WorldWideScience

Sample records for uranium treatment process

  1. Process for sewage biological treatment from uranium

    International Nuclear Information System (INIS)

    Popa, Karin; Cecal, Alexandru; Craciun, Iftimie Ionel; Rudic, Valeriu; Gulea, Aurelian; Cepoi, Liliana

    2004-01-01

    The invention relates to the sewage treatment, in particular to the sewage biological treatment from radioactive waste, namely from uranium. The process for sewage biological treatment from uranium includes cultivation in the sewage of the aquatic plants Lemna minor and Spirulina platensis. The plant cultivation is carried out in two stages. In the first stage for cultivation is used Lemna minor and in the second stage - Spirulina platensis. After finishing the plant cultivation it is carried out separation of their biomass. The result of the invention consists in increasing the uranyl ions accumulation by the biomass of plants cultivated in the sewage.

  2. Process for sewage biological treatment from uranium

    International Nuclear Information System (INIS)

    Popa, K.; Cecal, A.; Craciun, I.

    2004-01-01

    The invention relates to the sewage treatment, in particular to the sewage biological treatmen from radioactive waste, namely from uranium. The process dor sewage biological treatment from uranium includes cultivation in the sewage of the aquatic plants Lemna minor and Spirulina platensis. The plants cultivation is carried out in two stages. In the first stage for cultivation is used Lemna minor in the second stage - Spirulina platensis . After finishing the plant cultivation it is carried out separation of their biomass. The result of the invention consists in increasing the uranyl ions by the biomass of plants cultivated in the sewage

  3. Treatment of uranium turning with the controllable oxidizing process

    International Nuclear Information System (INIS)

    Shen Bingyi; Zhang Yonggang; Zhen Huikuan

    1989-02-01

    The concept, procedure and safety measures of the controllable oxidizing for uranium turning is described. The feasibility study on technological process has been made. The process provided several advantages such as: simplicity of operation, no pollution environment, safety, high efficiency and low energy consumption. The process can yield nuclear pure uranium dioxide under making no use of a great number of chemical reagent. It may supply raw material for fluoration and provide a simply method of treatment for safe store of uranium turning

  4. Some aspects of the processing development for uranium ores treatment

    International Nuclear Information System (INIS)

    Bruno, J.B.

    1982-01-01

    It is discussed the methodology adopted by NUCLEBRAS to the processing development for uranium ores treatment. The used methodology has the following steps: exploratories studies, preliminaries stiudies and optimization studies. The studies include physical and chemical contained in the solution. As examples are cited the uranium ores treatment in Lagoa Real and Itataia. (A.B.) [pt

  5. Treatment of uranium-containing effluent in the process of metallic uranium parts

    International Nuclear Information System (INIS)

    Yuan Guoqi

    1993-01-01

    The anion exchange method used in treatment of uranium-containing effluent in the process of metallic parts is the subject of the paper. The results of the experiments shows that the uranium concentration in created water remains is less than 10 μg/l when the waste water flowed through 10000 column volume. A small facility with column volume 150 litre was installed and 1500 m 3 of waste water can be cleaned per year. (1 tab.)

  6. The progress in the researches for uranium mill tailings cleaning treatment and no-waste uranium ore milling processes

    International Nuclear Information System (INIS)

    Wang Jintang

    1990-01-01

    The production of uranium mill tailings and their risk assessment are described. The moethods of uranium mill tailings disposal and management are criticized and the necessity of the researches for uranium mill tailings cleaning treatment and no-wasle uranium ore milling process are demonstrated. The progress for these researches in China and other countries with uranium production is reviewed, and the corresponding conclusions are reported

  7. Process water treatment at the Ranger uranium mine, Northern Australia.

    Science.gov (United States)

    Topp, H; Russell, H; Davidson, J; Jones, D; Levy, V; Gilderdale, M; Davis, S; Ring, R; Conway, G; Macintosh, P; Sertorio, L

    2003-01-01

    The conceptual development and piloting of an innovative water treatment system for process water produced by a uranium mine mill is described. The process incorporates lime/CO2 softening (Stage 1), reverse osmosis (Stage 2) and biopolishing (Stage 3) to produce water of quality suitable for release to the receiving environment. Comprehensive performance data are presented for each stage. The unique features of the proposed process are: recycling of the lime/CO2 softening sludge to the uranium mill as a neutralant, the use of power station off-gas for carbonation, the use of residual ammonia as the pH buffer in carbonation; and the recovery and recycling of ammonia from the RO reject stream.

  8. Study on the chemical treatment processes of the uranium pyrochlore of Araxa

    International Nuclear Information System (INIS)

    Batista, H.F.; Fernandes, M.D.

    Several processes are presented for the chemical treatment, in laboratory scale, of the uranium pyrochlore concentrates found in Araxa (Minas Gerais, Brazil), aiming to the extraction of uranium, thorium and rare earths, besides the recovery of niobium pentoxide [pt

  9. Improvement for waste water treatment process of a uranium deposite and its effect

    International Nuclear Information System (INIS)

    Huang Jimao

    2013-01-01

    Uranium was recovered from alkaline uranium ores by heap leaching and traditional agitation leaching methods at a uranium mine, and the waste water (including waste water produced in hydrometallurgy process and mine drainage) was treated by using chemical precipitation method and chemical precipitation loading method. It was found that the removal rate of uranium by the waste water treatment process was not satisfactory after one year's run. So, the waste water treatment process was improved. After the improvement, removal rate of CO 3 2- ,HCO 3 - , U and Ra was enhanced and the treated waste water reached the standard of discharge. (author)

  10. Evaluation of neutralization treatment processes and their use for uranium tailings solutions

    International Nuclear Information System (INIS)

    Sherwood, D.R.; Opitz, B.E.; Serne, R.J.

    1985-01-01

    The potential for groundwater contamination from the typically acidic mill wastes that are disposed of in tailings impoundments is of primary concern at uranium mill sites in the US. Solution-treatment processes provide a system for limiting the environmental impact from acidic seepage. Treatment of uranium tailings solutions from evaporation ponds, underdrains, and surface seeps could aid in decommissioning active sites or be used as an emergency measure to avert possible uncontrolled discharges. At present, neutralization processes appear to be best suited for treating uranium mill tailings solution because they can, at a reasonable cost, limit the solution concentration of many contaminants and thus reduce the potential for groundwater contamination. However, the effectiveness of the neutralization process depends on the reagent used as well as the chemistry of the waste stream. This article provides a description of neutralization processes, an assessment of their performance on acidic uranium tailings leachates, and recommendations for their use at US uranium mill sites

  11. Uranium processing and properties

    CERN Document Server

    2013-01-01

    Covers a broad spectrum of topics and applications that deal with uranium processing and the properties of uranium Offers extensive coverage of both new and established practices for dealing with uranium supplies in nuclear engineering Promotes the documentation of the state-of-the-art processing techniques utilized for uranium and other specialty metals

  12. The Development of Treatment Process Technology for Uranium Soil washing Leachate

    Energy Technology Data Exchange (ETDEWEB)

    Shon, Dong Bin; Kim, Gye Nam; Park, Hye Min; Kim, Ki Hong; Lee, Ki Won; Moon, Jeik won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    Electrokinetic treatment technology is a good method for removing radioactive substances such as U, Co, Cs: but it has a weakness. It takes a long time to get high removal efficiency. The Soil washing method compensates for this weak point with its short reaction time and with this method it is possible to remove a lot of uranium-contaminated soil. But a great deal of leachate is generated. That is, about more amounts of leachate are generated for the decontamination of the same volume of radioactive soil using the electrokinetic equipment. Therefore, the development of a treatment process for The Soil washing leachate is important so that there is a reduction of leachate waste volume and a choice of process. Previously, studies for liquid radioactive waste were in process at various nuclear facilities. Nuclear fuel plant survey appropriate cohesion quantity of liquid waste of radioactive. Nuclear power plants manage liquid radioactive waste with centrifugation equipment. In this study, the treatment technology for uranium Soil washing leachate generated on Soil washing decontamination for the soil contaminated with uranium was developed. A treatment process suitable to the contamination characteristics of Soil washing leachate was proposed

  13. PROCESS OF RECOVERING URANIUM

    Science.gov (United States)

    Carter, J.M.; Larson, C.E.

    1958-10-01

    A process is presented for recovering uranium values from calutron deposits. The process consists in treating such deposits to produce an oxidlzed acidic solution containing uranium together with the following imparities: Cu, Fe, Cr, Ni, Mn, Zn. The uranium is recovered from such an impurity-bearing solution by adjusting the pH of the solution to the range 1.5 to 3.0 and then treating the solution with hydrogen peroxide. This results in the precipitation of uranium peroxide which is substantially free of the metal impurities in the solution. The peroxide precipitate is then separated from the solution, washed, and calcined to produce uranium trioxide.

  14. Treatment of tailings water from uranium ore processing by reverse osmosis

    International Nuclear Information System (INIS)

    Georgescu, D.P.; Andrei, L.

    2000-01-01

    Mining and metallurgical waste waters are considered to be the major sources of heavy metal contamination. The need of economic and effective methods for metals removal have resulted in the development of new separation technologies. Precipitation, ion exchange, electrochemical processes, filtration and flotation are commonly applied for industrial effluents treatment. Occasionally, the application of such processes is limited because of technical or economical constraints. The search for new technologies regarding the recovery and removal of toxic metals from waste waters has directed attention to membrane processes. These processes are developed in the recent years due to the availability of many new types of membranes. This paper presents the laboratory test results for liquid radioactive effluent treatment from alkaline uranium ore processing by reverse osmosis. (author)

  15. Uranium 2000 : International symposium on the process metallurgy of uranium

    International Nuclear Information System (INIS)

    Ozberk, E.; Oliver, A.J.

    2000-01-01

    The International Symposium on the Process Metallurgy of Uranium has been organized as the thirtieth annual meeting of the Hydrometallurgy Section of the Metallurgical Society of the Canadian Institute of Mining, Metallurgy and Petroleum (CIM). This meeting is jointly organized with the Canadian Mineral Processors Division of CIM. The proceedings are a collection of papers from fifteen countries covering the latest research, development, industrial practices and regulatory issues in uranium processing, providing a concise description of the state of this industry. Topics include: uranium industry overview; current milling operations; in-situ uranium mines and processing plants; uranium recovery and further processing; uranium leaching; uranium operations effluent water treatment; tailings disposal, water treatment and decommissioning; mine decommissioning; and international regulations and decommissioning. (author)

  16. Deactivation and treatment of mine wastewaters and of aqueous solutions discharged in uranium ore treatment and processing

    International Nuclear Information System (INIS)

    Jilek, R.; Prochazka, H.; Fuska, J.; Nemec, P.; Katzer, J.

    1974-01-01

    A description is presented of decontamination and purification of mine wastewaters and water solutions discharged from uranium ore treatment and processing and of the simultaneous removal and concentration of uranium-radium daughters, mainly of 226 Ra and 210 Pb. These elements are incorporated in the mycelium of microorganisms, such as those of the Fungi imperfecti class or are sorbed on the mycelium surface. The mycelia are then mechanically separated from the decontaminated solution, e.g., by filtration, centrifugation or sedimentation. The mycelium may be cultivated in the purified solutions to which nutrients are added, such as carbon, nitrogen and phosphorus in concentrations necessary for the growth of the microorganisms used. The mycelium may be added to the purified solution either in the native or in the dried state. (B.S.)

  17. DUPoly process for treatment of depleted uranium and production of beneficial end products

    International Nuclear Information System (INIS)

    Kalb, P.D.; Adams, J.W.; Lageraaen, P.R.; Cooley, C.R.

    2000-01-01

    The present invention provides a process of encapsulating depleted uranium by forming a homogeneous mixture of depleted uranium and molten virgin or recycled thermoplastic polymer into desired shapes. Separate streams of depleted uranium and virgin or recycled thermoplastic polymer are simultaneously subjected to heating and mixing conditions. The heating and mixing conditions are provided by a thermokinetic mixer, continuous mixer or an extruder and preferably by a thermokinetic mixer or continuous mixer followed by an extruder. The resulting DUPoly shapes can be molded into radiation shielding material or can be used as counter weights for use in airplanes, helicopters, ships, missiles, armor or projectiles

  18. Uranium enrichment. Enrichment processes

    International Nuclear Information System (INIS)

    Alexandre, M.; Quaegebeur, J.P.

    2009-01-01

    Despite the remarkable progresses made in the diversity and the efficiency of the different uranium enrichment processes, only two industrial processes remain today which satisfy all of enriched uranium needs: the gaseous diffusion and the centrifugation. This article describes both processes and some others still at the demonstration or at the laboratory stage of development: 1 - general considerations; 2 - gaseous diffusion: physical principles, implementation, utilisation in the world; 3 - centrifugation: principles, elementary separation factor, flows inside a centrifuge, modeling of separation efficiencies, mechanical design, types of industrial centrifuges, realisation of cascades, main characteristics of the centrifugation process; 4 - aerodynamic processes: vortex process, nozzle process; 5 - chemical exchange separation processes: Japanese ASAHI process, French CHEMEX process; 6 - laser-based processes: SILVA process, SILMO process; 7 - electromagnetic and ionic processes: mass spectrometer and calutron, ion cyclotron resonance, rotating plasmas; 8 - thermal diffusion; 9 - conclusion. (J.S.)

  19. URANIUM LEACHING AND RECOVERY PROCESS

    Science.gov (United States)

    McClaine, L.A.

    1959-08-18

    A process is described for recovering uranium from carbonate leach solutions by precipitating uranium as a mixed oxidation state compound. Uranium is recovered by adding a quadrivalent uranium carbon;te solution to the carbonate solution, adjusting the pH to 13 or greater, and precipitating the uranium as a filterable mixed oxidation state compound. In the event vanadium occurs with the uranium, the vanadium is unaffected by the uranium precipitation step and remains in the carbonate solution. The uranium-free solution is electrolyzed in the cathode compartment of a mercury cathode diaphragm cell to reduce and precipitate the vanadium.

  20. Study of the Treatment of the Liquid Radioactive Waste Nong Son Uranium Ore Processing

    International Nuclear Information System (INIS)

    Nguyen Ba Tien; Trinh Giang Huong; Luu Cao Nguyen; Harvey, L.K.; Tran Van Quy

    2011-01-01

    Liquid waste from Nong Son uranium ore processing is treated with concentrated acid, agglomerated, leached, run through ion exchange and then treated with H 2 O 2 to precipitate yellowcake. The liquid radioactive waste has a pH of 1.86 and a high content of radioactive elements, such as: [U] 143.898 ppm and [Th] = 7.967 ppm. In addition, this waste contains many polluted chemical elements with high content, such as arsenic, mercury, aluminum, iron, zinc, magnesium, manganese and nickel. The application of the general method as one stage precipitation or precipitation in coordination with BaCl 2 is not effective. These methods generated a large amount of sludge with poor settling characteristics. The volume of final treated waste was large. This paper introduces the investigation of the treatment of this liquid radioactive waste by the method of two stage of precipitation in association with polyaluminicloride (PAC) and polymer. The impact of factors: pH, neutralizing agents, quantity of PAC and polymer to effect precipitation and improve the settling characteristics during processing was studied. The results showed that the processing of liquid radioactive waste treatment through two stages: first stage at pH = 3 and the second stage at pH = 8.0 with limited PAC and polymer (A 101) resulted in significant reduced volume of the treated waste. The discharged liquid satisfied the requirement of the National Technical Regulation on Industrial Waste Water (QCVN 24:2009). (author)

  1. Uranium ore processing

    International Nuclear Information System (INIS)

    Ritcey, G.M.; Haque, K.E.; Lucas, B.H.; Skeaff, J.M.

