuity and success. The repository development program has significantly benefited from international collaborations with other nations in the Americas

  4. 10 CFR 72.128 - Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for spent fuel, high-level radioactive waste, reactor-related greater than Class C waste, and other radioactive waste storage and handling. 72.128... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C...

  5. 10 CFR 72.108 - Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste...

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Spent fuel, high-level radioactive waste, or reactor... RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Siting Evaluation Factors § 72.108 Spent fuel, high-level radioactive waste, or reactor-related greater than Class C waste transportation. The...

  6. Study of the on line radioactive multicharged ion production

    International Nuclear Information System (INIS)

    Lecesne, N.

    1997-01-01

    This work is directly related to the SPIRAL project (Systeme de Production d'Ions Radioactifs Acceleres en Ligne) which will start at GANIL at the end of 1998. The aim of the thesis was to study the on line radioactive multicharged ion beam production stages, i.e. the production and diffusion of the radioactive nuclei in a thick target, their possible transfer up to an ECR ion source and their ionisation. Production cross sections of radioactive neutron rich nuclei, formed by fragmentation of a heavy ion beam in a thick target, were measured. An external target-ECR source system, dedicated to the radioactive noble gases production, and two internal target-ECR source systems, dedicated to the radioactive condensable element production, were built and tested on the SIRa tests bench (Separateur d'Ions Radioactifs). Different detection configurations were elaborated in order to identify the radioactive nuclei and estimate their production yields. Finally, a new method for measuring the overall efficiency of the separator was developed and allowed to study the diffusion properties of radioactive noble gases in various targets. (author)

  7. Radioactive spent resins conditioning by the hot super-compaction process

    International Nuclear Information System (INIS)

    Roth, Andreas; Centner, Baudouin; Lemmens, Alain

    2007-01-01

    Spent ion exchanger media are considered to be problematic waste that, in many cases, requires special approaches and precautions during its immobilization to meet the acceptance criteria for disposal. The waste acceptance criteria define, among others, the quality of waste forms for disposal, and therefore will sometimes define appropriate treatment options. The selection of treatment options for spent ion exchange materials must consider their physical and chemical characteristics. Basically, the main methods for the treatment of spent organic ion exchange materials, following to pretreatment methods are: - Direct immobilization, producing a stable end product by using Cement, Bitumen, Polymer or High Integrity Containers, - The destruction of the organic compounds by using Thermochemical processes or Oxidation to produce an inorganic intermediate product that may or may not be further conditioned for storage and/or disposal, - The complete removal of the resin inner structural water by a thermal process. After a thorough technical economical analysis, Tractebel Engineering selected the Resin Hot Compaction Process to be installed at Tihange Nuclear Power Plant. The Resin Hot Compaction Process is used to make dense homogenous organic blocks from a wide range of particulate waste. In this process spent resins are first dewatered and dried to remove the inner structural water content. The drying takes place in a drying vessel that holds the contents of two 200 L drums (Figure). In the oil heated drying and mixing unit, the resins are heated to the necessary process temperature for the hot pressing step and then placed into special metal drums, which are automatically lidded and immediately transferred to a high force compactor. After high force compaction the pellets are transferred to a measuring unit, where the dose rate, height and weight are automatically measured and recorded. A volume reduction factor of approximately up to four (depending on the type of

  8. Resistance to radiation and concretes thermal cycles for conditioning of spent radioactive sources

    International Nuclear Information System (INIS)

    Gonzalez N, M.; Monroy G, F.; Gonzalez D, R. C.; Corona P, I. J.; Ortiz A, G.

    2014-10-01

    In order to know the concrete type most suitable for use as a matrix of conditioning of spent radioactive sources, concrete test tubes using 4 different types of cement were prepared: CPC 30-Rs Extra, CPC 30-R Impercem, CPC 30-R Rs and CPC 30-R with two gravel sizes >30 mm and <10 mm. The concrete test tubes were subjected to testing compressive strength after 28 days of hardening and after being irradiated and subjected to thermal cycles. Subsequently they were characterized by X-ray diffraction and scanning electron microscopy, in order to evaluate whether these concretes accredited the tests set by the NOM-019-Nucl-1995. The results show that the compressive strength of the hardened concretes to 28 days presents values between 36 and 25 MPa; applying irradiation the resistance may decrease to 30% of its original strength; and if subjected to high and low temperatures the ettringite formation also causes a decrease in resistance. The results show that concretes made from cement Impercem, Cruz Azul with gravel <10 mm comply with the provisions of standard and they can be used for conditioning of spent radioactive sources. (Author)

  9. ISOL science at the Holifield Radioactive Ion Beam Facility

    Energy Technology Data Exchange (ETDEWEB)

    Beene, James R [ORNL; Bardayan, Daniel W [ORNL; Galindo-Uribarri, Alfredo {nmn} [ORNL; Gross, Carl J [ORNL; Jones, K. L. [University of Tennessee, Knoxville (UTK); Liang, J Felix [ORNL; Nazarewicz, Witold [ORNL; Stracener, Daniel W [ORNL; Tatum, B Alan [ORNL; Varner Jr, Robert L [ORNL

    2011-01-01

    The Holi eld Radioactive Ion Beam Facility, located in Oak Ridge, Tennessee, is operated as a National User Facility for the U.S. Department of Energy, producing high quality ISOL beams of short-lived, radioactive nuclei for studies of exotic nuclei, astrophysics research, and various societal applications. The primary driver, the Oak Ridge Isochronous Cyclotron, produces rare isotopes by bombarding highly refractory targets with light ions. The radioactive isotopes are ionized, formed into a beam, mass selected, injected into the 25-MV Tandem, accelerated, and used in experiments. This article reviews HRIBF and its science.

  10. Ion sources development at GANIL for radioactive beams and high charge state ions

    International Nuclear Information System (INIS)

    Leroy, R.; Barue, C.; Canet, C.; Dupuis, M.; Flambard, J.L.; Gaubert, G.; Gibouin, S.; Huguet, Y.; Jardin, P.; Lecesne, N.; Leherissier, P.; Lemagnen, F.; Pacquet, J.Y.; Pellemoine-Landre, F.; Rataud, J.P.; Saint-Laurent, M.G.; Villari, A.C.C.; Maunoury, L.

    2001-01-01

    The GANIL laboratory has in charge the production of ion beams for nuclear and non nuclear physics. This article reviews the last developments that are underway in the fields of radioactive ion beam production, increase of the metallic ion intensities and production of highly charges ion beams. (authors)

  11. Options for Management of Spent Fuel and Radioactive Waste for Countries Developing New Nuclear Power Programmes

    International Nuclear Information System (INIS)

    2013-01-01

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. Today, numerous countries are considering construction of their first nuclear power plant or the expansion of a small nuclear power programme, and many of these countries have limited experience in managing radioactive waste and spent nuclear fuel. They often have limited information about available technologies and approaches for safe and long term management of radioactive waste and spent nuclear fuel arising from power reactors. The lack of basic know-how and of a credible waste management strategy could present a major challenge or even an obstruction for countries wishing to

  12. Bituminous solidification, disposal, transport and burial of spent ion-exchange resins. Part of a coordinated programme on treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Mozes, G.; Kristof, M.

    1983-07-01

    The project dealing with the incorporation of spent ion-exchange resins into bitumen was performed within the Agency coordinated research programme on treatment of spent ion-exchange resins. Physical and chemical properties of commercial ion-exchange resins, bitumens and bituminized resins were studied. It was shown that bitumen with low oil content and with a softening point of 60-70 deg. C are applicable for the incorporation of resins. The final waste form is allowed to contain maximum 50% resin. The comprehensive study of the biological resistance of B-30 bitumen was performed. That showed that any bacteriological attack can be regarded as generally insignificant. A continuously operating technology was realized on a semi-plant scale. The best operating conditions of this technology were determined. On the basis of the experience gained from the experiments a design of the bituminization plant of 50m 3 dry resin/year treatment capacity was proposed

  13. A method for embedding granulated or spent ion-exchanging organic substances in concrete

    International Nuclear Information System (INIS)

    Christensen, H.; Jejes, P.

    1976-01-01

    A method of embedding in concrete a spent ion-exchanging organic substance originating from a scrubbing-circuit of a nuclear reactor Prior to solidification of the mixture, a substance is incorporated therewith said substance being capable of preventing water in the grains of the ion-exchanging substance from penetrating into the solidified product

  14. Collinear laser spectroscopy on radioactive praseodymium ions and cadmium ions

    International Nuclear Information System (INIS)

    Froemmgen, Nadja

    2013-01-01

    Collinear laser spectroscopy is a tool for the model independent determination of spins, charge radii and electromagnetic moments of nuclei in ground and long-lived isomeric states. In the context of this thesis a new offline ion source for high evaporating temperatures and an ion beam analysis system were implemented at the TRIGA-LASER Experiment at the Institute for Nuclear Chemistry at the University of Mainz. The main part of the thesis deals with the determination of the properties of radioactive praseodymium and cadmium isotopes by collinear laser spectroscopy at ISOLDE/CERN. The necessary test measurements for the spectroscopy of praseodymium ions have been conducted with the aforementioned offline ion source at the TRIGA-LASER experiment. The spectroscopy of the praseodymium ions was motivated by the observation of a modulation of the electron capture decay rates of hydrogen-like 140 Pr 58+ . The nuclear magnetic moment of the nucleus is, among others, required for the explanation of the so-called GSI Oscillations and has not been studied experimentally before. Additionally, the determined electron capture decay constant of hydrogen-like 140 Pr 58+ is lower than the one of helium-like 140 Pr 57+ . The explanation of this phenomenon requires a positive magnetic moment. During the experiment at the COLLAPS apparatus the magnetic moments of the neutron-deficient isotopes 135 Pr, 136 Pr and 137 Pr could be determined for the first time. Unfortunately, due to a too low production yield the desired isotope 140 Pr could not be studied.The systematic study of cadmium isotopes was motivated by nuclear physics in the tin region. With Z=48 two protons are missing for the shell closure and the isotopes extend from the magic neutron number N=50 to the magic neutron number N=82. The extracted nuclear properties allow tests of different nuclear models in this region. In this thesis the obtained results of the spectroscopy of the cadmium isotopes 106-124,126 Cd and their

  15. Radioactive characteristics of spent fuels and reprocessing products in thorium fueled alternative cycles

    International Nuclear Information System (INIS)

    Maeda, Mitsuru

    1978-09-01

    In order to provide one fundamental material for the evaluation of Th cycle, compositions of the spent fuels were calculated with the ORIGEN code on following fuel cycles: (1) PWR fueled with Th- enriched U, (2) PWR fueled with Th-denatured U, (3) CANDU fueled with Th-enriched U and (4) HTGR fueled with Th-enriched U. Using these data, product specifications on radioactivity for their reprocessing were calculated, based on a criterion that radioactivities due to foreign elements do not exceed those inherent in nuclear fuel elements, due to 232 U in bred U or 228 Th in recovered Th, respectively. Conclusions are as the following: (1) Because of very high contents of 232 U and 228 Th in the Th cycle fuels from water moderated reactors, especially from PWR, required decontamination factors for their reprocessing will be smaller by a factor of 10 3 to 10 4 , compared with those from U-Pu fueled LWR cycle. (2) These less stringent product specifications on the radioactivity of bred U and recovered Th will justify introduction of some low decontaminating process, with additional advantage of increased proliferation resistance. (3) Decontamination factors required for HTGR fuel will be 10 to 30 times higher than for the other fuels, because of less 232 U and 228 Th generation, and higher burn-up in the fuel. (author)

  16. Status of radioactive ion beams at the HRIBF

    CERN Document Server

    Stracener, D W

    2003-01-01

    Radioactive Ion Beams (RIBs) at the Holifield Radioactive Ion Beam Facility (HRIBF) are produced using the isotope separation on-line technique and are subsequently accelerated up to a few MeV per nucleon for use in nuclear physics experiments. The first RIB experiments at the HRIBF were completed at the end of 1998 using sup 1 sup 7 F beams. Since then other proton-rich ion beams have been developed and a large number of neutron-rich ion beams are now available. The neutron-rich radioactive nuclei are produced via proton-induced fission of uranium in a low-density matrix of uranium carbide. Recently developed RIBs include sup 2 sup 5 Al from a silicon carbide target and isobarically pure beams of neutron-rich Ge, Sn, Br and I isotopes from a uranium carbide target.

  17. Generation projection of solid and liquid radioactive wastes and spent radioactive sources in Mexico; Proyeccion de generacion de desechos radiactivos solidos, liquidos y fuentes radiactivas gastadas en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Garcia A, E.; Hernandez F, I. Y.; Fernandez R, E. [Universidad Politecnica del Valle de Toluca, Km 5.7 Carretera Almoloya de Juarez, Estado de Mexico (Mexico); Monroy G, F.; Lizcano C, D., E-mail: fabiola.monroy@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    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)

  18. Radioactive ion beam production by the ISOL method for SPIRAL

    International Nuclear Information System (INIS)

    Landre-Pellemoine, Frederique

    2001-01-01

    This work is directly related to the SPIRAL project (Systeme de Production d'Ions Radioactifs Acceleres en Lignes) of which the start up will begin in September 2001 at GANIL (Grand Accelerateur National d'Ions Lourds) in Caen. This thesis primarily concerns the development of radioactive ion production systems (target/ion source) by the thorough study of each production stage of the ISOL (Isotopic Separation On Line) method: target and/or projectile fragmentation production, diffusion out of target material, effusion into the ion source and finally the ionization of the radioactive atoms. A bibliographical research and thermal simulations allowed us to optimize materials and the shape of the production and diffusion targets. A first target was optimized and made reliable for the radioactive noble gases production (argon, neon...). A second target dedicated to the radioactive helium production was entirely designed and realised (from the specifications to the 'off line' and 'on line' tests). Finally, a third target source system was defined for singly-charged radioactive alkaline production. The intensities of secondary beams planned for SPIRAL are presented here. A detailed study of the diffusion effusion efficiency for these various targets showed that the use of a fine microstructure carbon (grain size of 1 μm) improved the diffusion and showed the importance of thickness of the lamella for the short lived isotope effusion. (author) [fr

  19. Study of shielding analysis methods for casks of spent fuel and radioactive waste

    International Nuclear Information System (INIS)

    Saito, Ai

    2017-01-01

    Casks are used for storage or transport spent fuels or radioactive waste. Because high shielding performances are required, it is very important to confirm the validity of shielding analysis methods in order to evaluate cask shielding abilities appropriately. For this purpose, following studies were carried out. 1) A series of parameter survey for several codes to evaluated the difference of the results. 2) Calculations using the MCNP code are effective and theoretically have better accuracy. However setting reasonable variance reduction parameters is indispensable. Therefore, effectiveness of the ADVANTG code which produces automatically reasonable variance reduction parameters is carried out by comparison with conventional method. As a result, the validity of shielding analysis methods for casks is confirmed. The results will be taken into consideration in our future shielding analysis. (author)

  20. Learning and education on environmental radioactivity by residents of Rokkasho Site for the spent fuel recycling facilities

    International Nuclear Information System (INIS)

    Kawauchi, Kiye; Itoh, Natsuko; Ishikawa, Tomiye; Nihonyanagi, Haruko; Aratani, Michi

    2005-01-01

    The neutron criticality accident at the JCO, a private company for nuclear fuel processing facilities in Tokai has drastically changed minds and attitudes of residents toward environmental radioactivity. The accident happened on September 30, 1999. Before the accident the residents of the Rokkasho Site were not anxious about environmental radioactivity, because they thought the facilities were safe enough concerning containment policy of the radioactivity inside the facilities. Residents, however, had not been taught on a neutron. It is an unfamiliar radiation for them. So, they promptly learnt on neutrons, and some of them began the fixed point measurement of neutrons at the nearest site of the Spent Fuel Recycling Facilities of Rokkasho by the help of Prof. Kazuhisa. Komura, Kanazawa University. Members of the Reading Cicle, Rokkasho Culture Society, mainly women, learnt measurements of environmental radioactivity using simplified counters for alpha-, beta-, and gamma-ray from natural radioactive elements and prepared various kinds of environmental samples. After learning of environmental radioactivity, they began educational activities on the environmental radioactivity for boys and girls in the region. Monitoring of environmental radioactivity is performed by different institutions and with their purposes. Here is reported learning of environmental radioactivity by the residents and education of environmental radioactivity toward the young. Even with the simplest counters, we think that the monitoring of environmental radioactivity by the residents themselves is the royal road to the safety of the regional society. (author)

  1. A closed loop process for recycling spent lithium ion batteries

    Science.gov (United States)

    Gratz, Eric; Sa, Qina; Apelian, Diran; Wang, Yan

    2014-09-01

    As lithium ion (Li-ion) batteries continue to increase their market share, recycling Li-ion batteries will become mandatory due to limited resources. We have previously demonstrated a new low temperature methodology to separate and synthesize cathode materials from mixed cathode materials. In this study we take used Li-ion batteries from a recycling source and recover active cathode materials, copper, steel, etc. To accomplish this the batteries are shredded and processed to separate the steel, copper and cathode materials; the cathode materials are then leached into solution; the concentrations of nickel, manganese and cobalt ions are adjusted so NixMnyCoz(OH)2 is precipitated. The precipitated product can then be reacted with lithium carbonate to form LiNixMnyCozO2. The results show that the developed recycling process is practical with high recovery efficiencies (∼90%), and 1 ton of Li-ion batteries has the potential to generate 5013 profit margin based on materials balance.

  2. Regulatory supervision of sites for spent fuel and radioactive waste storage in the Russian northwest.

    Science.gov (United States)

    Shandala, N K; Sneve, M K; Smith, G M; Kiselev, M F; Kochetkov, O A; Savkin, M N; Simakov, A V; Novikova, N Ya; Titov, A V; Romanov, V V; Seregin, V A; Filonova, A V; Semenova, M P

    2008-12-01

    In the 1960s two technical bases for the Northern Fleet were created in the Russian northwest at Andreeva Bay in the Kola Peninsula and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, receiving and storing radioactive waste and spent nuclear fuel. No further waste was received after 1985, and the technical bases have since been re-categorised as temporary storage sites. The handling of these materials to put them into a safe condition is especially hazardous because of their degraded state. This paper describes regulatory activities which have been carried out to support the supervision of radiological protection during recovery of waste and spent fuel, and to support regulatory decisions on overall site remediation. The work described includes: an assessment of the radiation situation on-site; the development of necessary additional regulatory rules and standards for radiation protection assurance for workers and the public during remediation; and the completion of an initial threat assessment to identify regulatory priorities. Detailed consideration of measures for the control of radiation exposure of workers and radiation exposure of the public during and after operations and emergency preparedness and response are complete and provided in sister papers. The continuing requirements for regulatory activities relevant to the development and implementation of on-going and future remediation activities are also outlined. The Norwegian Radiation Protection Authority supports the work, as part of the Norwegian Government's plan of action to promote improvements in radiation protection and nuclear safety in northwest Russia.

  3. Regulatory supervision of sites for spent fuel and radioactive waste storage in the Russian Northwest

    Energy Technology Data Exchange (ETDEWEB)

    Shandala, N K; Kochetkov, O A; Savkin, M N; Simakov, A V; Novikova, N Ya; Titov, A V; Seregin, V A; Filonova, A V; Semenova, M P [Burnasyan Federal Medical Biophysical Centre, Moscow (Russian Federation); Sneve, M K [Norwegian Radiation Protection Authority, Oslo (Norway); Smith, G M [GMS Abingdon Ltd (United Kingdom); Kiselev, M F; Romanov, V V [Federal Medical-Biological Agency, Moscow (Russian Federation)], E-mail: shandala@srcibph.ru

    2008-12-15

    In the 1960s two technical bases for the Northern Fleet were created in the Russian northwest at Andreeva Bay in the Kola Peninsula and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, receiving and storing radioactive waste and spent nuclear fuel. No further waste was received after 1985, and the technical bases have since been re-categorised as temporary storage sites. The handling of these materials to put them into a safe condition is especially hazardous because of their degraded state. This paper describes regulatory activities which have been carried out to support the supervision of radiological protection during recovery of waste and spent fuel, and to support regulatory decisions on overall site remediation. The work described includes: an assessment of the radiation situation on-site; the development of necessary additional regulatory rules and standards for radiation protection assurance for workers and the public during remediation; and the completion of an initial threat assessment to identify regulatory priorities. Detailed consideration of measures for the control of radiation exposure of workers and radiation exposure of the public during and after operations and emergency preparedness and response are complete and provided in sister papers. The continuing requirements for regulatory activities relevant to the development and implementation of on-going and future remediation activities are also outlined. The Norwegian Radiation Protection Authority supports the work, as part of the Norwegian Government's plan of action to promote improvements in radiation protection and nuclear safety in northwest Russia.

  4. Adsorption of radioactive ions on carnauba-wax aerosols

    Energy Technology Data Exchange (ETDEWEB)

    Paul, A.; Keyser, U. [Physikalisch-Technische Bundesanstalt, Braunschweig (Germany)

    1998-08-01

    A new method based on parallel aerosol size spectrometry and {gamma}-spectrometry is introduced for the measurement of short-lived radioactive ions, fission products or super-heavy elements produced at accelerators. Furthermore a new aerosol generator is presented.The possibility of controlling and changing the aerosol size distribution in the helium aerosol jet produced by the aerosol generator allows the process of the adsorption and transport of radioactive ions on aerosols to be examined for the first time. This is due to the fact that the distribution is surveyed on-line using a negligible part of its total volume and parallel to the transporting flow. The radioactivity of the transported ions is measured by a germanium detector in offline position. In principle, both an on- or offline position with narrow multi-detector geometry (e.g. {beta}{gamma}{gamma}) is possible. (orig.) With 8 figs., 14 refs.

  5. Physics with energetic radioactive ion beams

    International Nuclear Information System (INIS)

    Henning, W.F.

    1996-01-01

    Beams of short-lived, unstable nuclei have opened new dimensions in studies of nuclear structure and reactions. Such beams also provide key information on reactions that take place in our sun and other stars. Status and prospects of the physics with energetic radioactive beams are summarized

  6. Design of the radioactive ion beam facility at the LNS

    International Nuclear Information System (INIS)

    Migneco, E.; Alba, R.; Calabretta, L.; Ciavola, G.; Cuttone, G.; Di Giacomo, M.; Gammino, S.; Gmaj, P.; Moscatello, M.H.; Raia, G.

    1992-01-01

    At the Laboratorio Nazionale del Sud the existing 15 MV Tandem will be coupled to the Superconducting Cyclotron booster, which will provide light and heavy ion beams in the energy range 100-20 MeV/n. Using these beams, secondary radioactive beams can be produced by projectile fragmentation. A fragment separator will collect the secondary beam produced at energies near that of the projectile and deliver it into the experimental areas. The possibility of using an ECRIS source for the axial injection into the Cyclotron and producing radioactive ions on a thick source placed inside the Tandem preinjector is also discussed. (author) 7 refs.; 2 figs.; 1 tab

  7. Strategic environmental assessment of the national programme for the safe management of spent fuel and radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Steinhoff, Mathias; Kallenbach-Herbert, Beate; Claus, Manuel [Oeko-Institut e.V. Darmstadt (Germany); and others

    2015-03-27

    The report on the strategic environmental audit for the national waste disposal program covers the following issues: aim of the study, active factors, environmental objectives; description and evaluation of environmental impact including site selection criteria for final repositories of heat generating radioactive waste, intermediate storage of spent fuel elements and waste from reprocessing plants, disposal of wastes retrieved from Asse II; hypothetical zero variants.

  8. Use of natural materials from Northern Russia for the isolation of radioactive wastes and spent nuclear fuel

    International Nuclear Information System (INIS)

    Komlev, V.N.

    1998-01-01

    The application of natural materials to the isolation of radioactive waste and spent nuclear fuel is being assessed, together with possible isolation technologies. The operational requirements for such materials are identified and a proposal for an inter-regional ecological and technological project is discussed. 39 refs

  9. Strategic environmental assessment of the national programme for the safe management of spent fuel and radioactive waste

    International Nuclear Information System (INIS)

    Steinhoff, Mathias; Kallenbach-Herbert, Beate; Claus, Manuel

    2015-01-01

    The report on the strategic environmental audit for the national waste disposal program covers the following issues: aim of the study, active factors, environmental objectives; description and evaluation of environmental impact including site selection criteria for final repositories of heat generating radioactive waste, intermediate storage of spent fuel elements and waste from reprocessing plants, disposal of wastes retrieved from Asse II; hypothetical zero variants.

  10. Norwegian national report. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management

    International Nuclear Information System (INIS)

    2011-11-01

    This report contains the national report from Norway to the fourth review meeting of the JointConvention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management to be held 14-23 May 2012. (Author)

  11. Ion migration in ocean sediments: subseafloor radioactive waste disposal

    International Nuclear Information System (INIS)

    Nuttall, H.E.; Ray, A.K.; Davis, E.J.

    1980-01-01

    In this study of seabed disposal, analytical ion transport models were developed and used to elucidate ion migration through ocean sediments and to study the escape of ions from the ocean floor into the water column. An unsteady state isothermal diffusion model was developed for the region far from the canister to examine the effects of ion diffusion, adsorption, radioactive decay, sediment thickness and canister position. Analytical solutions were derived to represent the transient concentration profiles within the sediment, ion flux and the ion discharge rate to the water column for two types of initial conditions: instantaneous dissolution of the canister and constant canister leakage. Generalized graphs showing ion migration and behavior are presented

  12. The 1+ → n+ transformation for the radioactive ion acceleration

    International Nuclear Information System (INIS)

    Chauvin, N.; Lamy, T.; Bruandet, J.F.; Bouly, J.L.; Curdy, J.C.; Geller, R.; Sole, P.; Sortais, P.; Vieux-Rochaz, J.L.

    1999-01-01

    The radioactive ions are produced as single-charge ions either starting from nuclear reactions induced by a high energy primary beam, or by neutron bombarding of a target. However, in order to obtain beams of several MeV per nucleon, il will be convenient of transforming the mono-charged ions issued from the production source, in multicharged ions. Consequently, an operation should be implemented to transform the 1+ charge state into n+ state, with a double requirement of maximal yield and minimal response time. The objectives are a particle yield of several percents and a response time below 1 second, taking into account the low lifetimes of certain radioactive nuclei. The conjoint achievement of both high charged states and maximal beam intensity forced us to make a choice for an ECR (Electron Cyclotron Resonance) type source to realize the transformation 1+ → n+

  13. Current issues in the transport of radioactive waste and spent fuel: work by the World Nuclear Transport Institute

    Energy Technology Data Exchange (ETDEWEB)

    Neau, H-J.; Bonnardel-Azzarelli, B. [World Nuclear Transport Inst., London (United Kingdom)

    2014-07-01

    Various kinds of radioactive waste are generated from nuclear power and fuel cycle facilities. These materials have to be treated, stored and eventually sent to a repository site. Transport of wastes between these various stages is crucial for the sustainable utilization of nuclear energy. The IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6) have, for many decades, provided a safe and efficient framework for radioactive materials transport and continue to do so. However, some shippers have experienced that in the transport of certain specific radioactive wastes, difficulties can be encountered. For example, some materials produced in the decommissioning of nuclear facilities are unique in terms of composition or size and can be difficult to characterize as surface contaminated objects (SCO) or homogeneous. One way WNTI (World Nuclear Transport Institute) helps develop transport methodologies is through the use of Industry Working Groups, bringing together WNTI members with common interests, issues and experiences. The Back-End Transport Industry Working Group focuses on the following issues currently. - Characterization of Waste: techniques and methods to classify wastes - Large Objects: slightly contaminated large objects (ex. spent steam generators) transport - Dual Use Casks: transportable storage casks for spent nuclear fuels, including the very long term storage of spent fuel - Fissile Exceptions: new fissile exceptions provisions of revised TS-R-1 (SSR-6) The paper gives a broad overview of current issues for the packaging and transport of radioactive wastes and the associated work of the WNTI. (author)

  14. Ion sources for initial use at the Holifield radioactive ion beam facility

    International Nuclear Information System (INIS)

    Alton, G.D.

    1994-01-01

    The Holifield Radioactive Ion Beam Facility (HRIBF) now under construction at the Oak Ridge National Laboratory will use the 25-MV tandem accelerator for the acceleration of radioactive ion beams to energies appropriate for research in nuclear physics; negative ion beams are, therefore, required for injection into the tandem accelerator. Because charge exchange is an efficient means for converting initially positive ion beams to negative ion beams, both positive and negative ion sources are viable options for use at the facility; the choice of the type of ion source will depend on the overall efficiency for generating the radioactive species of interest. A high-temperature version of the CERN-ISOLDE positive ion source has been selected and a modified version of the source designed and fabricated for initial use at the HRIBF because of its low emittance, relatively high ionization efficiencies and species versatility, and because it has been engineered for remote installation, removal and servicing as required for safe handling in a high-radiation-level ISOL facility. Prototype plasma-sputter negative ion sources and negative surfaceionization sources are also under design consideration for generating negative radioactive ion beams from high electron-affinity elements. A brief review of the HRIBF will be presented, followed by a detailed description of the design features, operational characteristics, ionization efficiencies, and beam qualities (emittances) of these sources

  15. Radioactive ion beams and techniques for solid state research

    International Nuclear Information System (INIS)

    Correia, J.G.

    1998-01-01

    In this paper we review the most recent and new applications of solid state characterization techniques using radioactive ion beams. For such type ofresearch, high yields of chemically clean ion beams of radioactive isotopesare needed which are provided by the on-line coupling of high resolution isotope separators to particle accelerators, such as the isotope separator on-line (ISOLDE) facility at CERN. These new experiments are performed by an increasing number of solid state groups. They combine nuclear spectroscopic techniques such as Moessbauer, perturbed angular correlations (PAC) and emission channeling with the traditional non-radioactive techniques liked deep level transient spectroscopy (DLTS) and Hall effect measurements. Recently isotopes of elements, not available before, were successfully used in new PAC experiments, and the first photoluminescence (PL) measurements, where the element transmutation plays the essential role on the PL peak identification, have been performed. The scope of applications of radioactive ion beams for research in solid state physics will be enlarged in the near future, with the installation at ISOLDE of a post-accelerator device providing radioactive beams with energies ranging from a few keV up to a few MeV. (orig.)

  16. The prototype of radioactive ion source

    CERN Document Server

    Aleksandrov, A V; Kot, N K; Andrighetto, A; Stroe, L

    2001-01-01

    The design and experimental results of the RIB source prototype are presented.A source will have the container of sup 2 sup 3 sup 5 U compounds heated up to 2200-2500 degree C. Vapors of uranium fission obtained when the ion source is irradiated by the high-energy neutron flux, are then ionized and extracted from the source. In the experiments with the prototype loaded by sup 1 sup 2 C the source working temperature 2700 degree C was reached, the carbon ion current 10 nA was obtained. The total operation time of more than 100 hours with no performance degradation was demonstrated.

  17. One-step separation by thermal treatment and cobalt acid-leaching from spent lithium-ion batteries

    Science.gov (United States)

    Mu, Deying

    2017-10-01

    Lithium-ion batteries are extensively used in portable storage devices and automobiles, therefore the environment and resource problems caused by spent lithium ion batteries have become increasingly severe. This paper focuses on the recovery process of spent lithium cobalt oxide active material and comes up with reasonable processes and the best conditions for cobalt leaching ultimately.

  18. National Inventories and Management Strategies for Spent Nuclear Fuel and Radioactive Waste. Extended Methodology for the Common Presentation of Data

    International Nuclear Information System (INIS)

    Volckaert, Geert; Dionisi, Mario; Heath, Maurice; Kugel, Karin; Garamszeghy, Miklos; Leclaire, Arnaud; Deryabin, Sergey; Hedberg, Bengt; Dapei, Dominic; Lebedev, Vladimir; )

    2017-01-01

    Radioactive waste inventory data are an important element in the development of a national radioactive waste management program since these data affect the design and selection of the ultimate disposal methods. Inventory data are generally presented as an amount of radioactive waste under various waste classes, according to the waste classification scheme developed and adopted by the country or national program in question. Various waste classification schemes have evolved in most countries, and these schemes classify radioactive waste according to its origin, to criteria related to the protection of workers or to the physical, chemical and radiological properties of the waste and the planned disposal method(s). The diversity in classification schemes across countries has restricted the possibility of comparing waste inventories and led to difficulties in interpreting waste management practices, both nationally and internationally. To help improve this situation, the Nuclear Energy Agency developed a methodology that ensures consistency of national radioactive waste and spent fuel inventory data when presenting them in a common scheme in direct connection with accepted management strategy and disposal routes. This report is a follow up to the 2016 report that introduced the methodology and presenting scheme for spent fuel, and it now extends this methodology and presenting scheme to all types of radioactive waste and corresponding management strategies

  19. Improved management of SG BD demineralizer for reduced generation of low-level radioactive spent resin in Korean nuclear power plants

    International Nuclear Information System (INIS)

    Rhee, I.; Cho, D.; Yeon, J.

    2003-01-01

    Most nuclear power plants in Korea have adopted Ethanolamine(ETA) as a secondary pH control agent to increase the pH at the liquid phase, which may reduce the corrosion in steam generator tubes and moisture separator/reheat system. Along with its beneficial effect of SG protection from corrosion and degradation, the replacement of ammonia with ETA causes the increased generation of spent resin and the reduced run time of demineralizer in steam generator blowdown(SG BD) system. The composition ratio of cation- to anion- exchange resin in SG BD mixed bed should be increased in the ETA chemistry environment to meet the ratio of cation to anion in the aqueous solution, which results in the simultaneous exhaustion of cation and anion exchange resins. The utilization rate of mixed bed is greatest at the cation-to-anion ratio of 95:1 on the theoretical equivalent basis in the solution, but practically highest at that of 22:1 due to the possible inhomogeneous distribution of cation and anion exchange resins in SG BD bed. The run time of the bed could be extended by 30% such that, at that much, the purchase cost of new resin is saved and the production rate of spent resin is reduced. The guideline on the replacement of resin in SG BD bed is not necessary to secure the removal of radioactive particles without the leakage of the primary coolant into the secondary side since all the radioactive ions can be eliminated by SG BD bed with the sufficient time. They are retained during more than one month after their ingress into the SG BD bed without leakage. With the reduced replacement, thus, the SG BD spent resin that comprises 65% of low-level radioactive solid waste can be much cut down

  20. Long-term safety of radioactive waste disposal. Radioactive analysis of samples from spent fuel leaching experiments

    International Nuclear Information System (INIS)

    Geckeis, H.; Degering, D.; Goertzen, A.; Geyer, F.W.; Dressler, P.

    1995-09-01

    In order to assess the long-term performance of spent fuel during direct disposal, high burnup fuel (50 MWd/kg U) has been exposed to non-buffered brine solutions and to deionized water under static anaerobic conditions at 25 C. The leaching behaviour of several radionuclides has been observed over periods of approximately 500 d. Currently used radiometric methods (α-, β-, γ-spectrometry) were applied to the analysis of sample solutions. Due to its low specific activity, uranium was determined using ICP-mass-spectrometry (ICP-MS) or laser induced fluorescence spectrometry (LFS). In order to determine radionuclide concentrations without interferences a preceeding radiochemical separation by ion-exchange, solvent-extraction or extraction chromatography was necessary in most cases. The Sc-isotopes 134/137, which are present in a high excess over other γ-emitting nuclides, were separated using the inorganic ion exchanger ammonium molybdato phosphate (AMP). This step allowed the subsequent γ-spectrometric determination of Am-241, Ag-110m, Ru-106, Sb-125 and Eu-154/155. Activity concentrations of pure β-emitters like Sr-90, Tc-99, I-129 and Pu-241 were determined by liquid scintillation counting (LSC) after selective separation using extraction chromatography or solvent extraction. The actinides Am-241, Cm-242/244, Pu-238/239/240 and Np-237 were analysed by α-spectrometry again after selective separation. The direct analysis of uranium by LFS or ICP-MS was hampered by high salt concentrations. Therefore a separation by extraction chromatography turned out to be necessary, too. The analytical procedures used throughout this work are described in detail. (orig.) [de

  1. Selection and design of ion sources for use at the Holifield radioactive ion beam facility

    International Nuclear Information System (INIS)

    Alton, G.D.; Haynes, D.L.; Mills, G.D.; Olsen, D.K.

    1994-01-01

    The Holifield Radioactive Ion Beam Facility now under construction at the Oak Ridge National Laboratory will use the 25 MV tandem accelerator for the acceleration of radioactive ion beams to energies appropriate for research in nuclear physics; negative ion beams are, therefore, required for injection into the tandem accelerator. Because charge exchange is an efficient means for converting initially positive ion beams to negative ion beams, both positive and negative ion sources are viable options for use at the facility. The choice of the type of ion source will depend on the overall efficiency for generating the radioactive species of interest. Although direct-extraction negative ion sources are clearly desirable, the ion formation efficiencies are often too low for practical consideration; for this situation, positive ion sources, in combination with charge exchange, are the logical choice. The high-temperature version of the CERN-ISOLDE positive ion source has been selected and a modified version of the source designed and fabricated for initial use at the facility because of its low emittance, relatively high ionization efficiencies, and species versatility, and because it has been engineered for remote installation, removal, and servicing as required for safe handling in a high-radiation-level ISOL facility. The source will be primarily used to generate ion beams from elements with intermediate to low electron affinities. Prototype plasma-sputter negative ion sources and negative surface-ionization sources are under design consideration for generating radioactive ion beams from high-electron-affinity elements. The design features of these sources and expected efficiencies and beam qualities (emittances) will be described in this report

  2. Holifield Radioactive Ion Beam Facility Development and Status

    CERN Document Server

    Tatum, Alan

    2005-01-01

    The Holifield Radioactive Ion Beam Facility (HRIBF) is a national user facility dedicated to nuclear structure, reactions, and nuclear astrophysics research with radioactive ion beams (RIBs) using the isotope separator on-line (ISOL) technique. An integrated strategic plan for physics, experimental systems, and RIB production facilities have been developed and implementation of the plan is under way. Specific research objectives are defined for studying the nature of nucleonic matter, the origin of elements, solar physics, and synthesis of heavy elements. Experimental systems upgrade plans include new detector arrays and beam lines, and expansion and upgrade of existing devices. A multifaceted facility expansion plan includes a $4.75M High Power Target Laboratory (HPTL), presently under construction, to provide a facility for testing new target materials, target geometries, ion sources, and beam preparation techniques. Additional planned upgrades include a second RIB production system (IRIS2), an external axi...

  3. Removal of radioactive ions from nuclear waste solutions by electrodialysis

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, S [Radia Industries Co. Ltd., Takasaki, Gunma (Japan)

    1978-10-01

    Removal of radioactive ions was studied from low and medium level radioactive waste solutions by electrodialysis using ion exchange membranes. The test solutions contained /sup 137/Cs/sup +/, /sup 106/Ru/sup 3 +/ or fission products (F.P.) as active ions and NaCl, Na/sub 2/SO/sub 4/ or Ca(NO/sub 3/)/sub 2/ as inactive coexisting salts. The decontamination factor of the active ions was in the order: /sup 137/Cs/sup +/ (greater than 99%) > /sup 90/Sr/sup 2 +/ > F.P. > /sup 106/Ru/sup 3 +/. The dialysis time required to attain the saturation was the shortest for monovalent cations K/sup +/, Cs/sup +/ and Na/sup +/, intermediate for divalent cation Sr/sup 2 +/, and the longest for trivalent cation Ru/sup 3 +/. The ratio of the decontamination factor of an active ion eta sub( a) to the desalination factor of an inactive ion eta sub( b) was nearly equal to unity for /sup 24/Na, /sup 42/K, /sup 137/Cs and /sup 90/Sr. On the other hand, the apparent selective permeability of an active ion (A/sup +/) against Na/sup +/ ion, T sub(Na/sup +/) sup( a) was higher than unity for all the active ions tested, and was in the order of /sup 137/Cs > /sup 90/Sr > /sup 42/K > /sup 24/Na, where T sub(Na/sup +/) sup( a) is defined by the ratio of ..gamma..sub( a) to ..gamma..sub(Na/sup +/) with ..gamma..sub( a) being the ratio of dilution of A in the diluate the ..gamma..sub(Na/sup +/) being that of Na/sup +/ in the same diluate. The decontamination factor of the active ions did not depend significantly on the species and concentration of the coexistent salts or on the concentration of the active ions.

  4. Radioactive ion beam facilities in Europe

    International Nuclear Information System (INIS)

    Blumenfeld, Y.

    2008-01-01

    The past two decades have seen extraordinarily rapid development of radioactive beam physics throughout the world and in particular in Europe. The important scientific advances have stemmed from a large number of facilities. Previously existing stable beam machines have been adapted to produce rare isotope beams and dedicated facilities have come on-line. This talk gives an overview of the present European installations highlighting their complementary nature. The European roadmap calls for the construction of two next generation facilities: FAIR making use of projectile fragmentation and EURISOL based on the ISOL technique. The future FAIR facility will be described and the path towards EURISOL presented in the light of the construction of 'intermediate' generation facilities SPIRAL2, HIE ISOLDE and SPES and results from the ongoing EURISOL Design Study.

  5. Bridging nuclear safety, security and safeguards at geological disposl of high level radioactive waste and spent nuclear fuel

    International Nuclear Information System (INIS)

    Niemeyer, Irmgard; Deissmann, Guido; Bosbach, Dirk

    2016-01-01

    Findings and recommendations: • Further R&D needed to identify concepts, methods and technologies that would be best suited for the holistic consideration of safety, security and safeguards provisions of geological disposal. • 3S ‘toolbox’, including concepts, methods and technologies for: ■ material accountancy, ■ measurement techniques for spent fuel verification, ■ containment and surveillance, ■ analysis of open source information, ■ environmental sampling and monitoring, ■ continuity of knowledge, ■ design implications. •: Bridging safety, security and safeguards in research funding and research activities related to geological disposal of high-level radioactive waste and spent nuclear fuel.

  6. Novel approach to recover cobalt and lithium from spent lithium-ion battery using oxalic acid.

    Science.gov (United States)

    Zeng, Xianlai; Li, Jinhui; Shen, Bingyu

    2015-09-15

    With the booming of consumer electronics (CE) and electric vehicle (EV), a large number of spent lithium-ion battery (LIBs) have been generated worldwide. Resource depletion and environmental concern driven from the sustainable industry of CE and EV have motivated spent LIBs should be recovered urgently. However, the conventional process combined with leaching, precipitating, and filtering was quite complicated to recover cobalt and lithium from spent LIBs. In this work, we developed a novel recovery process, only combined with oxalic acid leaching and filtering. When the optimal parameters for leaching process is controlled at 150 min retention time, 95 °C heating temperature, 15 g L(-1) solid-liquid ratio, and 400 rpm rotation rate, the recovery rate of lithium and cobalt from spent LIBs can reach about 98% and 97%, respectively. Additionally, we also tentatively discovered the leaching mechanism of lithium cobalt oxide (LiCoO2) using oxalic acid, and the leaching order of the sampling LiCoO2 of spent LIBs. All the obtained results can contribute to a short-cut and high-efficiency process of spent LIBs recycling toward a sound closed-loop cycle. Copyright © 2015 Elsevier B.V. All rights reserved.

  7. Comprehensive evaluation on effective leaching of critical metals from spent lithium-ion batteries.

    Science.gov (United States)

    Gao, Wenfang; Liu, Chenming; Cao, Hongbin; Zheng, Xiaohong; Lin, Xiao; Wang, Haijuan; Zhang, Yi; Sun, Zhi

    2018-05-01

    Recovery of metals from spent lithium-ion batteries (LIBs) has attracted worldwide attention because of issues from both environmental impacts and resource supply. Leaching, for instance using an acidic solution, is a critical step for effective recovery of metals from spent LIBs. To achieve both high leaching efficiency and selectivity of the targeted metals, improved understanding on the interactive features of the materials and leaching solutions is highly required. However, such understanding is still limited at least caused by the variation on physiochemical properties of different leaching solutions. In this research, a comprehensive investigation and evaluation on the leaching process using acidic solutions to recycle spent LIBs is carried out. Through analyzing two important parameters, i.e. leaching speed and recovery rate of the corresponding metals, the effects of hydrogen ion concentration, acid species and concentration on these two parameters were evaluated. It was found that a leachant with organic acids may leach Co and Li from the cathode scrap and leave Al foil as metallic form with high leaching selectivity, while that with inorganic acids typically leach all metals into the solution. Inconsistency between the leaching selectivity and efficiency during spent LIBs recycling is frequently noticed. In order to achieve an optimal status with both high leaching selectivity and efficiency (especially at high solid-to-liquid ratios), it is important to manipulate the average leaching speed and recovery rate of metals to optimize the leaching conditions. Subsequently, it is found that the leaching speed is significantly dependent on the hydrogen ion concentration and the capability of releasing hydrogen ions of the acidic leachant during leaching. With this research, it is expected to improve understanding on controlling the physiochemical properties of a leaching solution and to potentially design processes for spent LIBs recycling with high industrial

  8. A singly charged ion source for radioactive 11C ion acceleration

    Science.gov (United States)

    Katagiri, K.; Noda, A.; Nagatsu, K.; Nakao, M.; Hojo, S.; Muramatsu, M.; Suzuki, K.; Wakui, T.; Noda, K.

    2016-02-01

    A new singly charged ion source using electron impact ionization has been developed to realize an isotope separation on-line system for simultaneous positron emission tomography imaging and heavy-ion cancer therapy using radioactive 11C ion beams. Low-energy electron beams are used in the electron impact ion source to produce singly charged ions. Ionization efficiency was calculated in order to decide the geometric parameters of the ion source and to determine the required electron emission current for obtaining high ionization efficiency. Based on these considerations, the singly charged ion source was designed and fabricated. In testing, the fabricated ion source was found to have favorable performance as a singly charged ion source.

  9. Surface and Interface Studies with Radioactive Ions

    CERN Multimedia

    Weber, A

    2002-01-01

    Investigations on the atomic scale of magnetic surfaces and magnetic multilayers were performed by Perturbed Angular Correlation (PAC) spectroscopy. The unique combination of the Booster ISOLDE facility equipped with a UHV beamline and the UHV chamber ASPIC (Apparatus for Surface Physics and Interfaces at CERN) is ideally suited for such microscopic studies. Main advantages are the choice of problem-oriented radioactive probes and the purity of mass-separated beams. The following results were obtained: $\\,$i) Magnetic hyperfine fields (B$_{hf}$) of Se on Fe, Co, Ni surfaces were determined. The results prompted a theoretical study on the B$_{hf}$ values of the 4sp-elements in adatom position on Ni and Fe, confirming our results and predicting unexpected behaviour for the other elements. $\\,$ii) Exemplarily we have determined B$_{hf}$ values of $^{111}$Cd at many different adsorption sites on Ni surfaces. We found a strong dependence on the coordination number of the probes. With decreasing coordination nu...

  10. Radioactive ion beam production challenges at the Holifield Heavy Ion Research Facility

    International Nuclear Information System (INIS)

    Meigs, M.J.; Alton, G.D.; Dowling, D.T.; Haynes, D.L.; Jones, C.M.; Juras, R.C.; Lane, S.N.; Mills, G.D.; Mosko, S.W.; Olsen, D.K.; Tatum, B.A.

    1992-01-01

    The radioactive ion beam (RIB) project at the Holifield Heavy Ion Research Facility (HHIRF) will provide for reconfiguration of the HHIRF accelerator system to enable provision of low-intensity RIBs for nuclear and astrophysics research. As we have progressed with the design of the reconfiguration, we have encountered several challenges that were not immediately obvious when first contemplating the project. The challenges do not seem insurmountable but should keep life interesting for those of us doing the work. A brief review of the project will allow a better understanding of the challenges in RIB production. Radioactive ion beams will be produced with the Isotope Separator On-Line (ISOL) postacceleration technique. In particular, radioactive atoms will be produced by reactions in the thick stopping target of an ISOL-type target-ion source assembly using intense beams from the Oak Ridge Isochronous Cyclotron equipped with a light-ion internal source. This ISOL target-ion source assembly will be mounted on a high-voltage platform with a mass separator. The target ion source will operate at potentials up to 50 kV with respect to the high voltage platform. The radioactive atoms produced by nuclear reactions in the target diffuse to the surface of the heated target material, desorb from this surface, and effuse through a heated transfer tube into an ion source where ionization and extraction take place. Two types of ion sources will be initially considered. A Forced Electron Beam Induced Arc Discharge source, similar to those used by the ISOLDE facility at CERN and by the UNISOR facility at ORNL, will be built to produce positive ions. These positive ions will be focused through an alkali vapor charge-exchange canal to produce negative ions for tandem injection. In addition, a direct negative surface ionization addition or modification to the above source will be built and investigated

  11. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 9

    Energy Technology Data Exchange (ETDEWEB)

    Klein, J.A.; Storch, S.N.; Ashline, R.C. [and others

    1994-03-01

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  12. Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10

    Energy Technology Data Exchange (ETDEWEB)

    1994-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

  13. Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10

    International Nuclear Information System (INIS)

    1994-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs

  14. Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 8

    Energy Technology Data Exchange (ETDEWEB)

    Payton, M. L.; Williams, J. T.; Tolbert-Smith, M.; Klein, J. A.