    1983-01-01

    The authors have developed a complete method of recovering separately uranium, thorium and radium from impure solids such as ores, concentrates, calcines or tailings containing these metals. The technique involves leaching, in at least one stage. The impure solids in finely divided form with an aqueous leachant containing HCl and/or Cl 2 until acceptable amounts of uranium, thorium and radium are dissolved. Uranium is recovered from the solution by solvent extraction and precipitation. Thorium may also be recovered in the same manner. Radium may be recovered by at least one ion exchange, absorption and precipitation. This amount of iron in the solution must be controlled before the acid solution may be recycled for the leaching process. The calcine leached in the first step is prepared in a two stage roast in the presence of both Cl 2 and a metal sulfide. The first stage is at 350-450 0 and the second at 550-700 0

  2. Process for recovering uranium

    Science.gov (United States)

    MacWood, G. E.; Wilder, C. D.; Altman, D.

    1959-03-24

    A process useful in recovering uranium from deposits on stainless steel liner surfaces of calutrons is presented. The deposit is removed from the stainless steel surface by washing with aqueous nitric acid. The solution obtained containing uranium, chromium, nickel, copper, and iron is treated with an excess of ammonium hydroxide to precipitnte the uranium, iron, and chromium and convert the nickel and copper to soluble ammonio complexions. The precipitated material is removed, dried and treated with carbon tetrachloride at an elevated temperature of about 500 to 600 deg C to form a vapor mixture of UCl/ sub 4/, UCl/sub 5/, FeCl/sub 3/, and CrCl/sub 4/. The UCl/sub 4/ is separated from this vapor mixture by selective fractional condensation at a temperature of about 500 to 400 deg C.

  3. Study on treatment of radioactive liquid waste from uranium ore processing by the use of nano oxide ferromagnetic

    International Nuclear Information System (INIS)

    Vuong Huu Anh; Nguyen Van Chinh; Nguyen Ba Tien; Doan Thi Thu Hien; Luu Cao Nguyen

    2015-01-01

    Nano oxide ferromagnetic Fe_3O_4 KT which was produced by the Military Institute of Science and Technology were used to adsorbed heavy metal elements in liquid waste. In this report, the nano oxide ferromagnetic Fe_3O_4 KT with the particle size of 80-100 nm and the specific surface area of 50-70 m"2/g was applied to study the adsorption of radioactive elements in the liquid waste of uranium ores processing. The effective parameters on adsorption process included temperature, stirring rate, stirring time, the pH value of the solution, the initial concentration of uranium in solution were investigated. The results showed that the maximum adsorption capacity for uranium of the nano Fe_3O_4 KT was 53.5 mgU/g with conditions such as: room temperature, stirring speed 120 rounds/minute, the pH value of solution was 8, stirring time about 2 hours . From the results obtained, nano Fe_3O_4 KT was tested to treatment real liquid waste of uranium ore processing after removing almost heavy metals and a part of radioactive elements by preliminary precipitation at pH 8. The results were analyzed on the ICP-MS and α, β total activity equipment, the solution concentration after treatment suitable for Vietnamese Technical Regulation on industrial wastewater QCVN 40: 2011 (concentrations of heavy metals; total activity of α and β). (author)

  4. Uranium-contaminated soil pilot treatment study

    International Nuclear Information System (INIS)

    Turney, W.R.J.R.; Mason, C.F.V.; Michelotti, R.A.

    1996-01-01

    A pilot treatment study is proving to be effective for the remediation of uranium-contaminated soil from a site at the Los Alamos National Laboratory by use of a two-step, zero-discharge, 100% recycle system. Candidate uranium-contaminated soils were characterized for uranium content, uranium speciation, organic content, size fractionization, and pH. Geochemical computer codes were used to forecast possible uranium leach scenarios. Uranium contamination was not homogenous throughout the soil. In the first step, following excavation, the soil was sorted by use of the ThemoNuclean Services segmented gate system. Following the sorting, uranium-contaminated soil was remediated in a containerized vat leach process by use of sodium-bicarbonate leach solution. Leach solution containing uranium-carbonate complexes is to be treated by use of ion-exchange media and then recycled. Following the treatment process the ion exchange media will be disposed of in an approved low-level radioactive landfill. It is anticipated that treated soils will meet Department of Energy site closure guidelines, and will be given open-quotes no further actionclose quotes status. Treated soils are to be returned to the excavation site. A volume reduction of contaminated soils will successfully be achieved by the treatment process. Cost of the treatment (per cubic meter) is comparable or less than other current popular methods of uranium-contamination remediation

  5. Advanced uranium enrichment processes

    International Nuclear Information System (INIS)

    Clerc, M.; Plurien, P.

    1986-01-01

    Three advanced Uranium enrichment processes are dealt with in the report: AVLIS (Atomic Vapour LASER Isotope Separation), MLIS (Molecular LASER Isotope Separation) and PSP (Plasma Separation Process). The description of the physical and technical features of the processes constitutes a major part of the report. If further presents comparisons with existing industrially used enrichment technologies, gives information on actual development programmes and budgets and ends with a chapter on perspectives and conclusions. An extensive bibliography of the relevant open literature is added to the different subjects discussed. The report was drawn up by the nuclear research Centre (CEA) Saclay on behalf of the Commission of the European Communities

  6. Crud treatment with 3 phase centrifuge in heap leach uranium process

    International Nuclear Information System (INIS)

    Hartmann, T.

    2010-01-01

    The presence of crud represents a permanent challenge for solvent extraction in the hydro-metal Uranium industry. The crud forms in the settlers of SX extraction. The crud is a stable emulsion which slowly spreads along the phase boundary between the aqueous and organic phase. Spreading of this intermediate phase is determined by the following influencing factors. Wind blows dust into the open settlers, some suspended solids coming with the pregnant leach solution (PLS) and wrong design of the mixers cause stable emulsions. Metallic solid residue is likewise responsible for the growth rate of the crud at the above-mentioned phase boundary. The crud can significantly impair the efficiency of hydro-metal extraction because the phase boundary between the aqueous and organic phases assumes substantial proportions, and the settlers cannot react flexibly. In a chain reaction, all settlers connected in series become infected with crud. The transfer of organic phase to the electrowinning (EW) cell can cause 'cathode burn'. The entrainment of electrolyte into the extraction stage can result in loss of pH control in the extraction circuit which will cause a drop in extraction efficiency. On the other hand, entrainment of the organic in the raffinate will result in organic losses to the leach circuit. Continuous treatment of the crud is extremely effective and reliable with a 3-phase separating solid bowl centrifuge. All three phases are separated distinctly from one another. All associated process steps exhibit a steady uniform efficiency. The main benefit for the customer is that process fluctuations in the extraction process will no longer occur. The 3-phase separating solid bowl centrifuge consists of an axial solid-wall bowl. The solid-wall bowl has a cylindrical section for simultaneous separation and clarification of the aqueous and organic liquid phase and a conical section for efficient solids dewatering. The 3-phase feed suspension is fed into the solid bowl

  7. Uranium Processing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — An integral part of Y‑12's transformation efforts and a key component of the National Nuclear Security Administration's Uranium Center of Excellence, the Uranium...

  8. Process for electrolytically preparing uranium metal

    Science.gov (United States)

    Haas, Paul A.

    1989-01-01

    A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

  9. PROCESS FOR PREPARING URANIUM METAL

    Science.gov (United States)

    Prescott, C.H. Jr.; Reynolds, F.L.

    1959-01-13

    A process is presented for producing oxygen-free uranium metal comprising contacting iodine vapor with crude uranium in a reaction zone maintained at 400 to 800 C to produce a vaporous mixture of UI/sub 4/ and iodine. Also disposed within the maction zone is a tungsten filament which is heated to about 1600 C. The UI/sub 4/, upon contacting the hot filament, is decomposed to molten uranium substantially free of oxygen.

  10. Uranium refining process using ion exchange membrane

    International Nuclear Information System (INIS)

    Yamaguchi, Akira

    1977-01-01

    As for the method of refining uranium ore being carried out in Europe and America at present, uranium ore is roughly refined at the mine sites to yellow cake, then this is transported to refineries and refined by dry method. This method has the following faults, namely the number of processes is large, it requires expensive corrosion-resistant materials because of high temperature treatment, and the impurities in uranium tend to increase. On the other hand, in case of EXCER method, treatment is carried out at low temperature, and high purity uranium can be obtained, but the efficiency of electrolytic reduction process is extremely low, and economically infeasible. In the wet refining method called PNC process, uranium tetrafluoride is produced from uranium ore without making yellow cake, therefore the process is rationalized largely, and highly economical. The electrolytic reduction process in this method was developed by Asahi Chemical Industry Co., Ltd. by constructing the pilot plant in Ningyotoge Mine. The ion exchange membrane, the electrodes, and the problems concerning the process and the engineering for commercial plants were investigated. The electrolytic reduction process, the pilot plant, the development of the elements of electrolytic cells, the establishment of analytical process, the measurement of the electrolytic characteristics, the demonstration operation, and the life time of the electrolytic diaphragm are reported. (Kako, I.)

  11. Uranium processing developments

    International Nuclear Information System (INIS)

    Jones, J.Q.

    1977-01-01

    The basic methods for processing ore to recover the contained uranium have not changed significantly since the 1954-62 period. Improvements in mill operations have been the result of better or less expensive reagents, changes in equipment, and in the successful resolvement of many environmental matters. There is also an apparent trend toward large mills that can profitably process lower grade ores. The major thrust in the near future will not be on process technology but on the remaining environmental constraints associated with milling. At this time the main ''spot light'' is on tailings dam and impoundment area construction and reclamation. Plans must provide for an adequate safety factor for stability, no surface or groundwater contamination, and minimal discharge of radionuclides to unrestricted areas, as may be required by law. Solution mining methods must also provide for plans to restore the groundwater back to its original condition as defined by local groundwater regulations. Basic flowsheets (each to finished product) plus modified versions of the basic types are shown

  12. Treatment of liquid wastes from uranium hydrometallurgy

    International Nuclear Information System (INIS)

    Moraga G, J.C.

    1988-01-01

    Different treatments for low activity liquid wastes, generated by the hidromettalurgy of uranium ore are studied. A process of treatment was chosen which includes a neutralization with lime and limestone and a selective removal of Ra-226, through ion-exchange resins. A plant, with a capacity of treatment of 1 m 3 /h of liquid effluents was scoped. (author)

  13. URANIUM SEPARATION PROCESS

    Science.gov (United States)

    Lyon, W.L.

    1962-04-17

    A method of separating uranium oxides from PuO/sub 2/, ThO/sub 2/, and other actinide oxides is described. The oxide mixture is suspended in a fused salt melt and a chlorinating agent such as chlorine gas or phosgene is sparged through the suspension. Uranium oxides are selectively chlorinated and dissolve in the melt, which may then be filtered to remove the unchlorinated oxides of the other actinides. (AEC)

  14. Uranium ore processing in Spain

    International Nuclear Information System (INIS)

    Josa, J.M.

    1976-01-01

    The paper presents a review of the Spanish needs of uranium concentrates and uranium ore processing technology and trends in Spain. Spain produces approximately 200t U 3 O 8 /a at two facilities. One plant in the south (Andujar, Jaen) can obtain 70t U 3 O 8 /a and uses a conventional acid leaching process with countercurrent solvent extraction. A second plant, situated in the west (Ciudad Rodrigo, Salamanca) has started in 1975 and has a capacity of 120-130t U 3 O 8 /a, using acid heap leaching and solvent extraction. There is another experimental facility (Don Benito, Badajoz) scheduled to start in 1976 and expected to produce about 25-35t U 3 O 8 /a as a by-product of the research work. For the near future (1978) it is hoped to increase the production with: (a) A new conventional acid leaching/solvent extraction plant in Ciudad Rodrigo; its tentative capacity is fixed at 550t U 3 O 8 /a. (b) A facility in the south, to recover about 130t U 3 O 8 /a from phosphoric acid. (c) Several small mobile plants (30t U 3 O 8 /a per plant); these will be placed near small and isolated mines. The next production increase (1979-1980) will come with the treatment of sandstones (Guadalajara and Cataluna) and lignites(Cataluna); this is being studied. There are also research programmes to study the recovery of uranium from low-grade ores (heap, in-situ and bacterial leaching) and from other industries. (author)

  15. Uranium resource processing. Secondary resources

    International Nuclear Information System (INIS)

    Gupta, C.K.; Singh, H.

    2003-01-01

    This book concentrates on the processing of secondary sources for recovering uranium, a field which has gained in importance in recent years as it is environmental-friendly and economically in tune with the philosophy of sustainable development. Special mention is made of rock phosphate, copper and gold tailings, uranium scrap materials (both natural and enriched) and sea water. This volume includes related area of ore mineralogy, resource classification, processing principles involved in solubilization followed by separation and safety aspects

  16. Distribution of natural radionuclides of uranium and thorium series in the process of artesian water treatment for drinking consumption

    International Nuclear Information System (INIS)

    Grashchenko, S.M.; Gritchenko, Z.G.; Shishkunova, L.V.

    1997-01-01

    Distribution of natural radionuclides of uranium and thorium series during the treatment of artesian water for drinking consumption is studied using vacuum-emanation and gamma spectrometry methods. During the water treatment hydroxide precipitates are produced at the station, which are isolated using a sand filter, radium isotopes being coprecipitated alongside with them. As a result of this radioactive waste is accumulated at the station, radium isotope concentration in it being equivalent to radium isotope concentration in uranium-thorium ores with 0:11% uranium and 0.56% thorium content. radium isotope concentration in water, delivered to the user do not exceed the established domestic normatives do not exceed the established domestic normatives

  17. Chemical treatment of uranium ores in France

    International Nuclear Information System (INIS)

    Mouret, P.; Sartorius, R.

    1958-01-01

    The various processes of chemical treatment of uranium ores, from the oldest to the more recent, are exposed, considering the following conditions: economics, geography, techniques and safety. The interest of obtaining a final concentrate as uranyl nitrate is discussed. (author) [fr

  18. The treatment of uranium ores

    International Nuclear Information System (INIS)

    Michel, P.

    1979-01-01

    After having described the main steps in the treatment of uranium ores, the author describes the treament activities for these ores, as they are organized in France and in the African countries having made cooperation agreements with France in this field [fr

  19. Studies on uranium ore processing

    International Nuclear Information System (INIS)

    Kim, C.H.; Park, S.W.; Lim, J.K.; Chung, M.K.

    1981-01-01

    Chemical and chemical engineering techniques of the uranium ore processing established by France COGEMA (Compagnie Generale des Matieres Nucleaires) have been comprehensively reviewed in preparation for successful test operation of the pilot plant to be completed by the end of 1981. It was found that the amount of sulfuric acid (75 Kg/t, ore) and sodium chlorate (2.5 Kg/t, ore) recommended by COGEMA should be increased up to 100 Kg/t, ore and 10 Kg/t, ore respectively to obtain satisfactory leach of uranium for some ore samples produced at the different pits of Goesan uranium mine. Conditions of the other processes such as solvent extraction, stripping, and precipitation of yellow cake were generally agreed with the results of intensive studies done by this laboratory

  20. Use of vacuum in processing of uranium

    International Nuclear Information System (INIS)

    Saify, M.T.; Rai, C.B.; Singh, S.P.; Singh, R.P.

    2003-01-01

    Full text: Natural uranium in the form of metal and alloys with suitable heat treatment are being used as fuel in research and some of the power reactors. The fuel is required to satisfy the purity specification from the criteria of neutron economy, corrosion resistance and fabricability. Uranium and its alloys fall under the category of reactive materials. They readily react with atmospheric air to form oxides. If molten uranium is exposed to atmosphere, it reacts violently with atmospheric gases and moisture, leading to explosion in extreme cases. Hence, protective inert atmosphere or high vacuum is required in processing of the materials especially during the melting and casting operation. Vacuum is preferred for melting and remelting of metals and alloys to remove the gaseous and high volatile impurities, to improve the mechanical properties of the material. Also, under vacuum sound castings are produced for further processing by mechanical working or use in casting forms. The addition of reactive alloying elements in uranium is efficiently carried out under vacuum. The paper highlights vacuum systems deployed and applications of vacuum in various operations involved in the processing of uranium and its alloys

  1. Treatment of acidic mine water at uranium mine No. 711 by barium chloride-sludge recycle-fractional neutralization process

    International Nuclear Information System (INIS)

    Yang Chaowen; Wang Benyi; Ding Tongsen; Zhong Pingru; Liao Yongbing; Li Xiaochu; Lu Guohua

    1994-01-01

    The barium chloride-sludge recycle-fractional neutralization process for disposal of acidic mine water at Uranium Mine No. 711 was checked through laboratory and enlarged tests and one-year industrial trial-run. The results showed that the presented technology can meet the requirements of production and environmental protection

  2. Distillation modeling for a uranium refining process

    Energy Technology Data Exchange (ETDEWEB)

    Westphal, B.R.

    1996-03-01

    As part of the spent fuel treatment program at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a salt electrolyte, containing a eutectic mixture of lithium and potassium chlorides, from uranium is achieved by a simple batch operation and is termed {open_quotes}cathode processing{close_quotes}. The incremental distillation of electrolyte salt will be modeled by an equilibrium expression and on a molecular basis since the operation is conducted under moderate vacuum conditions. As processing continues, the two models will be compared and analyzed for correlation with actual operating results. Possible factors that may contribute to aberrations from the models include impurities at the vapor-liquid boundary, distillate reflux, anomalous pressure gradients, and mass transport phenomena at the evaporating surface. Ultimately, the purpose of either process model is to enable the parametric optimization of the process.

  3. Study on treatment of radioactive liquid waste from uranium ore processing by the use of nano Fe_3O_4 KT particles

    International Nuclear Information System (INIS)

    Vuong Huu Anh; Nguyen Ba Tien; Doan Thi Thu Hien; Luu Cao Nguyen; Nguyen Van Chinh

    2015-01-01

    Nano Fe_3O_4 KT was produced from the Military Institute of Science and Technology were used to adsorbed heavy metal elements in liquid waste. In this report, the nano Fe_3O_4 KT particles sized 80-100 nm and specific surface area was 50-70 m"2/g was applied to study the adsorption of radioactive elements in the liquid waste of uranium ores processing. The effective parameters on adsorption process included temperature, stirring rate, stirring time, the pH value of the solution, the initial concentration of uranium in solution. The results showed the maximum adsorption capacity of the nano Fe_3O_4 KT was 53.5 mg/g with conditions such as room temperature, stirring speed 120 rounds/minute, the pH value of solution was 8, stirring time about 2 hours (Uranium/materials). From the results obtained, nano Fe_3O_4 KT tested to treatment liquid waste of uranium ore processing after preliminary precipitation removed almost heavy metals and a part of radioactive elements. The results were analyzed on the ICP-MS and α, β total counting, instrument. The solution concentration after treatment was suitable for Vietnam discharge standards into environment (QCVN 40:2011 on Industrial wastewater). (author)

  4. Distillation modeling for a uranium refining process

    International Nuclear Information System (INIS)

    Westphal, B.R.

    1996-01-01

    As part of the spent fuel treatment program at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a salt electrolyte, containing a eutectic mixture of lithium and potassium chlorides, from uranium is achieved by a simple batch operation and is termed open-quotes cathode processingclose quotes. The incremental distillation of electrolyte salt will be modeled by an equilibrium expression and on a molecular basis since the operation is conducted under moderate vacuum conditions. As processing continues, the two models will be compared and analyzed for correlation with actual operating results. Possible factors that may contribute to aberrations from the models include impurities at the vapor-liquid boundary, distillate reflux, anomalous pressure gradients, and mass transport phenomena at the evaporating surface. Ultimately, the purpose of either process model is to enable the parametric optimization of the process

  5. Photochemical process of laboratory uranium wastes recovery

    International Nuclear Information System (INIS)

    Borges, O.N.; Barros, M.P. de.

    1984-01-01

    A method for uranium extraction in presence of various aquometallic ions, based on selective photo-reduction of uranium is studied. Some economical advantages in relation with others conventional processes are analysed. (M.J.C.) [pt

  6. Uranium in mantle processes

    International Nuclear Information System (INIS)

    Cortini, M.

    1984-01-01

    (1) Metasomatism is an effective process in the mantle. It controls the distribution of U, Th and Pb in the mantle before the onset of magma formation. (2) Radioactive disequilibria demonstrate that magma formation is an open-system very fast process in which Ra, U and Th are extracted in large amounts from a mantle source that is geochemically distinct from the mantle fraction from which the melt is formed. (3) Because the enrichment of U, Th and Ra in the magma is so fast, the concept of mineral-melt partition coefficient is not valid for these elements during magma formation. (4) Metasomatism seems to generally produce an increase in μ and a decrease in K of the metasomatized mantle region. (5) Magma formation at oceanic ridges and islands seems to generally produce a decrease in K, in its mantle source region. (6) The major source of U, Th, Ra and Pb in a magma probably is the metasomatic mantle component. Instead, the major source of Sr and Nd in a magma is the non-metasomatic, more 'refractory' mantle component. (7) This proposed model is testable. It predicts isotopic disequilibrium of Pb between coexisting minerals and whole rocks, and a correlation of Pb with Th isotopes. (author)

  7. Studies on uranium ore processing

    International Nuclear Information System (INIS)

    Suh, I.S.; Chun, J.K.; Park, S.W.; Choi, S.J.; Lee, C.H.; Chung, M.K.; Lim, J.K.

    1983-01-01

    For the exploitation of domestic uranium ore deposit, comprehensive studies on uranium ore processing of the Geum-San pit ore are carried out. Physical and chemical characteristics of the Geum-San ore are similar to those of Goe-San ore and the physical beneficiation could not be applicable. Optimum operating conditions such as uranium leaching, solid-liquid separation, solvent extraction and precipitation of yellow cake are found out and the results are confirmed by the continous operation of the micro-plant with the capacity of 50Kg, ore/day. In order to improve the process of ore milling pilot plant installed recently, the feasibility of raffinate-recycle and the precipitation methods of yellow cake are intensively examined. It was suggested that the raffinate-recycle in the leaching of filtering stage could be reduced the environmental contamination and the peroxide precipitation technique was applicable to improve the purity of yellow cake. The mechanism and conditions the third phase formation are thoroughly studied and confirmed by chemical analysis of the third phase actually formed during the operation of pilot plant. The major constituents of the third phase are polyanions such as PMosub(12)Osub(40)sup(3-) or SiMosub(12)Osub(40)sup(4-). And the formation of these polyanions could be reduced by the control of redox potential and the addition of modifier. (Author)

  8. Uranium Processing Research in Australia [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, J R [Australian Atomic Energy Commission, Coogee, N.S.W. (Australia)

    1967-06-15

    Uranium processing research in Australia has included studies of flotation, magnetic separation, gravity separation, heavy medium separation, atmospheric leaching, multi-stage leaching, alkali leaching, solar heating of leach pulps, jigged-bed resin-in-pulp and solvent-in-pulp extraction. Brief details of the results obtained are given. In general, it can be said that gravity, magnetic and flotation methods are of limited usefulness in the treatment of Australian uranium ores. Alkali leaching seldom gives satisfactory recoveries and multi-stage leaching is expensive. Jigged-bed resin-in-pulp and packed tower solvent-in-pulp extraction systems both show promise, but plant-scale development work is required. Bacterial leaching may be useful in the case of certain low-grade ores. The main difficulties to be overcome, either singly or in combination, in the case of Australian uranium ores not currently considered economically exploitable, are the extremely finely divided state of the uranium mineral, the refractory nature of the uranium mineral and adverse effects due to the gangue minerals present. With respect to known low-grade ores, it would be possible in only a few cases to achieve satisfactory recovery of uranium at reasonable cost by standard treatment methods. (author)

  9. Uranium ore deposits: geology and processing implications

    International Nuclear Information System (INIS)

    Belyk, C.L.

    2010-01-01

    There are fifteen accepted types of uranium ore deposits and at least forty subtypes readily identified around the world. Each deposit type has a unique set of geological characteristics which may also result in unique processing implications. Primary uranium production in the past decade has predominantly come from only a few of these deposit types including: unconformity, sandstone, calcrete, intrusive, breccia complex and volcanic ones. Processing implications can vary widely between and within the different geological models. Some key characteristics of uranium deposits that may have processing implications include: ore grade, uranium and gangue mineralogy, ore hardness, porosity, uranium mineral morphology and carbon content. Processing difficulties may occur as a result of one or more of these characteristics. In order to meet future uranium demand, it is imperative that innovative processing approaches and new technological advances be developed in order that many of the marginally economic traditional and uneconomic non-traditional uranium ore deposits can be exploited. (author)

  10. Analysis of nuclear reaction products and materials; Preliminary treatment of uranium analysis

    International Nuclear Information System (INIS)

    Soedyartomo.

    1976-01-01

    Pre-treatment of samples is necessary to be done in order to achieve the efficient steps and accurate results of uranium analysis. The pre-treatment is particularly affected by the type of sample, the uranium concentration predicated in the sample, and the uranium analytical method which will be applied. A brief discussion about the pre-treatment of uranium analysis in the uranium ore processing and the reprocessing of spent fuel is given. (author)

  11. Treatment of effluents in uranium industry

    International Nuclear Information System (INIS)

    Ghosh, S.K.

    2009-01-01

    Uranium processing technology in India has matured in the last 50 years and is able to meet the country's requirement. Right from mining of the ore to milling and refining, effluents are generated and are being processed for their safe disposal. While the available technology is able to meet the regulatory limits of the effluents, the same may not be enough to meet the increased demand of uranium in the future. The increased population, urbanization and climate change are not only going to decrease the supply of process water but will also place increased restrictions on disposal to environment. This demands technologies that will generate less effluent for disposal and enable reuse and recycle concept to the extent possible. Presently used conventional physical-chemical methods, to contain the contaminants would, therefore, require further refinements. Contaminants like sulfates, chlorides etc in the effluent of uranium mill based on acid leach process are the concerns for the future plants. Hence, there is an urgent need for development of suitable methods for maximum recycle of the process effluents, which will also enable in minimizing the consumption of process water. A suitable membrane based process can be an option leaving a concentrated brine for reuse or for further treatment and disposal

  12. Filtration aids in uranium ore processing

    International Nuclear Information System (INIS)

    Ford, H.L.; Levine, N.M.; Risdon, A.R.

    1975-01-01

    A process of improving the filtration efficiency and separation of uranium ore pulps obtained by carbonate leaching of uranium ore which comprises treating said ore pulps with an aqueous solution of hydroxyalkyl guar selected from the group consisting of hydroxyethyl and hydroxypropyl guar in the amount of 0.1 and 2.0 pounds of hydroxyalkyl guar per ton of uranium ore

  13. Advances in uranium enrichment processes

    International Nuclear Information System (INIS)

    Rae, H.K.; Melvin, J.G.; Slater, J.B.

    1986-05-01

    Advances in gas centrifuges and development of the atomic vapour laser isotope separation process promise substantial reductions in the cost of enriched uranium. The resulting reduction in LWR fuel costs could seriously erode the economic advantage of CANDU, and in combination with LWR design improvements, shortened construction times and increased operational reliability could allow the LWR to overtake CANDU. CANDU's traditional advantages of neutron economy and high reliability may no longer be sufficient - this is the challenge. The responses include: combining neutron economy and dollar economy by optimizing CANDU for slightly enriched uranium fuel; developing cost-reducing improvements in design, manufacture and construction; and reducing the cost of heavy water. Technology is a renewable resource which must be continually applied to a product for it to remain competitive in the decades to come. Such innovation is a prerequisite to Canada increasing her share of the international market for nuclear power stations. The higher burn-up achievable with enriched fuel in CANDU can reduce the fuel cycle costs by 20 to 40 percent for a likely range of costs for yellowcake and separative work. Alternatively, some of the benefits of a higher fissile content can take the form of a cheaper reactor core containing fewer fuel channels and less heavy water, and needing only a single fuelling machine. An opportunity that is linked to this need to introduce an enriched uranium fuel cycle into CANDU is to build an enrichment business in Canada. This could offer greater value added to our uranium exports, security of supply for enriched CANDUs, technological growth in Canada and new employment opportunities. AECL has a study in progress to define this opportunity

  14. Process for producing uranium oxide rich compositions from uranium hexafluoride

    International Nuclear Information System (INIS)

    DeHollander, W.R.; Fenimore, C.P.