    1992-10-01

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1991. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  15. The development of a mobile hot cell facility for the conditioning of spent high activity radioactive sources

    International Nuclear Information System (INIS)

    Liebenberg, G.R.; Al-Mughrabi, M.

    2010-01-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Administration (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused Spent High Activity Radioactive Sources (SHARS) in obsolete irradiation devices such as teletherapy heads and dry irradiators. The project is initially targeting the African continent but expected soon to expand to Latin America and Asia. This hot cell allows source removal, characterization, consolidation, repackaging in modern storage shields, and secure storage of high risk SHARS at national radioactive waste storage facilities. (authors)

  16. Hydrometallurgical recovery of metal values from sulfuric acid leaching liquor of spent lithium-ion batteries.

    Science.gov (United States)

    Chen, Xiangping; Chen, Yongbin; Zhou, Tao; Liu, Depei; Hu, Hang; Fan, Shaoyun

    2015-04-01

    Environmentally hazardous substances contained in spent Li-ion batteries, such as heavy metals and nocuous organics, will pose a threat to the environment and human health. On the other hand, the sustainable recycling of spent lithium-ion batteries may bring about environmental and economic benefits. In this study, a hydrometallurgical process was adopted for the comprehensive recovery of nickel, manganese, cobalt and lithium from sulfuric acid leaching liquor from waste cathode materials of spent lithium-ion batteries. First, nickel ions were selectively precipitated and recovered using dimethylglyoxime reagent. Recycled dimethylglyoxime could be re-used as precipitant for nickel and revealed similar precipitation performance compared with fresh dimethylglyoxime. Then the separation of manganese and cobalt was conducted by solvent extraction method using cobalt loaded D2EHPA. And McCabe-Thiele isotherm was employed for the prediction of the degree of separation and the number of extraction stages needed at specific experimental conditions. Finally, cobalt and lithium were sequentially precipitated and recovered as CoC2O4 ⋅ 2H2O and Li2CO3 using ammonium oxalate solution and saturated sodium carbonate solution, respectively. Recovery efficiencies could be attained as follows: 98.7% for Ni; 97.1% for Mn, 98.2% for Co and 81.0% for Li under optimized experimental conditions. This hydrometallurgical process may promise a candidate for the effective separation and recovery of metal values from the sulfuric acid leaching liquor. Copyright © 2015 Elsevier Ltd. All rights reserved.

  17. Source study for a concept of spent nuclear fuel and radioactive waste management in the Czech Republic

    International Nuclear Information System (INIS)

    Vokal, A.; Trtilek, R.; Tarasova, J.; Popelova, E.; Podlaha, J.; Krmela, J.; Vojtechova, H.; Uhlir, J.

    2013-02-01

    The sections of the study are as follows: Purpose of the study; Basic principles of peaceful uses of nuclear energy and spent nuclear fuel and radioactive waste management; Basic starting facts and assumptions and assessment of the fulfilment of the targets; Low-level and medium-level radioactive waste management; Management of spent nuclear fuels and wastes not admitted to near-surface repositories; Description of / proposal for the methodology of transparency policy and proposal for a method to involve the entities affected, including municipalities and the public; Current financial arrangements (reserves for decommissioning, nuclear account); Refinement/modification of the plans and technical solutions; Proposed schedule and milestones for attaining the targets; Proposal for a research and development programme; Assessment of the costs of implementation of the targets. (P.A.)

  18. Degradation of ion spent resin using the Fenton's reagent; Degradacao da resina de troca ionica utilizando o reagente de Fenton

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Leandro Goulart de

    2013-07-01

    The most common method for spent radioactive ion exchange resin treatment is its immobilization in cement, which reduces the radionuclides release into the environment. Although this method is efficient, it increases considerably the final volume of the waste due to the low incorporation capacity. The objective of this work was to develop a degradation method of spent resins arising from the nuclear research reactor located at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), using an Advanced Oxidation Process (AOP) with Fenton's reagents. This method would allow a higher incorporation in cement. Three different resins were evaluated: cationic, anionic and a mixture of both resins. The reactions were conducted varying the catalyst concentration (25, 50, 100 and 150 mM), the volume of hydrogen peroxide (320 to 460 mL), and three different temperatures, 50, 60 and 70 deg C. Degradation of about 98% was achieved using a 50 mM catalyst solution and 330 mL of hydrogen peroxide solution. The most efficient temperature was 60 deg C. (author)

  19. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management

    International Nuclear Information System (INIS)

    1997-01-01

    The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management was adopted on 5 September 1997 by a Diplomatic Conference convened by the IAEA from 1 to 5 September 1997. The Joint Convention was opened for signature at Vienna on 29 September 1997 during the forty-first session of the General Conference of the IAEA. This document reproduces the text of the Convention

  20. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-24

    The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management was adopted on 5 September 1997 by a Diplomatic Conference convened by the IAEA from 1 to 5 September 1997. The Joint Convention was opened for signature at Vienna on 29 September 1997 during the forty-first session of the General Conference of the IAEA. This document reproduces the text of the Convention.

  1. Study of radio-active ions in the atmosphere

    International Nuclear Information System (INIS)

    Renoux, A.

    1965-01-01

    A comparative study is made of active, deposits of radon and thoron in suspension in the atmosphere by means of α radiation counting, using ZELENY tubes, scattering equipment, filter papers or membranes. It has been possible to show the existence of small and large ions which are negative and positive, as well as of neutral radio-active nuclei; their properties are studied. A theoretical interpretation of the results is presented. The average content of radon (using the Ra A concentration) and of Th B in the air has been determined. The radioactive equilibrium between radon and its daughter products in atmospheric air are examined. The techniques developed for active radon and thoron deposits are applied to the study of artificial radio-activity, the analyses being carried out by means of γ spectrometry. (author) [fr

  2. Nuclear astrophysics at the Holifield Radioactive Ion Beam Facility

    International Nuclear Information System (INIS)

    Smith, M.S.

    1994-01-01

    The potential for understanding spectacular stellar explosions such as novae, supernovae, and X-ray bursts will be greatly enhanced by the availability of the low-energy, high-intensity, accelerated beams of proton-rich radioactive nuclei currently being developed at the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Laboratory. These beams will be utilized in absolute cross section measurements of crucial (p, γ) capture reactions in efforts to resolve the substantial qualitative uncertainties in current models of explosive stellar hydrogen burning outbursts. Details of the nuclear astrophysics research program with the unique HRIBF radioactive beams and a dedicated experimental endstation--centered on the Daresbury Recoil Separator--will be presented

  3. Moessbauer Effect applications using intense radioactive ion beams

    International Nuclear Information System (INIS)

    Taylor, R.D.

    1990-01-01

    The Moessbauer Effect is reviewed as a promising tool for a number of new solid state studies when used in combination with radioactive beam/implantation facilities. The usual Moessbauer Effect involves long-lived radioactive parents (days to years) that populate low-lying nuclear excited states that subsequently decay to the ground state. Resonant emission/absorption of recoil-free gamma rays from these states provide information on a number of properties of the host materials. Radioactive ion beams (RIB) produced on-line allow new Moessbauer nuclei to be studied where there is no suitable parent. The technique allows useful sources to be made having extremely low local concentrations. The ability to separate the beams in both Z and A should provide high specific activity ''conventional'' sources, a feature important in some applications such as Moessbauer studies in diamond anvil high pressure cells. Exotic chemistry is proposed using RIB and certain Krypton and Xenon Moessbauer isotopes

  4. Modification of ion chromatograph for analyses of radioactive samples

    International Nuclear Information System (INIS)

    Curfman, L.L.; Johnson, S.J.

    1979-01-01

    In ion chromatographic analysis, the sample is injected through a sample loop onto an analytical column where separation occurs. The sample then passes through a suppressor column to remove or neutralize background ions. A flow-through conductivity cell is used as a detector. Depending upon column and eluent selection, ion chromatography can be used for anion or cation analyses. Ion chromatography has proven to be a versatile analytical tool for the analysis of anions in Hanford waste samples. These radioactive samples range from caustic high salt solutions to hydrochloric acid dissolutions of insoluble sludges. Instrument modifications which provide safe and convenient handling of these samples without lengthening analysis time or altering instrument performance are described

  5. Bunching and cooling of radioactive ions with REXTRAP

    CERN Document Server

    Schmidt, P; Bollen, G; Forstner, O; Huber, G; Oinonen, M; Zimmer, J

    2002-01-01

    The post-accelerator REX-ISOLDE at ISOLDE/CERN will deliver radioactive ion beams with energies up to 2.2 MeV/u. For this purpose, a Penning trap and an electron-beam ion source are combined with a linear accelerator. REXTRAP—a large gas-filled Penning trap—has started its commissioning phase. First tests have shown that REXTRAP is able to accumulate, cool and bunch stable ISOLDE ion beams covering a large mass range. Fulfilling the REX-ISOLDE demands, it can handle beam intensities from a few hundred up to 1×10 6 ions per pulse at repetition rates up to 50 Hz.

  6. Transport of radioactive ions in soil by electrokinetics

    International Nuclear Information System (INIS)

    Buehler, M.F.; Surma, J.E.; Virden, J.W.

    1994-10-01

    An electrokinetic approach is being evaluated for in situ soil remediation at the Hanford Site in Richland, Washington. This approach uses an applied electric field to induce transport of both radioactive and hazardous waste ions in soil. The work discussed in this paper involves the development of a new method to monitor the movement of the radioactive ions within the soil during the electrokinetic process. A closed cell and a gamma counter were used to provide iii situ measurements of 137 Cs and 60 Co movement in Hanford soil. Preliminary results show that for an applied potential of 200 V over approximately 200 hr, 137 Cs and 60 60 were transported a distance of 4 to 5 in. The monitoring technique demonstrated the feasibility of using electrokinetics for soil separation applications

  7. Bioleaching of valuable metals from spent lithium-ion mobile phone batteries using Aspergillus niger

    Science.gov (United States)

    Horeh, N. Bahaloo; Mousavi, S. M.; Shojaosadati, S. A.

    2016-07-01

    In this paper, a bio-hydrometallurgical route based on fungal activity of Aspergillus niger was evaluated for the detoxification and recovery of Cu, Li, Mn, Al, Co and Ni metals from spent lithium-ion phone mobile batteries under various conditions (one-step, two-step and spent medium bioleaching). The maximum recovery efficiency of 100% for Cu, 95% for Li, 70% for Mn, 65% for Al, 45% for Co, and 38% for Ni was obtained at a pulp density of 1% in spent medium bioleaching. The HPLC results indicated that citric acid in comparison with other detected organic acids (gluconic, oxalic and malic acid) had an important role in the effectiveness of bioleaching using A. niger. The results of FTIR, XRD and FE-SEM analysis of battery powder before and after bioleaching process confirmed that the fungal activities were quite effective. In addition, bioleaching achieved higher removal efficiency for heavy metals than the chemical leaching. This research demonstrated the great potential of bio-hydrometallurgical route to recover heavy metals from spent lithium-ion mobile phone batteries.

  8. Radioactive ion implantation as a tool for wear measurements

    International Nuclear Information System (INIS)

    Bagger, C.; Soerensen, G.

    1979-01-01

    The present paper deals with ion implantation of radioactive krypton ions in surfaces with aim of measuring wear of different magnetic materials in sound-heads. The technique is especially suited for a relatively fast comparison of wear-characteristics of materials of varying composition in small inaccessible areas. In the present case utilisation of a 60 KeV accelerator allows determination of a total wear as small as 0.05 μm with an accuracy of 10%. Further the technique yields information of the time dependence of the wear process with an accuracy less than 0.001 μm. (author)

  9. Techniques to produce and accelerate radioactive ion beams

    CERN Document Server

    Penescu, Liviu Constantin; Lettry, Jacques; Cata-Danil, Gheorghe

    The production and acceleration of the Radioactive Ion Beams (RIB) continues the long line of nuclear investigations started in the XIXth century by Pierre and Marie Curie, Henri Becquerel and Ernest Rutherford. The contemporary applications of the RIBs span a wide range of physics fields: nuclear and atomic physics, solid-state physics, life sciences and material science. ISOLDE is a world-leading Isotope mass-Separation On-Line (ISOL) facility hosted at CERN in Geneva for more than 40 years, offering the largest variety of radioactive ion beams with, until now, more than 1000 isotopes of more than 72 elements (with Z ranging from 2 to 88), with half-lives down to milliseconds and intensities up to 1011 ions/s. The post acceleration of the full variety of beams allows reaching final energies between 0.8 and 3.0 MeV/u. This thesis describes the development of a new series of FEBIAD (“Forced Electron Beam Induced Arc Discharge”) ion sources at CERN-ISOLDE. The VADIS (“Versatile Arc Discharge Ion Source�...

  10. Methods for estimating costs of transporting spent fuel and defense high-level radioactive waste for the civilian radioactive waste management program

    International Nuclear Information System (INIS)

    Darrough, M.E.; Lilly, M.J.

    1989-01-01

    The US Department of Energy (DOE), through the Office of Civilian Radioactive Waste Management, is planning and developing a transportation program for the shipment of spent fuel and defense high-level waste from current storage locations to the site of the mined geologic repository. In addition to its responsibility for providing a safe transportation system, the DOE will assure that the transportation program will function with the other system components to create an integrated waste management system. In meeting these objectives, the DOE will use private industry to the maximum extent practicable and in a manner that is cost effective. This paper discusses various methodologies used for estimating costs for the national radioactive waste transportation system. Estimating these transportation costs is a complex effort, as the high-level radioactive waste transportation system, itself, will be complex. Spent fuel and high-level waste will be transported from more than 100 nuclear power plants and defense sites across the continental US, using multiple transport modes (truck, rail, and barge/rail) and varying sizes and types of casks. Advance notification to corridor states will be given and scheduling will need to be coordinated with utilities, carriers, state and local officials, and the DOE waste acceptance facilities. Additionally, the waste forms will vary in terms of reactor type, size, weight, age, radioactivity, and temperature

  11. Inclusion of radioactive ion-exchange resins into inorganic binders

    International Nuclear Information System (INIS)

    Epimakhov, V.N.; Olejnik, M.S.

    2005-01-01

    The paper is devoted to inclusion of the radioactive ion-exchange resins into the portland, slag-portland and alumina cements. The degree of filling the solidified products achieves 7-10, 12 and 18.9-19.7% correspondingly under conservation of sufficient strength (not less 5 MPa). The coefficient of waste volume increasing during solidification does not exceed 1.5 under consideration of addition of 10 mass % of clay into aluminia cement [ru

  12. Application of Spent Li-Ion Batteries Cathode in Methylene Blue Dye Discoloration

    Directory of Open Access Journals (Sweden)

    Eric M. Garcia

    2017-01-01

    Full Text Available This paper aims to present the mechanism study of methylene blue (MB discoloration using spent Li-ion battery cathode tape and hydrogen peroxide. The recycled cathode used in this work is composed of 72% of LiCoO2, 18% of carbon, and 10% of Al. The value found for surface area is 8.9 m2/g and the ZCP value occurs in pH = 2.95. Different from what is proposed in the literature, the most likely mechanism of methylene blue discoloration is the oxidation/delitiation of LiCoO2 and the reduction of H2O2 forming OH∙. Thus, in this paper, an important and promising alternative for discoloration of textile industry dyes using spent Li-ion battery cathode is presented.

  13. Management of Spent Sealed Radioactive Sources in Bulgaria, Latvia, Lithuania, Romania and Slovakia

    International Nuclear Information System (INIS)

    Angus, M.; Cowley, M.; Moreton, T.; Wells, D.

    2003-01-01

    This study has been performed to consider the situation relating to the regulation and management of spent sealed radioactive sources (SSRS) in five of the Central and Eastern European (C and EE) countries that are being considered for admission to the EU, namely, Bulgaria, Latvia, Lithuania, Romania and Slovakia. Two previous studies have considered the situation in the current EU member states(1) and in the Czech Republic, Estonia, Hungary, Poland and Slovenia(2). The general aim of this study has been to acquire a thorough understanding of the management of SSRS in the five countries, in order to recommend improvements in management schemes and to establish whether the application of common disposal criteria would be advantageous. This report is structured in the following manner; following the Introduction (Section 1), there is a description of the current and proposed regulatory requirements in the EU, together with a summarised comparison of the regulatory systems in the five countries with EU standards (Section 2). Sections 3 to 7 are dedicated to each of the five countries. Each of these sections is similarly sub-divided to enable country-by-country and topic-by-topic comparison. In each of Sections 3 to 7 there is an overview, description of the sealed source inventory, regulations, current management practices, retrieval of unregistered (sometimes known as lost or orphan sources) SSRS, conclusions and a description of possible future technical assistance projects. In addition, there is a description in each of Sections 3 to 7 of the management of 226 Ra sources, which is receiving special attention in many countries (Table I provides a summary and comparison of the management of 226 Ra in the five countries).both country-specific and generic recommendations. A common concern in the five countries and many other countries, including the EU member states, is the problem of accidental inclusion of SSRS in consignments of scrap metal. The detection of

  14. Study of mechanical and physicochemical properties of cementated spent ion-exchange-resins

    International Nuclear Information System (INIS)

    Patek, P.

    1981-09-01

    As first part of a study on the possibilities, to immobilize spent ion exchange resins, for final disposal, the dependence of compressive strength from the composition of cement - resin mixtures was detected. Powdered resins, bead resins and ashes from the incinerator plant and several cement brands were examinated. As result an area was defined in the three-phase diagram of cement, resins and water, in which the following leach tests will be performed. (author)

  15. NDA measurements on spent fuel assemblies at Tihange 1 by means of the ION 1/FORK

    International Nuclear Information System (INIS)

    Carchon, R.; Smaers, G.; Verrecchia, G.P.D.; Arlt, R.; Stoyanova, I.; Satinet, J.

    1986-06-01

    This report describes field tests performed at Tihange 1 Nuclear Power Station on PWR spent fuel by means of the ION 1-FORK detector. Two detector systems and three electronics systems were used to investigate the same fuel assemblies with various burn-ups and cooling times. The purpose of the exercise was to test the performance of the instrument for as well inspection purposes as for fuel management. The results are presented and discussed. (Author)

  16. Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C., E-mail: edgar.buck@pnnl.gov; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2016-12-15

    Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO{sub 2}) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine ({sup 131/129}I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50–60 MW d/kgU, and with increasing burn-up, the IRF of {sup 131/129}I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO{sub 2} fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from {sup 239}Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO{sub 2} nuclear fuels may reduce the impact of long-lived {sup 129}I on the repository performance assessment calculations. - Highlights: • A Pd-Ag halide phase has been observed in a high burn-up UO{sub 2} reactor fuel. • The phases contains iodine and bromine. • Iodine release in high burnup fuels may be reduced through the formation of recalcitrant phases.

  17. Management of spent ion-exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    Information presented at the IAEA organized Technical Committee Meeting in December 1976 is given on the management of spent ion-exchange resins with respect to their treatment and conditioning. Currently available processes, methods and technologies such as volume reduction techniques, immobilization techniques, etc. for the treatment and conditioning are described on the basis of operating experiences. Economic aspects associated with the use, treatment, packaging and disposal of ion-exchange resins are dealt with the purpose to serve as an example of an appropriate economic evaluation. The current and prospective status of the resin disposal in USA, France, Federal Republic of Germany, United Kingdom and India is briefly discussed

  18. Engineering study for the treatment of spent ion exchange resin resulting from nuclear process applications

    International Nuclear Information System (INIS)

    Place, B.G.

    1990-09-01

    This document is an engineering study of spent ion exchange resin treatment processes with the purpose of identifying one or more suitable treatment technologies. Classifications of waste considered include all classes of low-level waste (LLW), mixed LLW, transuranic (TRU) waste, and mixed TRU waste. A total of 29 process alternatives have been evaluated. Evaluation parameters have included economic parameters (both total life-cycle costs and capital costs), demonstrated operability, environmental permitting, operational availability, waste volume reduction, programmatic consistency, and multiple utilization. The results of this study suggest that there are a number of alternative process configurations that are suitable for the treatment of spent ion exchange resin. The determinative evaluation parameters were economic variables (total life-cycle cost or capital cost) and waste volume reduction. Immobilization processes are generally poor in volume reduction. Thermal volume reduction processes tend to have high capital costs. There are immobilization processes and thermal volume reduction processes that can treat all classifications of spent ion exchange resin likely to be encountered. 40 refs., 19 figs., 17 tabs

  19. Determination of sulfur anions in spent oil shale leachates by ion chromatography

    Energy Technology Data Exchange (ETDEWEB)

    Niss, N.D.

    1989-07-01

    The leaching and transport of chemical constituents from spent oil shale disposal areas is an area of environmental concern at the present time. Sulfur-containing compounds are prevalent in spent oil shales and have the potential to leach into aqueous systems surrounding disposal sites. Computer modeling has been used in recent years to predict the transport of species in an aqueous environment. The quality of model predictions, however, depends on the validation steps taken in comparing model predictions with laboratory data on ion speciation. Further, the quality of the validation step depends on the reliability of laboratory methods in generating ion speciation data. The purpose of this study was to develop methods to separate and quantify sulfur-containing anions in spent oil shale leachates by suppressed ion chromatography. The anions studied were S{sup 2{minus}} (sulfide), SO{sup 2{minus}}{sub 3} (sulfite), SO{sup 2{minus}}{sub 4} (sulfate), SCN{sup {minus}} (thiocyanate), S{sub 2}O{sup 2{minus}}{sub 3} (thiosulfate), and S{sub 4}O{sup 2{minus}}{sub 6} (tetrathionate). After the separations were developed, a series of method-challenging experiments were performed to test the reliability of the methods and assure the development of an analytically sound product. 24 refs., 7 figs., 5 tabs.

  20. Preparation of titanates and zeolites and their uses in radioactive waste management, particularly in the treatment of spent resins

    International Nuclear Information System (INIS)

    Hultgren, Aa.; Airola, C.; Forsberg, S.; Faelt, L.

    1983-05-01

    Work on the preparation of titanates and zeolites was started and their possible uses in the management of radioactive wastes proposed in the first years of the 1970's by the Department of Nuclear Chemistry at the Royal Institute of Technology in Stockholm and the Chemical Center of the University of Lund, respectively. The major part of these efforts was sponsored by the National Council for Radioactive Waste (Prav), while the concluding phase including an over-all system design study adapted to the Swedish nuclear power programme for an economic evaluation was sponsored by the KBS Division of the Swedish Nuclear Fuel Supply Co. The preparation work resulted in processes ready for industrial scale production of sorbents of qualities adequate for applications in radioactive waste treatment. The essential effort was devoted to the treatment of spent resins from nuclear power plants by transfer to their radioactive contents to titanates and zeolites, which can then be sintered to stabel ceramic bodies (the PILO process). The economic evaluation indicated a significant economic incentive for the introduction of the PILO process if an incineration step is included for all types of spent resins. The essential efforts and results from this programme are summarized in this report. (Authors)

  1. Licensing procedures for a dedicated ship for carrying spent nuclear fuel and radioactive waste. Report from workshop held at GOSAOMNADZOR, Moscow 2 -3 July 2001

    Energy Technology Data Exchange (ETDEWEB)

    Sneve, Margorzata K.; Bergman, Curt; Markarov, Valentin

    2001-07-01

    The report describes information exchange and discussion about the licensing principles and procedures for spent nuclear fuel and radioactive waste transportation at sea. Russian health, environment and safety requirements for transportation of waste by ships. (Author)

  2. Appendix 4. Documentation of sufficient capacity facility for spent nuclear fuel and radioactive waste management and its compliance with the decommissioning strategy and schedule

    International Nuclear Information System (INIS)

    2007-01-01

    In this chapter the documentation of sufficient capacity facility for spent nuclear fuel and radioactive waste management and its compliance with the decommissioning strategy and schedule of the NPP A-1 are presented.

  3. Basic design and construction of a mobile hot cell for the conditioning of spent high activity radioactive sources

    International Nuclear Information System (INIS)

    An Hongxiang; Fan Zhiwen; Al-Mughrabi, M.

    2011-01-01

    The conditioning of spent high activity radioactive sources is one important step in sealed radioactive sources management strategies. Based on the practice on the designing of the immobilized hot cell, the handling of the sealed radioactive sources, and the reference of the mobile hot cell constructed in South Africa, SHARS conditioning process and the basic design of a mobile hot cell is developed. The mobile hot cell has been constructed and the tests including the cold test of the SRS conditioning, the hot cell assemble and disassemble and SRS recovery were done. The shielding capacity were tested by 3.8 x 10 13 Bq cobalt-60 sources and the dose rate of the equipment surface, below 2 m, is less than 0.016 mSv/h. It is proved that the designing requirement is meet and the function of the equipment is good. (authors)

  4. Review of disposal techniques for radioactively contaminated organic ion-exchange resins

    International Nuclear Information System (INIS)

    Inman, J.R.; Mack, J.

    1993-03-01

    Organic ion-exchange resins are used in the UK nuclear industry to remove radioactive nuclides from dilute aqueous solution. Resins represent a significant proportion of the organic content of ILW and LLW, particularly ILW. Spent resins are destined to be disposed of in the UK deep repository. There are concerns regarding the potential effects of organic materials on long-term repository performance, and these effects have been the subject of much recent research work. The object of this study has been to conduct a worldwide review of treatment and conditioning techniques available for spent organic ion-exchange resins with the intention of recommending the best option for dealing with the waste in the UK. Data on available techniques have been gathered together, and are presented in tabular form at the back of the report. The techniques have been reviewed and compared considering safety, practicality and cost, and a best option selected on the basis of current knowledge. On balance it would appear that wet oxidation using hydrogen peroxide with residue encapsulation in BFS/OPC is the most appropriate technique, probably implemented using a mobile plant. This conclusion and recommendation is not however clear cut and further advice regarding the acceptability of organic material in the repository is necessary before a definite recommendation can be made. (Author)

  5. National Inventories and Management Strategies for Spent Nuclear Fuel and Radioactive Waste. Methodology for Common Presentation of Data

    International Nuclear Information System (INIS)

    Volckaert, Geert; George, Mathews; Kugel, Karin; Garamszeghy, Miklos; Leclaire, Arnaud; Dionisi, Mario; Deryabin, Sergey; Lebedev, Vladimir; ); Lemmens, A.; Cairns, B.; Neri, E.G.

    2016-01-01

    Radioactive waste inventory data are an important element in the development of a national radioactive waste management programme since these data affect the design and selection of the ultimate disposal methods. Inventory data are generally presented as an amount of radioactive waste under various waste classes, according to the waste classification scheme developed and adopted by the country or national programme in question. Various waste classification schemes have thus evolved in most countries, and these schemes classify radioactive waste according to its origin, to criteria related to the protection of workers or to the physical, chemical and radiological properties of the waste and the planned disposal method(s). The diversity in classification schemes across countries has restricted the possibility of comparing waste inventories and led to difficulties in interpreting waste management practices, both nationally and internationally. To help improve this situation, the Nuclear Energy Agency proposed to develop a methodology that would ensure consistency of national radioactive waste inventory data when presenting them in a common scheme. This report provides such a methodology and presenting scheme for spent nuclear fuel and for waste arising from reprocessing. The extension of the methodology and presenting scheme to other types of radioactive waste and corresponding management strategies is envisaged in a second phase. (authors)

  6. Achievements and Perspectives of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

    International Nuclear Information System (INIS)

    Louvat, D.; Lacoste, A.C.

    2006-01-01

    The Joint Convention on the Safety of Spent Fuel management and on the Safety of Radioactive Waste Management is the first legal instrument to directly address the safety of spent fuel and radioactive waste management on a global scale. The Joint Convention entered into force in 2001. This paper describes its process and its main achievements to date. The perspectives to establish of a Global Waste Safety Regime based on the Joint Convention are also discussed. (authors)

  7. Novel approach to recover cobalt and lithium from spent lithium-ion battery using oxalic acid

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, Xianlai; Li, Jinhui, E-mail: jinhui@tsinghua.edu.cn; Shen, Bingyu

    2015-09-15

    Highlights: • Short-cut recovery of cobalt and lithium was directly obtained using oxalic acid. • Short-cut recovery process was optimized for a high recovery rate. • Leaching process was controlled by chemical reaction. • Leaching order of the sampling LiCoO{sub 2} using oxalic acid was first proposed. - Abstract: With the booming of consumer electronics (CE) and electric vehicle (EV), a large number of spent lithium-ion battery (LIBs) have been generated worldwide. Resource depletion and environmental concern driven from the sustainable industry of CE and EV have motivated spent LIBs should be recovered urgently. However, the conventional process combined with leaching, precipitating, and filtering was quite complicated to recover cobalt and lithium from spent LIBs. In this work, we developed a novel recovery process, only combined with oxalic acid leaching and filtering. When the optimal parameters for leaching process is controlled at 150 min retention time, 95 °C heating temperature, 15 g L{sup −1} solid–liquid ratio, and 400 rpm rotation rate, the recovery rate of lithium and cobalt from spent LIBs can reach about 98% and 97%, respectively. Additionally, we also tentatively discovered the leaching mechanism of lithium cobalt oxide (LiCoO{sub 2}) using oxalic acid, and the leaching order of the sampling LiCoO{sub 2} of spent LIBs. All the obtained results can contribute to a short-cut and high-efficiency process of spent LIBs recycling toward a sound closed-loop cycle.

  8. Development of geological disposal system for spent fuels and high-level radioactive wastes in Korea

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

    2013-01-01

    Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

  9. DEVELOPMENT OF GEOLOGICAL DISPOSAL SYSTEMS FOR SPENT FUELS AND HIGH-LEVEL RADIOACTIVE WASTES IN KOREA

    Directory of Open Access Journals (Sweden)

    HEUI-JOO CHOI

    2013-02-01

    Full Text Available Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

  10. Molten salt oxidation of ion-exchange resins doped with toxic metals and radioactive metal surrogates

    International Nuclear Information System (INIS)

    Yang, Hee-Chul; Cho, Yong-Jun; Yoo, Jae-Hyung; Kim, Joon-Hyung; Eun, Hee-Chul

    2005-01-01

    Ion-exchange resins doped with toxic metals and radioactive metal surrogates were test-burned in a bench-scale molten salt oxidation (MSO) reactor system. The purposes of this study are to confirm the destruction performance of the two-stage MSO reactor system for the organic ion-exchange resin and to obtain an understanding of the behavior of the fixed toxic metals and the sulfur in the cationic exchange resins. The destruction of the organics is very efficient in the primary reactor. The primarily destroyed products such as carbon monoxide are completely oxidized in the secondary MSO reactor. The overall collection of the sulfur and metals in the two-stage MSO reactor system appeared to be very efficient. Over 99.5% of all the fixed toxic metals (lead and cadmium) and radioactive metal surrogates (cesium, cobalt, strontium) remained in the MSO reactor bottom. Thermodynamic equilibrium calculations and the XRD patterns of the spent salt samples revealed that the collected metals existed in the form of each of their carbonates or oxides, which are non-volatile species at the MSO system operating conditions. (author)

  11. Plasma arc pyrolysis of radioactive ion exchange resin

    International Nuclear Information System (INIS)

    Pickles, C.A.; Toguri, J.M.

    1992-01-01

    This paper reports on two ion exchange resins (IRN 77 and IRN 78) which were pyrolysed in a plasma-arc furnace. Both continuous and batch tests were performed. Volume reduction ratios of 10 to 1 and 10 to 3.5 were achieved for IRN 78 and IRN 77 respectively. The product of the resin pyrolysis was a char which contained the radioactive elements such as cobalt. The off-gases consisted of mainly hydrogen and carbon monoxide. There was a relatively small amount of dust in the off-gases. At the present time radioactive ion exchange resign is being kept in storage. The volume of this waste is increasing and it is important that the volume be reduce. The volume reduction ratio should be of the order of ten-to-one. Also, it is required that the radioactive elements can be collected or fixed in a form which could easily be disposed of. Plasma arc treatment offers considerable potential for the processing of the waste

  12. Spent fuel storage requirements for nuclear utilities and OCRWM [Office of Civilian Radioactive Waste Management

    International Nuclear Information System (INIS)

    Wood, T.W.

    1990-03-01

    Projected spent fuel generation at US power reactors exceeds estimated aggregate pool storage capacity by approximately 30,000 metric tons of uranium (MTU). Based on the current repository schedule, little of the spent fuel inventory will be disposed of prior to shutdown of existing reactors, and a large additional capacity for surface storage of spent fuel will be required, either at reactors or at a centralized DOE storage site. Allocation of this storage requirement across the utility-DOE interface, and the resulting implications for reactor sites and the performance of the federal waste management system, were studied during the DOE MRS System Study and again subsequent to the reassessment of the repository schedule. Spent fuel logistics and cost results from these analyses will be used in definition of spent fuel storage capacity requirements for the federal system. 9 refs., 8 figs., 1 tab

  13. Dehydrating process experiment on spent ion-exchange resin sludge by Funda Filter

    International Nuclear Information System (INIS)

    Hasegawa, Tatsuo; Ishino, Kazuyuki

    1977-01-01

    In nuclear power plants, Funda Filters are employed to dehydrate spent powdery ion-exchange resin sludge. The Funda Filter is very effective for eliminating small rust components contained in spent powdery resin slurry; however, in the drying process, the complete drying of spent powdery resin is very difficult because the filter cake of resin on the horizontal filter leaf is likely to crack and let out steam and hot air through the cracks. This paper deals with the results of experiments conducted to clarify the detailed phenomena of dehydration so the above problem could be solved. The above experiments were made on the precoating and drying of granular ion-exchange resin slurry that had not yet been put to practical use. The experiments were composed of one fundamental and one operational stage. In the fundamental experiment, the dehydration properties and dehydration mechanism of resins were made clear, and the most effective operational method was established through the operational experiments conducted using large-scale Funda Filter test equipment under various conditions. (auth.)

  14. Hydrometallurgical recovery of metal values from sulfuric acid leaching liquor of spent lithium-ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiangping; Chen, Yongbin; Zhou, Tao, E-mail: zhoutao@csu.edu.cn; Liu, Depei; Hu, Hang; Fan, Shaoyun

    2015-04-15

    Highlights: • Selective precipitation and solvent extraction were adopted. • Nickel, cobalt and lithium were selectively precipitated. • Co-D2EHPA was employed as high-efficiency extraction reagent for manganese. • High recovery percentages could be achieved for all metal values. - Abstract: Environmentally hazardous substances contained in spent Li-ion batteries, such as heavy metals and nocuous organics, will pose a threat to the environment and human health. On the other hand, the sustainable recycling of spent lithium-ion batteries may bring about environmental and economic benefits. In this study, a hydrometallurgical process was adopted for the comprehensive recovery of nickel, manganese, cobalt and lithium from sulfuric acid leaching liquor from waste cathode materials of spent lithium-ion batteries. First, nickel ions were selectively precipitated and recovered using dimethylglyoxime reagent. Recycled dimethylglyoxime could be re-used as precipitant for nickel and revealed similar precipitation performance compared with fresh dimethylglyoxime. Then the separation of manganese and cobalt was conducted by solvent extraction method using cobalt loaded D2EHPA. And McCabe–Thiele isotherm was employed for the prediction of the degree of separation and the number of extraction stages needed at specific experimental conditions. Finally, cobalt and lithium were sequentially precipitated and recovered as CoC{sub 2}O{sub 4}⋅2H{sub 2}O and Li{sub 2}CO{sub 3} using ammonium oxalate solution and saturated sodium carbonate solution, respectively. Recovery efficiencies could be attained as follows: 98.7% for Ni; 97.1% for Mn, 98.2% for Co and 81.0% for Li under optimized experimental conditions. This hydrometallurgical process may promise a candidate for the effective separation and recovery of metal values from the sulfuric acid leaching liquor.

  15. Hydrometallurgical recovery of metal values from sulfuric acid leaching liquor of spent lithium-ion batteries

    International Nuclear Information System (INIS)

    Chen, Xiangping; Chen, Yongbin; Zhou, Tao; Liu, Depei; Hu, Hang; Fan, Shaoyun

    2015-01-01

    Highlights: • Selective precipitation and solvent extraction were adopted. • Nickel, cobalt and lithium were selectively precipitated. • Co-D2EHPA was employed as high-efficiency extraction reagent for manganese. • High recovery percentages could be achieved for all metal values. - Abstract: Environmentally hazardous substances contained in spent Li-ion batteries, such as heavy metals and nocuous organics, will pose a threat to the environment and human health. On the other hand, the sustainable recycling of spent lithium-ion batteries may bring about environmental and economic benefits. In this study, a hydrometallurgical process was adopted for the comprehensive recovery of nickel, manganese, cobalt and lithium from sulfuric acid leaching liquor from waste cathode materials of spent lithium-ion batteries. First, nickel ions were selectively precipitated and recovered using dimethylglyoxime reagent. Recycled dimethylglyoxime could be re-used as precipitant for nickel and revealed similar precipitation performance compared with fresh dimethylglyoxime. Then the separation of manganese and cobalt was conducted by solvent extraction method using cobalt loaded D2EHPA. And McCabe–Thiele isotherm was employed for the prediction of the degree of separation and the number of extraction stages needed at specific experimental conditions. Finally, cobalt and lithium were sequentially precipitated and recovered as CoC 2 O 4 ⋅2H 2 O and Li 2 CO 3 using ammonium oxalate solution and saturated sodium carbonate solution, respectively. Recovery efficiencies could be attained as follows: 98.7% for Ni; 97.1% for Mn, 98.2% for Co and 81.0% for Li under optimized experimental conditions. This hydrometallurgical process may promise a candidate for the effective separation and recovery of metal values from the sulfuric acid leaching liquor

  16. Geologic disposal as optimal solution of managing the spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Ilie, P.; Didita, L.; Ionescu, A.; Deaconu, V.

    2002-01-01

    To date there exist three alternatives for the concept of geological disposal: 1. storing the high-level waste (HLW) and spent nuclear fuel (SNF) on ground repositories; 2. solutions implying advanced separation processes including partitioning and transmutation (P and T) and eventual disposal in outer space; 3. geological disposal in repositories excavated in rocks. Ground storing seems to be advantageous as it ensures a secure sustainable storing system over many centuries (about 300 years). On the other hand ground storing would be only a postponement in decision making and will be eventually followed by geological disposal. Research in the P and T field is expected to entail a significant reduction of the amount of long-lived radioactive waste although the long term geological disposal will be not eliminated. Having in view the high cost, as well as the diversity of conditions in the countries owning power reactors it appears as a reasonable regional solution of HLW disposal that of sharing a common geological disposal. In Romania legislation concerning of radioactive waste is based on the Law concerning Spent Nuclear Fuel and Radioactive Waste Management in View of Final Disposal. One admits at present that for Romania geological disposal is not yet a stressing issue and hence intermediate ground storing of SNF will allow time for finding a better final solution

  17. Assessment of the Proposed Design of a New Spent Sealed Radioactive Sources Storage Facility at Novi Han

    International Nuclear Information System (INIS)

    Alardin, J.M.; Lacroix, J.P.; Glibert, R.; Marneffe, L. de

    2001-09-01

    The NOVI HAN radioactive waste repository (NHRWR) in Bulgaria, built according to a Soviet design, was commissioned in 1964. The State Committee on the Use of Atomic Energy for Peaceful Purposes (CUAEPP) temporarily stopped operations at the repository from October 1994 until measures for improvement of the facility are undertaken. Since 1994, the Spent Sealed Radioactive Sources (SSRS) have been temporarily stored at the facilities at IRT-2000 research reactor of the Bulgarian Academy of Sciences (BAS) in Sofia. In view of the importance of the radiological risks associated with the present management of the SSRS in Bulgaria, the present study contract has been launched to critically review the proposal to provide a new interim storage facility for SSRS at NHRWR. A comprehensive critical review was performed of the feasibility study for the construction of a new SSRS facility at Novi Han, carried out by the local consultant engineering company (EQE), and detailed recommendations were made concerning the proposed new development at the site. The authors think that new concepts and procedures in the management of all categories of SSRS including smoke detectors have to be introduced, taking into account the regulatory framework and the inventories of existing and anticipated SSRS. This should be the basis for the technical specification of the new facilities for conditioning and storage of spent sealed radioactive sources (not only SHARS). (author)

  18. Mössbauer and Raman spectroscopy characterization of concretes used in the conditioning of spent radioactive sources

    Energy Technology Data Exchange (ETDEWEB)

    Monroy-Guzman, F., E-mail: fabiola.monroy@inin.gob.mx; González-Neri, M.; González-Díaz, R. C.; Ortíz-Arcivar, G.; Corona-Pérez, I. J. [Instituto Nacional de Investigaciones Nucleares. Carretera México-Toluca s/n, La Marquesa, Ocoyoacac (Mexico); Nava, N. [Instituto Mexicano del Petroleo (Mexico); Cabral-Prieto, A.; Escobar-Alarcón, L. [Instituto Nacional de Investigaciones Nucleares. Carretera México-Toluca s/n, La Marquesa, Ocoyoacac (Mexico)

    2015-06-15

    Spent radioactive sources are considered a type of radioactive waste which must be stored properly. These sources are usually conditioned in concrete that functions as shield and physical barrier to prevent the potential migration of radionuclides, and must have suitable properties: mechanical, thermal or irradiation resistance. Concretes used in the conditioning of spent radioactive source in Mexico were tested, preparing concrete test specimens with Portland cement CPC 30RS EXTRA CEMEX and aggregates, and subjected to compression strength, γ-ray-irradiation and thermal resistance assays and subsequently analyzed by Mössbauer and Raman Spectroscopies as well as by Scanning Electron Microscopy, in order to correlate the radiation and temperature effects on the compressive strengths, the oxidation states of iron and the structural features of the concrete. Iron was found in the concrete in Fe {sup 2+} and Fe {sup 3+} in the tetrahedral (T) and two octahedral positions (O1, O2). Radiolysis of water causes the dehydratation (200-600 kGy) and rehydratation (1000-10000 kGy) of calcium silicate hydrates (C-S-H) and ferric hydrate phases in concretes and structural distortion around the iron sites in concretes. The compressive strength of concretes are not significantly affected by γ-radiation or heat.

  19. Positron emission medical measurements with accelerated radioactive ion beams

    International Nuclear Information System (INIS)

    Llacer, J.

    1988-01-01

    This paper reviews in some detail the process by which a heavy ion accelerator can be used to inject positron emitting radioactive particles into a human body for a range of possible medical measurements. The process of radioactive beam generation and injection is described, followed by a study of the relationship between activity that can be injected versus dose to the patient as a function of which of the positron emitting ions is used. It is found that 6 C 10 and 10 Ne 19 are the two isotopes that appear more promising for injection into humans. The design considerations for a non-tomographic instrument to obtain images from beam injections are outlined and the results of 10 Ne 19 preliminary measurements with human phantoms and actual patients for the determination of end-of-range of cancer therapy ion beams is reported. Accuracies in the order of ±1 mm in the measurements of stopping point of a therapy beam with safe doses to the patient are reported. The paper concludes with a simple analysis of requirements to extend the technique to on-line verification of cancer treatment and to nuclear medicine research and diagnostics measurements. 17 refs.; 16 figs.; 3 tabs

  20. Applications of inorganic ion-exchange materials in managing radioactivity wastewater

    International Nuclear Information System (INIS)

    He Jiaheng; Li Xingliang; Li Shoujian

    2007-01-01

    This article introduces the application of abio-ion exchange materials in managing radioactivity wastewater, which would be useful for latter research of new inorganic materials that used in managing radioactivity wastewater. (authors)

  1. Radioactive ion beam development for the SPIRAL 2 project

    International Nuclear Information System (INIS)

    Pichard, A.

    2010-01-01

    This thesis focuses on the study of radioactive ion beam production by the ISOL method for the SPIRAL 2 project. The production of light ion beams is studied and the potential in-target yields of two beams are appraised. The neutron-rich 15 C yield in an oxide target is estimated with simulations (MCNPx, EAF-07) and experimental data bases; the neutron-deficient 14 O yield is estimated thanks to a new measurement of the 12 C( 3 He, n) 14 O reaction excitation function. Based on thermal simulations, a first design of the production target is presented. This thermal study gives the necessary answers for the detailed design of the system able to reach a production yield 140 times higher than with SPIRAL 1. The production of radioactive ion beams coming from fissions in the UCx target is also studied and more particularly effusion and ionisation processes. A global study and an off-line tests campaign allow essential knowledge to the design of the surface ionisation source for SPIRAL 2 to be acquired. A first prototype of this ion source dedicated to alkali and alkaline-earth element production has been built and a thermal calibration performed. Ionisation efficiency and time response of the target-ion source system have been measured at different target temperatures and for different noble gases. These measurements allow evaluation of the impact of effusion and ionisation processes on the production efficiency of different alkali and noble gases isotopes as a function of their half-life. (author) [fr

  2. Selection of targets and ion sources for RIB generation at the Holifield Radioactive Ion Beam Facility

    International Nuclear Information System (INIS)

    Alton, G.D.

    1995-01-01

    In this report, the authors describe the performance characteristics for a selected number of target ion sources that will be employed for initial use at the Holifield Radioactive Ion Beam Facility (HRIBF) as well as prototype ion sources that show promise for future use for RIB applications. A brief review of present efforts to select target materials and to design composite target matrix/heat-sink systems that simultaneously incorporate the short diffusion lengths, high permeabilities, and controllable temperatures required to effect fast and efficient diffusion release of the short-lived species is also given

  3. Development of a radioactive ion beam test stand at LBNL

    International Nuclear Information System (INIS)

    Burke, J.; Freedman, S.J.; Fujikawa, B.; Gough, R.A.; Lyneis, C.M.; Vetter, P.; Wutte, D.; Xie, Z.Q.

    1998-01-01

    For the on-line production of a 14 O + ion beam, an integrated target--transfer line ion source system is now under development at LBNL. 14 O is produced in the form of CO in a high temperature carbon target using a 20 MeV 3 He beam from the LBNL 88'' Cyclotron via the reaction 12 C( 3 He,n) 14 O. The neutral radioactive CO molecules diffuse through an 8 m room temperature stainless steel line from the target chamber into a cusp ion source. The molecules are dissociated, ionized and extracted at energies of 20 to 30 keV and mass separated with a double focusing bending magnet. The different components of the setup are described. The release and transport efficiency for the CO molecules from the target through the transfer line was measured for various target temperatures. The ion beam transport efficiencies and the off-line ion source efficiencies for Ar, O 2 and CO are presented. Ionization efficiencies of 28% for Ar + , 1% for CO, 0.7% for O + , 0.33 for C + have been measured

  4. Targets for ion sources for RIB generation at the Holifield Radioactive Ion Beam Facility

    International Nuclear Information System (INIS)

    Alton, G.D.

    1995-01-01

    The Holifield Radioactive Ion Beam Facility (HRIBF), now under construction at the Oak Ridge National Laboratory, is based on the use of the well-known on-line isotope separator (ISOL) technique in which radioactive nuclei are produced by fusion type reactions in selectively chosen target materials by high-energy proton, deuteron, or He ion beams from the Oak Ridge Isochronous Cyclotron (ORIC). Among several major challenges posed by generating and accelerating adequate intensities of radioactive ion beams (RIBs), selection of the most appropriate target material for production of the species of interest is, perhaps, the most difficult. In this report, we briefly review present efforts to select target materials and to design composite target matrix/heat-sink systems that simultaneously incorporate the short diffusion lengths, high permeabilities, and controllable temperatures required to effect maximum diffusion release rates of the short-lived species that can be realized at the temperature limits of specific target materials. We also describe the performance characteristics for a selected number of target ion sources that will be employed for initial use at the HRIBF as well as prototype ion sources that show promise for future use for RIB applications

  5. Evaluation of the safe disposal of radioactive waste and spent fuel in the vicinity of Bashmachnaya Bay on the Novaya Zemlya Archipelago

    International Nuclear Information System (INIS)

    Melnikov, N.N.; Konukirn, V.P.; Komlev, V.N.