    1978-01-01

    Conversion of gaseous uranium hexafluoride to a uranium dioxide rich composition in the presence of an active flame in a reactor defining a reaction zone is achieved by separately introducing a first gaseous reactant comprising a mixture of uranium hexafluoride and a reducing carrier gas, and a second gaseous reactant comprising an oxygen-containing gas. The reactants are separated by a shielding gas as they are introduced to the reaction zone. The shielding gas temporarily separates the gaseous reactants and temporarily prevents substantial mixing and reacting of the gaseous reactants. The flame occurring in the reaction zone is maintained away from contact with the inlet introducing the mixture to the reaction zone. After suitable treatment, the uranium dioxide rich composition is capable of being fabricated into bodies of desired configuration for loading into nuclear fuel rods. Alternatively, an oxygen-containing gas as a third gaseous reactant is introduced when the uranium hexafluoride conversion to the uranium dioxide rich composition is substantially complete. This results in oxidizing the uranium dioxide rich composition to a higher oxide of uranium with conversion of any residual reducing gas to its oxidized form

  15. Treatment of uranium contaminated wastewater – a review

    International Nuclear Information System (INIS)

    Dulama, M.; Iordache, M.; Deneanu, N.

    2013-01-01

    The paper presents a study of the treatment techniques used for uranium recovery from aqueous solutions, such as: precipitation, ion exchange processes, sorption processes, solvent extractions, separation by liquid membrane, nanofiltration and reverse osmosis. The necessary elements for rigorous treatment experiments that can be used to define innovative procedure for uranium contaminated wastewater treatment are described in this review. The published data were summarized and the areas for further research were identified in order to be able to propose an environmental friendly technology in the field of uranium production and recovery cycle. (authors)

  16. PROCESS OF RECOVERING URANIUM FROM ITS ORES

    Science.gov (United States)

    Galvanek, P. Jr.

    1959-02-24

    A process is presented for recovering uranium from its ores. The crushed ore is mixed with 5 to 10% of sulfuric acid and added water to about 5 to 30% of the weight of the ore. This pugged material is cured for 2 to 3 hours at 100 to 110 deg C and then cooled. The cooled mass is nitrate-conditioned by mixing with a solution equivalent to 35 pounds of ammunium nitrate and 300 pounds of water per ton of ore. The resulting pulp containing 70% or more solids is treated by upflow percolation with a 5% solution of tributyl phosphate in kerosene at a rate equivalent to a residence time of about one hour to extract the solubilized uranium. The uranium is recovered from the pregnant organic liquid by counter-current washing with water. The organic extractant may be recycled. The uranium is removed from the water solution by treating with ammonia to precipitate ammonium diuranate. The filtrate from the last step may be recycled for the nitrate-conditioning treatment.

  17. Chemical treatment of ammonium fluoride solution in uranium reconversion plant

    International Nuclear Information System (INIS)

    Carvalho Frajndlich, E.U. de.

    1992-01-01

    A chemical procedure is described for the treatment of the filtrate, produced from the transformation of uranium hexafluoride (U F 6 ) into ammonium uranyl carbonate (AUC). This filtrate is an intermediate product in the U F 6 to uranium dioxide (U O 2 ) reconversion process. The described procedure recovers uranium as ammonium peroxide fluoro uranate (APOFU) by precipitation with hydrogen peroxide (H 2 O 2 ), and as later step, its calcium fluoride (CaF 2 ) co-precipitation. The recovered uranium is recycled to the AUC production plant. (author)

  18. A process for uranium recovery in phosphoric acid

    International Nuclear Information System (INIS)

    Duarte Neto, J.

    1984-01-01

    Results are presented about studies carried out envisaging the development of a process for uranium recovery from phosphoric acid, produced from the concentrate obtained from phosphorus-uraniferous mineral from Itataia mines (CE, Brazil). This process uses a mixture of DEPA-TOPO as extractant and the extraction cycle involves the following stages: acid pre-treatment; adjustment of the oxidation potential so to ensure that all uranium is hexavalent; extraction of uranium from the acid; screening of the solvent to remove undesirable impurities; uranium re-extraction and precipitation; solvent recovery. A micro-pilot plant for continuous processing was built up. Data collected showed that uranium can be recovered with an yield greater than 99%, thus proving the feasibility of the process and encouraging the construction of a bigger scale plant. (Author) [pt

  19. Depleted uranium processing and fluorine extraction

    International Nuclear Information System (INIS)

    Laflin, S.T.

    2010-01-01

    Since the beginning of the nuclear era, there has never been a commercial solution for the large quantities of depleted uranium hexafluoride generated from uranium enrichment. In the United States alone, there is already in excess of 1.6 billion pounds (730 million kilograms) of DUF_6 currently stored. INIS is constructing a commercial uranium processing and fluorine extraction facility. The INIS facility will convert depleted uranium hexafluoride and use it as feed material for the patented Fluorine Extraction Process to produce high purity fluoride gases and anhydrous hydrofluoric acid. The project will provide an environmentally friendly and commercially viable solution for DUF_6 tails management. (author)

  20. Liquid membrane process for uranium recovery

    International Nuclear Information System (INIS)

    Valint, P.L. Jr.

    1982-01-01

    An improved liquid membrane emulsion extraction process for recovering uranium from a WPPA feed solution containing uranyl cations wherein said feed is contacted with a water-in-oil emulsion which extracts and captures the uranium in the interior aqueous phase thereof, wherein the improvement comprises the presence of an alkane diphosphonic acid uranium complexing agent in the interior phase of the emulsion. This improvement results in greater extraction efficiency

  1. Yellowcake processing in uranium recovery

    International Nuclear Information System (INIS)

    Paul, J.M.

    1981-01-01

    This information relates to the recovery of uranium from uranium peroxide yellowcake produced by precipitation with hydrogen peroxide. The yellowcake is calcined at an elevated temperature to effect decomposition of the yellowcake to uranium oxide with the attendant evolution of free oxygen. The calcination step is carried out in the presence of a reducing agent which reacts with the free oxygen, thus retarding the evolution of chlorine gas from sodium chloride in the yellowcake. Suitable reducing agents include ammonia producing compounds such as ammonium carbonate and ammonium bicarbonate. Ammonium carbonate and/or ammonium bicarbonate may be provided in the eluant used to desorb the uranium from an ion exchange column

  2. Yellowcake processing in uranium recovery

    Energy Technology Data Exchange (ETDEWEB)

    Paul, J.M.

    1981-10-06

    This information relates to the recovery of uranium from uranium peroxide yellowcake produced by precipitation with hydrogen peroxide. The yellowcake is calcined at an elevated temperature to effect decomposition of the yellowcake to uranium oxide with the attendant evolution of free oxygen. The calcination step is carried out in the presence of a reducing agent which reacts with the free oxygen, thus retarding the evolution of chlorine gas from sodium chloride in the yellowcake. Suitable reducing agents include ammonia producing compounds such as ammonium carbonate and ammonium bicarbonate. Ammonium carbonate and/or ammonium bicarbonate may be provided in the eluant used to desorb the uranium from an ion exchange column.

  3. Zoujiashan uranium waste water treatment optimizaiton design

    International Nuclear Information System (INIS)

    Huang Lianjun

    2014-01-01

    Optimization design follows the decontamination triage, comprehensive management, such as wastewater treatment principle and from easy to difficult. increasing the slurry treatment, optimization design containing ρ (U) > defines I mg/L wastewater for higher uranium concentration wastewater, whereas low uranium concentration wastewater. Through the optimization design, solve the problem of water turbidity 721-15 wastewater treatment station of the lack of capacity and mine. (author)

  4. Uranium recovery from wet process phosphoric acid

    International Nuclear Information System (INIS)

    1980-01-01

    In the field of metallurgy, specifically processes for recovering uranium from wet process phosphoric acid solution derived from the acidulation of uraniferous phosphate ores, problems of imbalance of ion exchange agents, contamination of recycled phosphoric acid with process organics and oxidizing agents, and loss and contamination of uranium product, are solved by removing organics from the raffinate after ion exchange conversion of uranium to uranous form and recovery thereof by ion exchange, and returning organics to the circuit to balance mono and disubstituted ester ion exchange agents; then oxidatively stripping uranium from the agent using hydrogen peroxide; then after ion exchange recovery of uranyl and scrubbing, stripping with sodium carbonate and acidifying the strip solution and using some of it for the scrubbing; regenerating the sodium loaded agent and recycling it to the uranous recovery step. Economic recovery of uranium as a by-product of phosphate fertilizer production is effected. (author)

  5. Process for uranium recovery in phosphorus compounds

    International Nuclear Information System (INIS)

    Demarthe, J.M.; Solar, Serge.

    1980-01-01

    Process for uranium recovery in phosphorus compounds with an organic phase containing a dialkylphosphoric acid. A solubilizing agent constituted of an heavy alcohol or a phosphoric acid ester or a tertiary phosphine oxide or octanol-2, is added to the organic phase for solubilization of the uranium and ammonium dialkyl pyrophosphate [fr

  6. Processing of Sierra Albarrana uranium ores

    International Nuclear Information System (INIS)

    Gutierrez Jodra, L.; Perez Luina, A.; Perarnau, M.

    1960-01-01

    Uranium recovery by hydrometallurgy from brannerite, found in Hornachuelos (Cordoba) is described. It has been studied the acid and alkaline leaching and salt roasting, proving as more satisfactory the acid leaching. Besides the uranium solubilization by acid leaching, is described the further process to obtain pure uranyl nitrate. (Author)

  7. Recovery and treatment of uranium from uranium-containing solution by liquid membrane emulsion technology

    International Nuclear Information System (INIS)

    Xia Liangshu; Zhou Yantong; Xiao Yiqun; Peng Anguo; Xiao Jingshui; Chen Wei

    2014-01-01

    The recovery and treatment of uranium from uranium-containing solution using liquid membrane emulsion (LME) technology were studied in this paper, which contained the best volume ratio of membrane materials, stirring speed during emulsion process, the conditions of extracting, such as temperature, pH, initial concentration of uranium. Moreover, the mechanism for extracting uranium was also discussed. The best experimental conditions of emulsifying were acquired. The volume fractions of P 204 and liquid paraffin are 0.1 and 0.05, the volume ratios of Span80 and sulphonated kerosene to P 204 are 0.06 and 0.79 respectively, stirring speed is controlled in 2 000 r/min, and the concentration of inner phase is 4 mol/L. The recovery rate of uranium is up to 99% through the LME extracted uranium for 0.5 h at pH 2.5 and room temperature when the initial concentration is less than 400 mg/L and the volume ratio is 5 between the uranium-containing waste water and LME. The calculation results of Gibbs free energy show that the reaction process is spontaneous. (authors)

  8. Chemical treatment proceed of poor uranium content ores; Un procede de traitement chimique des minerais pauvres d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Mouret, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Pagny, P [Societe Potasse et Engrais Chimiques (France)

    1955-07-01

    The needs in uranium constantly increased inciting to develop new chemical processes for the treatment of uranium ores. we searched processes that permit to get this element from ores poor in uranium, to a reasonable cost price. We used a sulphuric attack and a precipitation of uranium as phosphate uranate or pyrophosphate uranate to separate its from the different impurities. The process permitted to process ores contents of about 0,05% of uranium and to get an end product of sodium carbonate uranate containing 60 to 65% of uranium, with an acceptable cost price and an extraction yield situated between 90 and 95%. (M.B.) [French] Les besoins sans cesse accrus en uranium ont incite de developper de nouveaux procedes chimiques pour le traitement de minerais uranifere. nous avons recherche des procedes qui permettent d'obtenir cet element a partir de minerais pauvres en uranium, a un prix de revient raisonnable. Nous nous sommes orientes vers une attaque sulfurique et une precipitation de l'uranium sous forme de phosphate uraneux ou de pyrophosphate uraneux pour le separer des differentes impuretes. Le procede a permis de descendre a des teneurs en uranium de l'ordre de 0,05 % et d'obtenir un produit final a l'etat d'uranate de soude contenant 60 e 65 % d'uranium, avec un prix de revient acceptable et avec un rendement global d'extraction situe entre 90 et 95 %. (M.B.)

  9. Uranium in a recent phosphorite formation process

    Energy Technology Data Exchange (ETDEWEB)

    Baturin, G N; Dubinchuk, V I; Kochenov, A V

    1986-01-01

    Uranium behaviour in the process of nowadays phosphorite formation in the sediments of Namibia shelf is considered. The material collected during the 3-d trip of the research vessel ''Akademik Kurchatov'' and 26-th trip of the research vessel ''Mikhail Lomonosov'' is used. The samples from three geological stations 2046, 2047 and 2048 from the depths of 78-87 m have been investigated. Each sample (mass from 0.2 to 0.3 kg) is composed of several samples representing unified genetic series: holocene diatomic silts enclosing phosphorites - phosphatized silts - phosphorite concretions. Uranium has been determined by the X-ray spectral method; phosphorus, organic carbon and other components - by the chemical analysis. Uranium forms investigated by the combination of methods of electron microscopy, microdiffraction, microradioautography and microsounding. Uranium content in nowadays phosphorites at the shelf is 3-106 g/t. Uranium accumulation in phosphorites at the initial stages of their formation is controlled by its content in host sediments. In the course of litification of diagenetic phosphate concretions the uranium content in them varies from 40 to 80 g/t. The uranium concentration process in phosphorites is accompanied by formation of independent mineral phases of uranium oxide and ningyoite type.

  10. Uranium

    International Nuclear Information System (INIS)

    Cuney, M.; Pagel, M.; Leroy, J.

    1992-01-01

    First, this book presents the physico-chemical properties of Uranium and the consequences which can be deduced from the study of numerous geological process. The authors describe natural distribution of Uranium at different scales and on different supports, and main Uranium minerals. A great place in the book is assigned to description and classification of uranium deposits. The book gives also notions on prospection and exploitation of uranium deposits. Historical aspects of Uranium economical development (Uranium resources, production, supply and demand, operating costs) are given in the last chapter. 7 refs., 17 figs

  11. Filtration aids in uranium ore processing

    International Nuclear Information System (INIS)

    Ford, H.L.; Levine, N.M.; Risdon, A.L.

    1975-01-01

    The patent describes a process whereby improved flocculation efficiency and filtration of carbonate leached uranium ore pulps are obtained by treating the filter feed slurry with an aqueous solution of hydroxyalkyl guar. (J.R.)

  12. PROCESSES OF CHLORINATION OF URANIUM OXIDES

    Science.gov (United States)

    Rosenfeld, S.

    1958-09-16

    An improvement is described in the process fur making UCl/sub 4/ from uranium oxide and carbon tetrachloride. In that process, oxides of uranium are contacted with carbon tetrachloride vapor at an elevated temperature. It has been fuund that the reaction product and yield are improved if the uranlum oxide charge is disposed in flat trays in the reaction zone, to a depth of not more than 1/2 centimeter.

  13. New processes for uranium isotope separation

    International Nuclear Information System (INIS)

    Vanstrum, P.R.; Levin, S.A.

    1977-01-01

    An overview of the status and prospects for processes other than gaseous diffusion, gas centrifuge, and separation nozzle for uranium isotope separation is presented. The incentive for the development of these processes is the increasing requirements for enriched uranium as fuel for nuclear power plants and the potential for reducing the high costs of enrichment. The latest nuclear power projections are converted to uranium enrichment requirements. The size and timing of the market for new enrichment processes are then determined by subtracting the existing and planned uranium enrichment capacities. It is estimated that to supply this market would require the construction of a large new enrichment plant of 9,000,000 SWU per year capacity, costing about $3 billion each (in 1976 dollars) about every year till the year 2000. A very comprehensive review of uranium isotope separation processes was made in 1971 by the Uranium Isotope Separation Review Ad Hoc Committee of the USAEC. Many of the processes discussed in that review are of little current interest. However, because of new approaches or remaining uncertainties about potential, there is considerable effort or continuing interest in a number of alternative processes. The status and prospects for attaining the requirements for competitive economics are presented for these processes, which include laser, chemical exchange, aerodynamic other than separation nozzle, and plasma processes. A qualitative summary comparison of these processes is made with the gaseous diffusion, gas centrifuge, and separation nozzle processes. In order to complete the overview of new processes for uranium isotope separation, a generic program schedule of typical steps beyond the basic process determination which are required, such as subsystem, module, pilot plant, and finally plant construction, before large-scale production can be attained is presented. Also the present value savings through the year 2000 is shown for various

  14. Solubility of airborne uranium samples from uranium processing plant

    International Nuclear Information System (INIS)

    Kravchik, T.; Oved, S.; Sarah, R.; Gonen, R.; Paz-Tal, O.; Pelled, O.; German, U.; Tshuva, A.

    2005-01-01

    Full text: During the production and machining processes of uranium metal, aerosols might be released to the air. Inhalation of these aerosols is the main route of internal exposure of workers. To assess the radiation dose from the intake of these uranium compounds it is necessary to know their absorption type, based on their dissolution rate in extracellular aqueous environment of lung fluid. The International Commission on Radiological Protection (ICRP) has assigned UF4 and U03 to absorption type M (blood absorption which contains a 10 % fraction with an absorption rate of 10 minutes and 90 % fraction with an absorption rate of 140 fays) and UO2 and U3O8 to absorption type S (blood absorption rate with a half-time of 7000 days) in the ICRP-66 model.The solubility classification of uranium compounds defined by the ICRP can serve as a general guidance. At specific workplaces, differences can be encountered, because of differences in compounds production process and the presence of additional compounds, with different solubility characteristics. According to ICRP recommendations, material-specific rates of absorption should be preferred to default parameters whenever specific experimental data exists. Solubility profiles of uranium aerosols were determined by performing in vitro chemical solubility tests on air samples taken from uranium production and machining facilities. The dissolution rate was determined over 100 days in a simultant solution of the extracellular airway lining fluid. The filter sample was immersed in a test vial holding 60 ml of simultant fluid, which was maintained at a 37 o C inside a thermostatic bath and at a physiological pH of 7.2-7.6. The test vials with the solution were shaken to simulate the conditions inside the extracellular aqueous environment of the lung as much as possible. The tests indicated that the uranium aerosols samples taken from the metal production and machining facilities at the Nuclear Research Center Negev (NRCN

  15. Process for recovery of uranium from wet process phosphoric acid

    International Nuclear Information System (INIS)

    Wiewiorowski, T.K.; Thornsberry, W.L. Jr.

    1978-01-01

    Process is claimed for the recovery of uranium from wet process phosphoric acid solution in which an organic extractant, containing uranium values and dissolved iron impurities and comprising a dialkylphosphoric acid and a trialkylphosphine oxide dissolved in a water immiscible organic solvent, is contacted with a substantially iron-free dilute aqueous phosphoric acid to remove said iron impurities. The removed impurities are bled from the system by feeding the resulting iron-loaded phosphoric acid to a secondary countercurrent uranium extraction operation from which they leave as part of the uranium-depleted acid raffinate. Also, process for recovering uranium in which the extractant, after it has been stripped of uranium values by aqueous ammonium carbonate, is contacted with a dilute aqueous acid selected from the group consisting of H 2 SO 4 , HCl, HNO 3 and iron-free H 3 PO 4 to improve the extraction efficiency of the organic extractant

  16. Development of uranium processing at Wiluna

    Energy Technology Data Exchange (ETDEWEB)

    Kenny, D., E-mail: dayle.kenny@toroenergy.com.au [Toro Energy Ltd., West Perth, WA (Australia); Dombrose, E. [Metallurgical Support Pty Ltd., Shelley, WA (Australia)

    2010-07-01

    Toro Energy Ltd. has identified a resource of 20.2 million tonnes at a grade of 548 ppm U{sub 3}O{sub 8} at Wiluna, Western Australia. Calcrete and clay delta formations host the uranium mineral carnotite. Initial studies indicate a mining operation is technically, environmentally and commercially viable. Increase in demand for uranium and a change in State Government policy on uranium mining have lead Toro to proceed with a bankable feasibility study and commence approvals with State and Federal Governments. This paper discusses how Toro arrived at the decision to utilise alkaline heap leach, a process not widely used, and how it is being developed. (author)

  17. The jet nozzle process for uranium 235 isotopic enrichment

    International Nuclear Information System (INIS)

    Jordan, I.; Umeda, K.; Brown, A.E.P.

    1979-01-01

    A general survey of the isotopic enrichment of Uranium - 235, principally by jet nozzle process, is made. Theoretical treatment of a single stage and cascade of separation stages of the above process with its development in Germany until 1976 is presented [pt

  18. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, Jr., Howard W.; Horton, James A.; Elliott, Guy R. B.

    1995-01-01

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

  19. Process for continuous production of metallic uranium and uranium alloys

    Science.gov (United States)

    Hayden, H.W. Jr.; Horton, J.A.; Elliott, G.R.B.

    1995-06-06

    A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

  20. Uranium

    International Nuclear Information System (INIS)

    1982-01-01

    The development, prospecting, research, processing and marketing of South Africa's uranium industry and the national policies surrounding this industry form the headlines of this work. The geology of South Africa's uranium occurences and their positions, the processes used in the extraction of South Africa's uranium and the utilisation of uranium for power production as represented by the Koeberg nuclear power station near Cape Town are included in this publication

  1. Uranium alloy forming process research

    International Nuclear Information System (INIS)

    Chow, T.S.; Biesiada, T.A.; Sunwoo, A.; Long, J.; Anklam, T.; Kang, S.W.

    1997-01-01

    The study of modern U-6Nb processes is motivated by the needs to reduce fabrication costs and to improve efficiency in material usage. We have studied two potential options: physical vapor deposition (PVD) for manufacturing near-net-shape U-6Nb, and kinetic-energy metallization (KEM) as a supplemental process for refurbishing recycled parts. In FY 1996, we completed two series of PVD runs and heat treatment analyses, the characterization of the microstructure and mechanical properties, a comparison of the results to data for wrought-processed material, and experimental demonstration of the KEM feasibility process with a wide range of variables (particle materials and sizes, gases and gas pressures, and substrate materials), and computer modeling calculations

  2. Impurities in uranium process solutions

    International Nuclear Information System (INIS)

    Boydell, D.W.

    1980-01-01

    Several uranium purification circuits are presented in tabular form together with the average major impurity levels associated with each. The more common unit operations in these circuits, namely strong- and weak-base ion-exchange, solvent extraction and the precipitation of impurities are then discussed individually. Particular attention is paid to the effect and removal of impurities in each of these four unit operations. (author)

  3. Mining and processing of uranium ores in the USSR

    International Nuclear Information System (INIS)

    Laskorin, B.N.; Mamilov, V.A.; Korejsho, Yu.A.

    1983-01-01

    Experience gained in uranium ore mining by modern methods in combination with underground and heap leaching is summarized. More intensive processing of low-grade ores has been achieved through the use of autoclave leaching, sorptive treatment of thick pulps, extractive separation of pure uranium compounds, automated continuous sorption devices of high efficiency for processing the underground- and heap-leaching liquors, natural and mine water, and recovery of molybdenum, vanadium, scandium, rare earths and phosphate fertilizers from low-grade ores. Production of ion-exchangers and extractants has been developed and processes for concomitant recovery of copper, gold, ionium, tungsten, caesium, zirconium, tantalum, nickel and cobalt have been designed. (author)

  4. Application of biohydrometallurgy to uranium ore processing

    International Nuclear Information System (INIS)

    Zhang Jiantang

    1989-01-01

    The development on application of biohydrometallargy to uranium ore processing is briefly introduced. The device designed for oxidizing ferrous ions in solution by using biomembrane, several bacterial leaching methods and the experimental results are given in this paper. The presented biohydrometallurgical process for recovering uranium includes bacterial leaching following by adsorption using tertiary amine resin 351 and oxidation of ferrous ions in the device with biomembranes. This process brings more economical benefits for treating silicate type original ores. The prospects on application of biogydrometallyurgy to solution mining is also discussed

  5. Uranium

    International Nuclear Information System (INIS)

    Stewart, E.D.J.

    1974-01-01

    A discussion is given of uranium as an energy source in The Australian economy. Figures and predictions are presented on the world supply-demand position and also figures are given on the added value that can be achieved by the processing of uranium. Conclusions are drawn about Australia's future policy with regard to uranium (R.L.)

  6. Treatment of mine-water from decommissioning uranium mines

    International Nuclear Information System (INIS)

    Fan Quanhui

    2002-01-01

    Treatment methods for mine-water from decommissioning uranium mines are introduced and classified. The suggestions on optimal treatment methods are presented as a matter of experience with decommissioned Chenzhou Uranium Mine

  7. Uranium recovery from wet process phosphoric acid

    International Nuclear Information System (INIS)

    Carrington, O.F.; Pyrih, R.Z.; Rickard, R.S.

    1981-01-01

    Improvement in the process for recovering uranium from wetprocess phosphoric acid solution derived from the acidulation of uraniferous phosphate ores by the use of two ion exchange liquidliquid solvent extraction circuits in which in the first circuit (A) the uranium is reduced to the uranous form; (B) the uranous uranium is recovered by liquid-liquid solvent extraction using a mixture of mono- and di-(Alkyl-phenyl) esters of orthophosphoric acid as the ion exchange agent; and (C) the uranium oxidatively stripped from the agent with phosphoric acid containing an oxidizing agent to convert uranous to uranyl ions, and in the second circuit (D) recovering the uranyl uranium from the strip solution by liquid-liquid solvent extraction using di(2ethylhexyl)phosphoric acid in the presence of trioctylphosphine oxide as a synergist; (E) scrubbing the uranium loaded agent with water; (F) stripping the loaded agent with ammonium carbonate, and (G) calcining the formed ammonium uranyl carbonate to uranium oxide, the improvement comprising: (1) removing the organics from the raffinate of step (B) before recycling the raffinate to the wet-process plant, and returning the recovered organics to the circuit to substantially maintain the required balance between the mono and disubstituted esters; (2) using hydogren peroxide as the oxidizing agent in step (C); (3) using an alkali metal carbonate as the stripping agent in step (F) following by acidification of the strip solution with sulfuric acid; (4) using some of the acidified strip solution as the scrubbing agent in step (E) to remove phosphorus and other impurities; and (5) regenerating the alkali metal loaded agent from step (F) before recycling it to the second circuit

  8. ALKALINE CARBONATE LEACHING PROCESS FOR URANIUM EXTRACTION

    Science.gov (United States)

    Thunaes, A.; Brown, E.A.; Rabbitts, A.T.

    1957-11-12

    A process for the leaching of uranium from high carbonate ores is presented. According to the process, the ore is leached at a temperature of about 200 deg C and a pressure of about 200 p.s.i.g. with a solution containing alkali carbonate, alkali permanganate, and bicarbonate ion, the bicarbonate ion functionlng to prevent premature formation of alkali hydroxide and consequent precipitation of a diuranate. After the leaching is complete, the uranium present is recovered by precipitation with NaOH.

  9. Dry uranium tetrafluoride process preparation using the uranium hexafluoride reconversion process effluents

    International Nuclear Information System (INIS)

    Silva Neto, Joao Batista da

    2008-01-01

    It is a well known fact that the use of uranium tetrafluoride allows flexibility in the production of uranium suicide and uranium oxide fuel. To its obtention there are two conventional routes, the one which reduces uranium from the UF 6 hydrolysis solution with stannous chloride, and the hydro fluorination of a solid uranium dioxide. In this work we are introducing a third and a dry way route, mainly utilized to the recovery of uranium from the liquid effluents generated in the uranium hexafluoride reconversion process, at IPEN/CNEN-SP. Working in the liquid phase, this route comprises the recuperation of ammonium fluoride by NH 4 HF 2 precipitation. Working with the solid residues, the crystallized bifluoride is added to the solid UO 2 , which comes from the U mini plates recovery, also to its conversion in a solid state reaction, to obtain UF 4 . That returns to the process of metallic uranium production unity to the U 3 Si 2 obtention. This fuel is considered in IPEN CNEN/SP as the high density fuel phase for IEA-R1m reactor, which will replace the former low density U 3 Si 2 -Al fuel. (author)

  10. Processing of uranium-containing coal

    International Nuclear Information System (INIS)

    Cordero Alvarez, M.

    1987-01-01

    A direct storage of uranium-bearing coal requires the processing of large amounts of raw materials while lacking guarantee of troublefree process cycles. With the example of an uranium-bearing bituminous coal from Stockheim, it was aimed at the production of an uranium ore concentrate by means of mechanical, thermal and chemical investigations. Above all, amorphous pitch blende was detected as a uranium mineralization which occurs homogeneously distributed in the grain size classes of the comminuted raw material with particle diameters of a few μm and, after the combustion, enriches in the field of finest grain of the axis. Heterogeneous and solid-state reactions in the thermal decarburization above 700deg C result in the development of hardly soluble uranium oxides and and calcium uranates as well as in enclosures in mineral glass. Thus, the pre-enrichment has to take place in a temperature range below 600deg C. By means of a sorting classification of the ash at ± 2.0 mm, it is possible to achieve an enrichment of up to factor 15 for a mineral of a mainly low carbonate content and, for a mineral of a rich carbonate content, up to the factor 4. The separation of the uranium from the concentrates produced is possible with a yield of 95% by means of leaching with sulphuric acid at a temperature of 20deg C. As far as their reproducibility was concerned, the laboratory tests were verified on a semi-industrial scale. A processing method is suggested on the basis of the data obtained. (orig.) [de

  11. Uranium

    International Nuclear Information System (INIS)

    Poty, B.; Cuney, M.; Bruneton, P.; Virlogeux, D.; Capus, G.