    1998-01-01

    The problem of disposing of the radioactive waste and spent fuel accumulated in north-west Russia, in the Novaya Zemlya Archipelago, is discussed in this paper. An analysis of risk factors associated with the construction of a pilot facility is given. The investigations carried out are considered to be insufficient in the context of guaranteeing long-term isolation of radioactive waste and spent fuel in permafrost carbonate rocks within Bashmachnaya Bay. Recommendations are proposed concerning the scientific approach to solving this radioactive waste management problem. 5 refs

  6. Safety aspects of the cleaning and conditioning of radioactive sludge from spent fuel storage pool on 'RA' Research reactor in the Vinca Institute

    International Nuclear Information System (INIS)

    Pavlovic, R; Pavlovic, S.; Plecas, I.

    1999-01-01

    Spent fuel elements from nuclear reactors in the Vinca Institute have been temporary stored in water filled storage pool. Due to the fact that the water in the spent fuel elements storage pool have not been purified for a long time, all metallic components submerged in the water have been hardly corroded and significant amount of the sludge has been settled on the bottom of the pool. As a first step in improving spent fuel elements storage conditions and slowing down corrosion in the storage spent fuel elements pool we have decided to remove the sludge from the bottom of the pool. Although not high, but slightly radioactive, this sludge had to be treated as radioactive waste material. Some safety aspects and radiation protection measures in the process of the spent fuel storage pool cleaning are presented in this paper

  7. Update on ASME rules for spent nuclear fuel and high level radioactive material and waste storage containments

    International Nuclear Information System (INIS)

    Ralph S. Hill III; Foster, G.M.

    2005-01-01

    In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) was published. The Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. The Code Case has been incorporated into Section III of the Code as Division 3, Subsection WC, Class SC Storage Containments, and will be published in the 2005 Addenda. This paper provides an informative background and insight for these rules to provide Owners, regulators, designers, and fabricators with a more comprehensive understanding of the technical basis for these rules. (authors)

  8. Direct Reaction Experimental Studies with Beams of Radioactive Tin Ions

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K. L. [University of Tennessee, Knoxville (UTK); Ahn, S.H. [University of Tennessee, Knoxville (UTK); Allmond, James M [ORNL; Ayres, A. [University of Tennessee, Knoxville (UTK); Bardayan, Daniel W [ORNL; Baugher, T. [Michigan State University, East Lansing; Bazin, D. [Michigan State University, National Superconducting Cyclotron Laboratory (NSCL); Beene, James R [ORNL; Berryman, J. S. [Michigan State University, East Lansing; Bey, A. [University of Tennessee, Knoxville (UTK); Bingham, C. R. [University of Tennessee, Knoxville (UTK); Cartegni, L. [University of Tennessee, Knoxville (UTK); Chae, K. Y. [University of Tennessee, Knoxville (UTK)/Sungkyunkwan University, Korea; Cizewski, J. A. [Rutgers University; Gade, A. [Michigan State University, National Superconducting Cyclotron Laboratory (NSCL); Galindo-Uribarri, Alfredo {nmn} [ORNL; Garcia-Ruiz, R.F. [Instituut voor Kernen Stralingsfysica, KU Leuven, B-3001, Leuven, Belgium; Grzywacz, Robert Kazimierz [ORNL; Howard, Meredith E [ORNL; Kozub, R. L. [Tennessee Technological University (TTU); Liang, J Felix [ORNL; Manning, Brett M [ORNL; Matos, M. [Louisiana State University; McDaniel, S. [Michigan State University, East Lansing; Miller, D. [University of Tennessee, Knoxville (UTK); Nesaraja, Caroline D [ORNL; O' Malley, Patrick [Rutgers University; Padgett, S [University of Tennessee, Knoxville (UTK); Padilla-Rodal, Elizabeth [Universidad Nacional Autonoma de Mexico (UNAM); Pain, Steven D [ORNL; Pittman, S. T. [University of Tennessee (UTK) and Oak Ridge National Laboratory (ORNL); Radford, David C [ORNL; Ratkiewicz, Andrew J [ORNL; Schmitt, Kyle [ORNL; Smith, Michael Scott [ORNL; Stracener, Daniel W [ORNL; Stroberg, S. [Michigan State University, East Lansing; Tostevin, Jeffrey A [ORNL; Varner Jr, Robert L [ORNL; Weisshaar, D. [Michigan State University, East Lansing; Wimmer, K. [Michigan State University, National Superconducting Cyclotron Laboratory (NSCL)/Central Michigan University; Winkler, R. [Michigan State University, East Lansing

    2015-01-01

    The tin chain of isotopes provides a unique region in which to investigate the evolution of single-particle structure, spreading from N = 50 at Sn-100, through 10 stable isotopes and the N = 82 shell closure at Sn-132 out into the r-process path. Direct reactions performed on radioactive ion beams are sensitive spectroscopic tools for studying exotic nuclei. Here we present one experiment knocking out neutrons from tin isotopes that are already neutron deficient and two reactions that add a neutron to neutron-rich Sn-130. Both techniques rely on selective particle identification and the measurement of gamma rays in coincidence with charged ions. We present the goals of the two experiments and the particle identification for the channels of interest. The final results will be presented in future publications.

  9. Behaviour of Aluminosilicate Inorganic Matrix Sial During and After Solidification of Radioactive Sludge And Radioactive Spent Resins and Their Mixtures

    International Nuclear Information System (INIS)

    Lichvar, P.; Rozloznik, M.; Sekely, S.

    2013-01-01

    SIAL matrix is the product of inorganic compounds by polycondensation. The inorganic SIAL matrix was subject of study from the atomic level to the level of technological application. Infrared spectroscopy measurement and X-ray analysis confirmed that SIAL matrix is mostly formed by amorphous phase. This satisfies the required criteria for disposal of immobilized radioactive waste of appropriate properties into surface repository (Slovak and Czech Republic). First of all, leaching of radioactive nuclide Cs-137 and compressive strength of the final product were found. Various remotely operated devices have been designed, manufactured and successfully used for the radioactive sludge and sludge/resins mixture solidification at the Nuclear power plant (NPP) A-1, V-2 in Jaslovske Bohunice, at Mochovce NPP (Slovak Republic), at Dukovany NPP and at Temelin NPP (Czech Republic). The SIAL matrix was approved for waste package for sludge/resins mixture by the Slovak and Czech Nuclear Regulatory Authorities. (author)

  10. Behaviour of Aluminosilicate Inorganic Matrix Sial During and After Solidification of Radioactive Sludge And Radioactive Spent Resins and Their Mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Lichvar, P.; Rozloznik, M.; Sekely, S. [Amec Nuclear Slovakia, 919 31 Jaslovske Bohunice (Slovakia)

    2013-09-15

    SIAL matrix is the product of inorganic compounds by polycondensation. The inorganic SIAL matrix was subject of study from the atomic level to the level of technological application. Infrared spectroscopy measurement and X-ray analysis confirmed that SIAL matrix is mostly formed by amorphous phase. This satisfies the required criteria for disposal of immobilized radioactive waste of appropriate properties into surface repository (Slovak and Czech Republic). First of all, leaching of radioactive nuclide Cs-137 and compressive strength of the final product were found. Various remotely operated devices have been designed, manufactured and successfully used for the radioactive sludge and sludge/resins mixture solidification at the Nuclear power plant (NPP) A-1, V-2 in Jaslovske Bohunice, at Mochovce NPP (Slovak Republic), at Dukovany NPP and at Temelin NPP (Czech Republic). The SIAL matrix was approved for waste package for sludge/resins mixture by the Slovak and Czech Nuclear Regulatory Authorities. (author)

  11. A linear radiofrequency ion trap for accumulation, bunching, and emittance improvement of radioactive ion beams

    International Nuclear Information System (INIS)

    Herfurth, F.; Dilling, J.; Kellerbauer, A.

    2000-05-01

    An ion beam cooler and buncher has been developed for the manipulation of radioactive ion beams. The gas-filled linear radiofrequency ion trap system is installed at the Penning trap mass spectrometer ISOLTRAP at ISOLDE/CERN. Its purpose is to accumulate the 60-keV continuous ISOLDE ion beam with high efficiency and to convert it into low-energy low-emittance ion pulses. The efficiency was found to exceed 10% in agreement with simulations. A more than 10-fold reduction of the ISOLDE beam emittance can be achieved. The system has been used successfully for first on-line experiments. Its principle, setup and performance will be discussed. (orig.)

  12. The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

    International Nuclear Information System (INIS)

    Risoluti, P.

    2004-01-01

    The Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management (the Joint Convention) is the only legally binding international treaty in the area of radioactive waste management. It was adopted by a Diplomatic Conference in September 1997 and opened for signature on 29 September 1997. The Convention entered into force on 18 June 1998, and to date (September 04) has been signed by 42 States, of which 34 have formally ratified, thus becoming Contracting Parties. The Joint Convention applies to spent fuel and radioactive waste resulting from civilian application. Its principal aim is to achieve and maintain a high degree of safety in their management worldwide. The Convention is an incentive instrument, not designed to ensure fulfillment of obligations through control and sanction, but by a peer pressure. The obligations of the Contracting Parties are mainly based on the international safety standards developed by the IAEA in past decades. The Convention is intended for all countries generating radioactive waste. Therefore it is relevant not only for those using nuclear power, but for any country where application of nuclear energy in medicine, conventional industry and research is currently used. Obligations of Contracting Parties include attending periodic Review Meetings and prepare National Reports for review by the other Contracting Parties. The National Reports should describe how the country is complying with the requirements of the Articles of the Convention. The first such meeting was held at the IAEA headquarters in November 2003. This paper will describe the origin of the Convention, present its content, the expected outcome for the worldwide safety, and the benefits for a country to be part of it

  13. A copper-catalyzed bioleaching process for enhancement of cobalt dissolution from spent lithium-ion batteries

    International Nuclear Information System (INIS)

    Zeng, Guisheng; Deng, Xiaorong; Luo, Shenglian; Luo, Xubiao; Zou, Jianping

    2012-01-01

    Highlights: ► Catalytic ion was first applied to the bioleaching process of spent lithium-ion batteries. ► The bioleaching efficiency was great improved from 43.1% to 99.9% in the presence of copper ion. ► A new reaction model was proposed to explain the catalytic mechanism. - Abstract: A copper-catalyzed bioleaching process was developed to recycle cobalt from spent lithium-ion batteries (mainly LiCoO 2 ) in this paper. The influence of copper ions on bioleaching of LiCoO 2 by Acidithiobacillus ferrooxidans (A.f) was investigated. It was shown that almost all cobalt (99.9%) went into solution after being bioleached for 6 days in the presence of 0.75 g/L copper ions, while only 43.1% of cobalt dissolution was obtained after 10 days without copper ions. EDX, XRD and SEM analyses additionally confirmed that the cobalt dissolution from spent lithium-ion batteries could be improved in the presence of copper ions. The catalytic mechanism was investigated to explain the enhancement of cobalt dissolution by copper ions, in which LiCoO 2 underwent a cationic interchange reaction with copper ions to form CuCo 2 O 4 on the surface of the sample, which could be easily dissolved by Fe 3+ .

  14. A copper-catalyzed bioleaching process for enhancement of cobalt dissolution from spent lithium-ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, Guisheng, E-mail: zengguisheng@hotmail.com [Key Laboratory of Jiangxi Province for Ecological Diagnosis-Remediation and Pollution Control, Nanchang Hangkong University, Nanchang 330063 (China); College of Chemistry and Chemical Engineering, Hunan University, Changsha 410082 (China); Deng, Xiaorong [Key Laboratory of Jiangxi Province for Ecological Diagnosis-Remediation and Pollution Control, Nanchang Hangkong University, Nanchang 330063 (China); School of Environment and Chemical Engineering, Nanchang Hangkong University, Nanchang 330063 (China); Luo, Shenglian, E-mail: sllou@hnu.edu.cn [Key Laboratory of Jiangxi Province for Ecological Diagnosis-Remediation and Pollution Control, Nanchang Hangkong University, Nanchang 330063 (China); College of Chemistry and Chemical Engineering, Hunan University, Changsha 410082 (China); School of Environment and Chemical Engineering, Nanchang Hangkong University, Nanchang 330063 (China); Luo, Xubiao; Zou, Jianping [Key Laboratory of Jiangxi Province for Ecological Diagnosis-Remediation and Pollution Control, Nanchang Hangkong University, Nanchang 330063 (China); School of Environment and Chemical Engineering, Nanchang Hangkong University, Nanchang 330063 (China)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer Catalytic ion was first applied to the bioleaching process of spent lithium-ion batteries. Black-Right-Pointing-Pointer The bioleaching efficiency was great improved from 43.1% to 99.9% in the presence of copper ion. Black-Right-Pointing-Pointer A new reaction model was proposed to explain the catalytic mechanism. - Abstract: A copper-catalyzed bioleaching process was developed to recycle cobalt from spent lithium-ion batteries (mainly LiCoO{sub 2}) in this paper. The influence of copper ions on bioleaching of LiCoO{sub 2} by Acidithiobacillus ferrooxidans (A.f) was investigated. It was shown that almost all cobalt (99.9%) went into solution after being bioleached for 6 days in the presence of 0.75 g/L copper ions, while only 43.1% of cobalt dissolution was obtained after 10 days without copper ions. EDX, XRD and SEM analyses additionally confirmed that the cobalt dissolution from spent lithium-ion batteries could be improved in the presence of copper ions. The catalytic mechanism was investigated to explain the enhancement of cobalt dissolution by copper ions, in which LiCoO{sub 2} underwent a cationic interchange reaction with copper ions to form CuCo{sub 2}O{sub 4} on the surface of the sample, which could be easily dissolved by Fe{sup 3+}.

  15. Practical experience for liquid radioactive waste treatment from spent fuel storage pool on RA reactor in Vinca Institute

    International Nuclear Information System (INIS)

    Plecas, I.; Pavlovic, R.; Pavlovic, S.

    2002-01-01

    The present paper reports the results of the preliminary removal of sludge from the bottom of the spent fuel storage pool in the RA reactor, mechanical filtration of the pool water and sludge conditioning and storage. Yugoslavia is a country without a nuclear power plant (NPP) on its territory. The law which strictly forbids NPP construction is still valid, but, nevertheless we must handle and dispose radioactive waste. This is not only because of radwaste originating from the use of radioactive materials in medicine and industry, but also because of the waste generated by research in the Nuclear Sciences Institute Vinca. In the last forty years, in the Vinca Institute, as a result of two research reactors being operational, named RA and RB, and as a result of the application of radionuclides in medicine, industry and agriculture, radioactive waste materials of different levels of specific activity were generated. As a temporary solution, radioactive waste materials are stored in two interim storages. Radwaste materials that were immobilized in the inactive matrices are to be placed in concrete containers, for further manipulation and disposal.(author)

  16. Ion exchangers in radioactive waste management: natural Iranian zeolites.

    Science.gov (United States)

    Nilchi, A; Maalek, B; Khanchi, A; Ghanadi Maragheh, M; Bagheri, A; Savoji, K

    2006-01-01

    Five samples of natural zeolites from different parts of Iran were chosen for this study. In order to characterize and determine their structures, X-ray diffraction and infrared spectrometry were carried out for each sample. The selective absorption properties of each zeolite were found by calculating the distribution coefficient (K(d)) of various simulated wastes which were prepared by spiking the radionuclides with (131)I, (99)Mo, (153)Sm, (140)La and (147)Nd. All the zeolite samples used in this study had extremely high absorption value towards (140)La; clinoptolite from Mianeh and analsite from Ghalehkhargoshi showed good absorption for (147)Nd; clinoptolite from Semnan and clinoptolite from Firozkoh showed high absorption for (153)Sm; mesolite from Arababad Tabas showed good absorption for (99)Mo; and finally mesolite from Arababad Tabas, clinoptolite from Semnan and clinoptolite from Firozkoh could be used to selectively absorb (131)I from the stimulated waste which was prepared. The natural zeolites chosen for these studies show a similar pattern to those synthetic ion exchangers in the literature and in some cases an extremely high selectivity towards certain radioactive elements. Hence the binary separation of radioactive elements could easily be carried out. Furthermore, these zeolites, which are naturally occurring ion exchangers, are viable economically and extremely useful alternatives in this industry.

  17. Catalytic oxidative pyrolysis of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.; Shirsat, A.N.; Bharadwaj, S.R.

    2005-08-01

    The spent IX resins from nuclear power reactors are highly active solid wastes generated during operations of nuclear reactors. Catalytic oxidative pyrolysis of these resins can lead to high volume reduction of these wastes. Low temperature pyrolysis of transition metal ion loaded IX resins in presence of nitrogen was carried out in order to optimize catalyst composition to achieve maximum weight reduction. Thermo gravimetric analysis of the pyrolysis residues was carried out in presence of air in order to compare the oxidative characteristics of transition metal oxide catalysts. Copper along with iron, chromium and nickel present in the spent IX resins gave the most efficient catalyst combination for catalytic and oxidative pyrolysis of the residues. During low temperature catalytic pyrolysis, 137 Cesium volatility was estimated to be around 0.01% from cationic resins and around 0.1% from anionic resins. During oxidative pyrolysis at 700 degC, nearly 10 to 40% of 137 Cesium was found to be released to off gases depending upon type of resin and catalyst loaded on to it. The oxidation of pyrolytic residues at 700 degC gave weight reduction of 15% for cationic resins and 93% for anionic resins. Catalytic oxidative pyrolysis is attractive for reducing weight and volume of spent cationic resins from PHWRs and VVERs. (author)

  18. Recycling of spent lithium-ion battery with polyvinyl chloride by mechanochemical process.

    Science.gov (United States)

    Wang, Meng-Meng; Zhang, Cong-Cong; Zhang, Fu-Shen

    2017-09-01

    In the present study, cathode materials (C/LiCoO 2 ) of spent lithium-ion batteries (LIBs) and waste polyvinyl chloride (PVC) were co-processed via an innovative mechanochemical method, i.e. LiCoO 2 /PVC/Fe was co-grinded followed by water-leaching. This procedure generated recoverable LiCl from Li by the dechlorination of PVC and also generated magnetic CoFe 4 O 6 from Co. The effects of different additives (e.g. alkali metals, non-metal oxides, and zero-valent metals) on (i) the conversion rates of Li and Co and (ii) the dechlorination rate of PVC were investigated, and the reaction mechanisms were explored. It was found that the chlorine atoms in PVC were mechanochemically transformed into chloride ions that bound to the Li in LiCoO 2 to form LiCl. This resulted in reorganization of the Co and Fe crystals to form the magnetic material CoFe 4 O 6 . This study provides a more environmentally-friendly, economical, and straightforward approach for the recycling of spent LIBs and waste PVC compared to traditional processes. Copyright © 2017. Published by Elsevier Ltd.

  19. High-level radioactive wastes storage characterization and behaviour of spent fuels in long-term

    International Nuclear Information System (INIS)

    Diaz Arocas, P.; Cobos, J.; Quinones, J.; Rodriguez Almazan, J. L.; Serrano, J.

    2001-01-01

    In order to understand the long term spent fuel dissolution under repository this report shows the study performed by considering spent fuel as a part of the multi barriers containment system. The study takes into account that the oxidative alteration/dissolution of spent fuel matrix is influenced by the intrinsic spent fuel physicochemical characteristics and the repository environmental parameters. Experimental and modelling results for granite and saline repositories are reported. Parameters considered in this work were pH, pCO 2 , S/V ratio, redox conditions and the influence of the container material in the redox conditions. The influence of alpha, beta and gamma radiation and the resulting radiolytic products formed remains as one of the main uncertainties to quantify the spent fuel behaviour under repository conditions. It was studied in a first approach through dose calculations, modelling of radiolytic products formation and leaching experiments in the presence of external gamma irradiation source and leaching experiments of alpha doped UO 2 pellets. Materials considered are LWR spent fuel (UO 2 and MOX fuel) and their chemical analogues non irradiated UO 2 , SIMFUEL and alpha doped UO 2 . Lea chants were granite groundwater, synthetic granite groundwater, synthetic granite groundwater saturated in bentonite, and high concentrated saline solutions. The matrix dissolution rate and release rate of key radionuclides (i. e. actinides and fission products) obtained through the several experimental techniques and methodologies (dissolution, co-dissolution, precipitation and co-precipitation) together with modelling studies supported in geochemical codes are proposed. Moreover, secondary phases formed that could control release and retention of key nuclides are identified. Maximum concentration values for these radionuclides are reported. The data provided by this study were used in ENRESA-2000 performance assessment. (Author)

  20. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management. Report of the Federal Republic of Germany for the sixth review meeting in May 2018

    International Nuclear Information System (INIS)

    2017-08-01

    The joint convention on the safety of spent fuel management and on the safety of radioactive waste management covers the following topics: historical development and actual status of the civil use of nuclear power, politics and the spent fuel management, inventories and listing, legislation and executive systems, other safeguard regulations, safety during spent fuel handling, safety during radioactive waste processing, transport across national borders, disused enclosed radioactive sources, general regulations for safety enhancement.

  1. The new revision of NPP Krsko decommissioning, radioactive waste and spent fuel management program: analyses and results

    International Nuclear Information System (INIS)

    Zeleznik, Nadja; Kralj, Metka; Lokner, Vladimir; Levanat, Ivica; Rapic, Andrea; Mele, Irena

    2010-01-01

    The preparation of the new revision of the Decommissioning and Spent Fuel (SF) and Low and Intermediate level Waste (LILW) Disposal Program for the NPP Krsko (Program) started in September 2008 after the acceptance of the Term of Reference for the work by Intergovernmental Committee responsible for implementation of the Agreement between the governments of Slovenia and Croatia on the status and other legal issues related to investment, exploitation, and decommissioning of the Nuclear power plant Krsko. The responsible organizations, APO and ARAO together with NEK prepared all new technical and financial data and relevant inputs for the new revision in which several scenarios based on the accepted boundary conditions were investigated. The strategy of immediate dismantling was analyzed for planned and extended NPP life time together with linked radioactive waste and spent fuel management to calculate yearly annuity to be paid by the owners into the decommissioning funds in Slovenia and Croatia. The new Program incorporated among others new data on the LILW repository including the costs for siting, construction and operation of silos at the location Vrbina in Krsko municipality, the site specific Preliminary Decommissioning Plan for NPP Krsko which included besides dismantling and decontamination approaches also site specific activated and contaminated radioactive waste, and results from the referenced scenario for spent fuel disposal but at very early stage. Important inputs for calculations presented also new amounts of compensations to the local communities for different nuclear facilities which were taken from the supplemented Slovenian regulation and updated fiscal parameters (inflation, interest, discount factors) used in the financial model based on the current development in economical environment. From the obtained data the nominal and discounted costs for the whole nuclear program related to NPP Krsko which is jointly owned by Slovenia and Croatia have

  2. Radioactive Ions Production Ring for Beta-Beams

    CERN Document Server

    Benedetto, E; Wehner, J

    2010-01-01

    Within the FP7 EUROnu program, Work Package 4 addresses the issues of production and acceleration of 8Li and 8B isotopes through the Beta-Beam complex, for the production of electron-neutrino. One of the major critical issues is the production of a high enougth ion ßux, to fulÞll the requirements for physics. In alternative to the direct ISOL production method, a new ap- proach is proposed in [1]. The idea is to use a compact ring for Litium ions at 25 MeV and an internal He or D target, in which the radioactive-isotopes production takes place. The beam is expected to survive for several thousands of turns, therefore cooling in 6D is required and, according this scheme, the ionization cooling provided by the target itself and a suitable RF system would be sufÞcient. We present some preliminary work on the Production ring lat- tice design and cooling issues, for the 7Li ions, and propose plans for future studies, within the EUROnu program.

  3. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.

    1999-10-21

    The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation.

  4. Radioactive Release from Aluminum-Based Spent Nuclear Fuel in Basin Storage

    International Nuclear Information System (INIS)

    Sindelar, R.L.

    1999-01-01

    The report provides an evaluation of: (1) the release rate of radionuclides through minor cladding penetrations (breaches) on aluminum-based spent nuclear fuel (AL SNF), and (2) the consequences of direct storage of breached AL SNF relative to the authorization basis for SRS basin operation

  5. Analysis of alternative transportation methods for radioactive materials shipments including the use of special trains for spent fuel and wastes

    International Nuclear Information System (INIS)

    Smith, D.R.; Luna, R.E.; Taylor, J.M.

    1978-01-01

    Two studies were completed which evaluate the environmental impact of radioactive material transport. The first was a generic study which evaluated all radioactive materials and all transportation modes; the second addressed spent fuel and fuel-cycle wastes shipped by truck, rail and barge. A portion of each of those studies dealing with the change in impact resulting from alternative shipping methods is presented in this paper. Alternatives evaluated in each study were mode shifts, operational constraints, and, in generic case, changes in material properties and package capabilities. Data for the analyses were obtained from a shipper survey and from projections of shipments that would occur in an equilibrium fuel cycle supporting one hundred 1000-MW(e) reactors. Population exposures were deduced from point source radiation formulae using separation distances derived for scenarios appropriate to each shipping mode and to each exposed population group. Fourteen alternatives were investigated for the generic impact case. All showed relatively minor changes in the overall radiological impact. Since the radioactive material transport is estimated to be fewer than 3 latent cancer fatalities (LCF) for each shipment year (compared to some 300,000 yearly cancer fatalities or 5000 LCF's calculated for background radiation using the same radiological effects model), a 15% decrease caused by shifting from passenger air to cargo air is a relatively small effect. Eleven alternatives were considered for the fuel cycle/special train study, but only one produced a reduction in total special train baseline LCF's (.047) that was larger than 5%

  6. Investigation of pharmaceuticals and medical devices containing 90Y extracted from high radioactive liquid waste in spent-fuel reprocessing

    International Nuclear Information System (INIS)

    Hosoma, Takashi

    2012-07-01

    Pharmaceuticals and medical devices containing radioactive 90 Y are realized, approved and placed on the international market where three products are available in Europe and the United States, and one product in Japan. These products are used not for diagnosis but for treatment by internal irradiation. It was estimated from the deliberative report of the approval in Japan that 90 Y was extracted in Europe from high radioactive liquid waste (HALW) yielded in spent-fuel reprocessing. In this report, products placed on the market and physical properties were reviewed, reasons of the realization and conditions to realize succeeding products were estimated, extraction method was compared with other methods, technical subjects, and relevant regulations were investigated. Although a medical device containing radioactive 90 Y has been studied in Japan and one pharmaceutical product was approved, a breakthrough would be necessary to put 90 Y utilization beyond alternative treatments. The breakthrough would become be promising; for example, if conventional treatments could be supported by technical development to deliver 90 Y more sharply to the target with shorter serum half-life. Extraction of 90 Y nuclide from HALW has advantages over thermal neutron irradiation of natural nuclide, a system is envisioned where 90 Sr as a parent nuclide is separated in the reprocessing then transported to and stored in a factory of radiopharmaceuticals followed by 90 Y extraction on demand. (author)

  7. Advances in the disposal of radioactive ion exchange resins

    International Nuclear Information System (INIS)

    McCoy, S.B.

    1983-01-01

    During the last several years, more stringent regulations have been imposed on the disposal of low-level radioactive wastes. In particular, the disposal of high-activity ion exchange resins has been affected by the recent requirements intended to enhance waste stability. High-activity resins must now be either solidified or placed in a ''high-integrity'' container. The allowable levels of free liquids in the containers have also been reduced. Solidification of resins has long been applied at nuclear power stations, but new designs in high-integrity containers and dewatering techniques to enhance the waste stability and ensure regulatory compliance have been developed and are being introduced for use at power stations

  8. High level radioactive wastes storage characterization and long-term behaviour of spent fuels

    International Nuclear Information System (INIS)

    Diaz Arocas, P.P.; Garcia Serrano, J.; Mendez Martin, F.J.; Quinones Diez, J.; Rodriguez Almazan, J.L.; Serrano Agejas, J.A.; Esteban Hernandez, J.A.

    1997-04-01

    The knowledge of long term spent fuel behaviour in a repository is one of the main goals in the waste management assessment due to its influence on repository design topics and on the performance assessment. At the moment, Spain has not selected a geological formation for a final repository. Therefore, R AND D activities are performed by considering granite, salt and clay as candidate options. This report summarises the activities carried out in CIEMAT from 1991 to 1995 in the frame of the Agreement between CIEMAT and ENRESA in the Area of spent fuel direct disposed. Experimental activities include leaching experiments of spent fuel, UO 2 and SIMFUEL and co-precipitation/solubility experiments of relevant secondary solid phases expected under repository conditions. The objective of leaching studies is to understand the processes which will occur when the underground water accede to the source term and to provide leaching rates of spent fuel and the influence of several variables as pH, Eh, etc. The co-precipitation/solubility experiments are focused on the knowledge of the formation conditions of relevant secondary phases, to characterise these phases and to determine their solubility, which could control the leaching of spent fuel. One of the main items to carry out the objectives before indicated in both leaching and co-precipitation/solubility experiments is to perform a extensive solid phases characterisation in order to facilitate the understanding of the processes involved. This report is structured in three parts, the first include experimental procedures, characterisation techniques and solid and solution analyses. The second shows the leaching results obtained by considering the effect of pH, complex formation, redox conditions, surface/volume ratio, etc. The third supply the results of the co-precipitation/solubility studies. The conclusions obtained in this work are considered as the start point of going on and more extensive studies on the mechanisms

  9. Experimental methods in radioactive ion-beam target/ion source development and characterization

    International Nuclear Information System (INIS)

    Welton, R.F.; Alton, G.D.; Cui, B.; Murray, S.N.

    1998-01-01

    We have developed off-line experimental techniques and apparatuses that permit direct measurement of effusive-flow delay times and ionization efficiencies for nearly any chemically reactive element in high-temperature target/ion sources (TIS) commonly used for on-line radioactive ion-beam (RIB) generation. The apparatuses include a hot Ta valve for effusive-flow delay-time measurements, a cooled molecular injection system for determination of ionization efficiencies, and a gas flow measurement/control system for introducing very low, well-defined molecular flows into the TIS. Measurements are performed on a test stand using molecular feed compounds containing stable complements of the radioactive nuclei of interest delivered to the TIS at flow rates commensurate with on-line RIB generation. In this article, the general techniques are described and effusive-flow delay times and ionization efficiency measurements are reported for fluorine in an electron-beam plasma target/ion source developed for RIB generation and operated in both positive- and negative-ion extraction modes. copyright 1998 American Institute of Physics

  10. Selection of RIB targets using ion implantation at the Holifield radioactive ion beam facility

    International Nuclear Information System (INIS)

    Alton, G.D.; Dellwo, J.

    1995-01-01

    Among several major challenges posed by generating and accelerating adequate intensities of RIBs, selection of the most appropriate target material is perhaps the most difficult because of the requisite fast and selective thermal release of minute amounts of the short-lived product atoms from the ISOL target in the presence of bulk amounts of target material. Experimental studies are under way at the Oak Ridge National Laboratory (ORNL) which are designed to measure the time evolution of implanted elements diffused from refractory target materials which are candidates for forming radioactive ion beams (RIBs) at the Holifield Radioactive Ion Beam Facility (HRIBF). The diffusion coefficients are derived by comparing experimental data with numerical solutions to a one-dimensional form of Fick's second law for ion implanted distributions. In this report, we describe the experimental arrangement, experimental procedures, and provide time release data and diffusion coefficients for releasing ion implanted 37 Cl from Zr 5 Si 3 and 75 As, 79 Br, and 78 Se from Zr 5 Ge 3 and estimates of the diffusion coefficients for 35 Cl, 63 Cu, 65 Cu, 69 Ga and 71 Ga diffused from BN; 35 Cl, 63 Cu, 65 Cu, 69 Ga, 75 As, and 78 Se diffused from C; 35 Cl, 68 Cu, 69 Ga, 75 As, and 78 Se diffused from Ta

  11. National report of the Slovak Republic - proposal. Compiled in terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. Jun 2008

    International Nuclear Information System (INIS)

    Jurina, V.; Viktory, D.; Kobzova, D.; Petrik, T.; Sovcik, J.; Hekel, P.; Suess, J.; Tomek, J.; Lukacovic, J.; Hekel, P.; Ivan, J.; Ziakova, M.; Metke, E.; Pospisil, M.; Turner, M.; Homola, J.; Konecny, L.; Parimucha, F.; Vaclav, J.; Horvath, J.; Soos, F.; Betak, A.; Pospisil, P.; Mihaly, B.; Kubala, M.; Schmidtova, B.; Orihel, M.; Vasina, D.; Balaz, J.; Ehn, L.; Micovicova, D.; Vrtoch, M.; Mlcuch, L.; Granak, P.; Meleg, J.; Sedliak, D.; Bardy, M.; Gogoliak, J.; Prazska, M.; Burslova, J.

    2008-06-01

    A brief national safety report of the Slovak Republic compiled in terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management in 2008 is presented. This safety report consists of following chapters: (A) Introduction; (B) Spent nuclear fuel (SNF) and radioactive waste (RAW) management conception; (C) Scope of application; (D) Spent nuclear fuel (SNF) and radioactive waste (RAW) management; (E) Legislation and regulatory framework; (F) General safety provisions; (G) Safety of spent nuclear fuel management; (H) Safety of radioactive waste management; (I) Transboundary movement of spent nuclear fuel and radioactive waste; (J) Discussed sealed radioactive sources; (K) Planned measures to improve safety; (L) Annexes

  12. Radioactive waste management, decommissioning, spent fuel storage. V.2. 3. ed.

    International Nuclear Information System (INIS)

    1986-10-01

    The aim of this book is to provide information on french technology achievements in the field of processing and conditioning of solid, liquid and gaseous radioactive wastes, through stationary or mobile units. It presents the facilities and services proposed by french industry, as well as the experience acquired. Separate abstracts were prepared for 16 papers [fr

  13. Soluble Graphene Nanosheets from Recycled Graphite of Spent Lithium Ion Batteries

    Science.gov (United States)

    Zhao, Liangliang; Liu, Xiya; Wan, Chuanyun; Ye, Xiangrong; Wu, Fanhong

    2018-02-01

    Soluble graphene nanosheets are fabricated from recycled graphite of spent lithium ion batteries through a modified Hammers process followed by deoxygenation with NaOH-KOH eutectic. Ultrasonic exfoliation in N-methyl-pyrrolidone indicates the loosened graphene layers in recycled graphite are prone to exfoliation. Reduction of the exfoliated graphene oxide sheets was conducted in molten NaOH-KOH eutectic at different temperatures. The results show that molten NaOH-KOH effectively eliminates the unsaturated oxygen-containing moieties from the exfoliated graphene oxide sheets while creating more hydroxyl functional groups. Higher temperature treatment is more prone to remove hydroxyls while producing the shrinkage on the surface of graphene sheets. Graphene sheet with a good solubility is produced when the graphene oxide is heat-treated at 220 °C for 10 h. After reduction, the graphene oxide sheets exhibit excellent dispersibility or solubility in water, ethanol and other polar solvents, therefore being highly desirable for solution processing of graphene materials. Such study not only identifies a high-quality stockpile to prepare soluble graphene but also paves a feasible alternative of graphite recycling from spent lithium batteries.

  14. Production of light radioactive ion beams (RIB) using inverse kinematics

    International Nuclear Information System (INIS)

    Das, J.J.; Sugathan, P.; Madhavan, N.; Madhusudhana Rao, P.V.; Jhingan, A.; Varughese, T.; Barua, S.; Nath, S.; Sinha, A.K.; Kumar, B.; Zacharias, J.

    2005-01-01

    At Nuclear Science Centre (NSC), New Delhi, we have implemented a facility to produce low energy light radioactive ion beams (RIBs) using (p,n) type of reactions in inverse kinematics. For this purpose primary beams from the 15-UD Pelletron accelerator impinged on a thin polypropylene foil mounted on a rotating/linearly moving target assembly. For efficiently separating the secondary beam from primary beam, the existing recoil mass spectrometer (RMS) HIRA was operated with new ion optics. Suitable hardware modifications were also made. Using this facility, we have extracted a 7 Be beam of purity better than 99% and spot-size ∼4 mm in diameter. This 7 Be beam has been utilized in a variety of experiments in the energy range of 15-22 MeV. Typical beam parameters are: intensity 10 4 pps, angular spread ±30 mrad and energy spread ±0.5 MeV. Development of appropriate detector setup/target arrangement were also made to perform these experiments. In this paper, we describe the implementation of this project

  15. High-spin nuclear structure studies with radioactive ion beams

    International Nuclear Information System (INIS)

    Baktash, C.

    1992-01-01

    Two important developments in the sixties, namely the advent of heavy-ion accelerators and fabrication of Ge detectors, opened the way for the experimental studies of nuclear properties at high angular momentum. Addition of a new degree of freedom, namely spin, made it possible to observe such fascinating phenomena as occurrences and coexistence of a variety of novel shapes, rise, fall and occasionally rebirth of nuclear collectivity, and disappearance of pairing correlations. Today, with the promise of development of radioactive ion beams (RIB) and construction of the third-generation Ge-detection systems (GAMMASPHERE and EUROBALL), nuclear physicists are poised to explore new and equally fascinating phenomena that have been hitherto inaccessible. With the addition of yet another dimension, namely the isospin, they will be able to observe and verify predictions for exotic shapes as varied as rigid triaxiality, hyperdeformation and triaxial-octupole shapes, or to investigate the T=O pairing correlations. In this paper, the author reviews, separately for neutron-deficient and neutron-rich nuclei, these and a few other new high-spin physics opportunities that may be realized with RIB. Following this discussion, a list of the beam species, intensities and energies that are needed to fulfill these goals is presented. The paper concludes with a description of the experimental techniques and instrumentations that are required for these studies

  16. Recovery of cobalt and lithium from spent Li-ion batteries

    International Nuclear Information System (INIS)

    Busnardo, Natalia Giovanini; Paulino, Jessica Frontino; Afonso, Julio Carlos

    2007-01-01

    The 'active mass' (cathode + anode + electrolyte) of spent Li-ion batteries was submitted to one of the following procedures: (a) it was calcined (500 deg C) and submitted to extraction with water to recover lithium salts. The residual solid was treated with sulfuric acid containing hydrogen peroxide. Cobalt was recovered as sulfate; (b) the 'active mass' was treated with potassium hydrogen sulfate (500 deg C) and dissolved in water. Cobalt was precipitated together with copper after addition of sodium hydroxide. Lithium was partially recovered as lithium fluoride. Co-processing of other battery components (aluminum and copper foils) affected negatively the behavior of the recovery procedures. Previous segregation of battery components is essential for an efficient and economical processing of the 'active mass'. (author)

  17. Detection systems for radioactive ion beams; Systeme de detection en ions radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Savajols, H

    2002-07-01

    Two main methods are used to produce radioactive ion beams: -) the ISOL method (isotope separation on-line) in which the stable beam interacts with a thick target, the reaction products diffuse outside the target and are transferred to a source where they are ionized, a mass separator and a post-accelerator drive the selected radioactive ions to the right energy; -) the in-flight fragmentation method in which the stable beam interacts with a thin target, the reaction products are emitted from the target with a restricted angular distribution and a velocity close to that of the incident beam, the experimenter has to take advantage from the reaction kinetics to get the right particle beam. Characteristic time is far longer with the ISOL method but the beam intensity is much better because of the use of a post-accelerator. In both cases, the beam intensity is lower by several orders of magnitude than in the case of a stable beam. This article presents all the constraints imposed by radioactive beams to the detection systems of the reaction products and gives new technical solutions according to the type of nuclear reaction studied. (A.C.)

  18. Recycling of spent lithium-ion battery cathode materials by ammoniacal leaching

    International Nuclear Information System (INIS)

    Ku, Heesuk; Jung, Yeojin; Jo, Minsang; Park, Sanghyuk; Kim, Sookyung; Yang, Donghyo; Rhee, Kangin; An, Eung-Mo; Sohn, Jeongsoo; Kwon, Kyungjung

    2016-01-01

    Highlights: • Ammoniacal leaching is used to recover spent Li-ion battery cathode materials. • Leaching agents consist of ammonia, ammonium sulfite and ammonium carbonate. • Ammonium sulfite is a reductant and ammonium carbonate acts as pH buffer. • Co and Cu can be fully leached while Mn and Al are not leached. • Co recovery via ammoniacal leaching is economical compared to acid leaching. - Abstract: As the production and consumption of lithium ion batteries (LIBs) increase, the recycling of spent LIBs appears inevitable from an environmental, economic and health viewpoint. The leaching behavior of Ni, Mn, Co, Al and Cu from treated cathode active materials, which are separated from a commercial LIB pack in hybrid electric vehicles, is investigated with ammoniacal leaching agents based on ammonia, ammonium carbonate and ammonium sulfite. Ammonium sulfite as a reductant is necessary to enhance leaching kinetics particularly in the ammoniacal leaching of Ni and Co. Ammonium carbonate can act as a pH buffer so that the pH of leaching solution changes little during leaching. Co and Cu can be fully leached out whereas Mn and Al are hardly leached and Ni shows a moderate leaching efficiency. It is confirmed that the cathode active materials are a composite of LiMn_2O_4, LiCo_xMn_yNi_zO_2_, Al_2O_3 and C while the leach residue is composed of LiNi_xMn_yCo_zO_2, LiMn_2O_4, Al_2O_3, MnCO_3 and Mn oxides. Co recovery via the ammoniacal leaching is believed to gain a competitive edge on convenitonal acid leaching both by reducing the sodium hydroxide expense for increasing the pH of leaching solution and by removing the separation steps of Mn and Al.

  19. Recent results on reactions with radioactive beams at RIBRAS (Radioactive Ion Beams in Brazil)

    Science.gov (United States)

    Lépine-Szily, A.; Lichtenthäler, R.; Guimarães, V.; Arazi, A.; Barioni, A.; Benjamim, E. A.; de Faria, P. N.; Descouvemont, P.; Gasques, L. R.; E; Leistenschneider; Mendes, D. R., Jr.; Morais, M. C.; Morcelle, V.; Moro, A. M.; Pampa Condori, R.; Pires, K. C. C.; Rodriguez-Gallardo, M.; Scarduelli, V.; Shorto, J. M. B.; Zamora, J. C.

    2015-04-01

    We present a quick description of RIBRAS (Radioactive Ion beams in Brazil), which is a superconducting double solenoid system, installed at the Pelletron Laboratory of the University of São Paulo and extends the capabilities of the original Pelletron Tandem Accelerator of 8MV terminal voltage (8UD) by producing secondary beams of unstable nuclei. The experimental program of the RIBRAS covers the study of elastic and inelastic scattering with the objective to study the interaction potential and the reaction mechanisms between weakly bound (RIB) and halo (6He and 8B) projectiles on light, medium and heavy mass targets. With highly purified beams, the study of resonant elastic scattering and resonant transfer reactions, using inverse kinematics and thick targets, have also been included in our recent experimental program.

  20. Conditioning of low level radioactive wastes, spent radiation sources and their transport at the interim storage building of the Institute of Nuclear Physics in Albania

    International Nuclear Information System (INIS)

    Qafmolla, L.

    2000-01-01

    Aspects of treatment and management of radioactive wastes resulting from the use of radiation sources and radioisotopes in research, medicine and industry, are described. The methods applied for the conditioning of low-level radioactive wastes and spent radiation sources are simple. Solid radioactive wastes with low-level activity, after accumulation, minimization, segregation and measurement, are burned or compressed in a simple compactor of the PGS type. Spent radiation sources are placed into 200 l drums, are cemented and conditioned. Conditioned drums from the Radiation Protection Division of the Institute of Nuclear Physics (INP), which is the responsible Institution for the treatment and management of radioactive wastes in Albania, are transported to the interim storage building of the Institute of Nuclear Physics in Tirana. Work to construct a new building for treatment and management of radioactive wastes and spent radiation sources within the territory of INP is underway. Funds have been allocated accordingly: based on the Law No. 8025 of 25.11.1995, it is the Albanian Government's responsibility to finance activities concerned with the treatment and management of radioactive wastes generating from the use of ionizing radiation in science, medicine and industry in the country. (author)

  1. Considerations in the transportation of spent fuel and other radioactive materials by sea

    International Nuclear Information System (INIS)

    Rhoads, R.E.; Heaberlin, S.W.; Brobst, W.A.

    1978-05-01

    A study was made of what would happen to casks and plutonium packages in the unlikely event of a serious sea accident involving shipments of spent fuel or plutonium, and what the radiological consequences might be to man through the marine food chain. Four postulated loss scenarios were considered: loss on the continental shelf and in the deep ocean of an undamaged package, and a package exposed to a severe fire. Conservative estimates of the maximum population doses are comparable to those expected from natural background sources. Areas where futher research is needed are identified. 3 tables

  2. DOE program for the management of radioactive waste and spent reactor fuel

    International Nuclear Information System (INIS)

    Cooley, C.R.

    1978-01-01

    The development of nuclear energy is seen by the Administration and the Department of Energy (DOE) as one of the important sources of energy for the country. Nuclear energy now provides a major fraction of the electrical power generation in some parts of the United States. In the northwest, with a wealth of hydroelectric power, nuclear power is expected to provide an increasing share of the total electrical energy. However, a great deal of public concern is being expressed about waste management associated with nuclear power. On current Research and Development programs, this paper considers several of the key activities which include work on disposal of spent fuel. 2 refs

  3. Charge Breeding of Radioactive Ions in an Electron Cyclotron Resonance Ion Source(ECRIS) at ISOLDE

    CERN Multimedia

    Lindroos, M

    2002-01-01

    The development of an efficient charge breeding scheme for the next generation of RIB facilities will have a strong impact on the post-accelerator for several Radioactive Ion Beam (RIB) projects at European large scale facilities. At ISOLDE/CERN there will be the unique possibility to carry out experiments with the two possible charge breeding set-ups with a large variety of radioactive isotopes using identical injection conditions. One charge breeding set-up is the Penning trap/EBIS combination which feeds the REX-ISOLDE linear accelerator and which is in commissioning now. The second charge breeder is a new ECRIS PHOENIX developed at the ISN ion source laboratory at Grenoble. This ECRIS is now under investigation with a 14 GHz amplifier to characterize its performance. The experiments are accompanied by theoretical studies in computer simulations in order to optimize the capture of the ions in the ECRIS plasma. A second identical PHOENIX ECRIS which is under investigation at the Daresbury Laboratory is avai...

  4. Ion beam production and study of radioactive isotopes with the laser ion source at ISOLDE

    Science.gov (United States)

    Fedosseev, Valentin; Chrysalidis, Katerina; Day Goodacre, Thomas; Marsh, Bruce; Rothe, Sebastian; Seiffert, Christoph; Wendt, Klaus

    2017-08-01

    At ISOLDE the majority of radioactive ion beams are produced using the resonance ionization laser ion source (RILIS). This ion source is based on resonant excitation of atomic transitions by wavelength tunable laser radiation. Since its installation at the ISOLDE facility in 1994, the RILIS laser setup has been developed into a versatile remotely operated laser system comprising state-of-the-art solid state and dye lasers capable of generating multiple high quality laser beams at any wavelength in the range of 210-950 nm. A continuous programme of atomic ionization scheme development at CERN and at other laboratories has gradually increased the number of RILIS-ionized elements. At present, isotopes of 40 different elements have been selectively laser-ionized by the ISOLDE RILIS. Studies related to the optimization of the laser-atom interaction environment have yielded new laser ion source types: the laser ion source and trap and the versatile arc discharge and laser ion source. Depending on the specific experimental requirements for beam purity or versatility to switch between different ionization mechanisms, these may offer a favourable alternative to the standard hot metal cavity configuration. In addition to its main purpose of ion beam production, the RILIS is used for laser spectroscopy of radioisotopes. In an ongoing experimental campaign the isotope shifts and hyperfine structure of long isotopic chains have been measured by the extremely sensitive in-source laser spectroscopy method. The studies performed in the lead region were focused on nuclear deformation and shape coexistence effects around the closed proton shell Z = 82. The paper describes the functional principles of the RILIS, the current status of the laser system and demonstrated capabilities for the production of different ion beams including the high-resolution studies of short-lived isotopes and other applications of RILIS lasers for ISOLDE experiments. This article belongs to the Focus on

  5. Logistics models for the transportation of radioactive waste and spent fuel

    International Nuclear Information System (INIS)

    Joy, D.S.; Holcomb, B.D.