    2010-01-01

    With the worldwide revival of nuclear energy comes the question of uranium reserves. For more than 20 years, nuclear energy has been neglected and uranium prospecting has been practically abandoned. Therefore, present day production covers only 70% of needs and stocks are decreasing. Production is to double by 2030 which represents a huge industrial challenge. The FBR-type reactors technology, which allows to consume the whole uranium content of the fuel, is developing in several countries and will ensure the long-term development of nuclear fission. However, the implementation of these reactors (the generation 4) will be progressive during the second half of the 21. century. For this reason an active search for uranium ores will be necessary during the whole 21. century to ensure the fueling of light water reactors which are huge uranium consumers. This dossier covers all the aspects of natural uranium production: mineralogy, geochemistry, types of deposits, world distribution of deposits with a particular attention given to French deposits, the exploitation of which is abandoned today. Finally, exploitation, ore processing and the economical aspects are presented. Contents: 1 - the uranium element and its minerals: from uranium discovery to its industrial utilization, the main uranium minerals (minerals with tetravalent uranium, minerals with hexavalent uranium); 2 - uranium in the Earth's crust and its geochemical properties: distribution (in sedimentary rocks, in magmatic rocks, in metamorphic rocks, in soils and vegetation), geochemistry (uranium solubility and valence in magmas, uranium speciation in aqueous solution, solubility of the main uranium minerals in aqueous solution, uranium mobilization and precipitation); 3 - geology of the main types of uranium deposits: economical criteria for a deposit, structural diversity of deposits, classification, world distribution of deposits, distribution of deposits with time, superficial deposits, uranium

  12. Uranium bed oxidation vacuum process system

    International Nuclear Information System (INIS)

    McLeland, H.L.

    1977-01-01

    Deuterium and tritium gases are occluded in uranium powder for release into neutron generator tubes. The uranium powder is contained in stainless steel bottles, termed ''beds.'' If these beds become damaged, the gases must be removed and the uranium oxidized in order not to be flammable before shipment to ERDA disposal grounds. This paper describes the system and methods designed for the controlled degassing and oxidation process. The system utilizes sputter-ion, cryo-sorption and bellows pumps for removing the gases from the heated source bed. Removing the tritium gas is complicated by the shielding effect of helium-3, a byproduct of tritium decay. This effect is minimized by incremental pressure changes, or ''batch'' processing. To prevent runaway exothermic reaction, oxidation of the uranium bed is also done incrementally, or by ''batch'' processing, rather than by continuous flow. The paper discusses in detail the helium-3 shielding effect, leak checks that must be made during processing, bed oxidation, degree of gas depletion, purity of gases sorbed from beds, radioactivity of beds, bed disposal and system renovation

  13. REMOVAL OF URANIUM FROM DRINKING WATER BY CONVENTIONAL TREATMENT METHODS

    Science.gov (United States)

    The USEPA currently does not regulate uranium in drinking water but will be revising the radionuclide regulations during 1989 and will propose a maximum contaminant level for uranium. The paper presents treatment technology information on the effectiveness of conventional method...

  14. Process for recovering uranium from wet process phosphoric acid

    International Nuclear Information System (INIS)

    Pyrih, R.Z.; Rickard, S.; Carrington, F.

    1982-01-01

    A process for recovering uranium from phosphoric acid solutions uses an acidified alkali metal carbonate solution for the second-stage strip of uranyl uranium from the ion-exchange solution. The stripped solution is then recycled to the ion-exchange circuit. In the first stripping stage the ion-exchange solution containing the recovered uranyl uranium and an inert organic diluent is stripped with ammonium carbonate, producing a slurry of ammonium uranyl tricarbonate. The second strip, with a solution of 50-200 grams per litre of sodium carbonate eliminates the problems of inadequate removal of phosphorus, iron and vanadium impurities, solids accumulation, and phase separation in the strip circuit

  15. Uranium manufacturing process employing the electrolytic reduction method

    International Nuclear Information System (INIS)

    Oda, Yoshio; Kazuhare, Manabu; Morimoto, Takeshi.

    1986-01-01

    The present invention related to a uranium manufacturing process that employs the electrolytic reduction method, but particularly to a uranium manufacturing process that employs an electrolytic reduction method requiring low voltage. The process, in which uranium is obtained by means of the electrolytic method and with uranyl acid as the raw material, is prior art

  16. Process for recovering uranium from wet process phosphoric acid (III)

    International Nuclear Information System (INIS)

    Pyrih, R.Z.; Rickard, R.S.; Carrington, O.F.

    1983-01-01

    Uranium is conventionally recovered from wet-process phosphoric acid by two liquid ion exchange steps using a mixture of mono- and disubstituted phenyl esters of orthophosphoric acid (OPPA). Efficiency of the process drops as the mono-OPPA is lost preferentially to the aqueous phase. This invention provides a process for the removal of the uranium process organics (OPPA and organic solvents) from the raffinate of the first liquid ion exchange step and their return to the circuit. The process organics are removed by a combination flotation and absorption step, which results in the recovery of the organics on beads of a hydrophobic styrene polymer

  17. Method of removing uranium and its compounds from mine wastewaters and from aqueous solutions discharged in hydrometallurgical uranium ore treatment

    International Nuclear Information System (INIS)

    Jilek, R.; Prochazka, H.; Kuhr, I.; Fuska, J.; Nemec, P.; Katzer, J.

    1974-01-01

    The separation of uranium and its compounds from mine wastewaters and from water solutions discharged from uranium ore hydrometallurgical treatment, and its eventual simultaneous concentration in the biomass during uranium ore technological processing are described. The solutions are replenished with nutrients necessary for the growth of microorganisms, mainly with nitrogen, carbon and phosphorus and inoculated with fungi. During submersion cultivation, uranium incorporates in the mycelium, or is bound physico-chemically to the mycelium components. Together with these components, uranium is mechanically separated, i.e., by filtration, centrifugation or sedimentation. Organisms of the Fungi imperfecti class, mainly the Aspergillus and Penicillium genera are used for cultivation which may be continuous or semicontinuous. (B.S.)

  18. Initial process development for uranium bioprecipitation

    International Nuclear Information System (INIS)

    Truex, M.; Peyton, B.; Gorby, Y.; Valentine, N.

    1994-01-01

    Some bacteria can destabilize soluble metal complexes by enzymatically reducing the metal to a valence state where insoluble compounds are formed. For instance, oxidized uranium (VI) is highly soluble, but it precipitates from solution as the U(IV) oxide uraninite after microbial reduction. The advantage of this technology is that the uranium is easily separated from the aqueous phase, resulting in a small volume of relatively pure uraninite waste. A dissimilatory iron-reducing bacterium capable of uranium reduction was found to have a maximum growth rate of 0.142/hr, a Monod half-saturation constant of 3.4 mg/L, and a cellular yield of 0.071 mg-biomass/mg-iron for iron reduction at 30 C and pH 6.8. The kinetics of iron reduction were used to predict the performance of several reactor configurations for reduction of metals of interest such as uranium. A stirred-tank reactor in series with a plug-flow reactor was determined to be the best configuration for application of the bioprecipitation technology in a continuous-flow process

  19. Aerodynamic isotope separation processes for uranium enrichment: process requirements

    International Nuclear Information System (INIS)

    Malling, G.F.; Von Halle, E.

    1976-01-01

    The pressing need for enriched uranium to fuel nuclear power reactors, requiring that as many as ten large uranium isotope separation plants be built during the next twenty years, has inspired an increase of interest in isotope separation processes for uranium enrichment. Aerodynamic isotope separation processes have been prominently mentioned along with the gas centrifuge process and the laser isotope separation methods as alternatives to the gaseous diffusion process, currently in use, for these future plants. Commonly included in the category of aerodynamic isotope separation processes are: (a) the separation nozzle process; (b) opposed gas jets; (c) the gas vortex; (d) the separation probes; (e) interacting molecular beams; (f) jet penetration processes; and (g) time of flight separation processes. A number of these aerodynamic isotope separation processes depend, as does the gas centrifuge process, on pressure diffusion associated with curved streamlines for the basic separation effect. Much can be deduced about the process characteristics and the economic potential of such processes from a simple and elementary process model. In particular, the benefit to be gained from a light carrier gas added to the uranium feed is clearly demonstrated. The model also illustrates the importance of transient effects in this class of processes

  20. Uranium removal from organic solutions of PUREX process

    International Nuclear Information System (INIS)

    Dell'Occhio, L.A.; Dupetit, G.A.; Pascale, A.A.; Vicens, H.E.

    1987-01-01

    During the uranium extraction process with tributyl phosphate (TBP) in nitric medium, a bi solvated, non hydrated complex is formed, of formula UO2(NO3)2TBP, which is soluble in the diluent, a paraffin hydrocarbon. As it is known that some uranium salts, for instance the nitrate, when dissolved in organic solvents, like isopropanol, can be discharged as complex molecules at the cathode of an electrodeposition cell, it was decided to apply this technique to uranium loaded TBP solutions. From preliminary experiments resulted a practical possibility for the analytical control through the alpha measurement of electro deposits. This technique could be applied as well to the treatment of depleted organic streams carrying undesirable alpha activity, because the so treated solutions become deprived of uranium. This work presents the curves obtained working at constant voltage with uranium-loaded TBP solutions, the determination of the optimal operation voltage in these conditions, the electrodeposition yield for electro polished copper and stainless steel cathodes and the tests of reproducibility of deposits. A summary of the results obtained operating the high voltage supply at constant power is also presented. (Author)

  1. Application of ion-exchange unit in uranium extraction process in China (to be continued)

    International Nuclear Information System (INIS)

    Gong Chuanwen

    2004-01-01

    The application conditions of five different ion exchange units in uranium milling plant and wastewater treatment plant of uranium mine in China are introduced, including working parameters, existing problems and improvements. The advantages and disadvantages of these units are reviewed briefly. The procedure points to be followed in selecting ion exchange unit are recommended in the engineering design. The primary views are presented upon the application prospects of some ion exchange units in uranium extraction process in China

  2. Uranium/plutonium and uranium/neptunium separation by the Purex process using hydroxyurea

    International Nuclear Information System (INIS)

    Zhu Zhaowu; He Jianyu; Zhang Zefu; Zhang Yu; Zhu Jianmin; Zhen Weifang

    2004-01-01

    Hydroxyurea dissolved in nitric acid can strip plutonium and neptunium from tri-butyl phosphate efficiently and has little influence on the uranium distribution between the two phases. Simulating the 1B contactor of the Purex process by hydroxyurea with nitric acid solution as a stripping agent, the separation factors of uranium/plutonium and uranium/neptunium can reach values as high as 4.7 x 10 4 and 260, respectively. This indicates that hydroxyurea is a promising salt free agent for uranium/plutonium and uranium/neptunium separations. (author)

  3. Process for the in-situ leaching of uranium

    International Nuclear Information System (INIS)

    Habib, E.T.; Vogt, T.C.

    1982-01-01

    Process for the in-situ leaching of uranium employing an alkaline lixiviant and an alkali metal or alkaline earth metal hypochlorite as an oxidizing agent. The use of the hypochlorite oxidant results in significantly higher uranium recoveries and leaching rates than those attained by the use of conventional oxidants. The invention is particularly suitable for use in subterranean deposits in which the uranium mineral is associated with carbonaceous material which retards access to the uranium by the lixiviant

  4. Aftermath of Uranium Ore Processing on Floodplains: Lasting Effects of Uranium on Soil and Microbes

    Science.gov (United States)

    Tang, H.; Boye, K.; Bargar, J.; Fendorf, S. E.

    2016-12-01

    A former uranium ore processing site located between the Wind River and the Little Wind River near the city of Riverton, Wyoming, has generated a uranium plume in the groundwater within the floodplain. Uranium is toxic and poses a threat to human health. Thus, controlling and containing the spread of uranium will benefit the human population. The primary source of uranium was removed from the processing site, but a uranium plume still exists in the groundwater. Uranium in its reduced form is relatively insoluble in water and therefore is retained in organic rich, anoxic layers in the subsurface. However, with the aid of microbes uranium becomes soluble in water which could expose people and the environment to this toxin, if it enters the groundwater and ultimately the river. In order to better understand the mechanisms controlling uranium behavior in the floodplains, we examined sediments from three sediment cores (soil surface to aquifer). We determined the soil elemental concentrations and measured microbial activity through the use of several instruments (e.g. Elemental Analyzer, X-ray Fluorescence, MicroResp System). Through the data collected, we aim to obtain a better understanding of how the interaction of geochemical factors and microbial metabolism affect uranium mobility. This knowledge will inform models used to predict uranium behavior in response to land use or climate change in floodplain environments.

  5. Selective arsenical purification of substances during an alkaline treatment process of an uranium and/or molybdenum bearing ore by means of a magnesium compound

    International Nuclear Information System (INIS)

    Maurel, Pierre; Lamerant, J.M.; Pallez, Francois.

    1983-01-01

    The ores is digested by means of an aqueous liquor of sodium or potassium carbonate and/or bicarbonate, the digestion being carried out under conditions of concentrations, temperatures and pressures bringing about the solubilization of the uranium and/or molybdenum and the arsenic present in the core. A solid phase suspension is lifted from a liquid phase and the phases are separated. The arsenic solubilized during the digestion is extracted as magnesium arsenate by treatment of the medium containing the arsenic by means of a magnesium compound [fr

  6. Study of the dry processing of uranium ores

    International Nuclear Information System (INIS)

    Guillet, H.

    1959-02-01

    A description is given of direct fluorination of pre-concentrated uranium ores in order to obtain the hexafluoride. After normal sulfuric acid treatment of the ore to eliminate silica, the uranium is precipitated by a load of lime to obtain: either impure calcium uranate of medium grade, or containing around 10% of uranium. This concentrate is dried in an inert atmosphere and then treated with a current of elementary fluorine. The uranium hexafluoride formed is condensed at the outlet of the reaction vessel and may be used either for reduction to tetrafluoride and the subsequent manufacture of uranium metal or as the initial product in a diffusion plant. (author) [fr

  7. Treatment of uranium mining and milling wastewater using biological adsorbents

    International Nuclear Information System (INIS)

    Tsezos, M.

    1983-01-01

    Selected samples of waste microbial biomass originating from various industrial fermentation processes and biological treatment plants have been screened for biosorbent properties in conjunction with uranium, thorium and radium in aqueous solutions. Biosorption isotherms were used for the evaluation of biosorptive uptake capacity of the biomass. The biomass was also compared to synthetic adsorbents such as activated carbon. Determined uranium, thorium and radium biosorption isotherms were independent of the initial solution concentrations. Solution pH affected uptake. Rhizopus arrhizus at pH 4 exhibited the highest uranium and thorium biosorptive uptake capacity in excess of 180 Mg/g. It removed about 2.5 and 3.3 times more uranium than the ion exchange resin and activated carbon tested. Penicillium chrysogenum adsorbed 50000 pCi/g radium at pH 7 and at an equilibrium radium concentration of 1000 pCi/L. The most effective biomass types studied exhibited removals in excess of 99% of the radium in solution

  8. Ranstad - A new uranium-processing plant

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, A [AB Atomenergi, Stockholm (Sweden)

    1967-06-15

    A short outline is given of the decisions concerning the erection and operation of the Ranstad mill which was recently taken into operation. It is followed by a brief description of the geological conditions and the planning of the mining system, plant location, and the factory. The main part of the paper describes processes and equipment of the plant which has a capacity to treat approx. 850 000 tons of low-grade ore (alum shale) per year. The operational experience so far is also reviewed. The economy of uranium production at Ranstad is discussed and some development possibilities are indicated. (author)

  9. Modelling a uranium ore bioleaching process

    International Nuclear Information System (INIS)

    Chien, D.C.H.; Douglas, P.L.; Herman, D.H.; Marchbank, A.

    1990-01-01

    A dynamic simulation model for the bioleaching of uranium ore in a stope leaching process has been developed. The model incorporates design and operating conditions, reaction kinetics enhanced by Thiobacillus ferroxidans present in the leaching solution and transport properties. Model predictions agree well with experimental data with an average deviation of about ± 3%. The model is sensitive to small errors in the estimates of fragment size and ore grade. Because accurate estimates are difficult to obtain a parameter estimation approach was developed to update the value of fragment size and ore grade using on-line plant information

  10. PROCESS FOR THE PURIFICATION OF URANIUM

    Science.gov (United States)

    Rosenfeld, S.

    1959-01-20

    A proccss is described for reclaiming uranium values from aqueous solutions containing U, Fe, Ni, Cu, and Cr comprising treating the solution with NH/sub 3/ to precipitate the: U, Fc, and Cr and leaving Cu and Ni in solution as ammonia complex ions. The precipitate is chlorinated with CCl/sub 4/ at an elevated temperature to convert the U, Tc, and Cr into their chlorides. The more volatile FeCl/sub 3/ and CrCl/sub 3/ are separated from the UCl/sub 4/. The process is used when U is treated in a calutron, and composite solutions are produccd which contain dissolved products of stainless steel.

  11. Alternative processes for uranium recovery from phosphoric acid

    International Nuclear Information System (INIS)

    Duarte Neto, J.; Santos Benedetto, J. dos; Aquino, J.A. de

    1987-01-01

    Two processes of solvent extraction using D 2 EHPATOPO synergistic mixture, in order to recover uranium from phosphoric acid proceeding from physical and chemical treatments of the phosphorus-uraniferous ore of Itataia-CE, Brazil, are studied. The steps of each process were studied in laboratory and pilot scales. The flow charts for both processes with detailed description of each step, the operational conditions, the mass balances, the results obtained and the description of pilot units, are presented. (M.C.K.) [pt

  12. Study On The Uranium Adsorption Capability Of Bone Black In Radioactive Waste Water Treatment

    International Nuclear Information System (INIS)

    Phan Dinh Tuan

    2008-01-01

    It has been found that bone black can adsorb uranium and radium from radioactive wastewater. Nevertheless, bone black is not so competitive for the low adsorption capability and the slow adsorption rate. The article describes the research results in increasing the uranium adsorption capability of bone black by treating it with hydrochloric acid. The influences of pH on adsorption capability and the results of batch- and column tests have been investigated. Column tests for elution process have pointed out that HCl is quite good eluent for uranium. It is recommended to apply the treated bone black for radioactive wastewater treatment and uranium recovery. (author)

  13. Status Report from the United Kingdom [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    North, A A [Warren Spring Laboratory, Stevenage, Herts. (United Kingdom)

    1967-06-15

    The invitation to present this status report could have been taken literally as a request for information on experience gained in the actual processing of low-grade uranium ores in the United Kingdom, in which case there would have been very little to report; however, the invitation naturally was considered to be a request for a report on the experience gained by the United Kingdom of the processing of uranium ores. Lowgrade uranium ores are not treated in the United Kingdom simply because the country does not possess any known significant deposits of uranium ore. It is of interest to record the fact that during the nineteenth century mesothermal vein deposits associated with Hercynian granite were worked at South Terras, Cornwall, and ore that contained approximately 100 tons of uranium oxide was exported to Germany. Now only some 20 tons of contained uranium oxide remain at South Terras; also in Cornwall there is a small number of other vein deposits that each hold about five tons of uranium. Small lodes of uranium ore have been located in the southern uplands of Scotland; in North Wales lower palaeozoic black shales have only as much as 50 to 80 parts per million of uranium oxide, and a slightly lower grade carbonaceous shale is found near the base of the millstone grit that occurs in the north of England. Thus the experience gained by the United Kingdom has been of the treatment of uranium ores that occur abroad.

  14. Correlation of radioactive-waste-treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: conversion of yellow cake to uranium hexafluoride. Part II. The solvent extraction-fluorination process

    Energy Technology Data Exchange (ETDEWEB)

    Sears, M.B.; Etnier, E.L.; Hill, G.S.; Patton, B.D.; Witherspoon, J.P.; Yen, S.N.

    1983-03-01

    A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials and chemicals from a model uranium hexafluoride (UF/sub 6/) production plant using the solvent extraction-fluorination process, and to evaluate the radiological impact (dose commitment) of the release materials on the environment. The model plant processes 10,000 metric tons of uranium per year. Base-case waste treatment is the minimum necessary to operate the process. Effluents meet the radiological requirements listed in the Code of Federal Regulations, Title 10, Part 20 (10 CFR 20), Appendix B, Table II, but may not be acceptable chemically at all sites. Additional radwaste treatment techniques are applied to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The costs for the added waste treatment operations and the corresponding dose committment are correlated with the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases will require development and demonstration, or else is proprietary and unavailable for immediate use. The methodology and assumptions for the radiological doses are found in ORNL-4992.

  15. Correlation of radioactive-waste-treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: conversion of yellow cake to uranium hexafluoride. Part II. The solvent extraction-fluorination process

    International Nuclear Information System (INIS)

    Sears, M.B.; Etnier, E.L.; Hill, G.S.; Patton, B.D.; Witherspoon, J.P.; Yen, S.N.

    1983-03-01

    A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials and chemicals from a model uranium hexafluoride (UF 6 ) production plant using the solvent extraction-fluorination process, and to evaluate the radiological impact (dose commitment) of the release materials on the environment. The model plant processes 10,000 metric tons of uranium per year. Base-case waste treatment is the minimum necessary to operate the process. Effluents meet the radiological requirements listed in the Code of Federal Regulations, Title 10, Part 20 (10 CFR 20), Appendix B, Table II, but may not be acceptable chemically at all sites. Additional radwaste treatment techniques are applied to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The costs for the added waste treatment operations and the corresponding dose committment are correlated with the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases will require development and demonstration, or else is proprietary and unavailable for immediate use. The methodology and assumptions for the radiological doses are found in ORNL-4992

  16. Ore-processing technology and the uranium supply outlook

    International Nuclear Information System (INIS)

    James, H.E.; Simonsen, H.A.

    1978-01-01

    The subject is covered in sections, as follows: the resource base (uranium content of rocks, regional distribution of Western World uranium); ore types (distribution of Western World uranium, by ore types, response to ore-processing); constraints on expansion in traditional uranium areas (defined for this paper as the sandstone deposits of the U.S.A. and the quartz-pebble conglomerates of the Witwatersrand and Elliot Bay areas, all other deposits being referred to as new uranium areas). Sections then follow dealing in detail with the processing of deposits in U.S.A., South Africa, Canada, Niger, Australia, South West Africa, Greenland. More general sections follow on: shale, lignite and coal deposits, calcrete deposits. Finally, there are sections on: uranium as a by-product; uranium from very low-grade resources; constraints on expansion rate for production facilities. (U.K.)

  17. Treatment of the acid mine drainage residue for uranium recovery

    International Nuclear Information System (INIS)

    Dias, M.M.; Horta, D.G.; Fukuma, H.T.; Villegas, R.A.S.; Carvalho, C.H.T. de; Silva, A.C. da

    2017-01-01

    Acid mine drainage (AMD) is a process that occurs in many mining that have sulfide ores. With water and oxygen, several metals are oxidized, one example being uranium. At the mine pit of the Osamu Utsumi Mine located at INB - Caldas and in two other boot-wastes (mining waste pile), AMD is present and currently, without a technological solution. The acidic water present in the pit is treated with hydrated lime, generating water for disposal and an alkaline residue called calcium diuranate - DUCA. The DUCA has a concentration of approximately 0.32% U 3 O 8 , which makes interesting the development of a process for extracting that metal. One of the processes that can be used is leaching. For this study, it was decided to evaluate the alkaline leaching to extract the uranium present in the residue. It is necessary to optimize operational parameters for the process: percentage of solids, concentration of leaching agent in solution, temperature and reaction time. With these parameters, it is possible to improve the leaching so that the largest amount of uranium is extracted from the sample, to help solve the environmental impact caused by the wastewater from the treatment of acid waters and, in addition, to give an economical destination for this metal that is contained in the deposited DUCA

  18. Rejuvenation processes applied to 'poisoned' anion exchangers in uranium processing

    International Nuclear Information System (INIS)

    Gilmore, A.J.

    1979-11-01

    The removal of 'poisons' from anion exchangers in uranium processing of Canadian radioactive ores is commonly called rejuvenation or regeneration. The cost of the ion exchange recovery of uranium is adversely affected by a decrease in the capacity and efficiency of the anion exchangers, due to their being 'poisoned' by silica, elemental sulphur, molybdenum and tetrathionates. These 'poisons' have a high affinity for the anion exchangers, are adsorbed in preference to the uranyl complex, and do not desorb with the reagents used normally in the uranyl desorption phase. The frequency of rejuvenation and the reagents required for rejuvenation are determined by the severity of the 'poisoning' accumulated by the exchanger in contact with the uranium leach liquor. Caustic soda (NaOH) at approximately equal to 18 cents/lb is commonly used to remove uranium anion exchangers of tetrathionate ((S 4 0 6 )/-/-) 'poisons'. A potential saving in operating cost would be of consequence if other reagents, e.g. sodium carbonate (Na 2 CO 3 ) at approximately equal to 3.6 cents/lb or calcium hydroxide (Ca(OH) 2 ) at approximately equal to 1.9 cents/lb, were effective in removing (S 4 0 6 )/-/-) from a 'poisoned' exchanger. A rejuvenation process for a test program was adopted after a perusal of the literature

  19. Development of Practical Remediation Process for Uranium-Contaminated Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. S.; Kim, W. S.; Kim, G. N.; Moon, J. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A volume reduction of the concrete waste by the appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a 100 drums/year decontamination process and facilities for the decontamination of radioactive concrete. This practical scale process is little known. A practical decontamination process was developed to remove uranium from concrete pieces generated from the decommissioning of a uranium conversion plant. The concrete pieces are divided into two groups: concrete coated with and without epoxy. For the removal of epoxy from the concrete, direct burning by an oil flame is preferable to an electric heating method. The concrete blocks are crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm are sequentially washed with a clear washing solution and 1.0 M of nitric acid, most of their radioactivity reaches below the limit value of uranium for self-disposal. The concrete pieces smaller than 1 mm are decontaminated in a rotary washing machine by nitric acid, and an electrokinetic equipment is also used if their radioactivity is high.

  20. Development of Practical Remediation Process for Uranium-Contaminated Concrete

    International Nuclear Information System (INIS)

    Kim, S. S.; Kim, W. S.; Kim, G. N.; Moon, J. K.

    2013-01-01

    A volume reduction of the concrete waste by the appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a 100 drums/year decontamination process and facilities for the decontamination of radioactive concrete. This practical scale process is little known. A practical decontamination process was developed to remove uranium from concrete pieces generated from the decommissioning of a uranium conversion plant. The concrete pieces are divided into two groups: concrete coated with and without epoxy. For the removal of epoxy from the concrete, direct burning by an oil flame is preferable to an electric heating method. The concrete blocks are crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm are sequentially washed with a clear washing solution and 1.0 M of nitric acid, most of their radioactivity reaches below the limit value of uranium for self-disposal. The concrete pieces smaller than 1 mm are decontaminated in a rotary washing machine by nitric acid, and an electrokinetic equipment is also used if their radioactivity is high

  1. Study on the uranium-cerium extraction and his application to the treatment of irradiated uranium

    International Nuclear Information System (INIS)

    Lobao, Afonso dos Santos Tome

    1979-01-01

    It was made a study on the behavior of uranium and cerium(IV) extraction, using the latter element as a plutonium simulator in a flowsheet of the treatment of irradiated uranium. Cerium(IV) was used under the same conditions as a plutonium in the Purex process because the admitted similar properties. An experimental work was initiated to determine the equilibrium curves of uranium, under the following conditions: concentration of 1 to 20 g U/1 and acidity varying from 1 to 5M in HNO 3 . Other parameters studied were the volumetric ratio of the phases and the influence of the concentration of TBP (tri-n-butyl phosphate). To guarantee the cerium(IV) extraction, the diluent (varsol) was previously treated with 10% potassium dichromate in perchloric acid, potassium permanganate in 1M sulphuric acid and concentrated sulphuric acid at 70 deg to eliminate reducing compounds. The results obtained for cerium extraction, allowed a better understanding of its behavior in solution. The results permitted to conclude that the decontamination for cerium are very high in the first Purex extraction cycle. The easy as cerium(IV) is reduced to the trivalent state contributes a great deal to its decontamination. (author)

  2. Phosphorus and uranium recovery process from phosphated rocks

    Energy Technology Data Exchange (ETDEWEB)

    Sze, M C.Y.; Long, R H

    1981-01-30

    Improvement of uranium recovery in phosphate rocks by treatment with nitric acid avoiding the formation of a precipitate including a part of the uranium. The separation of uranium from phosphoric acid is obtained by liquid-liquid extraction using dialkyl posphoric acid with at least 10 carbon atoms and a phosphoryl alkyl alkoxy compound with at least 10 carbon atoms and a non water miscible organic solvent.

  3. Process for recovering a uranium containing concentrate and purified phosphoric acid from a wet process phosphoric acid containing uranium

    International Nuclear Information System (INIS)

    Weterings, C.A.M.; Janssen, J.A.

    1985-01-01

    A process is claimed for recovering from a wet process phosphoric acid which contains uranium, a uranium containing concentrate and a purified phosphoric acid. The wet process phosphoric acid is treated with a precipitant in the presence of a reducing agent and an aliphatic ketone

  4. Process for recovering a uranium containing concentrate and purified phosphoric acid from a wet process phosphoric acid containing uranium

    Energy Technology Data Exchange (ETDEWEB)

    Weterings, C.A.M.; Janssen, J.A.

    1985-04-30

    A process is claimed for recovering from a wet process phosphoric acid which contains uranium, a uranium containing concentrate and a purified phosphoric acid. The wet process phosphoric acid is treated with a precipitant in the presence of a reducing agent and an aliphatic ketone.

  5. Developments on uranium enrichment processes in France

    International Nuclear Information System (INIS)

    Frejacques, C.; Gelee, M.; Massignon, D.; Plurien, P.