    1978-03-01

    Mathematical modeling of the logistics of waste shipment is an effective way to provide input to program planning and long-range waste management. Several logistics models have been developed for use in parametric studies, contingency planning, and management of transportation networks. These models allow the determination of shipping schedules, optimal routes, probable transportation modes, minimal costs, minimal personnel exposure, minimal transportation equipment, etc. Such information will permit OWI to specify waste-receiving rates at various repositories in order to balance work loads, evaluate surge capacity requirements, and estimate projected shipping cask fleets. The programs are tailored to utilize information on the types of wastes being received, location of repositories and waste-generating facilities, shipping distances, time required for a given shipment, availability of equipment, above-ground storage capabilities and locations, projected waste throughput rates, etc. Two basic models have been developed. The Low-Level Waste Model evaluates the optimal transportation policy for shipping waste directly from the source to a final destination without any intermediate stops. The Spent Fuel Logistics Model evaluates the optimal transportation policy for shipping unreprocessed spent fuel from nuclear power plants (1) indirectly, that is, to an Away-From-Reactor (AFR) storage facility, with subsequent transhipment to a repository, or (2) directly to a repository

  6. Operation and control of ion-exchange processes for treatment of radioactive wastes

    International Nuclear Information System (INIS)

    Emelity, L.A.

    1967-01-01

    A manual dealing with the application of ion-exchange materials to the treatment of radioactive wastes and reviewing the facilities currently using this method. This book is one of three commissioned by the IAEA on the principal methods of concentrating radioactive wastes. The content of this document is: (i) Historical review related to removal of radioactivity; (ii) Principles of ion exchange (iii) Ion-exchange materials; (iv) Limitations of ion exchangers; (v) Application of ion exchange to waste processing; (vi) Operational procedures and experiences; (vii) Cost-of-treatment by ion-exchange. The document also gives a list of producers of ion-exchange material and defines some relevant terms. 101 refs, 31 figs, 27 tabs

  7. Operation and control of ion-exchange processes for treatment of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Emelity, L A [Los Alamos National Lab., NM (United States)

    1967-12-01

    A manual dealing with the application of ion-exchange materials to the treatment of radioactive wastes and reviewing the facilities currently using this method. This book is one of three commissioned by the IAEA on the principal methods of concentrating radioactive wastes. The content of this document is: (i) Historical review related to removal of radioactivity; (ii) Principles of ion exchange (iii) Ion-exchange materials; (iv) Limitations of ion exchangers; (v) Application of ion exchange to waste processing; (vi) Operational procedures and experiences; (vii) Cost-of-treatment by ion-exchange. The document also gives a list of producers of ion-exchange material and defines some relevant terms. 101 refs, 31 figs, 27 tabs.

  8. Introduction to the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management and Canada's participation

    International Nuclear Information System (INIS)

    Mecke, J.L.

    2011-01-01

    The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention) is the first and the only legally binding international instrument to address safety issues concerning the management of spent fuel and radioactive waste on a global scale. It entered into force on June 18, 2001. The Government of Canada strongly supported international efforts to bring into force the Joint Convention and was the second country to ratify it. The Joint Convention is an 'incentive instrument' that is based on peer review (similar in that respect to the Convention on Nuclear Safety) and devised to encourage countries that are Contracting Parties to report and to foster open and frank discussions on the safety of spent fuel and radioactive waste management. Being an incentive convention, it is not designed to mandate Contracting Parties to fulfill its obligation through control and sanction, but it is based on the common objectives of Contracting Parties to achieve and maintain a high level of safety in spent fuel and radioactive waste management, protect individuals, society and the environment from ionizing radiation and prevent accidents and if necessary mitigating the consequences of such accidents. The following paper will provide an introduction to the Joint Convention and provide a summary of Canada's peer review at the most recent Review Meeting which was held on May 11-20, 2009, at the International Atomic Energy Agency (IAEA) headquarters in Vienna, Austria. (author)

  9. Development of a graded approach to natural phenomena hazard design and evaluation of radioactive waste and spent fuel stored at nuclear power plants

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    2001-01-01

    Nuclear safety related structures, systems and components, SSC, at large commercial nuclear power plants other than those applicable to reactor safety have in general not received the attention and detailed loading and behavior criteria use for reactor design safety. Such systems include spent fuel storage and radioactive waste storage and processing. In this paper is a suggested grading of design bases for natural hazards to be applied to such facilities commensurate with their radioactive risk. They are applicable to the full range of safety related SSC which are determined by the inventory of radioactive isotopes and the unmitigated doses at appropriate plant and site boundaries. (author)

  10. Conceptual design and cost inputs associated with co-disposal of the spent fuel and long lived radioactive wastes in the deep geologic disposal facility

    International Nuclear Information System (INIS)

    Fako, R.; Sociu, F.; Nicolae, R.; Barariu, G

    2013-01-01

    The paper aims to be an integrated approach for the containment and isolation of spent fuel and / or long lived radioactive wastes in a Deep Geologic Repository in Romania. Several scenarios could be defined for the management of spent fuel and long lived radioactive waste in Romania considering many specific constraints in Romania (political, geological, economic, demographic, etc.). This paper intends to be an upgrade of several Research, Development and Demonstration (RD&D) works performed by SITON specialists on this subject, taking into account also the conclusions of the Workshop ôCost estimation on spent nuclear fuel disposal in Romaniaö organized by IAEA in cooperation with ANDR at the beginning of this year in Romania.This paper is, also, addressed to decision makers with target on to adopt the best strategy for construction of Deep Geologic Repository in Romania. (authors)

  11. Responsible and safe management of spent fuel and radioactive waste in Germany

    International Nuclear Information System (INIS)

    Caspers, Mechthild; Schulte, Lukas; Rüger, Jörg

    2016-01-01

    Key points of the National Programme: • Radioactive Waste Management is to be carried out within German national responsibility and disposal is to be on German territory. • Disposal facilities are to be established at two sites: • Konrad facility for waste with negligible heat generation (commissioning expected in 2022, operation for 40 years); • Disposal facility for, in particular, high level waste according to Site Selection Act (site to be determined by 2031). • Dismantling of NPPs and other nuclear facilities is to be executed in due time so that arising LILW can be emplaced in Konrad facility

  12. The radioactive ion beams facility project for the legnaro laboratories

    Science.gov (United States)

    Tecchio, Luigi B.

    1999-04-01

    In the frame work of the Italian participation to the project of a high intensity proton facility for the energy amplifier and nuclear waste transmutations, LNL is involving in the design and construction of prototypes of the injection system of the 1 GeV linac that consists of a RFQ (5 MeV, 30 mA) followed by a 100 MeV linac. This program has been already financially supported and the work is actually in progress. In this context, the LNL has been proposed a project for the construction of a second generation facility for the production of radioactive ion beams (RIBs) by using the ISOL method. The final goal consists in the production of neutron rich RIBs with masses ranging from 80 to 160 by using primary beams of protons, deuterons and light ions with energy of 100 MeV and 100 kW power. This project is proposed to be developed in about 10 years from now and intermediate milestones and experiments are foreseen and under consideration for the next INFN five year plan (1999-2003). In such period of time is proposed the construction of a proton/deuteron accelerator of 10 MeV energy and 10 mA current, consisting of a RFQ (5 MeV, 30 mA) and a linac (10 MeV, 10 mA), and of a neutron area dedicated to the RIBs production, to the BNCT applications and to the neutron physics. Some remarks on the production methods will be presented. The possibility of producing radioisotopes by means of the fission induced by neutrons will be investigated and the methods of production of neutrons will be discussed.

  13. Vacuum pyrolysis and hydrometallurgical process for the recovery of valuable metals from spent lithium-ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Liang [College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China); Key Laboratory of Resources Chemistry of Nonferrous Metals, Central South University, Ministry of Education of the People' s Republic of China (China); Qiu, Keqiang, E-mail: qiuwhs@sohu.com [College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China); Key Laboratory of Resources Chemistry of Nonferrous Metals, Central South University, Ministry of Education of the People' s Republic of China (China)

    2011-10-30

    Highlights: {yields} The cathode active materials LiCoO{sub 2} from spent lithium-ion batteries peeled completely from aluminum foils by vacuum pyrolysis and hydrometallurgical process. {yields} The aluminum foils were excellent without damage after vacuum pyrolysis. {yields} The pyrolysis products organic fluorine compounds from organic electrolyte and binder were collected and enriched. {yields} High leaching efficiencies of cobalt and lithium were obtained with H{sub 2}SO{sub 4} and H{sub 2}O{sub 2}. - Abstract: Spent lithium-ion batteries contain lots of strategic resources such as cobalt and lithium together with other hazardous materials, which are considered as an attractive secondary resource and environmental contaminant. In this work, a novel process involving vacuum pyrolysis and hydrometallurgical technique was developed for the combined recovery of cobalt and lithium from spent lithium-ion batteries. The results of vacuum pyrolysis of cathode material showed that the cathode powder composing of LiCoO{sub 2} and CoO peeled completely from aluminum foils under the following experimental conditions: temperature of 600 {sup o}C, vacuum evaporation time of 30 min, and residual gas pressure of 1.0 kPa. Over 99% of cobalt and lithium could be recovered from peeled cobalt lithium oxides with 2 M sulfuric acid leaching solution at 80 {sup o}C and solid/liquid ratio of 50 g L{sup -1} for 60 min. This technology offers an efficient way to recycle valuable materials from spent lithium-ion batteries, and it is feasible to scale up and help to reduce the environmental pollution of spent lithium-ion batteries.

  14. Vacuum pyrolysis and hydrometallurgical process for the recovery of valuable metals from spent lithium-ion batteries

    International Nuclear Information System (INIS)

    Sun, Liang; Qiu, Keqiang

    2011-01-01

    Highlights: → The cathode active materials LiCoO 2 from spent lithium-ion batteries peeled completely from aluminum foils by vacuum pyrolysis and hydrometallurgical process. → The aluminum foils were excellent without damage after vacuum pyrolysis. → The pyrolysis products organic fluorine compounds from organic electrolyte and binder were collected and enriched. → High leaching efficiencies of cobalt and lithium were obtained with H 2 SO 4 and H 2 O 2 . - Abstract: Spent lithium-ion batteries contain lots of strategic resources such as cobalt and lithium together with other hazardous materials, which are considered as an attractive secondary resource and environmental contaminant. In this work, a novel process involving vacuum pyrolysis and hydrometallurgical technique was developed for the combined recovery of cobalt and lithium from spent lithium-ion batteries. The results of vacuum pyrolysis of cathode material showed that the cathode powder composing of LiCoO 2 and CoO peeled completely from aluminum foils under the following experimental conditions: temperature of 600 o C, vacuum evaporation time of 30 min, and residual gas pressure of 1.0 kPa. Over 99% of cobalt and lithium could be recovered from peeled cobalt lithium oxides with 2 M sulfuric acid leaching solution at 80 o C and solid/liquid ratio of 50 g L -1 for 60 min. This technology offers an efficient way to recycle valuable materials from spent lithium-ion batteries, and it is feasible to scale up and help to reduce the environmental pollution of spent lithium-ion batteries.

  15. Tungsten Recovery from Spent SCR Catalyst Using Alkaline Leaching and Ion Exchange

    Directory of Open Access Journals (Sweden)

    Wen-Cheng Wu

    2016-10-01

    Full Text Available The recovery of tungsten (W from a honeycomb-type spent selective catalytic reduction (SCR catalyst using an alkaline leaching–ion exchange method was investigated. Spent SCR catalyst mainly consists of TiO2 and other oxides (6.37% W, 1.57% vanadium (V, and 2.81% silicon (Si, etc.. The ground catalyst was leached at the optimal conditions, as follows: NaOH concentration of 0.3 kg/kg of catalyst, pulp density of 3%, leaching temperature of 70 °C, particle size of −74 μm, and leaching time of 30 min. In this study, the leaching rate values of V and W under the above conditions were 87 wt %, and 91 wt %, respectively. The pregnant solution was then passed through a strong base anion exchange resin (Amberlite IRA900. At high pH conditions, the use of strong base anion exchange resin led to selective loading of divalent WO42− from the solution, because the fraction of two adjacent positively-charged sites on the IRA900 resin was higher and separate from the coexisting VO43−. The adsorbed W could then be eluted with 1 M NaCl + 0.5 M NaOH. The final concentrated W solution had 8.4 g/L of W with 98% purity. The application of this process in industry is expected to have an important impact on the recovery of W from secondary sources of these metals.

  16. The Management of the Solid Radioactive Waste and Used (Spent) Fuel in South Africa: An Overview of Past, Present and Future Practices

    International Nuclear Information System (INIS)

    Maree, V.

    2015-01-01

    As a country with a nuclear program, the Republic of South Africa (RSA) generates radioactive waste through numerous activities. Radioactive waste, for legal and regulatory purposes, is defined as “material that contains or is contaminated with radio-nuclides at concentrations or activities greater than clearance levels as established by the regulatory body and for which no use is foreseen”. The RSA recognises the importance of the safe management of spent fuel and radioactive waste, for this reason the country is a contracting party to the International Atomic Energy Agency (IAEA) Joint Convention on the Safety of Spent Nuclear Fuel Management and Safety of Radioactive Waste Management. South Africa fulfils its obligations under the Joint Convention by the establishment of a Radioactive Waste Management Policy and Strategy for the Republic of South Africa (Policy and Strategy). It lists the principles and provides direction relating to solid radioactive waste management. Although all key players i.e. government agencies and the private sector are participating to implement the national commitment in a coordinated and cooperative manner, huge uncertainty remains. This poster presents the South African National Radioactive Waste Management Model with a description of – the radioactive waste generated, – the hierarchy of waste management options, – the waste classification scheme adopted, – the current disposal option, – the current management of used (spent) fuel. Good intentions have not always been matched by action and measures are still needed to improve safety especially to integrate the lessons learnt from the Fukushima accident, management of legacy waste, monitoring of disused sealed sources, recovery of orphan sources and additional waste due to operation of potential new nuclear power plants etc. This poster also addresses current discussions and ideas relating to the above challenges. (author)

  17. Low-energy radioactive ion beam production of 22Mg

    International Nuclear Information System (INIS)

    Duy, N.N.; Kubono, S.; Yamaguchi, H.; Kahl, D.; Wakabayashi, Y.; Teranishi, T.; Iwasa, N.; Kwon, Y.K.; Khiem, L.H.; Kim, Y.H.; Song, J.S.; Hu, J.; Ayyad, Y.

    2013-01-01

    The 22 Mg nucleus plays an important role in nuclear astrophysics, specially in the 22 Mg(α,p) 25 Al and proton capture 22 Mg(p,γ) 23 Al reactions. It is believed that 22 Mg is a waiting point in the αp-process of nucleosynthesis in novae. We proposed a direct measurement of the 22 Mg+α resonance reaction in inverse kinematics using a radioactive ion (RI) beam. A 22 Mg beam of 3.73 MeV/u was produced at CRIB (Center for Nuclear Study (CNS) low-energy RI Beam) facility of the University of Tokyo located at RIKEN (Japan) in 2011. In this paper we present the results about the production of the 22 Mg beam used for the direct measurement of the scattering reaction 22 Mg(α,α) 22 Mg, and the stellar reaction 22 Mg(α,p) 25 Al in the energy region concerning an astrophysical temperature of T 9 =1–3 GK

  18. Copper ferrocyanide - polyurethane foam as a composite ion exchanger for removal of radioactive cesium

    International Nuclear Information System (INIS)

    Rao, S.V.S.; Lal, K.B.; Ahmed, J.; Narasimhan, S.V.

    1999-01-01

    A method has been developed for the removal of cesium from the aqueous radioactive waste using a composite ion-exchanger consisting of Copper-Ferrocyanide Powder (CFC) and Polyurethane (PU) Foam. Polyvinyl acetate has been used as a binder in the preparation of CFC-PU foam. The physical properties of CFC such as density, surface area, IR stretching frequency and lattice parameters have been evaluated and also its potassium and copper(II) content have been estimated. Optimization of loading of CFC on PU foam has been studied. The CFC-PU was viewed under microscope to find out the homogeneity of distribution. Exchange capacities of the CFC-PU foam in different media have been determined and column studies have been carried out. Studies have been undertaken on extraction of cesium from CFC foam and also on digestion of spent CFC-PU foam and immobilization of digested solution in cement matrix. The cement matrices have been characterized with respect to density, bio-resistance and leaching resistance. (author)

  19. Recycling of spent lithium-ion battery cathode materials by ammoniacal leaching

    Energy Technology Data Exchange (ETDEWEB)

    Ku, Heesuk; Jung, Yeojin; Jo, Minsang; Park, Sanghyuk [Department of Energy & Mineral Resources Engineering, Sejong University, Seoul 05006 (Korea, Republic of); Kim, Sookyung [Urban Mine Department, Korea Institute of Geoscience and Mineral Resources, 124 Gwahang-no, Yuseong-gu, Daejeon (Korea, Republic of); Yang, Donghyo, E-mail: ydh@kigam.re.kr [Urban Mine Department, Korea Institute of Geoscience and Mineral Resources, 124 Gwahang-no, Yuseong-gu, Daejeon (Korea, Republic of); Rhee, Kangin; An, Eung-Mo; Sohn, Jeongsoo [Urban Mine Department, Korea Institute of Geoscience and Mineral Resources, 124 Gwahang-no, Yuseong-gu, Daejeon (Korea, Republic of); Kwon, Kyungjung, E-mail: kfromberk@gmail.com [Department of Energy & Mineral Resources Engineering, Sejong University, Seoul 05006 (Korea, Republic of)

    2016-08-05

    Highlights: • Ammoniacal leaching is used to recover spent Li-ion battery cathode materials. • Leaching agents consist of ammonia, ammonium sulfite and ammonium carbonate. • Ammonium sulfite is a reductant and ammonium carbonate acts as pH buffer. • Co and Cu can be fully leached while Mn and Al are not leached. • Co recovery via ammoniacal leaching is economical compared to acid leaching. - Abstract: As the production and consumption of lithium ion batteries (LIBs) increase, the recycling of spent LIBs appears inevitable from an environmental, economic and health viewpoint. The leaching behavior of Ni, Mn, Co, Al and Cu from treated cathode active materials, which are separated from a commercial LIB pack in hybrid electric vehicles, is investigated with ammoniacal leaching agents based on ammonia, ammonium carbonate and ammonium sulfite. Ammonium sulfite as a reductant is necessary to enhance leaching kinetics particularly in the ammoniacal leaching of Ni and Co. Ammonium carbonate can act as a pH buffer so that the pH of leaching solution changes little during leaching. Co and Cu can be fully leached out whereas Mn and Al are hardly leached and Ni shows a moderate leaching efficiency. It is confirmed that the cathode active materials are a composite of LiMn{sub 2}O{sub 4}, LiCo{sub x}Mn{sub y}Ni{sub z}O{sub 2,} Al{sub 2}O{sub 3} and C while the leach residue is composed of LiNi{sub x}Mn{sub y}Co{sub z}O{sub 2}, LiMn{sub 2}O{sub 4}, Al{sub 2}O{sub 3}, MnCO{sub 3} and Mn oxides. Co recovery via the ammoniacal leaching is believed to gain a competitive edge on convenitonal acid leaching both by reducing the sodium hydroxide expense for increasing the pH of leaching solution and by removing the separation steps of Mn and Al.

  20. Inter-modal Transportation of Spent Nuclear Fuel from Office of Civilian Radioactive Waste Management Sites

    International Nuclear Information System (INIS)

    Schmid, St.; Thrower, A.; Best, R.E.

    2009-01-01

    U.S. Department of Energy (DOE) plans to ship most commercial spent nuclear fuel (SNF) by rail in sealed transportation, aging and disposal (TAD) canisters. Based on current and projected dry SNF storage programs, DOE believes the majority of commercial nuclear sites would have the capacity to load and prepare large-capacity, canister-based dry storage canisters such as the proposed TAD systems. Thus, only a small fraction of SNF, such as odd lots and SNF not meeting TAD criteria, is projected for truck (highway) shipment. However, at some commercial facilities rail tracks do not extend to the site, or on-site rail does not extend to the site's proposed loading spot, and inter-modal transfers will be required. Advance coordination between DOE and commercial site operators, commercial carriers, specialized carriers and riggers, and state, tribal and local routing officials will be necessary to establish inter-modal transfer areas and obtain necessary permits to move heavy loads over highways. Although inter-modal transfers can involve a number of steps and several different entities acting in close coordination, such moves are commonly performed by industry and the system requirements are well-understood. (authors)

  1. Radiological criteria for the remediation of sites for spent fuel and radioactive waste storage in the Russian Northwest.

    Science.gov (United States)

    Shandala, N K; Sneve, M K; Titov, A V; Smith, G M; Novikova, N Ya; Romanov, V V; Seregin, V A

    2008-12-01

    In the 1960s, two technical bases of the Northern Fleet were created in Northwest Russia, at Andreeva Bay in the Kola Peninsula and Gremikha village on the coast of the Barents Sea. They maintained nuclear submarines, performing receipt and storage of radioactive waste and spent nuclear fuel, and are now designated sites of temporary storage (STSs). An analysis of the radiation situation at these sites demonstrates that substantial long-term remediation work will be required after the removal of the waste and spent nuclear fuel. Regulatory guidance is under development to support this work. Having in mind modern approaches to guaranteeing radiation safety, the primary regulatory focus is on a justification of dose constraints for determining acceptable residual contamination which might lead to exposure to workers and the public. For these sites, four principal options for remediation have been considered-renovation, conversion, conservation and liquidation. This paper describes a system of recommended dose constraints and derived control levels formulated for each option. The unconditional guarantee of long-term radioecological protection provides the basis for criteria development. Non-exceedance of these dose constraints and control levels implies compliance with radiological protection objectives related to the residual contamination. Dose reduction below proposed dose constraint values must also be carried out according to the optimisation principle. The developed criteria relate to the condition of the facilities and the STS areas after the termination of remediation activities. The proposed criteria for renovation, conversion, conservation and liquidation are entirely within the dose limits adopted in Russia for the management of man-made radiation sources, and are consistent with ICRP recommendations and national practice in other countries. The proposed criteria for STS remediation and new industrial (non-radiation-hazardous) facilities and buildings on

  2. Prospects for studies of ground-state proton decays with the Holifield Radioactive Ion Beam Facility

    International Nuclear Information System (INIS)

    Toth, K.S.

    1994-01-01

    By using radioactive ions from the Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory it should be possible to identify many new ground-state proton emitters in the mass region from Sn to Pb. During this production and search process the limits of stability on the proton-rich side of the nuclidic chart will be delineated for a significant fraction of medium-weight elements and our understanding of the proton-emission process will be expanded and improved

  3. Recycling of spent lithium-ion battery cathode materials by ammoniacal leaching.

    Science.gov (United States)

    Ku, Heesuk; Jung, Yeojin; Jo, Minsang; Park, Sanghyuk; Kim, Sookyung; Yang, Donghyo; Rhee, Kangin; An, Eung-Mo; Sohn, Jeongsoo; Kwon, Kyungjung

    2016-08-05

    As the production and consumption of lithium ion batteries (LIBs) increase, the recycling of spent LIBs appears inevitable from an environmental, economic and health viewpoint. The leaching behavior of Ni, Mn, Co, Al and Cu from treated cathode active materials, which are separated from a commercial LIB pack in hybrid electric vehicles, is investigated with ammoniacal leaching agents based on ammonia, ammonium carbonate and ammonium sulfite. Ammonium sulfite as a reductant is necessary to enhance leaching kinetics particularly in the ammoniacal leaching of Ni and Co. Ammonium carbonate can act as a pH buffer so that the pH of leaching solution changes little during leaching. Co and Cu can be fully leached out whereas Mn and Al are hardly leached and Ni shows a moderate leaching efficiency. It is confirmed that the cathode active materials are a composite of LiMn2O4, LiCoxMnyNizO2, Al2O3 and C while the leach residue is composed of LiNixMnyCozO2, LiMn2O4, Al2O3, MnCO3 and Mn oxides. Co recovery via the ammoniacal leaching is believed to gain a competitive edge on convenitonal acid leaching both by reducing the sodium hydroxide expense for increasing the pH of leaching solution and by removing the separation steps of Mn and Al. Copyright © 2016 Elsevier B.V. All rights reserved.

  4. Recovery process of cathode material of the spent lithium-ion batteries using Pechini methods

    International Nuclear Information System (INIS)

    Polo Fonseca, C.; Prado, R.M.; Santos Junior, G.A.; Marques, E.C.; Neves, S.

    2010-01-01

    This work proposes a new process of recovering LiCoO 2 from spent Li-ion batteries (LIBs) by a combination of acid leaching and Pechini synthesis, as an alternative process to improve the recovery efficiency of LiCoO 2 and reduce energy consumption and pollution. The effects of calcination temperature and lithium acetate addition in the synthesis on the structure and morphology of LiCoO 2 powders were characterized by X-ray diffraction, Raman spectroscopy and scanning electron microscopy. According to the analysis, the crystallinity of LiCoO 2 powders depends on the calcination temperature. The results indicate the layered HT-LiCoO 2 powders can be obtained at 750 deg C for 24 h in oxygen with lithium salt addition. Cyclic voltammograms showed one reversible redox process at 4.0/3.85 V for the LiCoO 2 obtained with lithium addition in the synthesis and irreversible redox process for the LiCoO 2 obtained without lithium addition. (author)

  5. Radiological protection regulation during spent nuclear fuel and radioactive waste management in the western branch of the Federal State Unitary Enterprise 'SevRAO'.

    Science.gov (United States)

    Simakov, A V; Sneve, M K; Abramov, Yu V; Kochetkov, O A; Smith, G M; Tsovianov, A G; Romanov, V V

    2008-12-01

    The site of temporary storage of spent nuclear fuel and radioactive waste, situated at Andreeva Bay in Northwest Russia, was developed in the 1960s, and it has carried out receipt and storage of fresh and spent nuclear fuel, and solid and liquid radioactive waste generated during the operation of nuclear submarines and nuclear-powered icebreakers. The site is now operated as the western branch of the Federal State Unitary Enterprise, SevRAO. In the course of operation over several decades, the containment barriers in the Spent Nuclear Fuel and Radioactive Waste storage facilities partially lost their containment effectiveness, so workshop facilities and parts of the site became contaminated with radioactive substances. This paper describes work being undertaken to provide an updated regulatory basis for the protection of workers during especially hazardous remediation activities, necessary because of the unusual radiation conditions at the site. It describes the results of recent survey work carried out by the Burnasyan Federal Medical Biophysical Centre, within a programme of regulatory cooperation between the Norwegian Radiation Protection Authority and the Federal Medical-Biological Agency of Russia. The survey work and subsequent analyses have contributed to the development of special regulations setting out radiological protection requirements for operations planned at the site. Within these requirements, and taking account of a variety of other factors, a continuing need arises for the implementation of optimisation of remediation at Andreeva Bay.

  6. Degradation of ion spent resin using the Fenton's reagent; Degradacao da resina de troca ionica utilizando o reagente de Fenton

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Leandro Goulart de

    2013-07-01

    The most common method for spent radioactive ion exchange resin treatment is its immobilization in cement, which reduces the radionuclides release into the environment. Although this method is efficient, it increases considerably the final volume of the waste due to the low incorporation capacity. The objective of this work was to develop a degradation method of spent resins arising from the nuclear research reactor located at the Nuclear and Energy Research Institute (IPEN-CNEN/SP), using an Advanced Oxidation Process (AOP) with Fenton's reagents. This method would allow a higher incorporation in cement. Three different resins were evaluated: cationic, anionic and a mixture of both resins. The reactions were conducted varying the catalyst concentration (25, 50, 100 and 150 mM), the volume of hydrogen peroxide (320 to 460 mL), and three different temperatures, 50, 60 and 70 deg C. Degradation of about 98% was achieved using a 50 mM catalyst solution and 330 mL of hydrogen peroxide solution. The most efficient temperature was 60 deg C. (author)

  7. Animal experimental investigations of the problem of the decorporation of radioactive metal ions

    International Nuclear Information System (INIS)

    Berner, W.

    1973-01-01

    Basically, it is possible to reduce the radiation exposure by excretion intensification of incorporated radioactive materials. Chelate agents have proved to be particularly effective for the accelerated elimination of radioactive metal ions. The kinetics of the distribution and excretion of 57 CoCl 2 , 65 ZnCl 2 and 203 HgCl 2 on the rat under the influence of the chelating agent penicillamin (D-ββ-dimethylcystein) was investigated and the reduction of the radiation exposure in man was calculated from the animal-experimentally gained data. At various times after the incorporation of metal ions, the whole body radioactivity and, after killing the animals, the radioactivity of the organs liver, kidney, spleen, skeleton, muscles and blood, were measured. From the course of the measured radioactivities with time, the biokinetic data of the radioactive metal ions (effective half-lives, selection factors and their components) were determined by means of regression analyses. The chelating agent was applied at different times before or after incorporation of the radioactive metal ions. (EK/LH) [de

  8. Cementation of secondary wastes generated from carbonisation of spent organic ion exchange resins from nuclear power plants

    International Nuclear Information System (INIS)

    Sathi Sasidharan, N.; Deshingkar, D.S.; Wattal, P.K.

    2004-07-01

    The spent IX resins containing radioactive fission and activation products from power reactors are highly active solid wastes generated during operations of nuclear reactors. Process for carbonization of IX resins to achieve weight and volume reduction has been optimized on 50 dm 3 /batch pilot test rig. The process generates carbonaceous residue, organic liquid condensates (predominantly styrene) and aqueous alkaline scrubber solutions as secondary wastes. The report discusses laboratory tests on leaching of 137 Cs from cement matrix incorporating carbonaceous residues and extrapolation of results to 200 liter matrix block. The cumulative fraction of 137 Cs leached from 200 liter cement matrix was estimated to be 0.0021 in 200 days and 0.0418 over a period of 30 years. Incorporation of organic liquid condensates into cement matrix has been tried out successfully. Thus two types of secondary wastes generated during carbonization of spent IX resins can be immobilized in cement matrix. (author)

  9. Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    International Nuclear Information System (INIS)

    1999-01-01

    The Proposed Action addressed in this EIS is to construct, operate and monitor, and eventually close a geologic repository at Yucca Mountain in southern Nevada for the disposal of spent nuclear fuel and high-level radioactive waste currently in storage at 72 commercial and 5 DOE sites across the United States. The EIS evaluates (1) projected impacts on the Yucca Mountain environment of the construction, operation and monitoring, and eventual closure of the geologic repository; (2) the potential long-term impacts of repository disposal of spent nuclear fuel and high-level radioactive waste; (3) the potential impacts of transporting these materials nationally and in the State of Nevada; and (4) the potential impacts of not proceeding with the Proposed Action

  10. Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. National Report of the Kingdom of the Netherlands

    International Nuclear Information System (INIS)

    2005-10-01

    On 10 March 1999, the Netherlands signed the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, which was subsequently formally ratified on 26 April 2000 and entered into force on 18 June 2001. The Joint Convention obliges each contracting party to apply widely recognized principles and tools in order to achieve and maintain high standards of safety during management of spent fuel and radioactive waste. The Joint Convention also requires each party to report on the national implementation of these principles to review meetings of the parties to this Convention. This report describes the manner in which the Netherlands is fulfilling its obligations under the Joint Convention

  11. Organic oxalate as leachant and precipitant for the recovery of valuable metals from spent lithium-ion batteries

    International Nuclear Information System (INIS)

    Sun Liang; Qiu Keqiang

    2012-01-01

    Graphical abstract: Display Omitted Highlights: ► Vacuum pyrolysis as a pretreatment was used to separate cathode material from aluminum foils. ► Cobalt and lithium can be leached using oxalate while cobalt can be directly precipitated as cobalt oxalate. ► Cobalt and lithium can be separated efficiently from each other only in the oxalate leaching process. ► High reaction efficiency of LiCoO 2 was obtained with oxalate. - Abstract: Spent lithium-ion batteries containing lots of strategic resources such as cobalt and lithium are considered as an attractive secondary resource. In this work, an environmentally compatible process based on vacuum pyrolysis, oxalate leaching and precipitation is applied to recover cobalt and lithium from spent lithium-ion batteries. Oxalate is introduced as leaching reagent meanwhile as precipitant which leaches and precipitates cobalt from LiCoO 2 and CoO directly as CoC 2 O 4 ·2H 2 O with 1.0 M oxalate solution at 80 °C and solid/liquid ratio of 50 g L −1 for 120 min. The reaction efficiency of more than 98% of LiCoO 2 can be achieved and cobalt and lithium can also be separated efficiently during the hydrometallurgical process. The combined process is simple and adequate for the recovery of valuable metals from spent lithium-ion batteries.

  12. Advance in the study of removal of cesium from radioactive wastewater by inorganic ion exchangers

    International Nuclear Information System (INIS)

    Wang Songping; Wang Xiaowei; Du Zhihui

    2014-01-01

    The excellent performance in the removal of cesium from radioactive wastewater by inorganic ion exchangers has received extensive attention due to their characteristic physico-chemical features. The paper summarized research progress of removal of cesium by different inorganic ion exchangers such as silicoaluminate, salts of hetero polyacid, hexacyanoferrate, insoluble salts of acid with multivalent metals, insoluble hydrous oxides of multivalent metals and silicotitanate and reviewed several removal systems of cesium by inorganic ion exchangers which might offer China some reference in treatment and disposal of radioactive wastewater. (authors)

  13. Ion current prediction model considering columnar recombination in alpha radioactivity measurement using ionized air transportation

    International Nuclear Information System (INIS)

    Naito, Susumu; Hirata, Yosuke; Izumi, Mikio; Sano, Akira; Miyamoto, Yasuaki; Aoyama, Yoshio; Yamaguchi, Hiromi

    2007-01-01

    We present a reinforced ion current prediction model in alpha radioactivity measurement using ionized air transportation. Although our previous model explained the qualitative trend of the measured ion current values, the absolute values of the theoretical curves were about two times as large as the measured values. In order to accurately predict the measured values, we reinforced our model by considering columnar recombination and turbulent diffusion, which affects columnar recombination. Our new model explained the considerable ion loss in the early stage of ion diffusion and narrowed the gap between the theoretical and measured values. The model also predicted suppression of ion loss due to columnar recombination by spraying a high-speed air flow near a contaminated surface. This suppression was experimentally investigated and confirmed. In conclusion, we quantitatively clarified the theoretical relation between alpha radioactivity and ion current in laminar flow and turbulent pipe flow. (author)

  14. Innovative application of ionic liquid to separate Al and cathode materials from spent high-power lithium-ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, Xianlai; Li, Jinhui, E-mail: jinhui@tsinghua.edu.cn

    2014-04-01

    Highlights: • Manual dismantling is superior in spent high-power LiBs recycling. • Heated ionic liquid can effectively separate Al and cathode materials. • Fourier’s law was adopted to determine the heat transfer mechanism. • The process of spent LiBs recycling with heated ionic liquid dismantling was proposed. - Abstract: Because of the increasing number of electric vehicles, there is an urgent need for effective recycling technologies to recapture the significant amount of valuable metals contained in spent lithium-ion batteries (LiBs). Previous studies have indicated, however, that Al and cathode materials were quite difficult to separate due to the strong binding force supplied by the polyvinylidene fluoride (PVDF), which was employed to bind cathode materials and Al foil. This research devoted to seek a new method of melting the PVDF binder with heated ionic liquid (IL) to separate Al foil and cathode materials from the spent high-power LiBs. Theoretical analysis based on Fourier’s law was adopted to determine the heat transfer mechanism of cathode material and to examine the relationship between heating temperature and retention time. All the experimental and theoretic results show that peel-off rate of cathode materials from Al foil could reach 99% when major process parameters were controlled at 180 °C heating temperature, 300 rpm agitator rotation, and 25 min retention time. The results further imply that the application of IL for recycling Al foil and cathode materials from spent high-power LiBs is highly efficient, regardless of the application source of the LiBs or the types of cathode material. This study endeavors to make a contribution to an environmentally sound and economically viable solution to the challenge of spent LiB recycling.

  15. Feasibility study of solidification for low-level liquid waste generated by sulfuric acid elution treatment of spent ion exchange resin

    International Nuclear Information System (INIS)

    Asano, Takashi; Kawasaki, Tooru; Higuchi, Natsuko; Horikawa, Yoshihiko

    2007-01-01

    Low-level liquid waste with relatively high levels of radioactivity is generated by the sulfuric acid elution treatment of spent ion exchange resin used in water purification systems of nuclear power plants. We studied cement-like solidification process for this type waste that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete, and this leads to cracking during repository storage. It is also preferable that the pH of pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Our solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO 4 precipitation occurs with addition of Ba(OH) 2 ·8H 2 O to the liquid waste when the Ba/SO 4 molar ratio > 1. In the FLS procedure, silica fume and blast furnace slag are added to the liquid wastes containing Ba S O 4 precipitate. The CIS reaction yield is over 98% and the pH of pore water of the solid waste is 11.5 or less. Therefore, we think that our solidification process is one of the best methods for treating liquid waste that contains a high concentration of sodium sulfate. (author)

  16. High polymer composites for containers for the long-term storage of spent nuclear fuel and high level radioactive waste

    International Nuclear Information System (INIS)

    Bonin, H.W.; Vui, V.T.; Legault, J.-F.

    1997-01-01

    sufficiently shielded from the radiations emitted by the spent nuclear fuel or other high level radioactive waste, this material may well be an interesting candidate for this application. More recent work at RMC on the effects of radiations on PEEK has demonstrated that this polymer thermoplastic material was even superior to epoxies under radiation environments. Part of this research concentrated on the estimation of the doses accumulated in the container wall over the years using three basic models for the container: one without filling material, one with glass beads as proposed by AECL, and one using thorium dioxide (ThO 2 ) as filling material. This choice is based on the excellent physical and chemical properties of this compound (resistance to corrosion in particular) and to the expected low cost since thorium is usually discarded in the tailings of uranium mine concentrating plants. The dose calculations were carried out using the Microshield software and showed that both the epoxy and the PEEK could maintain structural integrity provided that they are shielded sufficiently against the radiations emitted by the high level radioactive waste. This research investigated also the resistance to the mechanical forces to which the container walls would be submitted in the underground vaults and it was concluded that these materials displayed sufficient mechanical strength for such application. It is permitted the identification of several aspects of the design of the storage containers that needed closer investigation. (author)

  17. Radioecological condition assessment and remediation criteria for sites of spent fuel and radioactive waste storage in the russian northwest

    International Nuclear Information System (INIS)

    Shandala, Nataliya; Titov, Alex; Novikova, Natalia; Kiselev, Mikhail; Romanov, Vladimir; Sneve, Malgorzata; Smith, Graham

    2008-01-01

    The Norwegian Radiation Protection Authority and the Federal Medical-Biological Agency of the Russian Federation have a regulatory cooperation programme which is concerned with management of the nuclear legacy in northwest Russia, and, in particular, the remediation of facilities for spent fuel and radioactive waste management at the former Shore Technical Bases at Andreeva Bay and Gremikha Village. New regulatory guidance documents have been developed, necessary because of the special abnormal situation at these sites, now designated as Sites of Temporary Storage, but also because of the transition from military to civilian regulatory supervision and the evolving regulatory system in the Russian Federation. This paper presents the progress made and on-going projects in 2008 which involve development of the radio-ecological basis for identifying radiation supervision area boundaries and a system of recommended dose constraints and derived control levels for protection of workers and the public. Unconditional guarantee of long-term radioecological protection serves as the basis for criteria development. Non-exceedance of these dose constraints and control levels implies compliance with radiological protection objectives related to the residual contamination. Dose reduction below proposed dose constraint values must also be carried out according to the optimization principle. A number of remediation strategies are considered, corresponding to different future land use assumptions, including interim continued use in a nuclear context. The developed criteria relate to conditions of facilities and surrounding areas at the sites of temporary storage after completion of their remediation, and during the interim stages of remediation, depending upon the remediation strategy adopted. (author)

  18. Production of multicharged radioactive ion beams for spiral: studies and realization of the first target-ion source system

    International Nuclear Information System (INIS)

    Maunoury, L.

    1998-01-01

    In the framework of the SPIRAL project, which concerns the production and the acceleration of a multicharged radioactive ions beam, the following part has been studied: production and ionization of the radioactive ions beam. A first target-source (nanogan II), devoted exclusively to the production of multicharged radioactive ions gas type beams, has been studied and tested. The diffusion efficiency has been deduced from the diffusion equations (Fick laws). This efficiency is governed by the following parameters: the temperature, the grains size of the target, the Arrhenius parameters and the radioactive period. Another study concerning the production targets is presented. It deals with the temperature distribution allowing an utilization of more than one month at a temperature of 2400 K. Another development (SPIRAL II) is devoted to the production of high neutron content radioactive atoms created by the uranium fission, from fast neutrons. The neutrons beam is produced by the ''stripping break-up'' of a deutons beam in a converter. (A.L.B.)

  19. Surface modification of spinel λ-MnO2 and its lithium adsorption properties from spent lithium ion batteries

    International Nuclear Information System (INIS)

    Li, Li; Qu, Wenjie; Liu, Fang; Zhao, Taolin; Zhang, Xiaoxiao; Chen, Renjie; Wu, Feng

    2014-01-01

    Highlights: • A method is designed to synthesize a λ-MnO 2 ion-sieve for lithium ions adsorption. • Ultrasonic treatment with acid is highly efficient for lithium ions extraction. • Surface modification by CeO 2 is used to improve the adsorption capacity. • A 0.5 wt.% CeO 2 -coated ion-sieve shows the best adsorption properties. • λ-MnO 2 ion-sieves are promising for recovering scarce lithium resources. - Abstract: Spinel λ-MnO 2 ion-sieves are promising materials because of their high selectivity toward lithium ions, and this can be applied to the recovery of lithium from spent lithium ion batteries. However, manganese dissolution loss during the delithiation of LiMn 2 O 4 causes a decrease in adsorption capacity and poor cycling stability for these ion-sieves. To improve the lithium adsorption properties of λ-MnO 2 ion-sieves, surface modification with a CeO 2 coating was studied using hydrothermal-heterogeneous nucleation. The structure, morphology and composition of the synthesized materials were determined by XRD, SEM, TEM and EDS. The effect of hydrothermal synthesis conditions and the amount of CeO 2 coating on the adsorption performance of λ-MnO 2 were also investigated. A 0.5 wt.% CeO 2 -coated ion-sieve was synthesized by heating at 120 °C for 3 h and it had better adsorption properties than the bare samples. The effect of ultrasonic treatment on the lithium extraction ratio from LiMn 2 O 4 upon acid treatment at various temperatures was studied and the results were compared with conventional mechanical stirring. We found that ultrasonic treatment at lower temperature gave almost the same maximum lithium extraction ratio and was more efficient and economic

  20. Innovative application of ionic liquid to separate Al and cathode materials from spent high-power lithium-ion batteries.

    Science.gov (United States)

    Zeng, Xianlai; Li, Jinhui

    2014-04-30

    Because of the increasing number of electric vehicles, there is an urgent need for effective recycling technologies to recapture the significant amount of valuable metals contained in spent lithium-ion batteries (LiBs). Previous studies have indicated, however, that Al and cathode materials were quite difficult to separate due to the strong binding force supplied by the polyvinylidene fluoride (PVDF), which was employed to bind cathode materials and Al foil. This research devoted to seek a new method of melting the PVDF binder with heated ionic liquid (IL) to separate Al foil and cathode materials from the spent high-power LiBs. Theoretical analysis based on Fourier's law was adopted to determine the heat transfer mechanism of cathode material and to examine the relationship between heating temperature and retention time. All the experimental and theoretic results show that peel-off rate of cathode materials from Al foil could reach 99% when major process parameters were controlled at 180°C heating temperature, 300 rpm agitator rotation, and 25 min retention time. The results further imply that the application of IL for recycling Al foil and cathode materials from spent high-power LiBs is highly efficient, regardless of the application source of the LiBs or the types of cathode material. This study endeavors to make a contribution to an environmentally sound and economically viable solution to the challenge of spent LiB recycling. Copyright © 2014 Elsevier B.V. All rights reserved.

  1. Novel Approach for in Situ Recovery of Lithium Carbonate from Spent Lithium Ion Batteries Using Vacuum Metallurgy.

    Science.gov (United States)

    Xiao, Jiefeng; Li, Jia; Xu, Zhenming

    2017-10-17

    Lithium is a rare metal because of geographical scarcity and technical barrier. Recycling lithium resource from spent lithium ion batteries (LIBs) is significant for lithium deficiency and environmental protection. A novel approach for recycling lithium element as Li 2 CO 3 from spent LIBs is proposed. First, the electrode materials preobtained by mechanical separation are pyrolyzed under enclosed vacuum condition. During this process the Li is released as Li 2 CO 3 from the crystal structure of lithium transition metal oxides due to the collapse of the oxygen framework. An optimal Li recovery rate of 81.90% is achieved at 973 K for 30 min with a solid-to-liquid ratio of 25 g L -1 , and the purity rate of Li 2 CO 3 is 99.7%. The collapsed mechanism is then presented to explain the release of lithium element during the vacuum pyrolysis. Three types of spent LIBs including LiMn 2 O 4 , LiCoO 2 , and LiCo x Mn y Ni z O 2 are processed to prove the validity of in situ recycling Li 2 CO 3 from spent LIBs under enclosed vacuum condition. Finally, an economic assessment is taken to prove that this recycling process is positive.

  2. Measurement of residual radioactivity in cooper exposed to high energy heavy ion beam

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunjoo; Nakamura, Takashi [Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center; Uwamino, Yoshitomo; Ito, Sachiko; Fukumura, Akifumi

    1999-03-01

    The residual radioactivities produced by high energy heavy ions have been measured using the heavy ion beams of the Heavy Ion Medical Accelerator (HIMAC) at National Institute of Radiological Sciences. The spatial distribution of residual radioactivities in 3.5 cm, 5.5 cm and 10 cm thick copper targets of 10 cm x 10 cm size bombarded by 290 MeV/u, 400 MeV/u-{sup 12}C ion beams and 400 MeV/u-{sup 20}Ne ion beam, respectively, were obtained by measuring the gamma-ray activities of 0.5 mm thick copper foil inserted in the target with a high purity Ge detector after about 1 hour to 6 hours irradiation. (author)

  3. A combined thermal dissociation and electron impact ionization source for radioactive ion beam generationa

    International Nuclear Information System (INIS)

    Alton, G.D.; Williams, C.

    1996-01-01

    The probability for simultaneously dissociating and efficiently ionizing the individual atomic constituents of molecular feed materials with conventional, hot-cathode, electron-impact ion sources is low and consequently, the ion beams from these sources often appear as mixtures of several molecular sideband beams. This fragmentation process leads to dilution of the intensity of the species of interest for radioactive ion beam (RIB) applications where beam intensity is at a premium. We have conceived an ion source that combines the excellent molecular dissociation properties of a thermal dissociator and the high ionization efficiency characteristics of an electron impact ionization source that will, in principle, overcome this handicap. The source concept will be evaluated as a potential candidate for use for RIB generation at the Holifield Radioactive Ion Beam Facility, now under construction at the Oak Ridge National Laboratory. The design features and principles of operation of the source are described in this article. copyright 1996 American Institute of Physics

  4. Study of the Fixation and Migration of Radioactive Cations in a Natural Ion-Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Baetsle, L. [Centre d' Etudes de l' Energie Nucleaire, Mol (Belgium)

    1960-07-01

    The purpose of this study is to analyse the behaviour of Sr90 and Cs137 on natural ion-exchangers such as lignite and soil. Lignite is a substance which is found in large quantities near the Belgian Nuclear Energy Research Centre (CEN) at Mol and is particularly useful in the processing of radioactive liquid wastes because of its ion-exchange properties. The physical and chemical characteristics of lignite which have a bearing on ion exchange are given in section 1 of this paper. The various ion equilibria which affect the processing of.radioactive liquid wastes are studied in section 2, which also lists the basic factors required for calculating the rate of saturation of a lignite column. The speed of ion migration in the soil is studied along the same lines as for lignite.

  5. Constraints due to the production of radioactive ion beams in the SPIRAL project

    International Nuclear Information System (INIS)

    Leroy, R.; Huguet, Y.; Jardin, P.; Marry, C.; Pacquet, J.Y.; Villari, A.C.C.

    1997-01-01

    The radioactive ion beams that will be delivered by the SPIRAL facility will be produced by the interaction of a stable high energy and high intensity primary ion beam delivered by the GANIL cyclotrons with a carbon target heated to 2000 deg C. During this interaction, some radioactive atoms will be created and will diffuse out of the target before entering into an electron cyclotron resonance ion source where they will be ionized and extracted. The production of radioactive ion beams with this method implies high radiation fields that activate and can damage materials located in the neighborhood of the target. Therefore, the production system which is composed of the permanent magnet ECR ion source coupled to a graphite target will be changed after two weeks of irradiation. As this ensemble will be very radioactive, this operation has to be supervised by remote control. The radiation levels around the target-ion source system and a detailed description of the different precautions that have been taken for safety and for prevention of contamination and irradiation are presented. (author)

  6. Spent lithium-ion battery recycling - Reductive ammonia leaching of metals from cathode scrap by sodium sulphite.

    Science.gov (United States)

    Zheng, Xiaohong; Gao, Wenfang; Zhang, Xihua; He, Mingming; Lin, Xiao; Cao, Hongbin; Zhang, Yi; Sun, Zhi

    2017-02-01

    Recycling of spent lithium-ion batteries has attracted wide attention because of their high content of valuable and hazardous metals. One of the difficulties for effective metal recovery is the separation of different metals from the solution after leaching. In this research, a full hydrometallurgical process is developed to selectively recover valuable metals (Ni, Co and Li) from cathode scrap of spent lithium ion batteries. By introducing ammonia-ammonium sulphate as the leaching solution and sodium sulphite as the reductant, the total selectivity of Ni, Co and Li in the first-step leaching solution is more than 98.6% while it for Mn is only 1.36%. In detail understanding of the selective leaching process is carried out by investigating the effects of parameters such as leaching reagent composition, leaching time (0-480min), agitation speed (200-700rpm), pulp density (10-50g/L) and temperature (323-353K). It was found that Mn is primarily reduced from Mn 4+ into Mn 2+ into the solution as [Formula: see text] while it subsequently precipitates out into the residue in the form of (NH 4 ) 2 Mn(SO 3 ) 2 ·H 2 O. Ni, Co and Li are leached and remain in the solution either as metallic ion or amine complexes. The optimised leaching conditions can be further obtained and the leaching kinetics is found to be chemical reaction control under current leaching conditions. As a result, this research is potentially beneficial for further optimisation of the spent lithium ion battery recycling process after incorporating with metal extraction from the leaching solution. Copyright © 2016 Elsevier Ltd. All rights reserved.

  7. The Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management. An instrument to achieve a global safety

    International Nuclear Information System (INIS)

    Risoluti, P.

    2006-01-01

    The Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management (the Joint Convention) is the first legally binding international treaty in the area of radioactive material management. It was adopted by a Diplomatic Conference in September 1997 and opened for signature on 29 September 1997. The Convention entered into force on 18 June 1998, and to date (May 2006) has been ratified by 41 countries. The Joint Convention applies to spent fuel and radioactive waste resulting from civilian application. Its principal aim is to achieve and maintain a high degree of safety in their management worldwide. The Convention is an incentive instrument, not designed to ensure fulfilment of obligations through control and sanction, but by a volunteer peer review mechanism. The obligations of the Contracting Parties are mainly based on the international safety standards developed by the IAEA in past decades. The Convention is of interest of all countries generating radioactive waste. Therefore it is relevant not only for those using nuclear power, but for any country where application of nuclear energy in education, agriculture, medicine and industry is currently used. Obligations of Contracting Parties include attending a Review Meeting held every three years and prepare National Reports for review by the other Contracting Parties. In the National Reports basic information on inventory and facilities for management of radioactive materials has to be provided. Countries with small nuclear power and/or research programs or countries having radioactive materials only from nuclear application on medicine, agriculture or conventional industry, can benefit from the exchange of information and the technical knowledge gained by the reporting procedure set up by the Convention. The second Review Meeting is to be held at IAEA headquarters from 15 to 26 May 2006. This paper presents the objectives and the implementation status of the Convention, the

  8. Radioactive waste management, decommissioning, spent fuel storage. V. 1. Waste management principles, decommissioning, dismantling, operations in hot environment

    International Nuclear Information System (INIS)

    1985-01-01

    This book deals mainly with decommissioning problems concerning more particularly dismantling and decontamination techniques, and radioactive waste processing. Radioactive waste management in France and the French regulation are tackled. Equipments developed for works in hostile environment are also presented [fr

  9. Proceedings of the workshop on prospects for research with radioactive beams from heavy ion accelerators

    International Nuclear Information System (INIS)

    Nitschke, J.M.

    1984-04-01

    The SuperHILAC Users Executive Committee organized a workshop on Prospects for Research with Radioactive Beams from Heavy Ion Accelerators. The main purpose of the workshop was to bring together a diverse group of scientists who had already done experients with radioactive beams or were interested in their use in the future. The topics of the talks ranged from general nuclear physics, astrophysics, production of radioactive beams and high energy projectile fragmentation to biomedical applications. This publication contains the abstracts of the talks given at the workshop and copies of the viewgraphs as they were supplied to the editor

  10. A cylindrical Penning trap for capture, mass selective cooling, and bunching of radioactive ion beams

    International Nuclear Information System (INIS)

    Raimbault-Hartmann, H.; Bollen, G.; Beck, D.; Koenig, M.; Kluge, H.-J.; Schwarz, S.; Schark, E.; Stein, J.; Szerypo, J.

    1997-01-01

    A Penning trap ion accumulator, cooler, and buncher for low-energy ion beams has been developed for the ISOLTRAP mass spectrometer at ISOLDE/CERN. A cylindrical electrode configuration is used for the creation of a nested trapping potential. This is required for efficient accumulation of externally produced ions and for high-mass selectivity by buffer gas cooling. The design goal of a mass resolving power of about 1 x 10 5 has been achieved. Isobar separation has been demonstrated for radioactive rare-earth ion beams delivered by the ISOLDE on-line mass separator. (orig.)

  11. A cylindrical Penning trap for capture, mass selective cooling, and bunching of radioactive ion beams

    CERN Document Server

    Raimbault-Hartmann, H; Bollen, G; König, M; Kluge, H J; Schark, E; Stein, J; Schwarz, S; Szerypo, J

    1997-01-01

    A Penning trap ion accumulator, cooler, and buncher for low energy ion beams has been developed for the ISOLTRAP mass spectrometer at ISOLDE/CERN. A cylindrical electrode configuration is used for the creation of a nested trapping potential. This is required for efficient accumulation of externally produced ions and for high mass selectivity by buffer gas cooling. The design goal of a mass resolving power of about $1\\cdot 10^{5}$ has been achieved. Isobar separation has been demonstrated for radioactive rare earth ion beams delivered by the ISOLDE on-line mass separator.

  12. Latest developments at GANIL for stable and radioactive ion beam production

    International Nuclear Information System (INIS)

    Jardin, P.; Barue, C.; Bajeat, O.; Canet, C.; Clement, E.; Cornell, J. C.; Delahaye, P.; Dubois, M.; Dupuis, M.; Flambard, J. L.; Fraanberg, H.; Frigot, R.; Leboucher, C.; Lecesne, N.; Lecomte, P.; Leherissier, P.; Lemagnen, F.; Leroy, R.; Maunoury, L.; Mery, A.

    2010-01-01

    In the frame of the SPIRAL II (Systeme de Production d'Ions Radioactifs Acceleres en Ligne Partie II) project, several developments of stable and radioactive ion production systems have been started up. In parallel, GANIL has the ambition to preserve the existing stable and radioactive beams and also to increase its range by offering new ones. In order to identify the best directions for this development, a new group called GANISOL has been formed. Its preliminary conclusions and the latest developments at GANIL are presented.

  13. Apparatus for waste disposal of radioactive hazardous waste

    International Nuclear Information System (INIS)

    Burack, R.D.; Stenger, W.J.; Wolfe, C.R.

    1992-01-01

    This patent describes an apparatus for concentrating dissolved and solid radioactive materials carried in a waste water solution containing a hazardous chelating agent used for cleaning nuclear equipment. It comprises oxidizing chamber means, separator means coupled to the oxidizing chamber means; ion exchange means containing an ion exchange resin; dryer means for receiving the radioactive solids from the separator means and for producing dry solids; and packaging means for receiving the dry solids and spent ion exchange resins containing the removed dissolved radioactive materials and for packaging the dry solids and spent resins in solid form

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  15. Laser photodetachment of radioactive ions: towards the determination of the electronegativity of astatine

    CERN Multimedia

    Rothe, Sebastian; Welander, Jakob Emanuel; Chrysalidis, Katerina; Day Goodacre, Thomas; Fedosseev, Valentine; Fiotakis, Spyridon; Forstner, Oliver; Heinke, Reinhard Matthias; Johnston, Karl; Kron, Tobias; Koester, Ulli; Liu, Yuan; Marsh, Bruce; Ringvall Moberg, Annie; Rossel, Ralf Erik; Seiffert, Christoph; Studer, Dominik; Wendt, Klaus; Hanstorp, Dag

    2017-01-01

    Negatively charged ions are mainly stabilized through the electron correlation effect. A measure of the stability of a negative ion is the electron affinity, which the energy gain by attaching an electron to a neutral atom. This fundamental quantity is, due to the almost general lack of bound excited states, the only atomic property that can be determined with high accuracy for negative ions. We will present the results of the first laser photodetachment studies of radioactive negative ions at CERN-ISOLDE. The photodetachment threshold for the radiogenic iodine isotope 128I was measured successfully, demonstrating the performance of the upgraded GANDALPH experimental beam line. The first detection of photo-detached astatine atoms marks a milestone towards the determination of the EA of this radioactive element.

  16. Radioactive ion beam facility at Louvain-La-Neuve, Belgium and its features

    International Nuclear Information System (INIS)

    Chintalapudi, S.N.

    1991-01-01

    Use of radioactive ion beams for the study of nuclear structure as well as the astrophysical reaction cross sections become the current interest in physics. A full-fledged facility based on two coupled cyclotrons comprising a compact high current cyclotron and a medium energy cyclotron with an intermediate target and ion source system has been recently commissioned at the Louvain-La-Neuve University in Belgium by its accelerator group and has been successfully used for the measurement of cross sections for the primordial nucleosynthesis reactions of astrophysical interest, directly. A brief description of the system, its operational features together with some details of the target and the ion source arrangement for the production of the radioactive ion beams and their acceleration to energies required for the proposed studies is presented. Description of the reactions studied by the Louvain La Neuve group for astrophysical interest is also given. (author). 20 refs., 6 figs., 4 tabs

  17. Management of Spent Sealed Radioactive Sources in Central and Eastern Europe (Czech Rep., Estonia, Hungary, Poland and Slovenia)

    International Nuclear Information System (INIS)

    Angus, M.J.; Moreton, A.D.; Wells, D.A.

    2001-04-01

    This study has been performed to consider the situation relating to the regulation and management of spent sealed radioactive sources (SSRS) in five central and Eastern European (C and EE) countries currently being considered for admission to the EU: the Czech Republic, Estonia, Hungary, Poland and Slovenia. The general aim of this study has been to acquire a thorough understanding of the management of SSRS in these five countries, in order to recommend improvements in management schemes and to establish whether the application of common disposal criteria would be advantageous. This report is structured in the following manner; following the Introduction (Section 1), there is description of current and proposed regulatory requirements in the EU, together with a summarised comparison of the regulatory systems in C and EE countries with EU standards in Section 2. Sections 3 to 7 are dedicated to the situation in each of the five countries. Each of these sections is similarly sub-divided to enable country-by-country and topic-by-topic comparison. In each of Sections 3 to 7 there is an overview, description of the sealed source inventory, regulations, current management practices, retrieval of unregistered SSRS, conclusions and a description of possible future technical assistance projects. Section 8 brings together a summary of the situation in each country, with conclusions and both country-specific and generic recommendations. A common concern in the five countries and also in existing EU member states is the problem of accidental inclusion of SSRS in consignments of scrap-metal. The detection of radioactive material at entrances to scrap metal facilities and at national borders has therefore received considerable attention in recent years. Practical issues regarding the detection of SSRS in scrap metal are described in Appendix A. None of the five countries considered in this report have any plans to develop regional disposal facilities and no specific common

  18. Charge breeding of radioactive isotopes at the CARIBU facility with an electron beam ion source

    Science.gov (United States)

    Vondrasek, R. C.; Dickerson, C. A.; Hendricks, M.; Ostroumov, P.; Pardo, R.; Savard, G.; Scott, R.; Zinkann, G.

    2018-05-01

    An Electron Beam Ion Source Charge Breeder (EBIS-CB) has been developed at Argonne National Laboratory as part of the californium rare ion breeder upgrade. For the past year, the EBIS-CB has been undergoing commissioning as part of the ATLAS accelerator complex. It has delivered both stable and radioactive beams with A/Q 18% into a single charge state. The operation of this device, challenges during the commissioning phase, and future improvements will be discussed.

  19. Targets for production of high-intensity radioactive ion-beams

    International Nuclear Information System (INIS)

    Hagebo, E.; Hoff, P.; Steffensen, K.

    1991-01-01

    The recent developments of target systems for production of high intensity radioactive ion-beams at the ISOLDE mass separators is described. Methods for chemically selective production through separation of molecular ions are outlined and the effects of the addition of reactive gases has been studied. Results and further possible applications in the light element region are discussed. (author) 10 refs.; 9 figs.; 1 tab

  20. Quality control of concretes for conditioning of spent radioactive sources; Control de calidad de concretos para acondicionamiento de fuentes radiactivas gastadas

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez N, M.

    2015-07-01

    The spent sealed radioactive sources are considered as a specific type of radioactive wastes and should be properly stored to ensure their integrity and prevent or limit the release of radionuclides in the geosphere. For this, these sources can be put up in concrete matrices. This research presents the evaluation and characterization of five concretes prepared with 4 brands of commercial cements: CPC Extra RS, CPC 30R Impercem of Cemex, Cruz Azul CPC 30R and CPC 30R of Apasco; three sizes of coarse aggregate (<30 mm, 29-11 mm and <10 mm) and fine aggregate (0.0797 mm) used as matrices for conditioning of spent sealed radioactive sources, in order to verify if these specific concretes accredit the standard NOM-019-Nucl-1995. After hardening for 28 days the concrete specimens were subjected to the tests: compressive strength; thermal cycles, irradiation, leaching and permeability, later to be characterized by: 1) X-ray diffraction in order to meet their crystalline phases; 2) scanning electron microscopy, to determine changes in morphology; 3) infrared spectroscopy, to determine the structural changes of concrete from its functional groups; 4) Raman spectroscopy to determine their structural changes and 5) Moessbauer spectroscopy, which determines changes in the oxidation state of iron in the concrete. According to the results and the changes presented by each concrete after applying the tests set by NOM-019-Nucl-1995, is concluded that the concrete made with cement Cemex brand (CPC 30-RS Extra), gravel of particle size 11-29 mm and sieved sand (0.0797 mm) can be used as matrices of spent sealed sources conditioning. Is remarkable a morphological and structural change of the concrete due to gamma irradiation and heat treatment. (Author)

  1. Radioactive core ions of microclusters, ``snowballs`` in superfluid helium

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, N. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Shimoda, T. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Fujita, Y. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Miyatake, H. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Mizoi, Y. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Kobayashi, H. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Sasaki, M. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Shirakura, T. [Osaka Univ., Toyonaka (Japan). Dept. of Physics; Itahashi, T. [Research Center for Nuclear Physics, Osaka Univ., Ibaraki (Japan); Mitsuoka, S. [Research Center for Nuclear Physics, Osaka Univ., Ibaraki (Japan); Matsukawa, T. [Naruto Univ. of Education, Tokushima (Japan); Ikeda, N. [Kyushu Univ., Fukuoka (Japan). Dept. of Physics; Morinobu, S. [Kyushu Univ., Fukuoka (Japan). Dept. of Physics; Hinde, D.J. [Australian National Univ., Canberra, ACT (Australia). Research School of Physical Sciences; Asahi, K. [Tokyo Inst. of Tech. (Japan). Dept. of Physics; Ueno, H. [Tokyo Inst. of Tech. (Japan). Dept. of Physics; Izumi, H. [Tokyo Inst. of Tech. (Japan). Dept. of Physics

    1996-12-01

    Short-lived beta-ray emitters, {sup 12}B, sustaining nuclear spin polarization were introduced into superfluid helium. The nuclear polarization of {sup 12}B was observed via measurement of beta-ray asymmetry. It was found that the nuclear polarization was preserved throughout the lifetime of {sup 12}B (20.3 ms). This suggests that the ``snowball``, an aggregation of helium atoms produced around an alien ion, constitutes a suitable milieu for freezing-out the nuclear spin of the core ion and that most likely the solidification takes place at the interior of the aggregation. (orig.).

  2. Radioactive core ions of microclusters, ''snowballs'' in superfluid helium

    International Nuclear Information System (INIS)

    Takahashi, N.; Mitsuoka, S.; Matsukawa, T.; Ikeda, N.; Morinobu, S.; Hinde, D.J.; Asahi, K.; Ueno, H.; Izumi, H.

    1996-01-01

    Short-lived beta-ray emitters, 12 B, sustaining nuclear spin polarization were introduced into superfluid helium. The nuclear polarization of 12 B was observed via measurement of beta-ray asymmetry. It was found that the nuclear polarization was preserved throughout the lifetime of 12 B (20.3 ms). This suggests that the ''snowball'', an aggregation of helium atoms produced around an alien ion, constitutes a suitable milieu for freezing-out the nuclear spin of the core ion and that most likely the solidification takes place at the interior of the aggregation. (orig.)

  3. Treatment of radioactive wastewaters by chemical precipitation and ion exchange

    International Nuclear Information System (INIS)

    Robinson, S.M.; Begovich, J.M.; Brown, C.H. Jr.; Campbell, D.O.; Collins, E.D.

    1987-01-01

    Precipitation and ion exchange methods are being developed at Oak Ridge National Laboratory to decontaminate wastewaters containing small amounts of 90 Sr and 137 Cs while minimizing waste generation. Distribution coefficients have been determined for strontium and cesium as functions of Ca, Na, and Mg concentrations from bench- and pilot-scale data for ion exchange resins and zeolites using actual wastewaters. Models have been used to estimate the total amount of waste that would be generated at full-scale operation. Based on these data, four process flowsheets are being tested at full-scale. 14 refs., 8 figs., 7 tabs

  4. The Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, J.D. [Oak Ridge National Lab., TN (United States)

    1996-12-31

    The status of the new Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory (ORNL), which is slated to start its scientific program late this year is discussed, as is the new experimental equipment which is being constructed at this facility. Information on the early scientific program also is given.

  5. The Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Garrett, J.D.

    1996-01-01

    The status of the new Holifield Radioactive Ion Beam Facility at Oak Ridge National Laboratory (ORNL), which is slated to start its scientific program late this year is discussed, as is the new experimental equipment which is being constructed at this facility. Information on the early scientific program also is given

  6. Colorimetric detection and removal of radioactive Co ions using sodium alginate-based composite beads

    International Nuclear Information System (INIS)

    Kim, Daigeun; Jo, Ara; Yang, Hee-Man; Seo, Bum-Kyoung; Lee, Kune-Woo; Lee, Taek Seung

    2017-01-01

    Highlights: • Organic/inorganic hybridized alginate beads were newly synthesized via sol-gel chemistry. • Interaction between the azopyridine and metal ion is the main cause of Co ion detection. • The beads showed improved stability and least leakage of azopyridine during use. • Removal of Co ion was assessed by the ion-exchange of carboxylate groups in alginate. • The beads with dual functions of detection and removal of Co ion were successfully accomplished. - Abstract: We demonstrate a simple method for the visual determination and removal of Co ions using a bead-shaped, capturing probe based on hybridized sodium alginate. For Co ions, the designed protocol consisted of three main constituents: an azopyridine-based Co ion-probe for visual detection; sodium alginate as an adsorbent for the Co ion and a bead construct for removal and structure; silica as a linker for the probe and the alginate, leading to a robust structure. When the composite beads were exposed to Co ions, the yellow color of the beads turned to intensive violet and the color intensity was associated with the Co ion concentration. The color variation was quantified using red-green-blue (RGB) color values that were obtained with a scanner and evaluated with Photoshop. The technique achieved both visual recognition with obvious color change of the beads and efficient removal of the radioactive 60 Co ion. The sensing and removal of any radioactive isotope could be achieved with an appropriate sensing probe, to provide a simple and universal platform for remediation.

  7. Colorimetric detection and removal of radioactive Co ions using sodium alginate-based composite beads

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Daigeun; Jo, Ara [Organic and Optoelectronic Materials Laboratory, Department of Organic Materials Engineering, Chungnam National University, Daejeon 34134 (Korea, Republic of); Yang, Hee-Man; Seo, Bum-Kyoung; Lee, Kune-Woo [Decontamination and Decommissioning Research Division, Korea Atomic Energy Research Institute, Daejeon 34057 (Korea, Republic of); Lee, Taek Seung, E-mail: tslee@cnu.ac.kr [Organic and Optoelectronic Materials Laboratory, Department of Organic Materials Engineering, Chungnam National University, Daejeon 34134 (Korea, Republic of)

    2017-03-15

    Highlights: • Organic/inorganic hybridized alginate beads were newly synthesized via sol-gel chemistry. • Interaction between the azopyridine and metal ion is the main cause of Co ion detection. • The beads showed improved stability and least leakage of azopyridine during use. • Removal of Co ion was assessed by the ion-exchange of carboxylate groups in alginate. • The beads with dual functions of detection and removal of Co ion were successfully accomplished. - Abstract: We demonstrate a simple method for the visual determination and removal of Co ions using a bead-shaped, capturing probe based on hybridized sodium alginate. For Co ions, the designed protocol consisted of three main constituents: an azopyridine-based Co ion-probe for visual detection; sodium alginate as an adsorbent for the Co ion and a bead construct for removal and structure; silica as a linker for the probe and the alginate, leading to a robust structure. When the composite beads were exposed to Co ions, the yellow color of the beads turned to intensive violet and the color intensity was associated with the Co ion concentration. The color variation was quantified using red-green-blue (RGB) color values that were obtained with a scanner and evaluated with Photoshop. The technique achieved both visual recognition with obvious color change of the beads and efficient removal of the radioactive {sup 60}Co ion. The sensing and removal of any radioactive isotope could be achieved with an appropriate sensing probe, to provide a simple and universal platform for remediation.

  8. Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel and Transuranic Radioactive Wastes (40 CFR Part 191)

    Science.gov (United States)

    This regulation sets environmental standards for public protection from the management and disposal of spent nuclear fuel, high-level wastes and wastes that contain elements with atomic numbers higher than uranium (transuranic wastes).

  9. Preparation of the National Programme for the Spent Fuel and Radioactive Waste Management Taking Into Account Possibility of Potential Multinational/Regional Disposal Facilities Development

    International Nuclear Information System (INIS)

    Kegel, L.

    2016-01-01

    Conclusions: • Final disposal in deep geological repository (national, regional or multinational) is planed: → Implementation of disposal after NPP closure (>2065). • The strategy principle of international cooperation: → National responsibility for radioactive waste and spent fuel management is considered in parallel with active participation in international regional efforts to make progress in connection to joint regional programmes on disposal. • Implementation is challenging but technical feasible. • Timely and appropriate “nesting” of multinational solutions into national plans. • Although a multinational repository is likely not ripe for development today, actions taken now can be important to increase the likelihood of its future development

  10. Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. National Report from Norway

    International Nuclear Information System (INIS)

    2006-05-01

    This report is the Norwegian report to the second review meeting to the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The comments, questions and remarks given to Norway's initial national report and Norway's presentation given at the first review meeting have been incorporated in this report. The second report is a full revision of the first report. This report concludes that Norway meets the obligations of the Joint Convention. However, Norwegian authorities will aim for development in the waste management policy and Norway will continue to improve its existing systems to further enhance safety, in line with the aims of the Joint Convention

  11. Processing of spent fuels and radioactive waste coming from foreigner: application of the new decree L.542-2--1 from the code of environment

    International Nuclear Information System (INIS)

    Lahorgue, M.B.

    2008-01-01

    This decree clarifies the procedures applicable to the treatment of spent fuels and waste resulting from the foreigner. He has to have preliminary agreement between the governments and the obligation for the operator to set up devices of distribution of the waste as well as system of follow-up of entrances and exit out of the territory of radioactive substances. In any case, the storage of the waste has to exceed the technical period imposed by the reprocessing. An annual report must be put back by the operator and given to the public in two newspapers of national distribution. (N.C.)

  12. The Integrated Data Base program: An executive-level data base of spent fuel and radioactive waste inventories, projections, and characteristics

    International Nuclear Information System (INIS)

    Klein, J.A.

    1987-01-01

    The Integrated Data Base (IDB) is the official US Department of Energy (DOE) data base for spent fuel and radioactive waste inventories and projections. As such, it should be as convenient to utilize as is practical. Examples of summary-level tables and figures are presented, as well as more-detailed graphics describing waste-form distribution by site and line charts illustrating historical and projected volume (or mass) changes. This information is readily accessible through the annual IDB publication. Other presentation formats are also available to the DOE community through a simple request to the IDB Program

  13. Magnetic and magnetostrictive properties of RE-doped Cu-Co ferrite fabricated from spent lithium-ion batteries

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Guoxi, E-mail: wlnfu107@126.com; Wang, Lu, E-mail: hnsdwl314@163.com; Zhao, Tingting

    2017-02-15

    Magnetostrictive Cu{sub 0.1}Co{sub 0.9}RE{sub x}Fe{sub 2-x}O{sub 4} (RE=Ho, Gd or Sm) was fabricated by a sol-gel auto-combustion technique using spent lithium-ion batteries as raw materials. X-ray diffraction analysis confirmed the spinel structure of the RE-incorporated samples with limited RE solubility. Field-emission scanning electron microscopy and Fourier transform infrared spectroscopy revealed a layered structure composed of particles and the cation distribution. Magnetic hysteresis loops and magnetostriction strain curves showed that the saturation magnetization, magnetostriction coefficient and strain derivative were significantly modified due to the substitution of larger ionic radius RE{sup 3+} ions for Fe{sup 3+} ions, influencing the interaction between the tetrahedral and octahedral sites. - Highlights: • Magnetostrictive Cu{sub 0.1}Co{sub 0.9}RE{sub x}Fe{sub 2−x}O{sub 4} (RE=Ho, Gd or Sm, x=0.0–0.25) nanocomposites were fabricated via sol-gel auto-combustion route using spent lithium-ion batteries as raw materials. • The RE elements doping had limited solubility. • The saturation magnetization (M{sub s}) and maximum magnetostriction (λ{sub max}) were reduced and the lattice parameter (a) was increasing by increasing RE{sup 3+} substitution contents. • The relationship of maximum strain derivative (dλ/dH{sub max}) after the incorporation of RE was Ho>Gd>Sm.

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

    International Nuclear Information System (INIS)

    St John, C.M.

    1977-04-01

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

  15. Recent radioactive ion beam program at RIKEN and related topics

    Indian Academy of Sciences (India)

    Recent experimental programs at RIKEN concerning RI beams are reviewed. RIKEN has the ring cyclotron (RRC) with high intense heavy-ion beams and large acceptance fragment separator, RIPS. The complex can provide high intense RI-beams. By using the high intense RI-beams, a variety of experiments have been ...

  16. Implications of theories of asteroid and comet impact for policy options for management of spent nuclear fuel and high-level radioactive wastes

    Science.gov (United States)

    Trask, Newell J.

    1994-01-01

    Concern with the threat posed by terrestrial asteroid and comet impacts has heightened as the catastrophic consequences of such events have become better appreciated. Although the probabilities of such impacts are very small, a reasonable question for debate is whether such phenomena should be taken into account in deciding policy for the management of spent fuel and high-level radioactive waste. The rate at which asteroid or comet impacts would affect areas of surface storage of radioactive waste is about the same as the estimated rate at which volcanic activity would affect the Yucca Mountain area. The Underground Retrievable Storage (URS) concept could satisfactorily reduce the risk from cosmic impact with its associated uncertainties in addition to providing other benefits described by previous authors.

  17. Implications of theories of asteroid and comet impact for policy options for management of spent nuclear fuel and high-level radioactive wastes

    International Nuclear Information System (INIS)

    Trask, N.J.

    1994-01-01

    Concern with the threat posed by terrestrial asteroid and comet impacts has heightened as the catastrophic consequences of such events have become better appreciated. Although the probabilities of such impacts are very small, a reasonable question for debate is whether such phenomena should be taken into account in deciding policy for the management of spent fuel and high-level radioactive waste. The rate at which asteroid or comet impacts would affect areas of surface storage of radioactive waste is about the same as the estimated rate at which volcanic activity would affect the Yucca Mountain area. The Underground Retrievable Storage (URS) concept could satisfactorily reduce the risk from cosmic impact with its associated uncertainties in addition to providing other benefits described by previous authors

  18. A study on the direct use of spent PWR fuel in CANDU -A study on the radioactive waste management for DUPIC fuel cycle-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Jun, Kwan Sik; Nah, Jung Won; Park, Jang Jin; Kim, Jong Hoh; Cho, Yung Hyun; Baek, Seung Woo; Shin, Jin Myung; Yang, Seung Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    The immobilization materials for radioactive wastes resulting from the DUPIC fuel manufacturing process were selected and their characteristics were evaluated. To predict the trapping behavior of the Ruthenium, a semi-volatile nuclide, its volatility was measured and thermogravimetric analysis were performed with simulated fuel. New Ruthenium trapping material was developed which is deposited on ceramic honey-comb monolith of cordierite. The base glass was manufactured with fly ash added to the borosilicate glass. The composition of the scrap waste was calculated based on the PWR spent fuel which has initial {sup 235}U content of 3.5%, burnup of 35,000 MWD/MTU and cooling time of 10 years. Simulated waste glass was manufactured, and its chemical durability was evaluated by soxhlet leach test. Radioactivity of non-oxidized cladding material were measured. The preliminary design criteria were prepared for off-gas treatment system in IMEF. 31 figs, 42 tabs, 51 refs. (Author).

  19. A study on the direct use of spent PWR fuel in CANDU -A study on the radioactive waste management for DUPIC fuel cycle-

    International Nuclear Information System (INIS)

    Park, Hyun Soo; Jun, Kwan Sik; Nah, Jung Won; Park, Jang Jin; Kim, Jong Hoh; Cho, Yung Hyun; Baek, Seung Woo; Shin, Jin Myung; Yang, Seung Yung

    1994-07-01

    The immobilization materials for radioactive wastes resulting from the DUPIC fuel manufacturing process were selected and their characteristics were evaluated. To predict the trapping behavior of the Ruthenium, a semi-volatile nuclide, its volatility was measured and thermogravimetric analysis were performed with simulated fuel. New Ruthenium trapping material was developed which is deposited on ceramic honey-comb monolith of cordierite. The base glass was manufactured with fly ash added to the borosilicate glass. The composition of the scrap waste was calculated based on the PWR spent fuel which has initial 235 U content of 3.5%, burnup of 35,000 MWD/MTU and cooling time of 10 years. Simulated waste glass was manufactured, and its chemical durability was evaluated by soxhlet leach test. Radioactivity of non-oxidized cladding material were measured. The preliminary design criteria were prepared for off-gas treatment system in IMEF. 31 figs, 42 tabs, 51 refs. (Author)

  20. Production of a radioactive 18F ion beam for nuclear reaction studies

    Science.gov (United States)

    Roberts, A. D.; Nickles, R. J.; Paul, M.; Rehm, K. E.; Jiang, C. L.; Blumenthal, D. J.; Gehring, J.; Henderson, D.; Nolen, J.; Pardo, R. C.; Schiffer, J. P.; Segel, R. E.

    1995-12-01

    A two-stage method for generating a radioactive 18F ion beam has been developed. 18F is produced with a medical cyclotron by 11 MeV proton activation of [ 18O]water, then chemically processed off-line for use in a tandem accelerator ion source. Azeotropic distillation reduces the 18O component by 10 5, with a resulting 18O to 18F beam ratio of about 10 3. The average 18F - beam intensity per synthesis is 1 ppA over 120 min from a cesium vapor, sputter negative ion source (SNICS), with a peak intensity of 4.5 ppA.

  1. Use of glucose as reductant to recover Co from spent lithium ions batteries.

    Science.gov (United States)

    Meng, Qi; Zhang, Yingjie; Dong, Peng

    2017-06-01

    A hydrometallurgical leaching process has been developed for recovery of Co and Li from cathode material (LiCoO 2 ) collected from spent LIBs using a mix solution of glucose and phosphoric acid. The spent LiCoO 2 before and after leaching process are analyzed by scanning electron microscopy. A leaching rate of about 98% Co and nearly 100% Li is presented with 1.5mol/L phosphoric acid and 0.02mol/L glucose at 80°C in about 2h. During leaching process, glucose was oxidized into monocarboxylic acid with reduction of Co(III) to Co(II). Co in solution was recovered as Co-oxalate after leaching process. Using glucose as reductant to dissolve LiCoO 2 with chelating agent of phosphoric acid is achieved here. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. SETH 200: new mobile unit for spent ion-exchange resins embedding in polymers

    International Nuclear Information System (INIS)

    de Buzonniere, A.; Raibaud, J.; Augustin, X.

    1985-01-01

    The CEA embedding process for low- and medium-activity waste in thermosetting resins (polyester or epoxy) has been used industrially. Recent developments (elimination of chemical pretreatment thanks to a new epoxy formulation and technological breakthroughs in the operating techniques) have greatly increased the potential of the process and have allowed with Technicatome's industrial experience, the elaboration of a new mobile unit easily operated and very competitive, particularly in spent resin processing

  3. Preliminary shielding estimates for the proposed Oak Ridge National Laboratory (ORNL) Radioactive Ion Beam Facility (RIBF)

    International Nuclear Information System (INIS)

    Johnson, J.O.; Gabriel, T.A.; Lillie, R.A.

    1996-01-01

    The Oak Ridge National Laboratory (ORNL) has proposed designing and implementing a new target-ion source for production and injection of negative radioactive ion beams into the Hollifield tandem accelerator. This new facility, referred to as the Radioactive Ion Beam Facility (RIBF), will primarily be used to advance the scientific communities' capabilities for performing state-of-the-art cross-section measurements. Beams of protons or other light, stable ions from the Oak Ridge Isochronous Cyclotron (ORIC) will be stopped in the RIBF target ion source and the resulting radioactive atoms will be ionized, charge exchanged, accelerated, and injected into the tandem accelerator. The ORIC currently operates with proton energies up to 60 MeV and beam currents up to 100 microamps with a maximum beam power less than 2.0 kW. The proposed RIBF will require upgrading the ORIC to generate proton energies up to 200 MeV and beam currents up to 200 microamps for optimum performance. This report summarizes the results of a preliminary one-dimensional shielding analysis of the proposed upgrade to the ORIC and design of the RIBF. The principal objective of the shielding analysis was to determine the feasibility of such an upgrade with respect to existing shielding from the facility structure, and additional shielding requirements for the 200 MeV ORIC machine and RIBF target room

  4. Ion exchanger material based on Titanium phosphate for liquid radioactive waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Maslova, M.; Gerasimova, L. [Tananaev Institute of Chemistry and Technology of Rare Elements and Mineral Resources, Kola Scientific Center, Russian Academy of Sciences, Apatity (Russian Federation)

    2013-07-01

    A comparative study of the physicochemical and service properties of samples of Ti(OH) 1.36(HPO4) 1.32 * 2.3H2O sorbent in the finely dispersed and granulated forms, mastered for commercial production, was made. The sorption of Cs and Sr cations from solutions of various, compositions was studied in batch experiments, and the diffusion coefficients of the exchanging ions were determined. The hydrolytic stability of the. sorbents was examined with the aim to determine the optimal operation conditions. Experiments showed that the cation exchangers based on titanium phosphate are the most efficient in removal from liquid radioactive waste of induced radioactive isotopes of corrosion products, which is due to formation of weakly dissociating compounds of nonferrous metal ions with functional groups of the ion exchangers in the sorbent phase. (author)

  5. Development of target ion source systems for radioactive beams at GANIL

    Energy Technology Data Exchange (ETDEWEB)

    Bajeat, O., E-mail: bajeat@ganil.fr [GANIL, BP 55027, 14076 CAEN Cedex 05 (France); Delahaye, P. [GANIL, BP 55027, 14076 CAEN Cedex 05 (France); Couratin, C. [GANIL, BP 55027, 14076 CAEN Cedex 05 (France); LPC Caen, 6 bd Maréchal Juin, 14050 CAEN Cedex (France); Dubois, M.; Franberg-Delahaye, H.; Henares, J.L.; Huguet, Y.; Jardin, P.; Lecesne, N.; Lecomte, P.; Leroy, R.; Maunoury, L.; Osmond, B.; Sjodin, M. [GANIL, BP 55027, 14076 CAEN Cedex 05 (France)

    2013-12-15

    Highlights: • For Spiral 1, a febiad ion source has been connected to a graphite target. • For Spiral 2, an oven made with a carbon resistor is under development. • We made some measurement of effusion in the Spiral 2 target. • A laser ion source is under construction. -- Abstract: The GANIL facility (Caen, France) is dedicated to the acceleration of heavy ion beams including radioactive beams produced by the Isotope Separation On-Line (ISOL) method at the SPIRAL1 facility. To extend the range of radioactive ion beams available at GANIL, using the ISOL method two projects are underway: SPIRAL1 upgrade and the construction of SPIRAL2. For SPIRAL1, a new target ion source system (TISS) using the VADIS FEBIAD ion source coupled to the SPIRAL1 carbon target will be tested on-line by the end of 2013 and installed in the cave of SPIRAL1 for operation in 2015. The SPIRAL2 project is under construction and is being design for using different production methods as fission, fusion or spallation reactions to cover a large area of the chart of nuclei. It will produce among others neutron rich beams obtained by the fission of uranium induced by fast neutrons. The production target made from uranium carbide and heated at 2000 °C will be associated with several types of ion sources. Developments currently in progress at GANIL for each of these projects are presented.

  6. Recovery of lithium from the effluent obtained in the process of spent lithium-ion batteries recycling.

    Science.gov (United States)

    Guo, Xueyi; Cao, Xiao; Huang, Guoyong; Tian, Qinghua; Sun, Hongyu

    2017-08-01

    A novel process of lithium recovery as lithium ion sieve from the effluent obtained in the process of spent lithium-ion batteries recycling is developed. Through a two-stage precipitation process using Na 2 CO 3 and Na 3 PO 4 as precipitants, lithium is recovered as raw Li 2 CO 3 and pure Li 3 PO 4 , respectively. Under the best reaction condition (both the amounts of Na 2 CO 3 and Li 3 PO 4 vs. the theoretical ones are about 1.1), the corresponding recovery rates of lithium (calculated based on the concentration of the previous stage) are 74.72% and 92.21%, respectively. The raw Li 2 CO 3 containing the impurity of Na 2 CO 3 is used to prepare LiMn 2 O 4 as lithium ion sieve, and the tolerant level of sodium on its property is studied through batch tests of adsorption capacity and corrosion resistance. When the weight percentage of Na 2 CO 3 in raw Li 2 CO 3 is controlled less than 10%, the Mn corrosion percentage of LiMn 2 O 4 decreases to 21.07%, and the adsorption capacity can still keep at 40.08 mg g -1 . The results reveal that the conventional separation sodium from lithium may be avoided through the application of the raw Li 2 CO 3 in the field of lithium ion sieve. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. Acquisition of Co metal from spent lithium-ion battery using emulsion liquid membrane technology and emulsion stability test

    Science.gov (United States)

    Yuliusman; Wulandari, P. T.; Amiliana, R. A.; Huda, M.; Kusumadewi, F. A.

    2018-03-01

    Lithium-ion batteries are the most common type to be used as energy source in mobile phone. The amount of lithium-ion battery wastes is approximated by 200 – 500 ton/year. In one lithium-ion battery, there are 5 – 20% of cobalt metal, depend on the manufacturer. One of the way to recover a valuable metal from waste is leaching process then continued with extraction, which is the aim of this study. Spent lithium-ion batteries will be characterized with EDX and AAS, the result will show the amount of cobalt metal with form of LiCoO2 in the cathode. Hydrochloric acid concentration used is 4 M, temperature 80°C, and reaction time 1 hour. This study will discuss the emulsion stability test on emulsion liquid membrane. The purpose of emulsion stability test in this study was to determine optimum concentration of surfactant and extractant to produce a stable emulsion. Surfactant and extractant used were SPAN 80 and Cyanex 272 respectively with both concentrations varied. Membrane and feed phase ratios used in this experiment was 1 : 2. The optimum results of this study were SPAN 80 concentrations of 10% w/v and Cyanex 272 0.7 M.

  8. Recovery of lithium from the effluent obtained in the process of spent lithium-ion batteries recycling

    DEFF Research Database (Denmark)

    Guo, Xueyi; Cao, Xiao; Huang, Guoyong

    2017-01-01

    A novel process of lithium recovery as lithium ion sieve from the effluent obtained in the process of spent lithium-ion batteries recycling is developed. Through a two-stage precipitation process using Na2CO3 and Na3PO4 as precipitants, lithium is recovered as raw Li2CO3 and pure Li3PO4...... of Na2CO3 is used to prepare LiMn2O4 as lithium ion sieve, and the tolerant level of sodium on its property is studied through batch tests of adsorption capacity and corrosion resistance. When the weight percentage of Na2CO3 in raw Li2CO3 is controlled less than 10%, the Mn corrosion percentage of LiMn2......O4 decreases to 21.07%, and the adsorption capacity can still keep at 40.08 mg g-1. The results reveal that the conventional separation sodium from lithium may be avoided through the application of the raw Li2CO3 in the field of lithium ion sieve....

  9. Leaching of Co and Cs from spent ion exchange resins in cement ...

    Indian Academy of Sciences (India)

    Unknown

    2003-08-22

    Aug 22, 2003 ... active waste management for a future Serbian radioactive waste disposal centre. Keywords. Cement ... cost, easy operation and radiation and thermal stability. It is generally .... Initial activity, A0 = 8,0⋅107 Bq/per sample (60Co and 137Cs). Table 2. ... Hespe E D 1971 Atomic Energy Rev. 9 195. Plecas I ...

  10. Effects of gaseous radioactive nuclides on the design and operation of repositories for spent LWR fuel in rock salt

    International Nuclear Information System (INIS)

    Jenks, G.H.

    1979-12-01

    Information relating to the identities and amounts of gaseous radionuclides present in spent LWR fuel and to their release from canistered spent fuel under plausible storage and disposal conditions was assembled, reviewed, and analyzed. Information was also reviewed and analyzed on several other subjects that relate to the integrity of the carbon steel canister in which the spent fuel is to be encapsulated and to the expected rates of transfer of gaseous radionuclides through crushed salt backfill within a disposal room in a reference repository in rock salt. The advantages and disadvantages were considered for several different canister-backfill materials, and recommendations were made regarding preferred materials. Other recommendations relate to encapsulation procedures and specifications and to needs for additional experimental studies. The objective of this work was to provide reference information, conclusions, and recommendations that could be used to establish design and operating conditions and procedures for a bedded salt repository for spent LWR fuel and that could also be used to help evaluate the safety of the repository. The results of this work will also generally apply to spent fuel repositories in domal salt. However, because the domal salt may have little or no brine inclusions within it, there may be little or no possibility that brine will migrate into open spaces around an emplaced canister. Addordingly, some of the concerns that result from the possible occurrence of brine migration in bedded salt may be of no importance in domal salt

  11. Resistance to radiation and concretes thermal cycles for conditioning of spent radioactive sources; Resistencia a la irradiacion y ciclos termicos de concretos para acondicionamiento de fuentes radiactivas gastadas

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez N, M.; Monroy G, F.; Gonzalez D, R. C.; Corona P, I. J.; Ortiz A, G., E-mail: fabiola.monroy@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In order to know the concrete type most suitable for use as a matrix of conditioning of spent radioactive sources, concrete test tubes using 4 different types of cement were prepared: CPC 30-Rs Extra, CPC 30-R Impercem, CPC 30-R Rs and CPC 30-R with two gravel sizes >30 mm and <10 mm. The concrete test tubes were subjected to testing compressive strength after 28 days of hardening and after being irradiated and subjected to thermal cycles. Subsequently they were characterized by X-ray diffraction and scanning electron microscopy, in order to evaluate whether these concretes accredited the tests set by the NOM-019-Nucl-1995. The results show that the compressive strength of the hardened concretes to 28 days presents values between 36 and 25 MPa; applying irradiation the resistance may decrease to 30% of its original strength; and if subjected to high and low temperatures the ettringite formation also causes a decrease in resistance. The results show that concretes made from cement Impercem, Cruz Azul with gravel <10 mm comply with the provisions of standard and they can be used for conditioning of spent radioactive sources. (Author)

  12. Recycling of spent adsorbents for oxyanions and heavy metal ions in the production of ceramics.

    Science.gov (United States)

    Verbinnen, Bram; Block, Chantal; Van Caneghem, Jo; Vandecasteele, Carlo

    2015-11-01

    Spent adsorbents for oxyanion forming elements and heavy metals are classified as hazardous materials and they are typically treated by stabilization/solidification before landfilling. The use of lime or cement for stabilization/solidification entails a high environmental impact and landfilling costs are high. This paper shows that mixing spent adsorbents in the raw material for the production of ceramic materials is a valuable alternative to stabilize oxyanion forming elements and heavy metals. The produced ceramics can be used as construction material, avoiding the high economic and environmental impact of stabilization/solidification followed by landfilling. To study the stabilization of oxyanion forming elements and heavy metals during the production process, two series of experiments were performed. In the first series of experiments, the main pollutant, Mo was adsorbed onto iron-based adsorbents, which were then mixed with industrial sludge (3 w/w%) and heated at 1100°C for 30 min. Mo was chosen, as this element is easily adsorbed onto iron-based adsorbents and it is the element that is the most difficult to stabilize (i.e. the highest temperatures need to be reached before the concentrations in the leachate are reduced). Leaching concentration from the 97/3 sludge/adsorbent mixture before heating ranged between 85 and 154 mg/kg; after the heating process they were reduced to 0.42-1.48 mg/kg. Mo was actually stabilized, as the total Mo concentration after addition was not affected by the heat treatment. In the second series of experiments, the sludge was spiked with other heavy metals and oxyanion forming elements (Cr, Ni, Cu, Zn, As, Cd and Pb) in concentrations 5 times higher than the initial concentrations; after heat treatment the leachate concentrations were below the regulatory limit values. The incorporation of spent adsorbents in ceramic materials is a valuable and sustainable alternative to the existing treatment methods, saving raw materials in the

  13. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management. Report of the Federal Republic of Germany for the sixth review meeting in May 2018; Gemeinsames Uebereinkommen ueber die Sicherheit der Behandlung abgebrannter Brennelemente und ueber die Sicherheit der Behandlung radioaktiver Abfaelle. Bericht der Bundesrepublik Deutschland fuer die sechste Ueberpruefungskonferenz im Mai 2018

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2017-08-15

    The joint convention on the safety of spent fuel management and on the safety of radioactive waste management covers the following topics: historical development and actual status of the civil use of nuclear power, politics and the spent fuel management, inventories and listing, legislation and executive systems, other safeguard regulations, safety during spent fuel handling, safety during radioactive waste processing, transport across national borders, disused enclosed radioactive sources, general regulations for safety enhancement.

  14. Charge breeding of stable and radioactive ion beams with EBIS/T devices

    CERN Document Server

    Kester, Oliver; Becker, R

    2004-01-01

    Radioactive ion beams (RIBs) are an important tool for experiments at the foremost frontier of nuclear physics. The quasi-continuous radioactive beams from target ion sources of RIB-facilities have to be accelerated to energies at and beyond the Coulomb barrier. An efficient acceleration requires a suitable A/q of the ions determined by the accelerator design, which can be reached via the stripping method or by using a charge state breeder like the REX-ISOLDE system. In order to get comparable efficiencies for a charge state breeder with the stripping scheme, the breeding efficiency in one charge state has to be optimized by narrowing the charge state distribution. In addition good beam quality and thus small emittances are required to achieve best transmission in the following accelerator, which is mandatory for high intensity RIBs. For EBIS/T devices the maximum intensity of the radioactive ion beam is a critical issue, and high current EBIS/T devices will be necessary to deal with intensities of second gen...

  15. Removal of some ions from the radioactive liquid wastes by means of membrane techniques

    International Nuclear Information System (INIS)

    Roman, Gabriela; Garganciuc, Dana; Batrinescu, Gheorghe; Popescu, Georgeta

    2000-01-01

    The radioactive wastes imply important problems in the pollution control. Contrary to the case of other liquid wastes, which are specifically treated depending on the nature of pollutants, the liquid radioactive wastes are treated as a function of their activity (high, medium or low) and not depending on the nature of radioisotopes. The paper presents the advantages of the membrane processes as comparing with the classical processes in the removal of some ions from liquid radioactive waste up to values admissible of the current standards. Two types of radioactive liquid solutions were processed namely: one solution from the decontamination of the parts of an installation and other from the decontamination of primary circuit of the nuclear power plant. The first solution was treated with ultrafiltration and reverse osmosis, the retention for radioactive and toxic elements ranging between 14 - 69% for ultrafiltration and 63 - 99% for reverse osmosis. The second solution was processed only with reverse osmosis, a retention between 64 - 98% being obtained. The tests proved that by reverse osmosis membrane process a good removal efficiency of radioactive elements from liquid waste is obtained, corresponding to the requirements imposed by the current regulations. (author)

  16. Radioactive ion beams for biomedical research and nuclear medical application

    CERN Document Server

    Beyer, Gerd-Jürgen

    2002-01-01

    The ISOLDE facility at CERN is the world leading on On-Line Isotope Separator installation. The main aspects which makes ISOLDE produced radio-isotopes such valuable for use in biomedical research are: the availability of exotic or uncommon radioisotopes, the high purity and the ion beam quality. A short overview on research strategies, on experimental work and application of ISOLDE produced radionuclides used in the field of biomedicine over a period of more than 2 decades will be given. Special attention will be directed to the radio- lanthanides, because they can be seen as one single element providing the unique possibility to study systematically relationships between molecule parameters and a biological response without changes in the basic tracer molecule. Among those radionuclides we find any radiation properties we wish (single photon emission) suitable for SPECT, positron emission suitable for positron emission tomography (PET), alpha -, beta /sup -/- and Auger electron emission. (21 refs).

  17. Ion exchange in ultramarine blue. Studies using radioactive tracers (1963)

    International Nuclear Information System (INIS)

    Goenvec, H.

    1963-05-01

    A study has been made of the exchange reaction between the constituent sodium in ultramarine blue, and silver in the form of a silver nitrate solution; an attempt hat been made to define the influence of the solvent of the silver nitrate on the kinetics of the exchange reaction. Five solvents have been used: water, methanol, ethanol, n-propanol and acetone. The reaction rates are controlled by a diffusion process inside the ultramarine grains, whatever the solvent used. It seems that the solvated ionic radius of the diffusing ion is one of the factors influencing the rate of reaction. Furthermore, the solvated ionic radius varies with temperature; this variation is different for each solvent and should explain the differences observed in the activation energies. (author) [fr

  18. A review on management of spent lithium ion batteries and strategy for resource recycling of all components from them.

    Science.gov (United States)

    Zhang, Wenxuan; Xu, Chengjian; He, Wenzhi; Li, Guangming; Huang, Juwen

    2018-02-01

    The wide use of lithium ion batteries (LIBs) has brought great numbers of discarded LIBs, which has become a common problem facing the world. In view of the deleterious effects of spent LIBs on the environment and the contained valuable materials that can be reused, much effort in many countries has been made to manage waste LIBs, and many technologies have been developed to recycle waste LIBs and eliminate environmental risks. As a review article, this paper introduces the situation of waste LIB management in some developed countries and in China, and reviews separation technologies of electrode components and refining technologies of LiCoO 2 and graphite. Based on the analysis of these recycling technologies and the structure and components characteristics of the whole LIB, this paper presents a recycling strategy for all components from obsolete LIBs, including discharge, dismantling, and classification, separation of electrode components and refining of LiCoO 2 /graphite. This paper is intended to provide a valuable reference for the management, scientific research, and industrial implementation on spent LIBs recycling, to recycle all valuable components and reduce the environmental pollution, so as to realize the win-win situation of economic and environmental benefits.

  19. Single-column ion chromatography with determination of hydrazoic acid produced in spent nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Ma Guilan; Tan Shuping

    2006-01-01

    The reaction of hydrazine and its derivative with ammonium metavanadate may produce hydrazoic acid (HN 3 ). A single-column ion chromatography is used for the determination of HN 3 after neutralizing the rest acid in the sample with sodium hydroxide. Chromatography separation of HN 3 is carried out on a 25 cm x 0.46 cm (inside diameter) stainless steel column packed with Vydac IC302 ion Chromatography packing. The eluent is 1 mmol/L o-phthalic acid, and the ion is detected by conductivity detector. The detection limit in the presence chromatography is 5 μg/mL, the linear range is from 5 to 201 μg/mL, the linear correlation coefficient is 0.9994, respectively. The analysis accuracy is 2% for standard sample, and the detection limit is 51 μg/mL for HN 3 in the real sample. (authors)

  20. Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. Fourth National Report on Compliance with the Joint Convention Obligations. France

    International Nuclear Information System (INIS)

    2011-09-01

    The Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management, hereinafter referred to as the 'Joint Convention', is the result of international discussions that followed the adoption of the Convention on Nuclear Safety, in 1994. France signed the Joint Convention at the General Conference of the International Atomic Energy Agency (IAEA) held on 29 September 1997, the very first day the Joint Convention was opened for signature. She approved it on 22 February 2000 and filed the corresponding instruments with the IAEA on 27 April 2000. The Joint Convention entered into force on 18 June 2001. For many years, France has been taking an active part in the pursuit of international actions to reinforce nuclear safety and considers the Joint Convention to be a key step in that direction. The fields covered by the Joint Convention have long been part of the French approach to nuclear safety. This report is the fourth of its kind. It is published in accordance with Article 32 of the Joint Convention and presents the measures taken by France to meet each of her obligations set out in the Convention. The facilities and radioactive materials covered by the Joint Convention are much diversified in nature and are controlled in France by different regulatory authorities (see Section E). Over and above a specific threshold of radioactive content, a facility is referred to as a 'basic nuclear facility' (installation nucleaire de base - INB) and placed under the control of the French Nuclear Safety Authority (Autorite de surete nucleaire - ASN). Below that threshold and provided that the facility involved falls under a category of the nomenclature of classified facilities for other purposes than their radioactive materials, any facility may be considered as a 'classified facility on environmental-protection grounds' (installation classee pour la protection de l'environnement - ICPE) and placed under the control of the Ministry for the

  1. Physics and Technology for the Next Generation of Radioactive Ion Beam Facilities: EURISOL

    CERN Document Server

    Kadi, Y; Catherall, R; Giles, T; Stora, T; Wenander, F K

    2012-01-01

    Since the discovery of artificial radioactivity in 1935, nuclear scientists have developed tools to study nuclei far from stability. A major breakthrough came in the eighties when the first high energy radioactive beams were produced at Berkeley, leading to the discovery of neutron halos. The field of nuclear structure received a new impetus, and the major accelerator facilities worldwide rivalled in ingenuity to produce more intense, purer and higher resolution rare isotope beams, leading to our much improved knowledge and understanding of the general evolution of nuclear properties throughout the nuclear chart. However, today, further progress is hampered by the weak beam intensities of current installations which correlate with the difficulty to reach the confines of nuclear binding where new phenomena are predicted, and where the r-process path for nuclear synthesis is expected to be located. The advancement of Radioactive Ion Beam (RIB) science calls for the development of so-called next-generation facil...

  2. Research of radioactive waste storage cask/canister materials, spent nuclear fuels and various radioactive waste forms and development of their assessment methods. Final report for Stage 3

    International Nuclear Information System (INIS)

    Dobrev, D.; Balek, V.; Červinka, R.; Večerník, P.; Člupek, M.; Kouřil, M.; Novák, P.; Stoulil, J.; Silber, R.

    2013-08-01

    The main topics treated are: Research and development of methodologies for canister/cask material degradation assessment; Laboratory research of selected materials of canister/cask with radioactive waste; and Research and assessment of canister/cask materials in natural granite rocks. Two additional documents are appended: Corrosion rate determination for samples in compacted bentonite in anaerobic conditions (methodology), and Roll test for corrosion test in an occluded solution at the interface between a radioactive waste disposal canister and the bentonite cover. (P.A.)

  3. Project No.3 - Cement solidification facility for spent ion exchange resins

    International Nuclear Information System (INIS)

    2000-01-01

    The existing storage capacity remaining for radioactive liquid wastes at the Ignalina NPP site is approximately 800 m 3 . The condition of the tanks is not fully known; however, recent engineering assessments have indicated that the tanks are unsuitable for interim storage of the liquid waste. The liquid waste currently stored in the tanks will need to be immobilised and the storage tanks emptied before they begin to deteriorate. The potential environment impact of these facilities must be reduced significantly. Project activities includes the design, construction and commissioning of the proposed facility, including all licensing documentation

  4. A positive (negative) surface ionization source concept for radioactive ion beam generation

    International Nuclear Information System (INIS)

    Alton, G.D.; Mills, G.D.

    1996-01-01

    A novel, versatile, new concept, spherical-geometry, positive (negative) surface-ionization source has been designed and fabricated which will have the capability of generating both positive- and negative-ion beams without mechanical changes to the source. The source utilizes a highly permeable, high-work-function Ir ionizer (φ ≅ 5.29 eV) for ionizing highly electropositive atoms/molecules; while for negative-surface ionization, the work function is lowered by continually feeding a highly electropositive vapor through the ionizer matrix. The use of this technique to effect low work function surfaces for negative ion beam generation has the potential of overcoming the chronic poisoning effects experienced with LaB 6 while enhancing the probability for negative ion formation of atomic and molecular species with low to intermediate electron affinities. The flexibility of operation in either mode makes it especially attractive for radioactive ion beam (RIB) applications and, therefore, the source will be used as a complementary replacement for the high-temperature electron impact ionization sources presently in the use at the Holifield radioactive ion beam facility (HRIBF). The design features and operational principles of the source are described in this report. (orig.)

  5. Safe disposal of radioactive iodide ions from solutions by Ag2O grafted sodium niobate nanofibers.

    Science.gov (United States)

    Mu, Wanjun; Li, Xingliang; Liu, Guoping; Yu, Qianhong; Xie, Xiang; Wei, Hongyuan; Jian, Yuan

    2016-01-14

    Radioactive iodine isotopes are released into the environment by the nuclear industry and medical research institutions using radioactive materials, and have negative effects on organisms living within the ecosystem. Thus, safe disposal of radioactive iodine is necessary and crucial. For this reason, the uptake of iodide ions was investigated in Ag2O nanocrystal grafted sodium niobate nanofibers, which were prepared by forming a well-matched phase coherent interface between them. The resulting composite was applied as an efficient adsorbent for I(-) anions by forming an AgI precipitate, which also remained firmly attached to the substrates. Due to their one-dimensional morphology, the new adsorbents can be easily dispersed in liquids and readily separated after purification. This significantly enhances the adsorption efficiency and reduces the separation costs. The change in structure from the pristine sodium niobate to Ag2O anchored sodium niobate and to the used adsorbent was examined by using various characterization techniques. The effects of Ag(+) concentration, pH, equilibration time, ionic strength and competing ions on the iodide ion removal ability of the composite were studied. The Ag2O nanocrystal grafted sodium niobate adsorbent showed a high adsorption capacity and excellent selectivity for I(-) anions in basic solutions. Our results are useful for the further development of improved adsorbents for removing I(-) anions from basic wastewater.

  6. Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

    Energy Technology Data Exchange (ETDEWEB)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge, CA 91330 (United States)

    2013-07-01

    Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for the selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail

  7. Biodegradation of radioactive organic liquid waste from spent fuel reprocessing; Biodegradacao de rejeitos radioativos liquidos organicos provenientes do reprocessamento do combustivel nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Rafael Vicente de Padua

    2008-07-01

    The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab-scale hot cell, known as Celeste located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90's, and the laboratory was closed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: organic compounds characterization in the radioactive liquid waste; the occurrence of bacterial consortia from Pocos de Caldas uranium mine soil and Sao Sebastiao estuary sediments that are able to degrade organic compounds; and the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming the cementation. From the characterization analysis, TBP and ethyl acetate were chosen to be degraded. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradation level were 66% for ethyl acetate and 70% for the TBP. (author)

  8. A concept for emittance reduction of DC radioactive heavy-ion beams

    International Nuclear Information System (INIS)

    Nolen, J.A.; Dooling, J.C.

    1995-01-01

    Numerical simulations indicate that it should be possible to use an electron beam to strip 1+ DC radioactive ion beams to 2+ or higher charge states with on the order of 50% efficiency. The device, which the authors call an Electron-Beam Charge-State Amplifier, is similar to an Electron Beam Ion Source, except that it is not pulsed, the beams are continuous. The 2+ beams are obtained in a single pass through a magnetic solenoid while higher charge states may be reached via multiple passes. An unexpected result of the ion optics simulations is that the normalized transverse emittance of the ion beam is reduced in proportion to the charge-state gain. Ion beams with realistic emittances and zero angular momentum relative to the optic axis before entering the solenoid will travel though the solenoid on helical orbits which intercept the axis once per cycle. With an ion beam about 2 mm in diameter and an electron beam about 0.2 mm in diameter, the ion stripping only occurs very near the optic axis, resulting in the emittance reduction

  9. Thin and thick targets for radioactive ion beam production at SPIRAL1 facility

    Science.gov (United States)

    Jardin, P.; Bajeat, O.; Delahaye, P.; Dubois, M.; Kuchi, V.; Maunoury, L.

    2018-05-01

    The upgrade of the Système de Production d'Ions Radioactifs Accélérés en Ligne (SPIRAL1) facility will deliver its new Radioactive Ion Beams (RIB) by summer 2017. The goal of the upgrade is an improvement of the performances of the installation in terms of isotopes species and ion charge states [1]. Ion beams are produced using the Isotope Separator On Line Method, consisting in an association of a primary beam of stable ions, a hot target and an ion source. The primary beam impinges on the material of the target. Radioactive isotopes are produced by nuclear reactions and propagate up to the source, where they are ionized and accelerated to create a RIB. One advantage of SPIRAL1 driver is the variety of its available primary beams, from carbon to uranium with energies up to 95 MeV/A. Within the SPIRAL1 upgrade, they will be combined with targets made of a large choice of materials, extending in this way the number of possible nuclear reactions (fusion-evaporation, transfer, fragmentation) for producing a wider range of isotopes, up to regions of the nuclide chart still scarcely explored. Depending on the reaction process, on the collision energy and on the primary beam power, thin and thick targets are used. As their functions can be different, their design must cope with specific constraints which will be described. After a presentation of the goals of present and future SPIRAL1 Target Ion Source System, the main target features, studies and designs under progress are presented.

  10. An approach to beneficiation of spent lithium-ion batteries for recovery of materials

    Science.gov (United States)

    Marinos, Danai

    Lithium ion batteries are one of the most commonly used batteries. A large amount of these have been used over the past 25 years and the use is expected to rise more due to their use in automotive batteries. Lithium ion batteries cannot be disposed into landfill due to safety reasons and cost. Thus, over the last years, there has been a lot of effort to find ways to recycle lithium ion batteries. A lot of valuable materials are present in a lithium ion battery making their recycling favorable. Many attempts, including pyrometallurgical and hydrometallurgical methods, have been researched and some of them are already used by the industry. However, further improvements are needed to the already existing processes, to win more valuable materials, use less energy and be more environmentally benign. The goal of this thesis is to find a low-temperature, low-energy method of recovering lithium from the electrolyte and to develop pathways for complete recycling of the battery. The research consists of the following parts: Pure LiPF6 powder, which is the electrolyte material, was characterized using x- ray diffraction analysis and DSC/TGA analysis. The LiPF6 powder was titrated using acid (HCl, HNO3, H2SO4), bases (NH4 OH) and distilled water. It was concluded that distilled water was the best solvent to selectively leach lithium from lithium-ion batteries. Leaching conditions were optimized including time, temperature, solid/liquid ratio and stirring velocity. All the samples were tested using ICP for chemical composition. Because leaching could be performed at room temperature, leaching was conducted in a flotation machine that was able to separate plastics by creating bubbles with no excess reagents use. The solution that contained lithium had to be concentrated more in order for lithium to be able to precipitate and it was shown that the solution could be concentrated by using the same solution over and over again. The next set of experiments was composed of battery

  11. A Transportation Risk Assessment Tool for Analyzing the Transport of Spent Nuclear Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository

    International Nuclear Information System (INIS)

    Best, Ralph; Winnard, T.; Ross, S.; Best, R.

    2001-01-01

    The Yucca Mountain Transportation Database was developed as a data management tool for assembling and integrating data from multiple sources to compile the potential transportation impacts presented in the Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DEIS). The database uses the results from existing models and codes such as RADTRAN, RISKIND, INTERLINE, and HIGHWAY to estimate transportation-related impacts of transporting spent nuclear fuel and high-level radioactive waste from commercial reactors and U. S. Department of Energy (DOE) facilities to Yucca Mountain. The source tables in the database are compendiums of information from many diverse sources including: radionuclide quantities for each waste type; route and route characteristics for rail, legal-weight truck, heavy haul. truck, and barge transport options; state-specific accident and fatality rates for routes selected for analysis; packaging and shipment data by waste type; unit risk factors; the complex behavior of the packaged waste forms in severe transport accidents; and the effects of exposure to radiation or the isotopic specific effects of radionclides should they be released in severe transportation accidents. The database works together with the codes RADTRAN (Neuhauser, et al, 1994) and RISKlND (Yuan, et al, 1995) to calculate incident-free dose and accident risk. For the incident-free transportation scenario, the database uses RADTRAN and RISKIND-generated data to calculate doses to offlink populations, onlink populations, people at stops, crews, inspectors, workers at intermodal transfer stations, guards at overnight stops, and escorts, as well as non-radioactive pollution health effects. For accident scenarios, the database uses RADTRAN-generated data to calculate dose risks based on ingestion, inhalation, resuspension, immersion (cloudshine), and groundshine as

  12. Recovery of cobalt and lithium from spent Li-ion batteries; Recuperacao de cobalto e de litio de baterias ion-litio usadas

    Energy Technology Data Exchange (ETDEWEB)

    Busnardo, Natalia Giovanini; Paulino, Jessica Frontino; Afonso, Julio Carlos [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Inst. de Quimica. Dept. de Quimica Analitica]. E-mail: julio@iq.ufrj.br

    2007-07-15

    The 'active mass' (cathode + anode + electrolyte) of spent Li-ion batteries was submitted to one of the following procedures: (a) it was calcined (500 deg C) and submitted to extraction with water to recover lithium salts. The residual solid was treated with sulfuric acid containing hydrogen peroxide. Cobalt was recovered as sulfate; (b) the 'active mass' was treated with potassium hydrogen sulfate (500 deg C) and dissolved in water. Cobalt was precipitated together with copper after addition of sodium hydroxide. Lithium was partially recovered as lithium fluoride. Co-processing of other battery components (aluminum and copper foils) affected negatively the behavior of the recovery procedures. Previous segregation of battery components is essential for an efficient and economical processing of the 'active mass'. (author)

  13. Research and development for the production of radioactive ions for SPIRAL; Recherche et developpement concernant la production d'ions radioactifs dans le cadre de SPIRAL

    Energy Technology Data Exchange (ETDEWEB)

    Eleon, C

    2007-12-15

    This thesis is related to the research and development program for the production of radioactive ion beams by the ISOL method for SPIRAL at GANIL. Two studies concerning improvements to the performance of SPIRAL target-source system have been made, using a statistical approach to the atoms-to-ions transformation. The first study concerns the transformation time between the production of the radioactive atoms of Ar{sup 35} inside a target and the extraction of the radioactive ions from the source with the TARGISOL set-up (target + ECR source). The goal was to determine the diffusion coefficients of the Ar for the carbon target. The results that are presented illustrate the difficulty of this work. The second study is the application of the statistical approach to the surface ionization source. It allowed one to define and to build a new MonoNaKe set-up for the production of 1{sup +} radioactive alkaline ions. Radioactive ions of K{sup 37,47}, Na{sup 25,26,27,28,30}, Li{sup 8,9} and Al{sup 28,29,30,31} were produced. For the production of the multicharged radioactive alkali ions, the MonoNaKe target/ion-source system was coupled to the ECR source of SPIRAL-1 without a mass separator (1{sup +}/N{sup +} direct method). A first radioactive ion beam of {sup 47}K{sup 5+} was extracted at the SIRa test bench. A surface ionization test source based on the same technical characteristics of MonoNaKe has been built. The goal of this system will be to define a prototype of source adapted to the constraints of SPIRAL-2 (ionization efficiency and lifetime). (author)

  14. Study on preparing the absorbent of potassium nickel hexacyanoferrate (II) loaded zeolite for removal of cesium from radioactive waste solutions and followed method for stable solidification of spent composites

    International Nuclear Information System (INIS)

    Pham Quynh Luong; Nguyen Hoang Lan; Nguyen Van Chinh; Nguyen Thu Trang; Vuong Huu Anh; Le Xuan Huu; Nguyen Thi Xuan; Le Van Duong

    2017-01-01

    The selective adsorption and stable immobilization of radioactive cesium, K-Ni-hexacyanoferrate (II) loaded zeolite (FC-zeolite) prepared by impregnation / precipitation method were studied. The uptake equilibrium of Cs + for composites FC-zeolite was attained within 8 h and estimated to be above 97% in Cs + 100 mg/l solution at pH 4-10. Maximum ion exchange capacity of Cs + ions (Q max ) for FC-zeoliteX was 112.5 and 69.7 mg/g in pure water and sea water, respectively. Those values for FC-zeolite A was 85.7 and 42.7 mg/g. Decontamination factor (DF) of FC-zeolite X for 134 Cs was 149.7 and 107.5 in pure water and sea water respectively. Study on synthesized zeolites (A and X) made of HUST was also conducted in similar manner. The values of Q max were 98.6 and 39.9 mg/g for zeolite A, and 69.5 and 20.8 mg/g for zeolite X in pure water and sea water, respectively. Decontamination factor (DF) of zeolite A and X for 134 Cs showed lower values. The spent CsFC-zeolite was solidificated in optimal experimental conditions: 5% Na 2 B 4 O 7 additives; calcination temperature at 900 o C for 2 h in air. Solid form was determined some of parameters: immobilization of Cs, compressive strength, volume reduction after calcination (%) and leaching rate of Cs + ions in deionization water. (author)

  15. Report on financing the disposal of commercial spent nuclear fuel and processed high-level radioactive waste

    International Nuclear Information System (INIS)

    Benny, R.I.; Sprecher, W.M.

    1983-06-01

    Projected revenues generated from the 1.0 mill per kWh fee mandated by the Act are sufficient to cover the full range of reference case program costs, assuming 3% annual inflation and nuclear installed capacity of 165 gigawatts-electric by the year 2000. Total estimated costs of the reference waste disposal program, encompassing either spent nuclear fuel disposal or reprocessing waste disposal, range between $18 to 20 billion in constant 1982 dollars. Sensitivity case analyses established upper and lower program cost bounds of $28 billion and $16 billion, respectively (in 1982 dollars). In terms of discounted levelized unit costs, the disposal of spent fuel equates to $122 to 125 per kilogram (uranium) compared with $115 to 119 per kilogram for the reprocessing waste equivalent. The levelized unit costs for reprocessing exclude the solidification of liquid wastes. Such costs are estimated to be $8 per kilogram. Discounted levelized unit costs corresponding to the upper and lower limits of the sensitivity cases equate to $176 per kilogram and $107 per kilogram. The 1.0 mill per kWh fee will be reviewed annually and adjusted, if necessary, to accommodate changes in program costs due to inflation and program shifts. When adjustments are made for applicable discount rates, inflation, repository design changes, and other factors, levelized unit costs for the reference case presented in this analysis agree closely with the results of two previous Department of Energy studies concerning charges for spent fuel storage and disposal services provided by the Federal government. The cost estimates developed for the program were based on the best available data

  16. Identification and understanding the factors affecting the public and political acceptance of long term storage of spent fuel and high-level radioactive wastes

    International Nuclear Information System (INIS)

    Gorea, Valica

    2006-01-01

    In the end of 2004, according to the information available to the IAEA, there were 440 nuclear reactors operating worldwide, with a total net capacity of 366.3 GW(e), 6 of them being connected to the grid in 2004 ( 2 in Ukraine, one each in China, Japan and the Russian Federation and a reconnection in Canada) by comparison with 2 connections and 2 re-connections in 2003. Also, in the end of 2004, 26 nuclear power plants were under construction with a total net capacity of 20.8 GW(e). The conclusion accepted by common consent is that the nuclear power is still in progress and represents one of the options for power security on long and middle term. If we refer to the nuclear fusion which will produce commercial electric power, over 30 - 40 years, in practically unlimited quantities, the above underlining becomes even more evident. Fortunately, besides the beneficent characteristics, such as: clean, stable as engendering and price, has also a negative characteristic, which generally breathes fear into the people: radioactive waste. A classification of the radioactive waste is not the target of this presentation. I just want to point that a nuclear power plant produces during the time spent fuel - long life high radioactive, generating heat. Another high radioactive waste have similar characteristics (HLW = High Level Waste) for which reason these two categories of wastes are treated together. The spent fuel and the High Level Waste are interim stored for cooling, for around 50 years, afterwards it is transferred to the final repository where it will be kept for hundreds of years, in the case of an open fuel cycle and this is also the case of Cernavoda NPP. Taking into consideration that the Cernavoda Unit 1 reaches the age of 10 years of commercial running during December 2006, it results that the issue of the final disposal is not such urgent as it looks. The objectives of long term management of radioactive waste are public health and protection of the environment

  17. Sol - gel inorganic ion exchangers for conditioning of medium level radioactive waste

    International Nuclear Information System (INIS)

    Arcangeli, G.; Traverso, D.M.; Gerontopoulos, P.; Fava, R.

    1988-01-01

    Decontamination of high-level liquid wastes and medium activity wastes streams by inorganic ion exchange combined with the conversion of the spent inorganic ion exchange material to waste ceramics presents a considerable potential for utilisation in waste conditioning. Ceramic waste forms are found superior to other candidate waste immobilisation forms but practical implementation is hampered because of the complexity of the related fabrication technology. This report shows the possibility of improving this situation by resorting to sol gel techniques earlier developed for preparation of nuclear fuel ceramics. The principal findings are: - superior quality ion exchange xerogel titanates in the form of mechanically resistant, size controlled microspheres can be prepared using a simple sol-gel technique; - the titanate particles can be also used as precursors in Evaporative Deposition on Xerogel Particles (EDXP) a new waste solidification process based on physical impregnation of the xerogel material with the waste liquid followed by evaporation; - waste loaded ion exchange microspheres can be converted to leach resistant ceramics by firing and/or cold pressing and sintering at 900 0 -1100 0 C; - sol-gel inorganic ion exchange and EDXP may find useful application in conditioning MAW streams. 44 figs., 43 refs

  18. Development of a low-energy radioactive ion beam facility for the MARA separator

    Energy Technology Data Exchange (ETDEWEB)

    Papadakis, Philippos, E-mail: philippos.papadakis@jyu.fi; Moore, Iain; Pohjalainen, Ilkka; Sarén, Jan; Uusitalo, Juha [University of Jyväskylä, Department of Physics (Finland)

    2016-12-15

    A low-energy radioactive ion beam facility for the production and study of nuclei produced close to the proton drip line is under development at the Accelerator Laboratory of the University of Jyväskylä, Finland. The facility will take advantage of the mass selectivity of the recently commissioned MARA vacuum-mode mass separator. The ions selected by MARA will be stopped and thermalised in a small-volume gas cell prior to extraction and further mass separation. The gas cell design allows for resonance laser ionisation/spectroscopy both in-gas-cell and in-gas-jet. The facility will include experimental setups allowing ion counting, mass measurement and decay spectroscopy.

  19. Between atomic and nuclear physics: radioactive decays of highly-charged ions

    International Nuclear Information System (INIS)

    Atanasov, Dinko; Bosch, Fritz; Brandau, Carsten; Chen, Xiangcheng; Dillmann, Iris; Gao, Bingshui; Geissel, Hans; Hagmann, Siegbert; Hillenbrand, Pierre-Michel; Kozhuharov, Christophor; Litvinov, Sergey A; Litvinov, Yuri A; Münzenberg, Gottfried; Blaum, Klaus; Bühler, Paul; Faestermann, Thomas; Gernhäuser, Roman; Izumikawa, Takuji; Kurcewicz, Jan; Ma, Xinwen

    2015-01-01

    Highly charged radioactive ions can be stored for extended periods of time in storage rings which allows for precision measurements of their decay modes. The straightforward motivation for performing such studies is that fully ionised nuclei or few-electron ions can be viewed as clean quantum-mechanical systems, in which the interactions of the many electrons can be either excluded or treated precisely. Thus, the influence of the electron shell on the decay probability can be investigated. Another important motivation is stellar nucleosynthesis, which proceeds at high temperatures and the involved atoms are therefore highly ionised. Presented here is a compact review of the relevant experiments conducted at heavy-ion storage rings. Furthermore, we outline the perspectives for future experiments at new-generation storage-ring facilities. (paper)

  20. Is Yucca Mountain a long-term solution for disposing of US spent nuclear fuel and high-level radioactive waste?

    Science.gov (United States)

    Thorne, M C

    2012-06-01

    On 26 January 2012, the Blue Ribbon Commission on America's Nuclear Future released a report addressing, amongst other matters, options for the managing and disposal of high-level waste and spent fuel. The Blue Ribbon Commission was not chartered as a siting commission. Accordingly, it did not evaluate Yucca Mountain or any other location as a potential site for the storage or disposal of spent nuclear fuel and high-level waste. Nevertheless, if the Commission's recommendations are followed, it is clear that any future proposals to develop a repository at Yucca Mountain would require an extended period of consultation with local communities, tribes and the State of Nevada. Furthermore, there would be a need to develop generally applicable regulations for disposal of spent fuel and high-level radioactive waste, so that the Yucca Mountain site could be properly compared with alternative sites that would be expected to be identified in the initial phase of the site-selection process. Based on what is now known of the conditions existing at Yucca Mountain and the large number of safety, environmental and legal issues that have been raised in relation to the DOE Licence Application, it is suggested that it would be imprudent to include Yucca Mountain in a list of candidate sites for future evaluation in a consent-based process for site selection. Even if there were a desire at the local, tribal and state levels to act as hosts for such a repository, there would be enormous difficulties in attempting to develop an adequate post-closure safety case for such a facility, and in showing why this unsaturated environment should be preferred over other geological contexts that exist in the USA and that are more akin to those being studied and developed in other countries.

  1. Comparison of the intermediate storage periods and areas required for final storage of high-level radioactive waste and spent fuel in various types of host rock

    International Nuclear Information System (INIS)

    Mueller-Hoeppe, N.; Lerch, C.; Jobmann, M.; Filbert, W.

    2005-01-01

    The present new version of the German concept for radioactive waste and spent fuel management is based on the assumption that a repository for high-level waste and spent fuel will not be required until 2030. One reason frequently given for this date is the intermediate storage period of at least forty years to allow the very high initial heat generation to decay. However, calculations performed by the authors have shown that the minimum intermediate storage period for a repository in rock salt is only between four and nineteen years, depending on the final storage concept and the load of the waste package. In clay as a host rock, the minimum intermediate storage times were calculated to be between 31 and 142 years; the same time spans are expected to apply to final storage in magmatic rock, such as granite. The maximum permissible loads of a container holding spent fuel in salt are many times those in clay and granite, respectively. It was also seen that the area requirement for final storage of the same waste structures is roughly a factor of ten higher in clay than in salt. The differences between granite and salt are similar. The reasons for these grave differences, on the one hand, are the better thermal conductivity of salt and, on the other hand, the better heat tolerance of the crushed salt used as backfill material compared to that of bentonite used in the clay and granite concepts. While salt will allow temperatures of up to 200 C, the maximum temperature in bentonite is limited to 100 C. (orig.)

  2. Organic reductants based leaching: A sustainable process for the recovery of valuable metals from spent lithium ion batteries.

    Science.gov (United States)

    Chen, Xiangping; Guo, Chunxiu; Ma, Hongrui; Li, Jiazhu; Zhou, Tao; Cao, Ling; Kang, Duozhi

    2018-05-01

    It is significant to recover metal values from spent lithium ion batteries (LIBs) for the alleviation or prevention of potential risks towards environmental pollution and public health, as well as for the conservation of valuable metals. Herein a hydrometallurgical process was proposed to explore the possibility for the leaching of different metals from waste cathodic materials (LiCoO 2 ) of spent LIBs using organics as reductant in sulfuric acid medium. According to the leaching results, about 98% Co and 96% Li can be leached under the optimal experimental conditions of reaction temperature - 95 °C, reaction time - 120 min, reductive agent dosage - 0.4 g/g, slurry density - 25 g/L, concentration of sulfuric acid-3 mol/L in H 2 SO 4  + glucose leaching system. Similar results (96% Co and 100% Li) can be obtained in H 2 SO 4  + sucrose leaching system under optimized leaching conditions. Despite a complete leaching of Li (∼100%), only 54% Co can be dissolved in the H 2 SO 4  + cellulose leaching system under optimized leaching conditions. Finally, different characterization methods, including UV-Vis, FT-IR, SEM and XRD, were employed for the tentative exploration of reductive leaching reactions using organic as reductant in sulfuric acid medium. All the leaching and characterization results confirm that both glucose and sucrose are effective reductants during leaching, while cellulose should be further degraded to organics with low molecular weights to achieve a satisfactory leaching performance. Copyright © 2018 Elsevier Ltd. All rights reserved.

  3. Development and properties of crystalline silicotitanate (CST) ion exchangers for radioactive waste applications

    International Nuclear Information System (INIS)

    Miller, J.E.; Brown, N.E.

    1997-04-01

    Crystalline silicotitanates (CSTs) are a new class of ion exchangers that were jointly invented by researchers at Sandia National Laboratories and Texas A ampersand M University. One particular CST, known as TAM-5, is remarkable for its ability to separate parts-per-million concentrations of cesium from highly alkaline solutions (pH> 14) containing high sodium concentrations (>5M). It is also highly effective for removing cesium from neutral and acidic solutions, and for removing strontium from basic and neutral solutions. Cesium isotopes are fission products that account for a large portion of the radioactivity in waste streams generated during weapons material production. Tests performed at numerous locations with early lab-scale TAM-5 samples established the material as a leading candidate for treating radioactive waste volumes such as those found at the Hanford site in Washington. Thus Sandia developed a Cooperative Research and Development Agreement (CRADA) partnership with UOP, a world leader in developing, commercializing, and supplying adsorbents and associated process technology to commercialize and further develop the material. CSTs are now commercially available from UOP in a powder (UOP IONSIV reg-sign IE-910 ion exchanger) and granular form suitable for column ion exchange operations (UOP IONSIV reg-sign IE-911 ion exchanger). These materials exhibit a high capacity for cesium in a wide variety of solutions of interest to the Department of Energy, and they are chemically, thermally, and radiation stable. They have performed well in tests at numerous sites with actual radioactive waste solutions, and are being demonstrated in the 100,000 liter Cesium Removal Demonstration taking place at Oak Ridge National Laboratory with Melton Valley Storage Tank waste. It has been estimated that applying CSTs to the Hanford cleanup alone will result in a savings of more than $300 million over baseline technologies

  4. Development and properties of crystalline silicotitanate (CST) ion exchangers for radioactive waste applications

    Energy Technology Data Exchange (ETDEWEB)

    Miller, J.E.; Brown, N.E.

    1997-04-01

    Crystalline silicotitanates (CSTs) are a new class of ion exchangers that were jointly invented by researchers at Sandia National Laboratories and Texas A&M University. One particular CST, known as TAM-5, is remarkable for its ability to separate parts-per-million concentrations of cesium from highly alkaline solutions (pH> 14) containing high sodium concentrations (>5M). It is also highly effective for removing cesium from neutral and acidic solutions, and for removing strontium from basic and neutral solutions. Cesium isotopes are fission products that account for a large portion of the radioactivity in waste streams generated during weapons material production. Tests performed at numerous locations with early lab-scale TAM-5 samples established the material as a leading candidate for treating radioactive waste volumes such as those found at the Hanford site in Washington. Thus Sandia developed a Cooperative Research and Development Agreement (CRADA) partnership with UOP, a world leader in developing, commercializing, and supplying adsorbents and associated process technology to commercialize and further develop the material. CSTs are now commercially available from UOP in a powder (UOP IONSIV{reg_sign} IE-910 ion exchanger) and granular form suitable for column ion exchange operations (UOP IONSIV{reg_sign} IE-911 ion exchanger). These materials exhibit a high capacity for cesium in a wide variety of solutions of interest to the Department of Energy, and they are chemically, thermally, and radiation stable. They have performed well in tests at numerous sites with actual radioactive waste solutions, and are being demonstrated in the 100,000 liter Cesium Removal Demonstration taking place at Oak Ridge National Laboratory with Melton Valley Storage Tank waste. It has been estimated that applying CSTs to the Hanford cleanup alone will result in a savings of more than $300 million over baseline technologies.

  5. Study of radio-active ions in the atmosphere; Etude des ions radioactifs de l'atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Renoux, A [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1965-01-01

    A comparative study is made of active, deposits of radon and thoron in suspension in the atmosphere by means of {alpha} radiation counting, using ZELENY tubes, scattering equipment, filter papers or membranes. It has been possible to show the existence of small and large ions which are negative and positive, as well as of neutral radio-active nuclei; their properties are studied. A theoretical interpretation of the results is presented. The average content of radon (using the Ra A concentration) and of Th B in the air has been determined. The radioactive equilibrium between radon and its daughter products in atmospheric air are examined. The techniques developed for active radon and thoron deposits are applied to the study of artificial radio-activity, the analyses being carried out by means of {gamma} spectrometry. (author) [French] On effectue une etude comparative entre les depots actifs du radon et du thoron en suspension dont l'atmosphere a l'aide de comptages de rayonnement {alpha}, en utilisant des tubes de ZELENY, une batterie de diffusion, des papiers filtres ou des membranes. On met ainsi en evidence la presence de petits et gros ions negatifs et positifs, ainsi que celle de noyaux neutres radioactifs, et on etudie leurs proprietes. Une Interpretation theorique des resultats obtenus est developpee. On determine la teneur moyenne de l'air en radon (a partir de la concentration en Ra A) et en Th B. L'equilibre radioactif entre le radon et ses descendants, dans l'air atmospherique, est examine. Les Techniques mises au point pour les depots actifs du radon et du thoron sont appliquees a l'etude de la radioactivite artificielle, les depouillements s'effectuant par spectrometrie {gamma}. (auteur)

  6. Recoil separators for radiative capture using radioactive ion beams. Recent advances and detection techniques

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz, Chris [TRIUMF, Vancouver, BC (Canada); Greife, Uwe; Hager, Ulrike [Colorado School of Mines, Golden, CO (United States)

    2014-06-15

    Radiative capture reactions involving the fusion of hydrogen or helium are ubiquitous in the stellar history of the universe, and are some of the most important reactions in the processes that govern nucleosynthesis and energy generation in both static and explosive scenarios. However, radiative capture reactions pose some of the most difficult experimental challenges due to extremely small cross sections. With the advent of recoil separators and techniques in inverse kinematics, it is now possible to measure radiative capture reactions on very short-lived radioactive nuclei, and in the presence of high experimental backgrounds. In this paper we review the experimental needs for making measurements of astrophysical importance on radiative capture reactions. We also review some of the important historical advances in the field of recoil separators as well as describe current techniques and performance milestones, including descriptions of some of the separators most recently working at radioactive ion beam facilities, such as DRAGON at TRIUMF and the DRS at the Holifield Radioactive Ion Beam Facility. We will also summarize some of the scientific highlight measurements at the RIB facilities. (orig.)

  7. Environmental standards for management and disposal of spent nuclear fuel, high-level and transuranic radioactive wastes, 40 CFR part 191: draft environmental impact statement

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The establishment of environmental standards for management and disposal of spent nuclear reactor fuel and high-level and transuranic radioactive wastes is proposed. The standards would require that maximum individual doses from all normal operations be limited to 25 millirem to the whole body, 75 millirem to the thyroid, and 25 millirem to any other organ. Regarding disposal of subject materials in geologic sites, the standards would include numerical containment requirements for the first 10,000 years following disposal, assurance requirements, and procedural requirements. The assurance requirements would provide seven principles necessary for developing confidence that long-term containment requirements would be upheld. The principles would call for well-designed, multiple-barrier disposal systems that would not rely on future generations for maintenance and would not be located near potential valuable resources. The principles would also require that future generations be provided with information about the location and dangers of the wastes and an option to recover the wastes if necessary. Procedural requirements would be developed to assure that the containment requirements were upheld. The implementation of the standards would protect public health and the environment against emissions of radioactivity. The maximum impact expected from a disposal system complying with the proposed standards would be less than 1000 premature cancer deaths over the first 10,000 years for disposal of high-level wastes produced by all currently operating reactors over their lifetime

  8. Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P.

    2014-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2008 of the performance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment-specific laboratory experiments, in-situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site-specific characterization. Because the relationship is important to understanding the evolution of the Yucca Mountain Project, the tabulation also shows the interaction between four broad categories of political bodies and government agencies/institutions: (a) technical milestones of the implementing institutions, (b) development of the regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives and decisions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste.

  9. Mining and engineering aspects and variants for the underground construction of a deep geological repository for radioactive waste and spent nuclear fuel

    International Nuclear Information System (INIS)

    Milchev, M.; Michailov, B.; Nanovska, E.; Harizanov, A.

    2003-01-01

    The aim of the present report is to investigate and to describe systematically the foreign experience, scientific and technical achievements and stages of development concerning the mining and engineering aspects and variants for underground construction of a deep geological repository for radioactive waste (RAW) and spent nuclear fuel (SNF). The ideal solution in managing the problems with harmful wastes seems to be either to remove them permanently from Earth (which is related with high risks and high costs) or to transform long-lived radionuclides to short-lived radionuclides using nuclear transmutation processes in a reactor or a particle accelerator. The latter is also a complex and immensely costly process and it can only reduce the quantities of some long-lived radionuclides, which can be then disposed in a geological repository. At present, the deep geological disposal remains the only solution for solving the problem with the hazard of storing radioactive wastes. The report submits a brief description and systematization of the performed investigations, accompanied by analysis of the scientific and technical level on world scale. The analysis is related with the particular geological conditions and the existing scientific studies available so far in Bulgaria. The main conclusions are that the complex scientific-technical and engineering problems related with the construction of a deep geological repository for RAW and SNF require long-term scientific investigations and preliminary complex works and it is high time to launch them in Bulgaria. (authors)

  10. Radioactivity

    International Nuclear Information System (INIS)

    Chelet, Y.

    2006-01-01

    The beginning of this book explains the why and how of the radioactivity, with a presentation of the different modes of disintegration. Are tackled the reports between radioactivity and time before explaining how the mass-energy equivalence appears during disintegrations. Two chapters treat natural radioisotopes and artificial ones. This book makes an important part to the use of radioisotopes in medicine (scintigraphy, radiotherapy), in archaeology and earth sciences (dating) before giving an inventory of radioactive products that form in the nuclear power plants. (N.C.)

  11. Radioactivity

    International Nuclear Information System (INIS)

    2002-01-01

    This pedagogical document presents the origin, effects and uses of radioactivity: where does radioactivity comes from, effects on the body, measurement, protection against radiations, uses in the medical field, in the electric power industry, in the food (ionization, radio-mutagenesis, irradiations) and other industries (radiography, gauges, detectors, irradiations, tracers), and in research activities (dating, preservation of cultural objects). The document ends with some examples of irradiation levels (examples of natural radioactivity, distribution of the various sources of exposure in France). (J.S.)

  12. On-line detection of small radioactive ions by capillary electrophoresis

    International Nuclear Information System (INIS)

    Klunder, G.L.; Andrews, J.E. Jr.; Russo, R.E.

    1994-01-01

    Worldwide environmental interests have placed a great demand on developing techniques for rapid characterization of contaminated soil and groundwater. Detection of radioactive contaminants is necessary for monitoring effluents from nuclear processes or to assure proper long term storage of radioactive waste. The authors have been investigating the chemistry required to separate representative radioactive small cations and anions by capillary electrophoresis. In order to evaluate the separation chemistry, detection of stable isotopes of the representative ions was achieved by indirect absorption for cations and direct absorption for anions. Several buffer systems which have been considered in the optimization of the separations will be discussed. The authors have designed and tested two on-line radioactivity detectors for capillary electrophoresis. An on-line solid state CdTe detector was constructed for this study and a scintillation detector has been designed using a high gain photodiode light sensor. Different scintillation materials have been tested. Comparison of the detectors, design considerations, efficiency and limits of detection will be presented

  13. Norwegian national report. Joint convention on the safety of spent fuel management and on the safety of radioactive waste management. [National report from Norway, fourth review meeting, 14-23 May 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    This report contains the national report from Norway to the fourth review meeting of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management to be held 14 to 23 May 2012. (Author)

  14. Answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. April 2009

    International Nuclear Information System (INIS)

    2009-01-01

    In this publication detail answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management of April 2009 are published

  15. The announcement of the Ministry of Foreign Affairs of the Slovak Republic about Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management

    International Nuclear Information System (INIS)

    2002-01-01

    The Ministry of Foreign Affairs of the Slovak Republic amends that in 30 September 1997 in Vienna was undersigned the Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management. This Joint Convention came into force on June 18, 2001

  16. Answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. April 2014

    International Nuclear Information System (INIS)

    2014-04-01

    In this publication detail answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management of April 2014 are published.

  17. Answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. April 2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-04-15

    In this publication detail answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management of April 2011 are published.

  18. Answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. April 2011

    International Nuclear Information System (INIS)

    2011-04-01

    In this publication detail answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management of April 2011 are published.

  19. Answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. April 2012

    International Nuclear Information System (INIS)

    2012-04-01

    In this publication detail answers to questions on National report of the Slovak Republic compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management of April 2011 are published.

  20. A linear radiofrequency quadrupole ion trap for the cooling and bunching of radioactive ion beams

    CERN Document Server

    Kellerbauer, A G; Dilling, J; Henry, S; Herfurth, F; Kluge, H J; Lamour, E; Moore, R B; Scheidenberger, C; Schwarz, S; Sikler, G; Szerypo, J

    2002-01-01

    A linear radiofrequency quadrupole ion guide and beam buncher has been installed at the ISOLTRAP mass spectrometry experiment at the ISOLDE facility at CERN. The apparatus is being used as a beam cooling, accumulation, and bunching system. It operates with a buffer gas that cools the injected ions and converts the quasicontinuous 60- keV beam from the ISOLDE facility to 2.5-keV beam pulses with improved normalized transverse emittance. Recent measurements suggest a capture efficiency of the ion guide of up to 40% and a cooling and bunching efficiency of at least 12% which is expected to still be increased. The improved ISOLTRAP setup has so far been used very successfully in three on-line experiments. (12 refs).

  1. First radioactive ions charge bred in REXEBIS at the REX-ISOLDE accelerator

    CERN Document Server

    Wolf, B H; Fostner, O; Wenander, F; Ames, F; Reisinger, K; Liljeby, L; Skeppstedt, Ö; Jonson, B; Nyman, G H

    2003-01-01

    REXEBIS is the charge breeder of the REX-ISOLDE post accelerator. The radioactive 1$^{+}$ ions produced at ISOLDE are accumulated, phase-space cooled and bunched in the REXTRAP, and thereafter injected into the EBIS with an energy up to 60 keV. The REXEBIS produced the first charge bred ions in August 2001 and has been running nearly non-stop during September to December 2001. It has delivered stable $^{39}$K$^{10+}$ and $^{23}$Na$^{6+}$ beams generated in the ion source in front of REXTRAP with a Na$^{7+}$ current exceeding 70 pA (6x10$^{7}$ p/s). Stable $^{27}$Al$^{7+}$ and $^{23}$Na$^{6+}$ from ISOLDE and also the first radioactive $^{26}$Na$^{7+}$ and $^{24}$Na$^{7+}$ beams (just 5x10$^{5}$ p/s) have been charge bred and accelerated for tests of the experimental setup. Despite some problems with the electron gun, which had one breakdown after about 1500 hours of operation and displays slow changes of the emission conditions, the EBIS is working remarkably stable (24 hours / 7 days a week). We will report ...

  2. A high-efficiency positive (negative) surface ionization source for radioactive ion beam (abstract)a

    International Nuclear Information System (INIS)

    Alton, G.D.; Mills, G.D.

    1996-01-01

    A versatile, new concept, spherical-geometry, positive (negative) surface-ionization source has been designed and fabricated which will have the capability of generating both positive- and negative-ion beams without mechanical changes to the source. The source utilizes a highly permeable, high-work-function Ir ionizer (φ≡5.29 eV) for ionizing highly electropositive atoms/molecules; while for negative-surface ionization, the work function is lowered to φ≡1.43 eV by continually feeding cesium vapor through the ionizer matrix. The use of this technique for negative ion beam generation has the potential of overcoming the chronic poisoning effects experienced with LaB 6 while enhancing considerably the efficiency for negative surface ionization of atoms and molecules with intermediate electron affinities. The flexibility of operation in either mode makes it especially attractive for radioactive ion beam applications and, therefore, the source will be used as a complementary replacement for the high-temperature electron impact ionization sources presently in use at the Holifield radioactive beam facility. The design features and operational principles of the source will be described in this report. copyright 1996 American Institute of Physics

  3. Upgrade of the facility EXOTIC for the in-flight production of light Radioactive Ion Beams

    Energy Technology Data Exchange (ETDEWEB)

    Mazzocco, M., E-mail: marco.mazzocco@pd.infn.it [Dipartimento di Fisica e Astronomia, Universitá di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); INFN-Sezione di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); Torresi, D.; Strano, E. [Dipartimento di Fisica e Astronomia, Universitá di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); INFN-Sezione di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); Boiano, A. [INFN-Sezione di Napoli, Via Cinthia, I-80126 Napoli (Italy); Boiano, C. [INFN-Sezione di Milano, Via Celoria 16, I-20133 Milano (Italy); Costa, L. [INFN-LNL, Viale dell’Università 2, I-35020 Legnaro, PD (Italy); Glodariu, T. [NIPNE, 407 Atomistilor Street, 077125 Magurele (Romania); Guglielmetti, A. [INFN-Sezione di Milano, Via Celoria 16, I-20133 Milano (Italy); Dipartimento di Fisica,Università di Milano, Via Celoria 16, I-20133 Milano (Italy); La Commara, M. [INFN-Sezione di Napoli, Via Cinthia, I-80126 Napoli (Italy); Dipartimento di Scienze Fisiche, Università di Napoli, Via Cinthia, I-80126 Napoli (Italy); Parascandolo, C. [Dipartimento di Fisica e Astronomia, Universitá di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); INFN-Sezione di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); Pierroutsakou, D. [INFN-Sezione di Napoli, Via Cinthia, I-80126 Napoli (Italy); Signorini, C.; Soramel, F. [Dipartimento di Fisica e Astronomia, Universitá di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); INFN-Sezione di Padova, Via F. Marzolo 8, I-35131 Padova (Italy); Stroe, L. [NIPNE, 407 Atomistilor Street, 077125 Magurele (Romania)

    2013-12-15

    Highlights: • Production of in-flight Radioactive Ion Beams via two-body reactions. • Development of a cryogenic gas target. • Event-by-event tracking via Parallel Plate Avalanche Counters (PPACs). -- Abstract: The facility EXOTIC for the in-flight production of light weakly-bound Radioactive Ion Beams (RIBs) has been operating at INFN-LNL since 2004. RIBs are produced via two-body reactions induced by high intensity heavy-ion beams impinging on light gas targets and selected by means of a 30°-dipole bending magnet and a 1-m long Wien filter. The facility has been recently upgraded (i) by developing a cryogenic gas target, (ii) by replacing the power supplies of the middle lenses of the two quadrupole triplets, (iii) by installing two y-steerers and (iv) by placing two Parallel Plate Avalanche Counters upstream the secondary target to provide an event-by-event reconstruction of the position hit on the target. So far, RIBs of {sup 7}Be, {sup 8}B and {sup 17}F in the energy range 3–5 MeV/u have been produced with intensities about 3 × 10{sup 5}, 1.6 × 10{sup 3} and 10{sup 5} pps, respectively. Possible light RIBs (up to Z = 10) deliverable by the facility EXOTIC are also reviewed.

  4. Multistage leaching of metals from spent lithium ion battery waste using electrochemically generated acidic lixiviant.

    Science.gov (United States)

    Boxall, N J; Adamek, N; Cheng, K Y; Haque, N; Bruckard, W; Kaksonen, A H

    2018-04-01

    Lithium ion battery (LIB) waste contains significant valuable resources that could be recovered and reused to manufacture new products. This study aimed to develop an alternative process for extracting metals from LIB waste using acidic solutions generated by electrolysis for leaching. Results showed that solutions generated by electrolysis of 0.5 M NaCl at 8 V with graphite or mixed metal oxide (MMO) electrodes were weakly acidic and leach yields obtained under single stage (batch) leaching were poor (leaching with the graphite electrolyte solution improved leach yields overall, but the electrodes corroded over time. Though yields obtained with both electrolyte leach solutions were low when compared to the 4 M HCl control, there still remains potential to optimise the conditions for the generation of the acidic anolyte solution and the solubilisation of valuable metals from the LIB waste. A preliminary value proposition indicated that the process has the potential to be economically feasible if leach yields can be improved, especially based on the value of recoverable cobalt and lithium. Copyright © 2018 Elsevier Ltd. All rights reserved.

  5. Recovery of cobalt from spent lithium-ion batteries using supercritical carbon dioxide extraction.

    Science.gov (United States)

    Bertuol, Daniel A; Machado, Caroline M; Silva, Mariana L; Calgaro, Camila O; Dotto, Guilherme L; Tanabe, Eduardo H

    2016-05-01

    Continuing technological development decreases the useful lifetime of electronic equipment, resulting in the generation of waste and the need for new and more efficient recycling processes. The objective of this work is to study the effectiveness of supercritical fluids for the leaching of cobalt contained in lithium-ion batteries (LIBs). For comparative purposes, leaching tests are performed with supercritical CO2 and co-solvents, as well as under conventional conditions. In both cases, sulfuric acid and H2O2 are used as reagents. The solution obtained from the supercritical leaching is processed using electrowinning in order to recover the cobalt. The results show that at atmospheric pressure, cobalt leaching is favored by increasing the amount of H2O2 (from 0 to 8% v/v). The use of supercritical conditions enable extraction of more than 95wt% of the cobalt, with reduction of the reaction time from 60min (the time employed in leaching at atmospheric pressure) to 5min, and a reduction in the concentration of H2O2 required from 8 to 4% (v/v). Electrowinning using a leach solution achieve a current efficiency of 96% and a deposit with cobalt concentration of 99.5wt%. Copyright © 2016 Elsevier Ltd. All rights reserved.

  6. A Novel Data-Driven Fast Capacity Estimation of Spent Electric Vehicle Lithium-ion Batteries

    Directory of Open Access Journals (Sweden)

    Caiping Zhang

    2014-12-01

    Full Text Available Fast capacity estimation is a key enabling technique for second-life of lithium-ion batteries due to the hard work involved in determining the capacity of a large number of used electric vehicle (EV batteries. This paper tries to make three contributions to the existing literature through a robust and advanced algorithm: (1 a three layer back propagation artificial neural network (BP ANN model is developed to estimate the battery capacity. The model employs internal resistance expressing the battery’s kinetics as the model input, which can realize fast capacity estimation; (2 an estimation error model is established to investigate the relationship between the robustness coefficient and regression coefficient. It is revealed that commonly used ANN capacity estimation algorithm is flawed in providing robustness of parameter measurement uncertainties; (3 the law of large numbers is used as the basis for a proposed robust estimation approach, which optimally balances the relationship between estimation accuracy and disturbance rejection. An optimal range of the threshold for robustness coefficient is also discussed and proposed. Experimental results demonstrate the efficacy and the robustness of the BP ANN model together with the proposed identification approach, which can provide an important basis for large scale applications of second-life of batteries.

  7. Capture of toxic radioactive and heavy metal ions from water by using titanate nanofibers

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Jiasheng, E-mail: jiashengxu@bhu.edu.cn [Liaoning Province Key Laboratory for Synthesis and Application of Functional Compounds, College of Chemistry, Chemical Engineering and Food Safety, Center of Science and Technology Experiment, Bohai University, 19 Sci-tech Road, Jinzhou 121013 (China); Zhang, He; Zhang, Jie [Liaoning Province Key Laboratory for Synthesis and Application of Functional Compounds, College of Chemistry, Chemical Engineering and Food Safety, Center of Science and Technology Experiment, Bohai University, 19 Sci-tech Road, Jinzhou 121013 (China); Kim, Eui Jung [School of Chemical Engineering and Bioengineering, University of Ulsan, Ulsan 680-749 (Korea, Republic of)

    2014-11-25

    Highlights: • Three types of titanate nanofibers were prepared via a hydrothermal porcess. • These nanofibers show availability for removal of the toxic ions from water. • The equilibrium data were fitted well with the Langmuir model. - Abstract: Three types of titanate nanofibers (sodium titanate nanofibers (TNF-A), potassium/sodium titanate nanofibers (TNF-B), potassium titanate nanofibers (TNF-C)) were prepared via a hydrothermal treatment of anatase powders in different alkali solutions at 170 °C for 96 h, respectively. The as-prepared nanofibers have large specific surface area and show availability for the removal of radioactive and heavy metal ions from water system, such as Ba{sup 2+} (as substitute of {sup 226}Ra{sup 2+}) and Pb{sup 2+} ions. The TNF-A shows a better capacity in the removal of Ba{sup 2+} and Pb{sup 2+} than TNF-B and TNF-C. Structural characterization of the materials was performed with powder X-ray diffraction (XRD), scanning electron microscopy (SEM) equipped with energy dispersive spectrometer (EDS) and with inductively coupled plasma optical emission spectrometry (ICP-OES). It is found that the equilibrium data fit well with the Langmuir model. This study highlights that nanoparticles of inorganic ion exchangers with layered structure are potential materials for efficient removal of the toxic ions from contaminated water.

  8. Solidification of saturated radioactive organic ion exchangers and of ash from incineration plant

    International Nuclear Information System (INIS)

    Timulak, J.; Krejci, F.; Pekar, A.; Gulis, G.; Breza, M.

    1985-01-01

    The study of bituminization of saturated radioactive organic ion exchangers was centred on finding the effect of the water content of ion exchangers on the process of solidification and on the water content of bituminization products, the optimization of temperature conditions in the process of bituminization, on seeking a suitable bitumen, on testing the radiation and thermal stability of the bituminization product, on finding its properties as well as the effects of nuclear radiation on these properties. Ion exchangers of Czechoslovak and Soviet make were used in all experiments. It was found that solidified ion exchangers must have a maximum moisture of 10%, and the temperature during solidification must not exceed 130 degC. The negative effect of boric acid on cement solidification may be removed by neutralization of esterification of this acid following its release from the ion exchangers by hydrochloric acid. Some other results of the experiments are tabulated. The obtained results describe the behaviour of the product only during a brief period of time as compared with the long time of long-term disposal. It will therefore be necessary to devote attention to finding the characteristics of long-term behaviour of products during disposal. (Z.M.)

  9. Studies on the incorporation of spent ion exchange resins from nuclear power plants into bitumen and cement

    International Nuclear Information System (INIS)

    Bonnevie-Svendsen, M.; Tallberg, K.; Aittola, P.; Tollbaeck, H.

    1976-01-01

    The joint Nordic incorporation experiments should provide technical data needed for the assessment of solidification techniques for wastes from nuclear reactors in the Nordic countries. Spent ion exchange resins are a main fraction of such wastes, and more knowledge about their incorporation is wanted. The effects of simulated and real ion exchange wastes on the quality of bitumen and cement incorporation products were studied. Blown and distilled bitumen and three Portland cement qualities were used. Product characterizations were based on properties relevant for safe waste management, storage, transport and disposal. The applicability and relevance of established and suggested tests is discussed. Up to 40-60% dry resin could be incorporated into bitumen without impairing product qualities. Products with higher resin contents were found to swell in contact with water. The products had a high leach resistance. Their form stability was improved by incorporated resins. Product qualities appeared to be less affected by physico-chemical variables than by mechanical process parameters. Pure resin-cement products tend to decompose in water. Product qualities were strongly affected by a variety of physico-chemical process parameters, and integer products were only obtained within narrow tolerance limits. Caesium was rapidly leached out. To attain integer products and improved leach resistance within technically acceptable tolerance limits it was necessary to utilize stabilizing and caesium-retaining additives such as Silix and vermiculite. Under the present conditions the water content of the resins limited the amounts that could be incorporated in 40-50wt% or about 70vol.% water-saturated (containing 20-40% dry) resin. (author)

  10. Conception of spent fuel and radioactive wastes management in Poland based on the results of the previous work performed in the frames of Governmental Strategic Programme realised under patronate of National Atomic Energy Agency

    International Nuclear Information System (INIS)

    Wlodarski, J.; Chwaszczewski, S.; Slizowski, K.; Frankowski, Z.

    1999-01-01

    About 300 cubic meters of solid and solidified radioactive wastes of low and medium activity are produced each year in Poland. Such materials, after processing, are stored in the Institute of Atomic Energy at Swierk or in the National Repository for Radioactive Wastes in Rozan. About 6000 spent fuel elements are temporarily stored in water pools at Swierk. Assumptions and strategy of future spent fuel and radioactive wastes management are presented taking into account operation of the first nuclear power plants (after 2010). Then Governmental Strategic Programme (GSP) for the year 1997-1999 concerning such topic is described and some results of the work performed in the frames of the GSP are given

  11. Lithium Carbonate Recovery from Cathode Scrap of Spent Lithium-Ion Battery: A Closed-Loop Process.

    Science.gov (United States)

    Gao, Wenfang; Zhang, Xihua; Zheng, Xiaohong; Lin, Xiao; Cao, Hongbin; Zhang, Yi; Sun, Zhi

    2017-02-07

    A closed-loop process to recover lithium carbonate from cathode scrap of lithium-ion battery (LIB) is developed. Lithium could be selectively leached into solution using formic acid while aluminum remained as the metallic form, and most of the other metals from the cathode scrap could be precipitated out. This phenomenon clearly demonstrates that formic acid can be used for lithium recovery from cathode scrap, as both leaching and separation reagent. By investigating the effects of different parameters including temperature, formic acid concentration, H 2 O 2 amount, and solid to liquid ratio, the leaching rate of Li can reach 99.93% with minor Al loss into the solution. Subsequently, the leaching kinetics was evaluated and the controlling step as well as the apparent activation energy could be determined. After further separation of the remaining Ni, Co, and Mn from the leachate, Li 2 CO 3 with the purity of 99.90% could be obtained. The final solution after lithium carbonate extraction can be further processed for sodium formate preparation, and Ni, Co, and Mn precipitates are ready for precursor preparation for cathode materials. As a result, the global recovery rates of Al, Li, Ni, Co, and Mn in this process were found to be 95.46%, 98.22%, 99.96%, 99.96%, and 99.95% respectively, achieving effective resources recycling from cathode scrap of spent LIB.

  12. Hydrometallurgical process for the recovery of metal values from spent lithium-ion batteries in citric acid media.

    Science.gov (United States)

    Chen, Xiangping; Zhou, Tao

    2014-11-01

    In this paper, a hydrometallurgical process has been proposed to recover valuable metals from spent lithium-ion batteries in citric acid media. Leaching efficiencies as high as 97%, 95%, 94%, and 99% of Ni, Co, Mn, and Li were achieved under the optimal leaching experimental conditions of citric acid concentration of 2 mol L(-1), leaching temperature of 80 °C, leaching time of 90 min, liquid-solid ratio of 30 ml g(-1), and 2 vol. % H2O2. For the metals recovery process, nickel and cobalt were selectively precipitated by dimethylglyoxime reagent and ammonium oxalate sequentially. Then manganese was extracted by Na-D2EHPA and the manganese-loaded D2EHPA was stripped with sulfuric acid. The manganese was recovered as MnSO4 in aqueous phase and D2EHPA could be reused after saponification. Finally, lithium was precipitated by 0.5 mol L(-1) sodium phosphate. Under their optimal conditions, the recovery percentages of Ni, Co, Mn, and Li can reach 98%, 97%, 98%, and 89%, respectively. This is a relatively simple route in which all metal values could be effectively leached and recovered in citric acid media. © The Author(s) 2014.

  13. Nuclear structure and astrophysics with accelerated beams of radioactive ions: A new multidisciplinary research tool

    International Nuclear Information System (INIS)

    Garrett, J.D.

    1995-01-01

    After a brief discussion of the techniques for producing accelerated radioactive ion beams (RIBs), several recent scientific applications are mentioned. Three general nuclear structure topics, which can be addressed using RIBs, are discussed in some detail: possible modifications of the nuclear shell structure near the particle drip lines; various possibilities for decoupling the proton and neutron mass distributions for weakly bound nuclei; and tests of fundamental nuclear symmetries for self-conjugate and nearly self-conjugate nuclei. The use of RIBs to study r- and rp-process nucleosynthesis also is discussed

  14. Gamma-Ray Spectroscopy at TRIUMF-ISAC: the New Frontier of Radioactive Ion Beam Research

    Science.gov (United States)

    Ball, G. C.; Andreoiu, C.; Austin, R. A. E.; Bandyopadhyay, D.; Becker, J. A.; Bricault, P.; Brown, N.; Chan, S.; Churchman, R.; Colosimo, S.; Coombes, H.; Cross, D.; Demand, G.; Drake, T. E.; Dombsky, M.; Ettenauer, S.; Finlay, P.; Furse, D.; Garnsworthy, A.; Garrett, P. E.; Green, K. L.; Grinyer, G. F.; Hyland, B.; Hackman, G.; Kanungo, R.; Kulp, W. D.; Lassen, J.; Leach, K. G.; Leslie, J. R.; Mattoon, C.; Melconian, D.; Morton, A. C.; Pearson, C. J.; Phillips, A. A.; Rand, E.; Sarazin, F.; Svensson, C. E.; Sumithrarachchi, S.; Schumaker, M. A.; Triambak, S.; Waddington, J. C.; Walker, P. M.; Williams, S. J.; Wood, J. L.; Wong, J.; Zganjar, E. F.

    2009-03-01

    High-resolution gamma-ray spectroscopy is essential to fully exploit the unique scientific opportunities at the next generation radioactive ion beam facilities such as the TRIUMF Isotope Separator and Accelerator (ISAC). At ISAC the 8π spectrometer and its associated auxiliary detectors is optimize for β-decay studies while TIGRESS an array of segmented clover HPGe detectors has been designed for studies with accelerated beams. This paper gives a brief overview of these facilities and also presents recent examples of the diverse experimental program carried out at the 8π spectrometer.

  15. Combined in-beam gamma-ray and conversion electron spectroscopy with radioactive ion beams

    Directory of Open Access Journals (Sweden)

    Konki J.

    2013-12-01

    Full Text Available In-beam gamma-ray and electron spectroscopy have been widely used as tools to study the broad variety of phenomena in nuclear structure. The SPEDE spectrometer is a new device to be used in conjunction with the MINIBALL germanium detector array to enable the detection of internal conversion electrons in coincidence with gamma rays from de-exciting nuclei in radioactive ion beam experiments at the upcoming HIE-ISOLDE facility at CERN, Switzerland. Geant4 simulations were carried out in order to optimise the design and segmentation of the silicon detector to achieve good energy resolution and performance.

  16. Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor

    International Nuclear Information System (INIS)

    Garcia M, H.; Emeterio H, M.; Canizal S, C.

    1999-01-01

    This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

  17. Dose and dose commitment calculations from groundwaterborne radio-active elements released from a repository for spent nuclear fuel

    International Nuclear Information System (INIS)

    Bergstroem, U.

    1983-05-01

    The turnover of radioactive matter entering the biosphere with groundwater has been studied with regard to exposure and doses to critical groups and populations. Two main recipients, a well and a lake, have been considered for the inflow of groundwaterborne nuclides. Mathematical models of a set of coupled ecosystems on regional, intermediate and global levels have been used for calculations of doses. The intermediate system refers to the Baltic Sea. The mathematical treatment of the model is based upon compartment theory with first order kinetics and also includes products in decay chains. The time-dependent exposures have been studied for certain long-lived nuclides of radiological interest in waste from disposed fuel. Dose and dose commitment have been calculated for different episodes for inflow to the biosphere. (author)

  18. The beam diagnostic instruments in Beijing radioactive ion-beam facilities isotope separator on-line

    International Nuclear Information System (INIS)

    Ma, Y.; Cui, B.; Ma, R.; Tang, B.; Chen, L.; Huang, Q.; Jiang, W.

    2014-01-01

    The beam diagnostic instruments for Beijing Radioactive Ion-beam Facilities Isotope Separator On-Line are introduced [B. Q. Cui, Z. H. Peng, Y. J. Ma, R. G. Ma, B. Tang, T. Zhang, and W. S. Jiang, Nucl. Instrum. Methods 266, 4113 (2008); T. J. Zhang, X. L. Guan, and B. Q. Cui, in Proceedings of APAC 2004, Gyeongju, Korea, 2004, http://www.jacow.org , p. 267]. For low intensity ion beam [30–300 keV/1 pA–10 μA], the beam profile monitor, the emittance measurement unit, and the analyzing slit will be installed. For the primary proton beam [100 MeV/200 μA], the beam profile scanner will be installed. For identification of the nuclide, a beam identification unit will be installed. The details of prototype of the beam diagnostic units and some experiment results will be described in this article

  19. Method for monitoring the course of oxidation of iodide ion during radioactive iodination operation

    Energy Technology Data Exchange (ETDEWEB)

    Xuezhong, Luo

    1986-05-01

    A micro-iodine ion selective electrode is developed to follow and monitor the course of oxidation of iodide ion during radioactive iodination operation. The experimental results indicate that this method can quickly respond to the course of oxidation if the reacting liquid is greater than 30 micro liter in volume. Therefore it can be used for accurate controlling the amount of oxidzing reagent used for the reaction, for example, the amount of chloroamine T can be reduced to 1/40 of the amount ordinarily used for the preparation of angiotonin II and insulin. The effect of pH and concentration of phosphate of the reacting liquid to the oxidation reaction of I with chloroamine T is also studied.

  20. Brief draft on surface and subsurface storage of high level and long-lived radioactive wastes. Spent fuels synthesis file

    International Nuclear Information System (INIS)

    Dumas, C.; Jaecki, P.

    2002-01-01

    This document makes a synthesis of the results of two brief draft studies performed in 2002 about the surface and subsurface storage of spent fuels. These studies stress on the long duration aspect of the disposal: feasibility of a secular disposal facility, potential risks and safety level of such a facility, estimation of the initial investment and of operation and maintenance costs. The main points of the specifications and the input data are presented first, and then the subsurface and surface draft studies are described. Content: specifications (imposed design principles and options, dry corrosion, input data); subsurface storage (description and design options, thermal dimensioning and ventilation, geotechnical stability of the facility, subsurface water management, dry corrosion, infrastructure durability, safety, monitoring, security and physical protection, technical-economical aspects, case of Mox fuel, case of glass packages); surface storage (description and design options, thermal dimensioning and ventilation, mechanical dimensioning of the facility, dry corrosion, infrastructure durability, safety, monitoring, security and physical protection, technical-economical aspects, case of Mox fuel, case of glass packages); conclusions and perspectives. (J.S.)

  1. Recovery of valuable metals from waste cathode materials of spent lithium-ion batteries using mild phosphoric acid

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiangping, E-mail: chenxiangping101@163.com [School of Environmental Science and Engineering, Shaanxi University of Science & Technology, Xi’an 710021 (China); College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China); Ma, Hongrui, E-mail: mahr@sust.edu.cn [School of Environmental Science and Engineering, Shaanxi University of Science & Technology, Xi’an 710021 (China); Luo, Chuanbao; Zhou, Tao [College of Chemistry and Chemical Engineering, Central South University, Changsha 410083 (China)

    2017-03-15

    Graphical abstract: Cobalt can be directly recovered as Co{sub 3}(PO{sub 4}){sub 2} from waste LiCoO{sub 2} using H{sub 3}PO{sub 4} as leaching and precipitating agent. - Highlights: • Phosphoric acid was innovatively used as leaching and precipitating agent. • Over 99% Co and Li can be separated and recovered in a single leaching step. • Co and Li can be separated under mild conditions of 40 °C and 0.7 M H{sub 3}PO{sub 4}. • Activation energy values for Co and Li are 7.3 and 10.168 kJ/mol. • Cobalt phosphate (97.1% in purity) can be obtained as the leaching product. - Abstract: Sustainable recycling of valuable metals from spent lithium-ion batteries (LIBs) may be necessary to alleviate the depletion of strategic metal resources and potential risk of environmental pollution. Herein a hydrometallurgical process was proposed to explore the possibility for the recovery of valuable metals from the cathode materials (LiCoO{sub 2}) of spent LIBs using phosphoric acid as both leaching and precipitating agent under mild leaching conditions. According to the leaching results, over 99% Co can be separated and recovered as Co{sub 3}(PO{sub 4}){sub 2} in a short-cut process involved merely with leaching and filtrating, under the optimized leaching conditions of 40 °C (T), 60 min (t), 4 vol.% H{sub 2}O{sub 2}, 20 mL g{sup −1} (L/S) and 0.7 mol/L H{sub 3}PO{sub 4}. Then leaching kinetics was investigated based on the logarithmic rate kinetics model and the obtained results indicate that the leaching of Co and Li fits well with this model and the activation energies (Ea) for Co and Li are 7.3 and 10.2 kJ/mol, respectively. Finally, it can be discovered from characterization results that the obtained product is 97.1% pure cobalt phosphate (Co{sub 3}(PO{sub 4}){sub 2}).

  2. Performance and life cycle environmental benefits of recycling spent ion exchange brines by catalytic treatment of nitrate.

    Science.gov (United States)

    Choe, Jong Kwon; Bergquist, Allison M; Jeong, Sangjo; Guest, Jeremy S; Werth, Charles J; Strathmann, Timothy J

    2015-09-01

    Salt used to make brines for regeneration of ion exchange (IX) resins is the dominant economic and environmental liability of IX treatment systems for nitrate-contaminated drinking water sources. To reduce salt usage, the applicability and environmental benefits of using a catalytic reduction technology to treat nitrate in spent IX brines and enable their reuse for IX resin regeneration were evaluated. Hybrid IX/catalyst systems were designed and life cycle assessment of process consumables are used to set performance targets for the catalyst reactor. Nitrate reduction was measured in a typical spent brine (i.e., 5000 mg/L NO3(-) and 70,000 mg/L NaCl) using bimetallic Pd-In hydrogenation catalysts with variable Pd (0.2-2.5 wt%) and In (0.0125-0.25 wt%) loadings on pelletized activated carbon support (Pd-In/C). The highest activity of 50 mgNO3(-)/(min - g(Pd)) was obtained with a 0.5 wt%Pd-0.1 wt%In/C catalyst. Catalyst longevity was demonstrated by observing no decrease in catalyst activity over more than 60 days in a packed-bed reactor. Based on catalyst activity measured in batch and packed-bed reactors, environmental impacts of hybrid IX/catalyst systems were evaluated for both sequencing-batch and continuous-flow packed-bed reactor designs and environmental impacts of the sequencing-batch hybrid system were found to be 38-81% of those of conventional IX. Major environmental impact contributors other than salt consumption include Pd metal, hydrogen (electron donor), and carbon dioxide (pH buffer). Sensitivity of environmental impacts of the sequencing-batch hybrid reactor system to sulfate and bicarbonate anions indicate the hybrid system is more sustainable than conventional IX when influent water contains reuse cycles. The study showed that hybrid IX/catalyst reactor systems have potential to reduce resource consumption and improve environmental impacts associated with treating nitrate-contaminated water sources. Copyright © 2015 Elsevier Ltd. All rights

  3. Radiation situation dynamics at the Andreeva Bay site for temporary storage of spent nuclear fuel and radioactive waste over the period 2002-2016.

    Science.gov (United States)

    Chizhov, K; Sneve, M K; Shandala, N; Siegien-Iwaniuk, K; Smith, G M; Krasnoschekov, A; Kosnikov, A; Grigoriev, A; Simakov, A; Kemsky, I; Kryuchkov, V

    2018-02-01

    The Coastal Technical Base (CTB) №569 at Andreeva Bay was established in the early 1960s and intended for the refueling of nuclear submarine reactors and temporary storage of spent nuclear fuel (SNF) and radioactive waste (RW). In 2001, the base was transferred to the Russian Ministry for Atomic Energy and the site remediation began. The paper describes in detail the radiation situation change at the technical site in Andreeva Bay from 2002-2016, the period of preparation for the most critical phase of remedial work: removal of spent fuel assemblies. The analysis of aggregated indicators and data mining were used. The article suggests the best number and location of checkpoints needed to ensure sufficient accuracy of the radiation situation description. The fractal properties of the radiation field are studied using the Hurst index. The relationship between checkpoints was assessed using the method of searching for checkpoint communities. The decrease in the integral of the ambient dose equivalent rate (ADER) at the technical site was evaluated by the method of time series decomposition. Three components of time series were identified: trend, seasonal and residual. The trend of the ADER integral over the technical site is a monotonic decreasing function, where the initial and final values differ tenfold. Taking into account that 137 Cs dominates the radiation situation on-site, it is clear that the ADER due to the radionuclide decay will have decreased by 1.4 times. It is estimated that only a small proportion of 137 Cs has migrated off-site. Therefore, approximately a sevenfold decrease in dose rate is mainly due to remediation activities of personnel. During the year, the seasonal component varies the ADER integral by a factor of two, due to snowfall. The residual component reflects the uncertainty of the ADER integral calculation and phases of active SNF and RW management. The methods developed are used to support the optimization of remediation work as well

  4. Dynamics of atoms-ions transformation processes in the radioactive ion production systems ISOL

    International Nuclear Information System (INIS)

    Jardin, Pascal

    2013-01-01

    The aims of this work were 1)to study the effect of diffusion, effusion and ionization processes in the atom-ion transformation, 2)to better understand the temporal behaviour of ISOL devices and to apply it to the developments of the ISOL production systems. These aims were partially reached: the results obtained with 'ECS ECR' of SPIRAL 1 and SPIRAL 2 and their confrontation have allowed to analytically described their temporal behaviour and to reveal under which conditions it is possible to consider the processes of diffusion, effusion and ionization as separable processes and consequently to consider them as consecutive. (O.M.) [fr

  5. Metallic ions catalysis for improving bioleaching yield of Zn and Mn from spent Zn-Mn batteries at high pulp density of 10.

    Science.gov (United States)

    Niu, Zhirui; Huang, Qifei; Wang, Jia; Yang, Yiran; Xin, Baoping; Chen, Shi

    2015-11-15

    Bioleaching of spent batteries was often conducted at pulp density of 1.0% or lower. In this work, metallic ions catalytic bioleaching was used for release Zn and Mn from spent ZMBs at 10% of pulp density. The results showed only Cu(2+) improved mobilization of Zn and Mn from the spent batteries among tested four metallic ions. When Cu(2+) content increased from 0 to 0.8 g/L, the maximum release efficiency elevated from 47.7% to 62.5% for Zn and from 30.9% to 62.4% for Mn, respectively. The Cu(2+) catalysis boosted bioleaching of resistant hetaerolite through forming a possible intermediate CuMn2O4 which was subject to be attacked by Fe(3+) based on a cycle of Fe(3+)/Fe(2+). However, poor growth of cells, formation of KFe3(SO4)2(OH)6 and its possible blockage between cells and energy matters destroyed the cycle of Fe(3+)/Fe(2+), stopping bioleaching of hetaerolite. The chemical reaction controlled model fitted best for describing Cu(2+) catalytic bioleaching of spent ZMBs. Copyright © 2015 Elsevier B.V. All rights reserved.

  6. Production of chemically reactive radioactive ion beams through on-line separation

    International Nuclear Information System (INIS)

    Joinet, A.

    2003-10-01

    The ISOL (isotope separation on line) allows the production of secondary radioactive ion beams through spallation or fragmentation or fission reactions that take place in a thick target bombarded by a high intensity primary beam. The challenge is to increase the intensity and purity of the radioactive beam. The optimization of the system target/source requires the right choice of material for the target by taking into account the stability of the material, its reactivity and the ionization method used. The target is an essential part of the system because radioactive elements are generated in it and are released more or less quickly. Tests have been made in order to select the best fitted material for the release of S, Se, Te, Ge and Sn. Materials tested as target filling are: ZrO 2 , Nb, Ti, V,TiO 2 , CeO x , ThO 2 , C, ZrC 4 and VC). Other molecules such as: COSe, COS, SeS, COTe, GeS, SiS, SnS have been studied to ease the extraction of recoil nuclei (Se, S, Te, Ge and Sn) produced inside the target

  7. The Holifield Radioactive Ion Beams Facility (HRIBF) - getting ready to do experiments

    International Nuclear Information System (INIS)

    Shapira, D.; Lewis, T.A.

    1998-01-01

    The conversion of the HHIRF facility to a Radioactive Ion Beam facility started in 1994. In this ISOL type facility the Cyclotron has been re-fitted as a driver providing high intensity proton beams which react with the target from which the radioactive products are extracted and then accelerated in the Tandem Electrostatic Accelerator to the desired energy for nuclear science studies. Facilities for nuclear physics experiments are at different stages of development: A Recoil Mass Spectrometer (RMS) with a complement of detectors at the focal plane and around the target is used primarily for nuclear structure studies. A large recoil separator combining velocity and momentum selection, with its complement of focal plane detectors, will be dedicated to measurements relevant to nuclear astrophysics. The Enge Split Pole spectrograph is being re-fitted for operation in a gas filled mode, making it a more versatile tool for nuclear reaction studies. With the new experimental equipment being commissioned and the prospects of running experiments with low intensity radioactive beams a significant effort to develop equipment for beam diagnostics is underway. Some of the efforts and results in developing beam diagnostic tools will be described

  8. The development of a mobile hot cell facility for the conditioning of spent high activity radioactive sources (SHARS)

    International Nuclear Information System (INIS)

    Liebenberg, G.R.; Al-Mughrabi, M.

    2008-01-01

    The International Atomic Energy Agency (IAEA) Waste Technology Section with additional support from the U.S. National Nuclear Security Agency (NNSA) through the IAEA Nuclear Security Fund has funded the design, fabrication, evaluation, and testing of a portable hot cell intended to address the problem of disused SHARS in obsolete irradiation devices such as teletherapy heads and dry irradiators. The project is initially targeting the African continent but expected soon to expand to Latin America and Asia. This hot cell would allow source removal, characterization, consolidation, repackaging in modern storage shields, and secure storage of high risk SHARS at single sites in each IAEA Member State. The mobile hot cell and related equipment is transported in two shipping containers to a specific country where the following process takes place: 1-) Assembly of hot cell; 2-) Removal of SHARS from working shields, encapsulation into a stainless steel capsule and placement into a long term storage shield; 3-) Conditioning of any other spent sources the country may require; 4-) Dismantling of the hot cell; 5-) Shipping equipment out of country. The operation in a specific country is planned to be executed over a three week period. This presentation will discuss the development of the mobile hot cell facility as well as the demonstration of the state of readiness of the system for manipulation of SHARS and the planned execution of the conditioning operations. As a result of this project, excess SHARS could be managed safely and securely and possibly be more easily repatriated to their country of origin for appropriate final disposition. (author)

  9. A combined thermal dissociation and electron impact ionization source for radioactive ion beam generation (abstract)a

    International Nuclear Information System (INIS)

    Alton, G.D.; Williams, C.

    1996-01-01

    The probability for simultaneously dissociating and efficiently ionizing the individual atomic constituents of molecular feed materials with conventional, hot-cathode, electron-impact ion sources is low and consequently, the ion beams from these sources often appear as mixtures of several molecular sideband beams. This fragmentation process leads to dilution of the intensity of the species of interest for radioactive ion beam (RIB) applications where beam intensity is at a premium. We have conceived an ion source that combines the excellent molecular dissociation properties of a thermal dissociator and the high ionization efficiency characteristics of an electron impact ionization source that will, in principle, overcome this handicap. The source concept will be evaluated as a potential candidate for use for RIB generation at the Holifield Radioactive Ion Beam Facility, now under construction at the Oak Ridge National Laboratory. The design features and principles of operation of the source are described in this article. copyright 1996 American Institute of Physics

  10. Development of composite ion exchangers and their use in treatment of liquid radioactive wastes

    International Nuclear Information System (INIS)

    Sebesta, F.; John, J.; Motl, A.

    1997-01-01

    Composite inorganic-organic absorbers represent a group of inorganic ion exchangers modified by using binding organic material for preparation of larger size particles having higher granular strength. Such modification of originally powdered or microcrystalline inorganic ion exchangers makes their application in packed beds possible. Modified polyacrylonitrile (PAN) has been used as a universal binding polymer for a number of inorganic ion exchangers. The kinetics of ion exchange and sorption capacity of such composite absorbers is not influenced by the binding polymer mentioned above. The contents of active component in composite absorber can be varied over a very broad range (5-95% of the dry weight of the composite absorber). These composite absorbers have been tested for separation and concentration of various contaminants from aqueous solutions. Their high selectivity and sorption efficiency are advantageous for treatment of various radioactive and/or industrial waste waters, removal of natural and/or artificial radionuclides and heavy or toxic metals from underground water, determination of radionuclides ( 137 Cs, 60 Co, U, Ra) in the environment, etc. Examples of some of these applications are presented in this report. (author). 21 refs, 9 figs, 10 tabs

  11. Development of a surface ionization source for the production of radioactive alkali ion beams in SPIRAL

    International Nuclear Information System (INIS)

    Eleon, C.; Jardin, P.; Gaubert, G.; Saint-Laurent, M.-G.; Alcantara-Nunez, J.; Alves Conde, R.; Barue, C.; Boilley, D.; Cornell, J.; Delahaye, P.; Dubois, M.; Jacquot, B.; Leherissier, P.; Leroy, R.; Lhersonneau, G.; Marie-Jeanne, M.; Maunoury, L.; Pacquet, J.Y.; Pellemoine, F.; Pierret, C.

    2008-01-01

    In the framework of the production of radioactive alkali ion beams by the isotope separation on-line (ISOL) method in SPIRAL I, a surface ionization source has been developed at GANIL to produce singly-charged ions of Li, Na and K. This new source has been designed to work in the hostile environment whilst having a long lifetime. This new system of production has two ohmic heating components: the first for the target oven and the second for the ionizer. The latter, being in carbon, offers high reliability and competitive ionization efficiency. This surface ionization source has been tested on-line using a 48 Ca primary beam at 60.3 A MeV with an intensity of 0.14 pA. The ionization efficiencies obtained for Li, Na and K are significantly better than the theoretical values of the ionization probability per contact. The enhanced efficiency, due to the polarization of the ionizer, is shown to be very important also for short-lived isotopes. In the future, this source will be associated with the multicharged electron-cyclotron-resonance (ECR) ion source NANOGAN III for production of multicharged alkali ions in SPIRAL. The preliminary tests of the set up are also presented in this contribution.

  12. Electron Cyclotron Resonance (ECR) Ion Source Development at the Holified Radioactive Ion Beam Facility

    Science.gov (United States)

    Bilheux, Hassina; Liu, Yuan; Alton, Gerald; Cole, John; Williams, Cecil; Reed, Charles

    2004-11-01

    Performance of ECR ion sources can be significantly enhanced by increasing the physical size of their ECR zones in relation to the size of their plasma volumes (spatial and frequency domain methods).^3-5 A 6 GHz, all-permanent magnet ECR ion source with a large resonant plasma volume has been tested at ORNL.^6 The magnetic circuit can be configured for creating both flat-β (volume) and conventional minimum-β (surface) resonance conditions. Direct comparisons of the performance of the two source types can be made under similar operating conditions. In this paper, we clearly demonstrate that the flat-β source outperforms its minimum-β counterpart in terms of charge state distribution and intensity within a particular charge state. ^1bilheuxhn@ornl.gov ^2Managed by UT-Battelle, LLC, for the U.S. Department of Energy under contract DE-AC05-00OR22725. ^3G.D. Alton, D.N. Smithe, Rev. Sci. Instrum. 65 (1994) 775. ^4G.D. Alton et al., Rev. Sci. Instrum. 69 (1998) 2305. ^5Z.Q. Xie, C.M. Lyneis, Rev. Sci. Instrum. 66 (1995) 4218. ^6Y. Liu et al., Rev. Sci. Instrum. 69 (1998) 1311.

  13. Proceedings of the workshop on the science of intense radioactive ion beams

    International Nuclear Information System (INIS)

    McClelland, J.B.; Vieira, D.J.

    1990-10-01

    This report contains the proceedings of a 2-1/2 day workshop on the Science of Intense Radioactive Ion Beams which was held at the Los Alamos National Laboratory on April 10--12, 1990. The workshop was attended by 105 people, representing 30 institutions from 10 countries. The thrust of the workshop was to develop the scientific opportunities which become possible with a new generation intense Radioactive Ion Beam (RIB) facility, currently being discussed within North America. The workshop was organized around five primary topics: (1) reaction physics; (2) nuclei far from stability/nuclear structure; (3) nuclear astrophysics; (4) atomic physics, material science, and applied research; and (5) facilities. Overview talks were presented on each of these topics, followed by 1-1/2 days of intense parallel working group sessions. The final half day of the workshop was devoted to the presentation and discussion of the working group summary reports, closing remarks and a discussion of future plans for this effort

  14. The Effusive-Flow Properties of Target/Vapor-Transport Systems for Radioactive Ion Beam Applications

    CERN Document Server

    Kawai, Yoko; Liu, Yuan

    2005-01-01

    Radioactive atoms produced by the ISOL technique must diffuse from a target, effusively flow to an ion source, be ionized, be extracted, and be accelerated to research energies in a time commensurate with the lifetime of the species of interest. We have developed a fast valve system (closing time ~100 us) that can be used to accurately measure the effusion times of chemically active or inactive species through arbitrary geometry and size vapor transport systems with and without target material in the reservoir. The effusive flow times are characteristic of the system and thus serve as figures of merit for assessing the quality of a given vapor transport system as well as for assessing the permeability properties of a given target design. This article presents effusive flow data for noble gases flowing through a target reservoir and ion source system routinely used to generate radioactive species at the HRIBF with and without disks of 6 times and 10 times compressed Reticulated Vitreous Carbon Foam (RVCF) with...

  15. Proceedings of the workshop on the science of intense radioactive ion beams

    Energy Technology Data Exchange (ETDEWEB)

    McClelland, J.B.; Vieira, D.J. (comps.)

    1990-10-01

    This report contains the proceedings of a 2-1/2 day workshop on the Science of Intense Radioactive Ion Beams which was held at the Los Alamos National Laboratory on April 10--12, 1990. The workshop was attended by 105 people, representing 30 institutions from 10 countries. The thrust of the workshop was to develop the scientific opportunities which become possible with a new generation intense Radioactive Ion Beam (RIB) facility, currently being discussed within North America. The workshop was organized around five primary topics: (1) reaction physics; (2) nuclei far from stability/nuclear structure; (3) nuclear astrophysics; (4) atomic physics, material science, and applied research; and (5) facilities. Overview talks were presented on each of these topics, followed by 1-1/2 days of intense parallel working group sessions. The final half day of the workshop was devoted to the presentation and discussion of the working group summary reports, closing remarks and a discussion of future plans for this effort.

  16. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation

    Energy Technology Data Exchange (ETDEWEB)

    None

    1987-12-01

    The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in the mined geologic disposal system. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document. This document is backed up by five PC-compatible data bases, written in a user-oriented, menu-driven format, which were developed for this purpose. The data bases are the LWR Assemblies Data Base; the LWR Radiological Data Base; the LWR Quantities Data Base; the LWR NFA Hardware Data Base; and the High-Level Waste Data Base. The above data bases may be ordered using the included form. An introductory information diskette can be found inside the back cover of this report. It provides a brief introduction to each of these five PC data bases. 116 refs., 18 figs., 67 tabs.

  17. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation

    International Nuclear Information System (INIS)

    1987-12-01

    The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in the mined geologic disposal system. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document. This document is backed up by five PC-compatible data bases, written in a user-oriented, menu-driven format, which were developed for this purpose. The data bases are the LWR Assemblies Data Base; the LWR Radiological Data Base; the LWR Quantities Data Base; the LWR NFA Hardware Data Base; and the High-Level Waste Data Base. The above data bases may be ordered using the included form. An introductory information diskette can be found inside the back cover of this report. It provides a brief introduction to each of these five PC data bases. 116 refs., 18 figs., 67 tabs

  18. Materials science and biophysics applications at the ISOLDE radioactive ion beam facility

    Energy Technology Data Exchange (ETDEWEB)

    Wahl, U., E-mail: uwahl@itn.pt [Instituto Tecnologico e Nuclear, Estrada Nacional 10, 2686-953 Sacavem (Portugal); Centro de Fisica Nuclear da Universidade de Lisboa, Av. Prof. Gama Pinto 2, 1649-003 Lisboa (Portugal)

    2011-12-15

    The ISOLDE isotope separator facility at CERN provides a variety of radioactive ion beams, currently more than 800 different isotopes from {approx}70 chemical elements. The radioisotopes are produced on-line by nuclear reactions from a 1.4 GeV proton beam with various types of targets, outdiffusion of the reaction products and, if possible, chemically selective ionisation, followed by 60 kV acceleration and mass separation. While ISOLDE is mainly used for nuclear and atomic physics studies, applications in materials science and biophysics account for a significant part (currently {approx}15%) of the delivered beam time, requested by 18 different experiments. The ISOLDE materials science and biophysics community currently consists of {approx}80 scientists from more than 40 participating institutes and 21 countries. In the field of materials science, investigations focus on the study of semiconductors and oxides, with the recent additions of nanoparticles and metals, while the biophysics studies address the toxicity of metal ions in biological systems. The characterisation methods used are typical radioactive probe techniques such as Moessbauer spectroscopy, perturbed angular correlation, emission channeling, and tracer diffusion studies. In addition to these 'classic' methods of nuclear solid state physics, also standard semiconductor analysis techniques such as photoluminescence or deep level transient spectroscopy profit from the application of radioactive isotopes, which helps them to overcome their chemical 'blindness' since the nuclear half life of radioisotopes provides a signal that changes in time with characteristic exponential decay or saturation curves. In this presentation an overview will be given on the recent research activities in materials science and biophysics at ISOLDE, presenting some of the highlights during the last five years, together with a short outlook on the new developments under way.

  19. Incineration of ion-exchange resins in fluidized bed. Part of a coordinated programme on treatment of spent ion exchange resins

    International Nuclear Information System (INIS)

    Valkianinen, M.

    1980-10-01

    Incineration of ion-exchange resins in a fluidized bed was studied on the pilot plant scale. The test programme performed consisted of the testing of various bed materials and finding the optimal conditions of incineration of spent resins. Granular resins were incinerated in an ethanol-water mixture. Incinernation converts the organic resin into inert oxide material, which can be solidified for instance with cement. The weight of the ash was 1...20% and the volume 2...30% of the original resins, which contained 15...25% moisture. When solidified with cement the volume of the ash-concrete is 4...22% of the concrete of equal compressive strength acquired by direct solidification. Water immersion and heat tests of solidified ash showed satisfactory results. The absorption of Cs and Co in various bed materials was studied by means of inactive tracer materials. Biotite and chamotte absorbed significantly, but this absorption does not drastically help on the off gas side. The sintering of the bed materials in the presence of sodium was studied. Corundum, chamotte and biotite have a safety limit of 5% sodium of the bed's weight at 850 0 C

  20. Recovery process of cathode material of the spent lithium-ion batteries using Pechini methods; Reprocessamento de catodos de baterias de ions litio descartadas utilizando sintese Pechini

    Energy Technology Data Exchange (ETDEWEB)

    Polo Fonseca, C.; Prado, R.M.; Santos Junior, G.A.; Marques, E.C.; Neves, S., E-mail: carla.polo@yahoo.com.b [Universidade Sao Francisco (LCAM/USF), Itatiba, SP (Brazil). Lab. de Caracterizacao e Aplicacao de Materiais; Amaral, F. [Universidade Federal de Uberlandia (UFU), MG (Brazil)

    2010-07-01

    This work proposes a new process of recovering LiCoO{sub 2} from spent Li-ion batteries (LIBs) by a combination of acid leaching and Pechini synthesis, as an alternative process to improve the recovery efficiency of LiCoO{sub 2} and reduce energy consumption and pollution. The effects of calcination temperature and lithium acetate addition in the synthesis on the structure and morphology of LiCoO{sub 2} powders were characterized by X-ray diffraction, Raman spectroscopy and scanning electron microscopy. According to the analysis, the crystallinity of LiCoO{sub 2} powders depends on the calcination temperature. The results indicate the layered HT-LiCoO{sub 2} powders can be obtained at 750 deg C for 24 h in oxygen with lithium salt addition. Cyclic voltammograms showed one reversible redox process at 4.0/3.85 V for the LiCoO{sub 2} obtained with lithium addition in the synthesis and irreversible redox process for the LiCoO{sub 2} obtained without lithium addition. (author)

  1. Photocatalytic properties of Co_3O_4/LiCoO_2 recycled from spent lithium-ion batteries using citric acid as leaching agent

    International Nuclear Information System (INIS)

    Santana, I.L.; Moreira, T.F.M.; Lelis, M.F.F.; Freitas, M.B.J.G.

    2017-01-01

    In this work, cobalt and lithium from the cathodes of spent lithium-ion batteries were recycled to synthesize a mixture of Co_3O_4 and LiCoO_2. The positive electrode was leached with citric acid in the green recycling. After being heated to 85 °C, the leaching solution formed a pink sol, and after being dried at 120 °C for 24 h, it formed a gel, which is a precursor material for Co_3O_4 and LiCoO_2 synthesis. A mixture of Co_3O_4 and LT-LiCoO_2 was obtained after the calcination of the precursor material at 450 °C for 3 h. The photocatalytic properties of the Co_3O_4 and LiCoO_2 were tested in the discoloration of methylene blue dye. The discoloration efficiency of methylene blue dye in the presence of Co_3O_4 and LiCoO_2 was 90% after 10 h and 100% after 24 h of heterogeneous catalysis. The contribution of this work is that it presents a means to produce valuable materials with photocatalytic properties from recycled batteries through a spent Li-ion battery recycling process without polluting the environment. - Highlights: • Synthesis a mixture of Co_3O_4/LiCoO_2 from spent Li-ion batteries. • Citric acid for leaching of the cathodes of the spent Li-ion batteries. • Co_3O_4/LiCoO_2 as catalysts in the photodegradation of the methylene blue dye.

  2. Universal method for effusive-flow characterization target ion source/vapor transport systems for radioactive ion beam generation (abstract)

    International Nuclear Information System (INIS)

    Alton, G.D.; Bilheux, J.-C.; Liu, Y.; Cole, J. A.; Williams, C.

    2004-01-01

    Worldwide interest in the use of accelerated radioactive ion beams (RIBs) for exploring reactions important in understanding the structure of the nucleus and nuclear astrophysical phenomena has motivated the construction of facilities dedicated to their production and acceleration. Many facilities utilize the isotope-separator-on-line (ISOL) method in which species of interest are generated within a solid or liquid target matrix. Experimentally useful RIBs are often difficult to generate by this technique because of the times required for diffusion from the interior of the target material, and to effusively transport the species of interest to the ion source following diffusion release in relation to its lifetime. Therefore, these delay times must be minimized. We have developed an experimental method that can be used to determine effusive-flow times of arbitrary geometry target/vapor transport systems. The technique utilizes a fast valve to measure effusive-flow times as short as 0.1 ms for any chemically active or inactive species through any target system, independent of size, geometry and materials of construction. In this report, we provide a theoretical basis for effusive flow through arbitrary geometry vapor transport systems, describe a universal experimental apparatus for measuring effusive-flow times, and provide time spectra for noble gases through prototype RIB target/vapor-transport systems

  3. Handling, conditioning and storage of spent sealed radioactive sources. Technical manual for the management of low and intermediate level wastes generated at small nuclear research centres and by radioisotope users in medicine, research and industry

    International Nuclear Information System (INIS)

    2000-05-01

    This report is intended to provide reference material, guidance and know-how on handling, conditioning and storage of spent sealed radioactive sources (SRS) to both users of SRS and operators of waste management facilities. The scope of this report covers all types of SRS except those exempted from regulatory control. The report contains in some detail technical procedures for the conditioning of spent SRS, describes the means required to assure the quality of the resulting package and discusses the measures to prepare waste packages with a certain flexibility to accommodate possible future disposal requirements

  4. Answers to questions on National Report of the Slovak Republic. Compiled according to the terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. October 2003

    International Nuclear Information System (INIS)

    2003-10-01

    Slovakia is pleased to present to the State Parties of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management the Answers to questions received on the National Report of the Slovak Republic compiled according to the terms of the Joint Convention (April 2003). Slovakia is ready to provide additional explanations to these Answers during the 1 st Review Meeting. In the Annexes the 254/1994 Coll. LL. Act of the National Council of the Slovak Republic of 25 August 1994 on State Fund of Decommissioning of Nuclear Installations and Handling of Spent Nuclear Fuels and Nuclear Wastes is included

  5. National report of the Slovak Republic. Compiled in terms of the joint convention on the safety of spent fuel management and on the safety of radioactive waste management. 2005

    International Nuclear Information System (INIS)

    Hekel, P.; Ivan, J.; Lukacovic, K.; Parimucha, F.; Suss, J; Tomek, J.; Jurina, V.; Kobzova, D.; Konecny, L.; Homola, J.; Zavazanova, A.; Metke, E.; Turner, M.; Vaclav, J.; Ziakova, M.; Pospisil, M.; Petrik, T.

    2005-09-01

    The National Report (Compiled in terms of the Joint Convention - September 2005) contains information how each of the obligations of the Joint Convention have been implemented. A account of activities carried out by the Nuclear Regulatory Authority of the Slovak Republic (UJD) and other Slovak organisations are presented. These activities are reported under the headings: (A) Introduction; (B) Spent fuel and radioactive waste management; C) Scope of application; (D) Inventories and lists; (E) Legislation and regulation; (F) Other general safety provisions; (G) Safety of spent fuel management; (H) Safety of RAW management; (I) Transboundary movement of SF and RAW; (J) Disused sealed sources; (K) planned activiries to improve safety; (L) Annexes

  6. Thermal treatment and ammoniacal leaching for the recovery of valuable metals from spent lithium-ion batteries.

    Science.gov (United States)

    Chen, Yongming; Liu, Nannan; Hu, Fang; Ye, Longgang; Xi, Yan; Yang, Shenghai

    2018-05-01

    The recycling of spent commercial lithium-ion batteries (LIBs) generates numerous environmental and economic benefits. In this research, a thermal treatment-ammoniacal leaching process is proposed to recover valuable metals from cathode active powder. Based on the thermal behavior by TG-DSC analysis, the cathode active powder is calcined at 300 °C and 550 °C in air atmosphere, and the crystalline phase characterization indicates that a new phase of Co 3 O 4 appears in the cathode active powder calcined at 550 °C, which signifies that the layer structure of LiCoO 2 collapses. The valence of manganese increases to form Li 4 Mn 5 O 12 in spinel structure of LiMn 2 O 4 . Using calcined cathode powder as feed material, ammoniacal leaching is carried out in (NH 4 ) 2 SO 4 -(NH 4 ) 2 SO 3 solution. Under the optimum conditions, Ni, Co, Mn and Li can be completely leached out with efficiencies of 98%, 81%, 92% and 98%, respectively. However, with the increase of ammonia concentration, the leaching efficiency of Mn decreases dramatically to 4% due to the formation of double salts. It is found that Co and Mn can be precipitated into residues in the form of (NH 4 ) 2 Co(SO 4 ) 2 ·H 2 O, (NH 4 ) 2 Mn(SO 3 ) 2 ·H 2 O and (NH 4 ) 2 Mn(SO 4 ) 2 ·6H 2 O under different leaching parameters. Based on the corresponding relationship between the leaching efficiency and phase evolution of object element, selective leaching can be achieved by controlling the formation of double salts. Copyright © 2018 Elsevier Ltd. All rights reserved.

  7. Economical recycling process for spent lithium-ion batteries and macro- and micro-scale mechanistic study

    Science.gov (United States)

    Li, Li; Bian, Yifan; Zhang, Xiaoxiao; Xue, Qing; Fan, Ersha; Wu, Feng; Chen, Renjie

    2018-02-01

    An economical effective method is developed for recycling spent LiNi1/3Co1/3Mn1/3O2 cathodes, where more than 98% Li, Co, Ni and Mn can be leached out with different organic acids, and resynthesized to LiNi1/3Co1/3Mn1/3O2. The leaching mechanism is investigated at macro- and micro-scales. The particles undergo a loosening-breaking-shrinking change for two acids, while the FTIR and UV-vis spectra indicate different coordination reactions. The performance of LiNi1/3Co1/3Mn1/3O2 resynthesized from the leachate of the acetic acid leaching (NCM-Ac) and maleic acid leaching (NCM-Ma) are compared. The first discharge capacity of NCM-Ma and NCM-Ac at 0.2C are 151.6 and 115.0 mA h g-1, respectively. The much better performance of NCM-Ma than NCM-Ac results from the different coordination of the two acids in the sol-gel process, where the maleic acid can esterify to establish a stable network to chelate metal ions, while the weak chelation of acetic acid leads to the formation of impurities. The economics analysis including the cost of leaching acid and energy consumption shows that the price of organic acids and reducing agents, short leaching time, low temperature and high-valued products are the effective way to increase recycling and environmental benefits, which shows advantages in terms of resources cost and added value.

  8. The Radioactive Ion Beams in Brazil (RIBRAS) facility. Description, program, main results, future plans

    Energy Technology Data Exchange (ETDEWEB)

    Lepine-Szily, A.; Lichtenthaeler, R.; Guimaraes, V. [Instituto de Fisica, Universidade de Sao Paulo (Brazil)

    2014-08-15

    RIBRAS (Radioactive Ion Beams in Brazil) is a facility installed at the Institute of Physics of the University of Sao Paulo (IFUSP), Brazil. The RIBRAS system consists of two superconducting solenoids and uses the ''in-flight method'' to produce radioactive ion beams using the primary beam provided by the 8UD Pelletron Tandem of IFUSP. The ion beams produced so far by RIBRAS are {sup 6}He, {sup 8}Li, {sup 7}Be, {sup 10}Be, {sup 8}B, {sup 12}B with intensities that can vary from 10{sup 4} to 10{sup 6} pps. Initially the experimental program covered the study of elastic and inelastic scattering with the objective to study the interaction potential and the reaction mechanisms between weakly bound (RIB) and halo ({sup 6}He and {sup 8}B) projectiles on light, medium and heavy mass targets. With highly purified beams, the study of resonant elastic scattering and resonant transfer reactions, using inverse kinematics and thick targets, has also been included in our experimental program. Also, transfer reactions of astrophysical interest and fusion reactions induced by halo nuclei are part of the near-future research program. Our recent results on elastic scattering, alpha-particle production and total reaction cross sections, as well as the resonant elastic and transfer reactions, are presented. Our plans for the near future are related to the installation of a new beam line and a cave for gamma-ray detection. We intend to place in operation a large area neutron detector available in our laboratory. The long-range plans could be the move of the RIBRAS system to the more energetic beam line of the LINAC post-accelerator (10MeV/nucleon primary beams) still in construction in our laboratory. (orig.)

  9. The Radioactive Ion Beams in Brazil (RIBRAS) facility. Description, program, main results, future plans

    Science.gov (United States)

    Lépine-Szily, A.; Lichtenthäler, R.; Guimarães, V.

    2014-08-01

    RIBRAS (Radioactive Ion Beams in Brazil) is a facility installed at the Institute of Physics of the University of São Paulo (IFUSP), Brazil. The RIBRAS system consists of two superconducting solenoids and uses the "in-flight method" to produce radioactive ion beams using the primary beam provided by the 8UD Pelletron Tandem of IFUSP. The ion beams produced so far by RIBRAS are 6He, 8Li, 7Be, 10Be, 8B, 12B with intensities that can vary from 104 to 106 pps. Initially the experimental program covered the study of elastic and inelastic scattering with the objective to study the interaction potential and the reaction mechanisms between weakly bound (RIB) and halo (6He and 8B projectiles on light, medium and heavy mass targets. With highly purified beams, the study of resonant elastic scattering and resonant transfer reactions, using inverse kinematics and thick targets, has also been included in our experimental program. Also, transfer reactions of astrophysical interest and fusion reactions induced by halo nuclei are part of the near-future research program. Our recent results on elastic scattering, alpha-particle production and total reaction cross sections, as well as the resonant elastic and transfer reactions, are presented. Our plans for the near future are related to the installation of a new beam line and a cave for gamma-ray detection. We intend to place in operation a large area neutron detector available in our laboratory. The long-range plans could be the move of the RIBRAS system to the more energetic beam line of the LINAC post-accelerator (10MeV/nucleon primary beams) still in construction in our laboratory.

  10. Unsaturated flow modeling in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Rechard, Rob P.; Birkholzer, Jens T.; Wu, Yu-Shu; Stein, Joshua S.; Houseworth, James E.

    2014-01-01

    This paper summarizes the progression of modeling efforts of infiltration, percolation, and seepage conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository in the unsaturated zone for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. Scientific understanding of infiltration in a desert environment, unsaturated percolation flux in fractures and matrix of the volcanic tuff, and seepage into an open drift in a thermally perturbed environment was initially lacking in 1984. As understanding of the Yucca Mountain disposal system increased through site characterization and in situ testing, modeling of infiltration, percolation, and seepage evolved from simple assumptions in a single model in 1984 to three modeling modules each based on several detailed process models in 2008. Uncertainty in percolation flux through Yucca Mountain was usually important in explaining the observed uncertainty in performance measures:cumulative release in assessments prior to 1995 and individual dose, thereafter. - Highlights: • Progression of modeling of infiltration, percolation, and seepage conducted is described for a geological repository at Yucca Mountain. • Progression from 1-D in single equivalent to 3-D model of percolation in dual permeability continuum is described. • Introduction of an infiltration boundary condition in 1998 and the refinement for evaluating uncertainty for the license application is described. • Introduction of a seepage module that included calibration to in-site measurements and separating uncertainty and variability is described

  11. Report on the proposal for a Council directive (EURATOM) on the management of spent nuclear fuel and radioactive waste - Committee on Industry, External Trade, Research and Energy

    International Nuclear Information System (INIS)

    2006-01-01

    By letter of 13 May 2003 the Council consulted Parliament, pursuant to Articles 31 and 32 of the EURATOM Treaty, on the proposal for a Council directive (EURATOM) on the management of spent nuclear fuel and radioactive waste (COM(2003) 32 - 2003/0022(CNS)). At the sitting of 15 May 2003 the President of Parliament announced that he had referred the proposal to the Committee on Industry, External Trade, Research and Energy as the committee responsible and the Committee on the Environment, Public Health and Consumer Policy for its opinion (C5-0229/2003). The Committee on Industry, External Trade, Research and Energy appointed Alejo Vidal-Quadras Roca Rapporteur at its meeting of 22 May 2003. The committee considered the Commission proposal and draft report at its meetings of 9 July, 9 September, 6 October, 3 November and 27 November 2003. At the last meeting it adopted the draft legislative resolution by 36 votes to 7, with 2 abstentions. The opinion of the Committee on the Environment, Public Health and Consumer Policy is attached. The report was tabled on 1 December 2003

  12. Waste degradation and mobilization in performance assessments for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Rechard, Rob P.; Stockman, Christine T.

    2014-01-01

    This paper summarizes modeling of waste degradation and mobilization in performance assessments (PAs) conducted between 1984 and 2008 to evaluate feasibility, viability, and assess compliance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain in southern Nevada. As understanding of the Yucca Mountain disposal system increased, the waste degradation module, or succinctly called the source-term, evolved from initial assumptions in 1984 to results based on process modeling in 2008. In early PAs, waste degradation had significant influence on calculated behavior but as the robustness of the waste container was increased and modeling of the container degradation improved, waste degradation had much less influence in later PAs. The variation of dissolved concentrations of radionuclides progressed from simple probability distributions in early PAs to functions dependent upon water chemistry in later PAs. Also, transport modeling of radionuclides in the waste, container, and invert were added in 1995; and, colloid-facilitated transport of radionuclides was added in 1998. - Highlights: • Progression of modeling of waste degradation in performance assessments is discussed for the proposed repository at Yucca Mountain. • Progression of evaluating dissolved concentrations of radionuclides in the source-term is discussed. • Radionuclide transport modeling in the waste, container, and invert in 1995 and thereafter is discussed. • Colloid-facilitated transport in the waste, container, and invert in 1998 and thereafter is discussed

  13. Criteria for selection of target materials and design of high-efficiency-release targets for radioactive ion beam generation

    CERN Document Server

    Alton, G D; Liu, Y

    1999-01-01

    In this report, we define criteria for choosing target materials and for designing, mechanically stable, short-diffusion-length, highly permeable targets for generation of high-intensity radioactive ion beams (RIBs) for use at nuclear physics and astrophysics research facilities based on the ISOL principle. In addition, lists of refractory target materials are provided and examples are given of a number of successful targets, based on these criteria, that have been fabricated and tested for use at the Holifield Radioactive Ion Beam Facility (HRIBF).

  14. Modeling of multi-species ion transport in cement-based materials for radioactive waste container

    International Nuclear Information System (INIS)

    Pang, X.Y.; Li, K.F.; Dangla, P.

    2015-01-01

    Through the conservations of heat and ions mass, a thermo-hydro-ionic model is established for radionuclide ions transport in cement-based porous barrier materials in radwaste disposal. This model is applied to the design and the safety assessment of a high-integrity container (HIC) used for near surface disposal of low- and intermediate-level radwaste. Five working cases are investigated in the safety assessment considering the internal nuclide ion release, internal heating and pressure accumulation, and external leaching. Comparative analysis shows that leaching increases concrete porosity from external side of container, internal heating of 10 K increase can considerably accelerate the nuclide transport process, and the internal pressure increases the transport rate to limited extent. It is shown that each increment of 10 mm in wall thickness will reduce the radioactivity release by 1.5 to 2 times. Together with the mechanical resistance of HIC under impact actions, the thickness of 100 mm is finally retained for design

  15. Technical-and-economic analysis and optimization of the full flow charts of processing of radioactive wastes on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of fast reactors

    Science.gov (United States)

    Gupalo, V. S.; Chistyakov, V. N.; Kormilitsyn, M. V.; Kormilitsyna, L. A.; Osipenko, A. G.

    2015-12-01

    When considering the full flow charts of processing of radioactive wastes (RAW) on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of NIIAR fast reactors, we corroborate optimum technical solutions for the preparation of RAW for burial from a standpoint of heat release, dose formation, and technological storage time with allowance for technical-and-economic and ecological indices during the implementation of the analyzed technologies and equipment for processing of all RAW fluxes.

  16. Resonant ionization by laser beams: application to ions sources and to study the nuclear structure of radioactive tellurium isotopes

    International Nuclear Information System (INIS)

    Sifi, R.

    2007-07-01

    The radioactive ion beams that are produced through current isotope separators are well separated according to the A mass but not according to the Z parameter. The resonant ionization through laser beams applied to ion sources allows the production of radioactive ion beam in a very selective and efficient way by eliminating the isobaric contamination. The first chapter is dedicated to the resonant ionization by laser beams, we describe the principle, the experimental setting, the lasers used, the ionization schemes and the domain of application. The second chapter deals with the application of resonant ionization to laser ion sources for the production of radioactive ion beams. We present experimental tests performed for getting copper ion beams. Resonant ionization through laser is also used in the spectroscopy experiments performed at the Isolde (isotope separation on-line device) installation in CERN where more than 20 elements are ionized very efficiently. The technique is based on a frequency scanning around the excitation transition of the atoms in order to probe the hyperfine structure. Laser spectroscopy allows the determination of the hyperfine structure as well as the isotopic shift of atoms. In the third chapter the method is applied to the spectroscopy of tellurium atoms. First, we define the 2 parameters on which the extraction is based: charge radius and nuclear moments, then we present several theoretical models that we have used to assess our experimental results. (A.C.)

  17. Application of ion exchange in liquid radioactive waste management of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Puskar; Chopra, S K; Sharma, P D [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The operation of nuclear power plants would necessarily result in generation of gaseous, liquid and solid radioactive wastes. The wastes are treated/conditioned to ensure that the permissible discharge limits laid down by Atomic Energy Regulatory Board of India are complied with. The wastes are segregated on activity levels, types of radioisotopes present and chemical nature of liquid streams. The basic philosophy of various treatment techniques is to concentrate and contain as much activity as possible. It is of utmost importance that the wastes are effectively treated by proven methods/processes. The radiochemical nature of waste generated is one of the parameters to select a treatment/conditioning method. The paper presents an outline of various processes adopted for treatment of liquid waste and ion exchange processes, their application in liquid waste management in detail. Projected quantities of liquid wastes for the current designs are included. (author). 2 tabs.

  18. Beam Dynamics Design Studies of a Superconducting Radioactive Ion Beam Post-accelerator

    CERN Document Server

    Fraser, MA; Pasini, M

    2011-01-01

    The HIE-ISOLDE project at CERN proposes a superconducting upgrade to increase the energy range and quality of the radioactive ion beams produced at ISOLDE, which are currently post- accelerated by the normal conducting REX linac. The specification and design choices for the HIE-ISOLDE linac are outlined along with a comprehensive beam dynamics study undertaken to understand and mitigate the sources of beam emittance dilution. The dominant cause of transverse emittance growth was attributed to the coupling between the transverse and longitudinal motions through the phase dependence of the rf defocusing force in the accelerating cavities. A parametric resonance induced by the coupling was observed and its excitation surveyed as a function of trans- verse phase advance using numerical simulations and analytic models to understand and avoid the regions of transverse beam instability. Other sources of emittance growth were studied and where necessary ameliorated, including the beam steering force in the quarter-wa...

  19. A Monte Carlo code to optimize the production of Radioactive Ion Beams by the ISOL technique

    CERN Document Server

    Santana-Leitner, M

    2005-01-01

    Currently the nuclear chart includes around 3000 nuclides, distributed as ${\\beta}^+$, ${\\beta}^-$ and $\\alpha$-emitters, stable and spontaneously fissioning isotopes. A similar amount of unknown nuclei belongs to the so-called \\textit{terra incognita}, the uncertain region contained also within the proton, neutron and (fast) fission driplines and thereby stable against nucleon emission. The exploration of this zone is to be assisted by the use of radioactive ion beams (RIB) and could provide a new understanding of several nuclear properties. Moreover, besides pointing at crucial questions such as the validity of the shell model, the dilute matter and the halo structure, challenging experiments outside nuclear physics are also attended, e.g., explanations of the nucleosythesis processes that may justify why the matter in the universe has evolved to present proportions of elements, and which represents a major challenge to nuclear physics. These, together with other fascinating research lines in particle physi...

  20. Lifetime measurements using radioactive ion beams at intermediate energies and the Doppler shift method

    Energy Technology Data Exchange (ETDEWEB)

    Dewald, A.; Melon, B.; Pissulla, T.; Rother, W.; Fransen, C.; Moeller, O.; Zell, K.O.; Jolie, J. [IKP, Univ. zu Koeln (Germany); Petkov, P. [Bulg. Acad. of Science, INRNE, Solfia (Bulgaria); Starosta, K.; Przemyslaw, A.; Miller, D.; Chester, A.; Vaman, C.; Voss, P.; Gade, A.; Glasmacher, T.; Stolz, A.; Bazin, D.; Weisshaar, D. [NSCL, MSU, East Lansing (United States)

    2007-07-01

    Absolute transition probabilities are crucial quantities in nuclear structure physics. Therefore, it is important to establish Doppler shift (plunger) techniques also for the measurement of level lifetimes in radioactive ion beam experiments. After a first successful test of the Doppler Shift technique at intermediate energy (52MeV/u) with a stable {sup 124}Xe beam, a plunger has been built and used in two experiments, performed at the NSCL/MSU with the SEGA Ge-array and the S800 spectrometer. The aim of the first experiment was to investigate the plunger technique after a knock-out reaction using a radioactive {sup 65}Ge beam at 100 MeV/u for populating excited states in {sup 64}Ge. The second experiment aimed to measure the lifetimes of the first 2{sup +} states in {sup 110,114}Pd with the plunger technique after Coulomb excitation at beam energies of 54 MeV/u. First results of both experiments will be presented and discussed. (orig.)

  1. Disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Dlouhy, Z.

    1982-01-01

    This book provides information on the origin, characteristics and methods of processing of radioactive wastes, as well as the philosophy and practice of their storage and disposal. Chapters are devoted to the following topics: radioactive wastes, characteristics of radioactive wastes, processing liquid and solid radioactive wastes, processing wastes from spent fuel reprocessing, processing gaseous radioactive wastes, fixation of radioactive concentrates, solidification of high-level radioactive wastes, use of radioactive wastes as raw material, radioactive waste disposal, transport of radioactive wastes and economic problems of radioactive wastes disposal. (C.F.)

  2. Exchangers of inorganic ions in the administration of radioactive wastes; Intercambiadores de iones inorganicos en la gestion de desechos radioactivos

    Energy Technology Data Exchange (ETDEWEB)

    Badillo A, V. E.; Lopez R, C., E-mail: veronica.badillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    The prediction of the radionuclide migration in geologic mean requires of a quantitative knowledge of the physiochemical phenomenon of retention in the surface of mineral phases. With this purpose, is necessary the study of the radionuclides retention in solids named model solids are the oxides and phosphates of polyvalent cations. This work presents experimental evidence of the convenience of using two exchangers of inorganic ions, alumina and apatite, in the administration of radioactive wastes due to its selectivity for the main products of divalent fission, Pd and Sr with regard to the anion species represented by Tc. The retention of Sr(III), Pd(II) and Tc(-I) in hydroxyapatite and alumina, in NaCl O.02 M in function of the ph is studied. The likeness of retention of the solids for the fission products is expressed in terms of the distribution coefficient kD which is obtained using the homologous radionuclides {sup 109}Pd and {sup 87m}Sr as well as the {sup 99m}Tc. The retention of Pd was of 100% and the Tc near to 0%. (Author)

  3. Nuclear reactions with 11C and 14O radioactive ion beams

    International Nuclear Information System (INIS)

    Guo, Fanqing

    2004-01-01

    Radioactive ion beams (RIBs) have been shown to be a useful tool for studying proton-rich nuclides near and beyond the proton dripline and for evaluating nuclear models. To take full advantage of RIBs, Elastic Resonance Scattering in Inverse Kinematics with Thick Targets (ERSIKTT), has proven to be a reliable experimental tool for investigations of proton unbound nuclei. Following several years of effort, Berkeley Experiments with Accelerated Radioactive Species (BEARS), a RIBs capability, has been developed at the Lawrence Berkeley National Laboratory's 88-Inch Cyclotron. The current BEARS provides two RIBs: a 11C beam of up to 2x108 pps intensity on target and an 14O beam of up to 3x104 pps intensity. While the development of the 11C beam has been relatively easy, a number of challenges had to be overcome to obtain the 14O beam. The excellent 11C beam has been used to investigate several reactions. The first was the 197Au(11C,xn)208-xnAt reaction, which was used to measure excitation functions for the 4n to 8n exit channels. The measured cross sections were generally predicted quite well using the fusion-evaporation code HIVAP. Possible errors in the branching ratios of ?? decays from At isotopes as well as the presence of incomplete fusion reactions probably contribute to specific overpredictions. 15F has been investigated by the p(14O,p)14O reaction with the ERSIKTT technology. Several 14O+p runs have been performed. Excellent energy calibration was obtained using resonances from the p(14N,p)14N reaction in inverse kinematics, and comparing the results to those obtained earlier with normal kinematics. The differences between 14N+p and 14O+p in the stopping power function have been evaluated for better energy calibration. After careful calibration, the energy levels of 15F were fitted with an R-matrix calculation. Spins and parities were assigned to the two observed resonances. This new measurement of the 15F ground state supports the disappearance of the Z = 8

  4. A TRANSPORTATION RISK ASSESSMENT TOOL FOR ANALYZING THE TRANSPORT OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE TO THE PROPOSED YUCCA MOUNTAIN REPOSITORY

    International Nuclear Information System (INIS)

    2001-01-01

    The Yucca Mountain Draft Environmental Impact Statement (DEIS) analysis addressed the potential for transporting spent nuclear fuel and high-level radioactive waste from 77 origins for 34 types of spent fuel and high-level radioactive waste, 49,914 legal weight truck shipments, and 10,911 rail shipments. The analysis evaluated transportation over 59,250 unique shipment links for travel outside Nevada (shipment segments in urban, suburban or rural zones by state), and 22,611 links in Nevada. In addition, the analysis modeled the behavior of 41 isotopes, 1091 source terms, and used 8850 food transfer factors (distinct factors by isotope for each state). The analysis also used mode-specific accident rates for legal weight truck, rail, and heavy haul truck by state, and barge by waterway. This complex mix of data and information required an innovative approach to assess the transportation impacts. The approach employed a Microsoft(reg s ign) Access database tool that incorporated data from many sources, including unit risk factors calculated using the RADTRAN IV transportation risk assessment computer program. Using Microsoft(reg s ign) Access, the analysts organized data (such as state-specific accident and fatality rates) into tables and developed queries to obtain the overall transportation impacts. Queries are instructions to the database describing how to use data contained in the database tables. While a query might be applied to thousands of table entries, there is only one sequence of queries that is used to calculate a particular transportation impact. For example, the incident-free dose to off-link populations in a state is calculated by a query that uses route segment lengths for each route in a state that could be used by shipments, populations for each segment, number of shipments on each segment, and an incident-free unit risk factor calculated using RADTRAN IV. In addition to providing a method for using large volumes of data in the calculations, the

  5. The evolution and future direction of the Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management

    International Nuclear Information System (INIS)

    Siraky, Gabriela

    2008-01-01

    Full text: The Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management, namely the Joint Convention, had been established in 1997. The objective of the Convention is to achieve and maintain a high level of safety worldwide in that fields to ensure that there are effective defenses against potential hazards so that individuals, society and the environment are protected from the harmful effects of ionizing radiation and to prevent accidents with radiological consequences and to mitigate their consequences. The Parties to the Convention intend to achieve this objective by international cooperation, peer reviews of each other's performance, assistance when needed for states with less developed programmes and capabilities and the use of internationally accepted standards of safety. The Joint Convention is rooted in the discussions held previously to the establishment of the Convention of the Nuclear Safety that had taken place in the period 1993-1994. The idea of a Convention on the Safety of waste management evolved since then, getting its final status in November 1997 after seven meetings of a specially appointed working group of outstanding specialists on the subject. After that, it was needed three years more until the number of the ratifying States, that then became Contracting Parties, to get the condition that let the Convention to enter into force: ninety days after the date of deposit with the Depository of the twenty-fifth instrument of ratification, acceptance or approval, including the instruments of fifteen States having an operational nuclear power plant. The First Review Meeting was held three years after it entered into force, in November 2003, with the attendance of 33 Parties that presented their National Reports. The second review meeting was held in May 2006. Forty-one Contracting Parties participated in the Second Review Meeting. There is then, a constant evolution in the number of Contracting Parties

  6. Report on the ANSTO application for a licence to construct a Replacement Research Reactor, addressing seismic analysis and seismic design accident analysis, spent fuel and radioactive wastes

    International Nuclear Information System (INIS)

    2002-02-01

    The Report of the Nuclear Safety Committee (NSC) covers specific terms of reference as requested by the Chief Executive Officer of ARPANSA. The primary issue for the Working Group(WG) consideration was whether ANSTO had demonstrated: (i) that the overall approach to seismic analysis and its implementation in the design is both conservative and consistent with the international best practice; (ii) whether the full accident analysis in the Probabilistic Safety Assesment Report (PSAR) satisfies the radiation dose/frequency criteria specified in ARPANSA's regulatory assessment principle 28 and the assumptions used in the Reference Accident for the siting assessment have been accounted for in the PSAR; and (iii) the adequacy of the strategies for managing the spent fuel as proposed to be used in the Replacement Research Reactor and other radioactive waste (including emissions, taking into account the ALARA criterion) arising from the operation of the proposed replacement reactor and radioisotope production. The report includes a series of questions that were asked of the Applicant in the course of working group deliberations, to illustrate the breadth of inquiries that were made. The Committee noted that replies to some questions remain outstanding at the date of this document. The NSC makes a number of recommendations that appear in each section of the document, which has been compiled in three parts representing the work of each group. The NSC notes some lack of clarity in what was needed to be considered at this approval stage of the project, as against information that would be required at a later stage. While not in the original work plan, recent events of September 11, 2001 also necessitated some exploration of issues relating to construction security. Copyright (2002) Commonwealth of Australia

  7. Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management. Third review meeting. Questions asked to France and answers

    International Nuclear Information System (INIS)

    2009-01-01

    The Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management, referred to as the 'Joint Convention', is the result of international discussions that followed the adoption of the Convention on Nuclear Safety, in 1994. France signed the Joint Convention at the General Conference of the International Atomic Energy Agency (IAEA) held on 29 September 1997, the very first day the Joint Convention was opened for signature. She approved it on 22 February 2000 and filed the corresponding instruments with the IAEA on 27 April 2000. The Joint Convention entered into force on 18 June 2001. For many years, France has been taking an active part in the pursuit of international actions to reinforce nuclear safety and considers the Joint Convention to be a key step in that direction. The fields covered by the Joint Convention have long been part of the French approach to nuclear safety. For his third report, France presented a document reflecting the viewpoints of the various stakeholders (regulatory authorities and operators). Thus, for each of the chapters in which the regulatory authority is not the only party to express its point of view, a three-stage structure was adopted: first of all a description by the regulatory authority of the regulations, followed by a presentation by the operators of the steps taken to meet the regulations and finally, an analysis by the regulatory authority of the steps taken by the operators. This third report was distributed in October 2008 to all Contracting Parties who asked 213 questions on the French report. France answered each of them in the present document

  8. Remote automatic plasma arc-closure welding of a dry-storage canister for spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Sprecace, R.P.; Blankenship, W.P.

    1982-01-01

    A carbon steel storage canister has been designed for the dry encapsulation of spent nuclear fuel assemblies or of logs of vitrified high level radioactive waste. The canister design is in conformance with the requirements of the ASME Code, Section III, Division 1 for a Class 3 vessel. The canisters will be loaded and sealed as part of a completely remote process sequence to be performed in the hot bay of an experimental encapsulation facility at the Nevada Test Site. The final closure to be made is a full penetration butt weld between the canister body, a 12.75-in O.D. x 0.25-in wall pipe, and a mating semiellipsoidal closure lid. Due to a combination of design, application and facility constraints, the closure weld must be made in the 2G position (canister vertical). The plasma arc welding system is described, and the final welding procedure is described and discussed in detail. Several aspects and results of the procedure development activity, which are of both specific and general interest, are highlighted; these include: The critical welding torch features which must be exactly controlled to permit reproducible energy input to, and gas stream interaction with, the weld puddle. A comparison of results using automatic arc voltage control with those obtained using a mechanically fixed initial arc gap. The optimization of a keyhole initiation procedure. A comparison of results using an autogenous keyhole closure procedure with those obtained using a filler metal addition. The sensitivity of the welding process and procedure to variations in joint configuration and dimensions and to variations in base metal chemistry. Finally, the advantages and disadvantages of the plasma arc process for this application are summarized from the current viewpoint, and the applicability of this process to other similar applications is briefly indicated

  9. A Study on the Establishment of Radiation Dose Estimation Procedure for Accumulated Radioactive Ions for RAON ISOL System

    Directory of Open Access Journals (Sweden)

    KIM Do Hyun

    2017-01-01

    Full Text Available For purposes of various experiments, RAON heavy ion accelerator facility has been designed in Korea. ISOL is one system of RAON accelerators to generate and separate rare isotopes. Radioactive ions generated from target-proton reactions are separated and accumulated at separation devices. The accumulated isotopes release the gamma radiations; therefore, the radiation safety must be clearly estimated. In this study, a process to evaluate radiations from the accumulated ions was proposed by modifying FISPACT code. The proposed process was validated by comparing a solution of single element decay problem. Using the process, a preliminary study for radiation doses were performed in a virtual separation devise.

  10. A novel experimental scheme of electron scattering off unstable nuclei with a self-confining radioactive ion target (SCRIT)

    International Nuclear Information System (INIS)

    Wakasugi, Masanori

    2005-01-01

    We proposed a new experimental scheme of an electron scattering off unstable nuclei using a Self-Confining Radioactive Ion Target (SCRIT). The SCRIT is an unstable ion target formed in the electron storage ring, and is based on completely new idea. We constructed prototype of the SCRIT device and installed it in the electron storage ring KSR in Kyoto University. In the test experiment, 10 7 -Cs ions are confined in the SCRIT with the lifetime of about 2 s and the feasibility of the SCRIT as the target has been confirmed. (author)

  11. Study of chemically synthesized ZnO nano particles under a bio template using radioactive ion beam

    CERN Multimedia

    This is a project proposal to study nano sized semiconductor ZnO system, useful in biology and medicinal purposes, using radioactive ion beam from ISOLDE. Doping of the nano particles with Cu, Cd and Ga ions (in their variable valancy states) are expected to impart changes in the electrical structure and properties in the said system under study. The morphological changes, chemical environment, micro structure, electrical and optical properties of the nano size particles of ZnO system (developed under a bio template of folic acid) after the interaction with radioactive ion beam will be studied. The provision of perturbed angular correlation (PAC) study with respect to the changes in chemical environment, where ever possible will be attempted.

  12. The Holifield Radioactive Ion Beam Facility at the Oak Ridge National Laboratory: Present status and future plans

    International Nuclear Information System (INIS)

    Alton, G.D.; Beene, J.R.

    1998-01-01

    The Holifield Radioactive Ion Beam Facility (HRIBF) is a first generation national user facility for nuclear physics and nuclear astrophysics research with radioactive ion beams (RIBs). The reconfiguration, construction, and equipment commissioning phases have been completed and the beam development program is in progress. In this article, descriptions of the facility and newly implemented experimental equipment for use in the nuclear and astrophysics programs will be given and an outline of the initial experimental program will be presented. Special target ion source related problems, endemic to the production of specific short lived RIBs will be discussed. In addition, plans, which involve either a 200 MeV or a 1 GeV proton linac driver for a second generation ISOL facility, will be presented

  13. Effect of doping rare earths on magnetostriction characteristics of CoFe2O4 prepared from spent Li-ion batteries

    Science.gov (United States)

    Xi, Guoxi; Zhao, Tingting; Wang, Lu; Dun, Changwei; Zhang, Ye

    2018-04-01

    Recovering spent Li-ion batteries is beneficial to the economy and environment. Therefore, this study synthesized nanoparticles of cobalt ferrite doped with different rare earth ions (Nd, Ce, and Pr) by a sol-gel auto-combustion method using spent Li-ion batteries. The effect of the different doping elements on grain sizes, structure, magnetic and magnetostrictive properties, and strain derivative were confirmed by X-ray diffraction, scanning election microscopy, vibrating sample magnetometer, and a magnetostrictive coefficient measuring system. Substitution of a small amount of Fe3+ with RE3+ in CoRExFe2-xO4 (x = 0.025, 0.05, and 0.1) had a large effect on magnetostrictive properties and strain derivative, which was improved compared with pure cobalt ferrite at low magnetic field. The maximum strain derivative (dλ/dH = -1.49 × 10-9 A-1 m at 18 kA m-1) was obtained for Nd, x = 0.05. Changes in the magnetostriction coefficients and strain derivatives were correlated with changes in cation distribution, microstructure, and magnetic anisotropy, which depended strongly on RE3+ substitution and distribution in the spinel structure.

  14. Collinear laser spectroscopy on radioactive praseodymium ions and cadmium ions; Kollineare Laserspektroskopie an radioaktiven Praseodymionen und Cadmiumatomen

    Energy Technology Data Exchange (ETDEWEB)

    Froemmgen, Nadja

    2013-11-21

    Collinear laser spectroscopy is a tool for the model independent determination of spins, charge radii and electromagnetic moments of nuclei in ground and long-lived isomeric states. In the context of this thesis a new offline ion source for high evaporating temperatures and an ion beam analysis system were implemented at the TRIGA-LASER Experiment at the Institute for Nuclear Chemistry at the University of Mainz. The main part of the thesis deals with the determination of the properties of radioactive praseodymium and cadmium isotopes by collinear laser spectroscopy at ISOLDE/CERN. The necessary test measurements for the spectroscopy of praseodymium ions have been conducted with the aforementioned offline ion source at the TRIGA-LASER experiment. The spectroscopy of the praseodymium ions was motivated by the observation of a modulation of the electron capture decay rates of hydrogen-like {sup 140}Pr{sup 58+}. The nuclear magnetic moment of the nucleus is, among others, required for the explanation of the so-called GSI Oscillations and has not been studied experimentally before. Additionally, the determined electron capture decay constant of hydrogen-like {sup 140}Pr{sup 58+} is lower than the one of helium-like {sup 140}Pr{sup 57+}. The explanation of this phenomenon requires a positive magnetic moment. During the experiment at the COLLAPS apparatus the magnetic moments of the neutron-deficient isotopes {sup 135}Pr, {sup 136}Pr and {sup 137}Pr could be determined for the first time. Unfortunately, due to a too low production yield the desired isotope {sup 140}Pr could not be studied.The systematic study of cadmium isotopes was motivated by nuclear physics in the tin region. With Z=48 two protons are missing for the shell closure and the isotopes extend from the magic neutron number N=50 to the magic neutron number N=82. The extracted nuclear properties allow tests of different nuclear models in this region. In this thesis the obtained results of the spectroscopy of

  15. Collective processing device for spent fuel

    International Nuclear Information System (INIS)

    Irie, Hiroaki; Taniguchi, Noboru.

    1996-01-01

    The device of the present invention comprises a sealing vessel, a transporting device for transporting spent fuels to the sealing vessel, a laser beam cutting device for cutting the transported spent fuels, a dissolving device for dissolving the cut spent fuels, and a recovering device for recovering radioactive materials from the spent fuels during processing. Reprocessing treatments comprising each processing of dismantling, shearing and dissolving are conducted in the sealing vessel can ensure a sealing barrier for the radioactive materials (fissionable products and heavy nuclides). Then, since spent fuels can be processed in a state of assemblies, and the spent fuels are easily placed in the sealing vessel, operation efficiency is improved, as well as operation cost is saved. Further, since the spent fuels can be cut by a remote laser beam operation, there can be prevented operator's exposure due to radioactive materials released from the spent fuels during cutting operation. (T.M.)

  16. A complete process for the treatment of low level radioactive wastes using composite ion exchange resins based on polyurethane foam

    International Nuclear Information System (INIS)

    Rao, S.V.S.; Mani, A.G.S.; Lakshmi, R.; Sinha, P.K.

    2007-01-01

    The CFC-PU foam and HMO-PU foams have been employed for the removal of radioactive cesium and strontium respectively from actual Cat. III radioactive liquid waste received from reprocessing plant. Batch studies have been carried out in order to optimize the loading of above chemicals. The results of column and batch studies have indicated that there is a good agreement between them. After passing about 1000 bed volumes, the average DF and volume reduction factor obtained were 20 and 200 respectively. A method has been developed for the wet digestion of spent resins in alkaline KMnO 4 medium. The digested foam has been immobilized in cement matrix and the matrices were characterized with respect to compressive strength and leach resistance. (author)

  17. The determination of radioactive strontium in environmental samples by preferential ion exchange separation from calcium

    International Nuclear Information System (INIS)

    Fukushima, Hiroto; Nonaka, Nobuhiro; Honda, Tetsutaro; Kimura, Toshimasa; Higuchi, Hideo

    1979-01-01

    The determination of radioactive strontium in environmental samples involves a difficulty in the separation and purification of strontium from large amounts of calcium. The principal technique employed in this study is based on a cation exchange separation after preferential chelation of calcium over strontium by EDTA. The alkaline earth fraction separated from samples is dissolved in 1.2 M HCl and an equivalent amount of EDTA is added. The solution is passed through a Dowex 50W-X8 ion exchange column at a flow rate of 20 ml/min after the pH has been adjusted to 5.1. A fraction of calcium retained on the resin is eluted out with 1.8 l of 2 w/v% EDTA solution at pH 5.1. Then the strontium is eluted with 450 ml of 2 M HCl. The strontium carbonate is precipitated, and the chemical yield is determined. After standing for 14 days to reach the radioactive equilibrium of 90 Sr - 90 Y, the activity of 89 Sr + 90 Sr + 90 Y is measured. Then the activity of 90 Sr + 90 Y is measured by milking the 90 Y. The proposed procedure has been applied to sea water, soil and various ash samples of vegetables, seaweeds and daily foods. The results obtained are in good agreement with those by a classical fuming nitric acid precipitation method. The detection limits are 0.1 pCi/l for 10 l of sea water, 10 pCi/kg for 100 g of soil and 1 pCi/kg-fresh for 20 g ash samples, respectively. (author)

  18. Transport of radioactive ion beams and related safety issues: The {sup 132}Sn{sup +} case study

    Energy Technology Data Exchange (ETDEWEB)

    Osswald, F., E-mail: francis.osswald@iphc.cnrs.fr; Bouquerel, E.; Boutin, D.; Dinkov, A.; Sellam, A. [IPHC/IN2P3/CNRS, University of Strasbourg, 67037 Strasbourg (France); Kazarinov, N. [JINR/FLNR, 141980 Dubna (Russian Federation); Perrot, L. [IPNO/IN2P3/CNRS, University of Paris-Sud-11, 91406 Orsay (France)

    2014-12-15

    The transport of intense radioactive ion beam currents requires a careful design in order to limit the beam losses, the contamination and thus the dose rates. Some investigations based on numerical models and calculations have been performed in the framework of the SPIRAL 2 project to evaluate the performance of a low energy beam transport line located between the isotope separation on line (ISOL) production cell and the experiment areas. The paper presents the results of the transverse phase-space analysis, the beam losses assessment, the resulting contamination, and radioactivity levels. They show that reasonable beam transmission, emittance growth, and dose rates can be achieved considering the current standards.

  19. Process for recycling mixed-cathode materials from spent lithium-ion batteries and kinetics of leaching.

    Science.gov (United States)

    Li, Li; Bian, Yifan; Zhang, Xiaoxiao; Guan, Yibiao; Fan, Ersha; Wu, Feng; Chen, Renjie

    2018-01-01

    A "grave-to-cradle" process for the recycling of spent mixed-cathode materials (LiCoO 2 , LiCo 1/3 Ni 1/3 Mn 1/3 O 2 , and LiMn 2 O 4 ) has been proposed. The process comprises an acid leaching followed by the resynthesis of a cathode material from the resulting leachate. Spent cathode materials were leached in citric acid (C 6 H 8 O 7 ) and hydrogen peroxide (H 2 O 2 ). Optimal leaching conditions were obtained at a leaching temperature of 90 °C, a H 2 O 2 concentration of 1.5 vol%, a leaching time of 60 min, a pulp density of 20 g L -1 , and a citric acid concentration of 0.5 M. The leaching efficiencies of Li, Co, Ni, and Mn exceeded 95%. The leachate was used to resynthesize new LiCo 1/3 Ni 1/3 Mn 1/3 O 2 material by using a sol-gel method. A comparison of the electrochemical properties of the resynthesized material (NCM-spent) with that synthesized directly from original chemicals (NCM-syn) indicated that the initial discharge capacity of NCM-spent at 0.2 C was 152.8 mA h g -1 , which was higher than the 149.8 mA h g -1 of NCM-syn. After 160 cycles, the discharge capacities of the NCM-spent and NCM-syn were 140.7 mA h g -1 and 121.2 mA h g -1 , respectively. After discharge at 1 C for 300 cycles, the NCM-spent material remained a higher capacity of 113.2 mA h g -1 than the NCM-syn (78.4 mA h g -1 ). The better performance of the NCM-spent resulted from trace Al doping. A new formulation based on the shrinking-core model was proposed to explain the kinetics of the leaching process. The activation energies of the Li, Co, Ni, and Mn leaching were calculated to be 66.86, 86.57, 49.46, and 45.23 kJ mol -1 , respectively, which indicates that the leaching was a chemical reaction-controlled process. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Process controls for improving bioleaching performance of both Li and Co from spent lithium ion batteries at high pulp density and its thermodynamics and kinetics exploration.

    Science.gov (United States)

    Niu, Zhirui; Zou, Yikan; Xin, Baoping; Chen, Shi; Liu, Changhao; Li, Yuping

    2014-08-01

    Release of Co and Li from spent lithium ion batteries (LIBs) by bioleaching has attracted growing attentions. However, the pulp density was only 1% or lower, meaning that a huge quantity of media was required for bioleaching. In this work, bioleaching behavior of the spent LIBs at pulp densities ranging from 1% to 4% was investigated and process controls to improve bioleaching performance at pulp density of 2% were explored. The results showed that the pulp density exerted a considerable influence on leaching performance of Co and Li. The bioleaching efficiency decreased respectively from 52% to 10% for Co and from 80% to 37% for Li when pulp density rose from 1% to 4%. However, the maximum extraction efficiency of 89% for Li and 72% for Co was obtained at pulp density of 2% by process controls. Bioleaching of the spent LIBs has much greater potential to occur than traditional chemical leaching based on thermodynamics analysis. The product layer diffusion model described best bioleaching behavior of Co and Li. Copyright © 2014 Elsevier Ltd. All rights reserved.