    1977-01-01

    Gaseous diffusion has so far been the main source of supply for enriched uranium and it is only recently that the gas centrifuge came into the picture. Numerous other isotope separation processes have been considered or are being assessed, and there is nothing to exclude the future use of a new process. Developments on likely new processes have been carried out by many organizations both governmental and private. The French Commissariat a l'energie atomique, besides their very extensive endeavours already devoted to gaseous diffusion, have studied and developed the gas centrifuge, chemical exchange, aerodynamic and selective photoexcitation processes. The gaseous diffusion process, selected by Eurodif for the Tricastin plant, and which will also be used by Coredif, is discussed in another paper in these Proceedings. This process is the technico-economic yardstick on which our comparisons are based. Within the limits of their development level, processes are compared on the basis of the separative work cost components: specific investment, specific power consumption and power cost, and specific operating and maintenance costs. (author)

  6. Influence of attrition scrubbing, ultrasonic treatment, and oxidant additions on uranium removal from contaminated soils

    International Nuclear Information System (INIS)

    Timpson, M.E.; Elless, M.P.; Francis, C.W.

    1994-01-01

    As part of the Uranium in Soils Integrated Demonstration Project being conducted by the US Department of Energy, bench-scale investigations of selective leaching of uranium from soils at the Fernald Environmental Management Project site in Ohio were conducted at Oak Ridge National Laboratory. Two soils (storage pad soil and incinerator soil), representing the major contaminant sources at the site, were extracted using carbonate- and citric acid-based lixiviants. Physical and chemical processes were used in combination with the two extractants to increase the rate of uranium release from these soils. Attrition scrubbing and ultrasonic dispersion were the two physical processes utilized. Potassium permanganate was used as an oxidizing agent to transform tetravalent uranium to the hexavalent state. Hexavalent uranium is easily complexed in solution by the carbonate radical. Attrition scrubbing increased the rate of uranium release from both soils when compared with rotary shaking. At equivalent extraction times and solids loadings, however, attrition scrubbing proved effective only on the incinerator soil. Ultrasonic treatments on the incinerator soil removed 71% of the uranium contamination in a single extraction. Multiple extractions of the same sample removed up to 90% of the uranium. Additions of potassium permanganate to the carbonate extractant resulted in significant changes in the extractability of uranium from the incinerator soil but had no effect on the storage pad soil

  7. Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R M

    1976-01-01

    Evidence of expanding markets, improved prices and the short supply of uranium became abundantly clear in 1975, providing the much needed impetus for widespread activity in all phases of uranium operations. Exploration activity that had been at low levels in recent years in Canada was evident in most provinces as well as the Northwest Territories. All producers were in the process of expanding their uranium-producing facilities. Canada's Atomic Energy Control Board (AECB) by year-end had authorized the export of over 73,000 tons of U/sub 3/0/sub 8/ all since September 1974, when the federal government announced its new uranium export guidelines. World production, which had been in the order of 25,000 tons of U/sub 3/0/sub 8/ annually, was expected to reach about 28,000 tons in 1975, principally from increased output in the United States.

  8. Behavior of radioactive elements (uranium and thorium) in Bayer process

    International Nuclear Information System (INIS)

    Sato, C.; Kazama, S.; Sakamoto, A.; Hirayanagi, K.

    1986-01-01

    It is essential that alumina used for manufacturing electronic devices should contain an extremely low level of alpha-radiation. The principal source of alpha-radiation in alumina is uranium, a minor source being thorium. Uranium in bauxite dissolves into the liquor in the digestion process and is fixed to the red mud as the desilication reaction progresses. A part of uranium remaining in the liquor precipitates together with aluminum hydroxide in the precipitation process. The uranium content of aluminum hydroxide becomes lower as the precipitation velocity per unit surface area of the seed becomes slower. Organic matters in the Bayer liquor has an extremely significant impact on the uranium content of aluminum hydroxide. Aluminum hydroxide free of uranium is obtainable from the liquor that does not contain organic matters

  9. Process of recovering uranium from wet process acid

    International Nuclear Information System (INIS)

    York, W.R.

    1983-01-01

    Entrainment of contaminated water in the organic phase and poor phase disengagement is prevented in the second cycle scrubber, in a two cycle uranium recovery process, by washing the organic solvent stream containing entrained H 3 PO 4 from the second cycle extractor, with a dilute aqueous sulfuric or nitric acid solution in an acid scrubber, prior to passing the solvent stream into the second cycle stripper. (author)

  10. Uranium ore processing minimizing reagent losses

    International Nuclear Information System (INIS)

    Shaogiang, Chen; Moret, J.; Lyaudet, G.

    1989-01-01

    The uranium ore is treated by sodium carbonates and the solution is divided in two parts: a production solution which is decarbonated by an acid before uranium precipitation with sodium hydroxide and a recycling solution directly treated by sodium hydroxide for precipitation of about 85% of uranium and total transformation of sodium bicarbonate into sodium carbonate, the quantity of sodium hydroxide used on the recycling solution brings sodium ions required for attack of the ore [fr

  11. Treatment of pit water from uranium mining operation

    International Nuclear Information System (INIS)

    Mouton, A.; Lafforgue, P.; Lyaudet, G.

    1984-01-01

    The pit water from uranium mines is normally treated to eliminate the soluble radium and suspended solids. The radium is precipitated together with the barium sulphate. The latter results from the reaction of barium chloride with an excess of sulphate ions. The suspended solids are flocculated by aluminium salts (chloride, polychloride). If necessary, synthetic flocculants are also used. Certain grades of pit water contain, sometimes incidentally, a few milligrams of uranium per litre. These quantities always remain too low for any direct recovery (treatment by ion exchange resins). By applying certain measures, the preceding processes can also be used to eliminate uranium. The latter is carried away by aluminium hydroxide in a very narrow zone of pH (6 to 7,4) which corresponds to the minimum solubility of the hydroxide. Depending on the characteristic of the water (pH, salinity), use is made either of aluminium sulphate or of sodium aluminate, with an addition of a base in extreme cases. This article gives various examples of applications in the Haute-Vienne, Chardon in Vendee, the Commanderie mine in Vendee, at Cerilly in Allier and at Lodeve in Herault [fr

  12. Uranium and sulphate values from carbonate leach process

    International Nuclear Information System (INIS)

    Berger, B.

    1983-01-01

    The process concerns the recovery of uraniferous and sulphur values from liquor resulting from the attack of sulphur containing uraniferous ores by an alkaline solution of sodium carbonate and/or bicarbonate. Ammonia is introduced into the liquor to convert any HCO 3 - to CO 3 2- . The neutralised liquor from this step is then contacted with an anion exchange resin to fix the uranium and sulphate ions, leaving a liquor containing ammonia, sodium carbonate and/or bicarbonate in solution. Uranium and sulphate ions are eluted with an ammonia carbonate and/or bicarbonate solution to yield a solution of ammonium uranyl carbonate complex and ammonium sulphate. The solution is subjected to thermal treatment until a suspension of precipitated ammonium uranate and/or diuranate is obtained in a solution of the ammonium sulphate. Carbon dioxide, ammonia and water vapor are driven off. The precipitated ammonium uranate and/or diuranate is then separated from the solution of ammonium sulphate and the precipitate is calcined to yield uranium trioxide and ammonia

  13. PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

    Science.gov (United States)

    Moore, R.H.

    1962-10-01

    A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

  14. Process development study on production of uranium metal from monazite sourced crude uranium tetra-fluoride

    International Nuclear Information System (INIS)

    Chowdhury, S; Satpati, S.K.; Hareendran, K.N.; Roy, S.B.

    2014-01-01

    Development of an economic process for recovery, process flow sheet development, purification and further conversion to nuclear grade uranium metal from the crude UF 4 has been a technological challenge and the present paper, discusses the same.The developed flow-sheet is a combination of hydrometallurgical and pyrometallurgical processes. Crude UF 4 is converted to uranium di-oxide (UO 2 ) by chemical conversion route and UO 2 produced is made fluoride-free by repeated repulping, followed by solid liquid separation. Uranium di-oxide is then purified by two stages of dissolution and suitable solvent extraction methods to get uranium nitrate pure solution (UNPS). UNPS is then precipitated with air diluted ammonia in a leak tight stirred vessel under controlled operational conditions to obtain ammonium di-uranate (ADU). The ADU is then calcined and reduced to produce metal grade UO 2 followed by hydro-fluorination using anhydrous hydrofluoric acid to obtain metal grade UF 4 with ammonium oxalate insoluble (AOI) content of 4 is essential for critical upstream conversion process. Nuclear grade uranium metal ingot is finally produced by metallothermic reduction process at 650℃ in a closed vessel, called bomb reactor. In the process, metal-slag separation plays an important role for attaining metal purity as well as process yield. Technological as well economic feasibility of indigenously developed process for large scale production of uranium metal from the crude UF 4 has been established in Bhabha Atomic Research Centre (BARC), India

  15. Waste monitoring of the uranium ore processing activities in Romania

    International Nuclear Information System (INIS)

    Nica, L.

    2002-01-01

    The uranium ore processing activities at the Feldioara site produce a range of liquid and solid waste that are monitored. Liquids are treated through decantation, pH correction and uranium precipitation before their release into the environment. The solid waste is gathered into ore specific area and are covered regularly with clay materials. (author)

  16. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: conversion of yellow cake to uranium hexafluoride. Part I. The fluorination-fractionation process

    Energy Technology Data Exchange (ETDEWEB)

    Sears, M.B.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1977-07-01

    A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials and chemicals from a model uranium hexafluoride (UF/sub 6/) production plant using the fluorination-fractionation (dry hydrofluor) process, and to evaluate the radiological impact (dose commitment) of the released materials on the environment. This study is designed to assist in defining the term as low as is reasonably achievable (ALARA) in relation to limiting the release of radioactive materials from nuclear facilities. The model plant processes 10,000 metric tons of uranium per year. Base-case waste treatment is the minimum necessary to operate the process. Effluents meet the radiological requirements listed in the Code of Federal Regulations, Title 10, Part 20 (10 CFR 20), Appendix B, Table II, but may not be acceptable chemically at all sites. Additional radwaste treatment techniques are applied to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The costs for the added waste treatment operations and the corresponding dose commitment are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases will require development and demonstration or else is proprietary and unavailable for immediate use. The methodology and assumptions for the radiological doses are found in ORNL-4992.

  17. Correlation of radioactive waste treatment costs and the environmental impact of waste effluents in the nuclear fuel cycle: conversion of yellow cake to uranium hexafluoride. Part I. The fluorination-fractionation process

    International Nuclear Information System (INIS)

    Sears, M.B.; Blanco, R.E.; Finney, B.C.; Hill, G.S.; Moore, R.E.; Witherspoon, J.P.

    1977-07-01

    A cost/benefit study was made to determine the cost and effectiveness of radioactive waste (radwaste) treatment systems for decreasing the release of radioactive materials and chemicals from a model uranium hexafluoride (UF 6 ) production plant using the fluorination-fractionation (dry hydrofluor) process, and to evaluate the radiological impact (dose commitment) of the released materials on the environment. This study is designed to assist in defining the term as low as is reasonably achievable (ALARA) in relation to limiting the release of radioactive materials from nuclear facilities. The model plant processes 10,000 metric tons of uranium per year. Base-case waste treatment is the minimum necessary to operate the process. Effluents meet the radiological requirements listed in the Code of Federal Regulations, Title 10, Part 20 (10 CFR 20), Appendix B, Table II, but may not be acceptable chemically at all sites. Additional radwaste treatment techniques are applied to the base-case plant in a series of case studies to decrease the amounts of radioactive materials released and to reduce the radiological dose commitment to the population in the surrounding area. The costs for the added waste treatment operations and the corresponding dose commitment are calculated for each case. In the final analysis, radiological dose is plotted vs the annual cost for treatment of the radwastes. The status of the radwaste treatment methods used in the case studies is discussed. Much of the technology used in the advanced cases will require development and demonstration or else is proprietary and unavailable for immediate use. The methodology and assumptions for the radiological doses are found in ORNL-4992

  18. Some potential strategies for the treatment of waste uranium metal and uranium alloys

    International Nuclear Information System (INIS)

    Burns, C.J.; Frankcom, T.M.; Gordon, P.L.; Sauer, N.N.

    1993-01-01

    Large quantities of uranium metal chips and turnings stored throughout the DOE Complex represent a potential hazard, due to the reactivity of this material toward air and water. Methods are being sought to mitigate this by conversion of the metal, via room temperature solutions routes, to a more inert oxide form. In addition, the recycling of uranium and concomitant recovery of alloying metals is a desirable goal. The emphasis of the authors' research is to explore a variety of oxidation and reduction pathways for uranium and its compounds, and to investigate how these reactions might be applied to the treatment of bulk wastes

  19. Uranium

    International Nuclear Information System (INIS)

    Battey, G.C.; McKay, A.D.

    1988-01-01

    Production for 1986 was 4899 t U 3 O 8 (4154 t U), 30% greater than in 1985, mainly because of a 39% increase in production at Ranger. Exports for 1986 were 4166 t U 3 O 8 at an average f.o.b. unit value of $40.57/lb U 3 O 8 . Private exploration expenditure for uranium in Australia during the 1985-86 fiscal year was $50.2 million. Plans were announced to increase the nominal capacity of the processing plant at Ranger from 3000 t/year U 3 O 8 to 4500 t and later to 6000 t/year. Construction and initial mine development at Olympic Dam began in March. Production is planned for mid 1988 at an annual rate of 2000 t U 3 O 8 , 30 000 t Cu, and 90 000 oz (2800 kg) Au. The first long-term sales agreement was concluded in September 1986. At the Manyingee deposit, testing of the alkaline solution mining method was completed, and the treatment plant was dismantled. Spot market prices (in US$/lb U 3 O 8 ) quoted by Nuexco were generally stable. From January-October the exchange value fluctuated from US$17.00-US$17.25; for November and December it was US$16.75. Australia's Reasonably Assured Resources of uranium recoverable at less than US$80/kg U at December 1986 were estimated as 462 000 t U, 3000 t U less than in 1985. This represents 30% of the total low-cost RAR in the WOCA (World Outside the Centrally Planned Economy Areas) countries. Australia also has 257 000 t U in the low-cost Estimated Additional Resources Category I, 29% of the WOCA countries' total resources in this category

  20. Status Report from Czechoslovakia [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Civin, V; Belsky, M [Research and Development Laboratory No.3 of the Uranium Industry, Prague, Czechoslovakia (Czech Republic)

    1967-06-15

    The present paper deals with the fundamental problems and the main routes followed in processing low-grade uranium ores in CSSR. In this connection it may be useful to discuss the definition of low-grade ore. In our country this term is applied to uraniferous material with a very low content of uranium (of the order of 0.01%) whose treatment causes no particular difficulty. However, the same term is also used to designate those materials whose processibility lies on the verge of economic profitability. In our view, this classification, of an ore using two independent criteria (i.e. uranium content and processing economy) is useful from the standpoint of technology. The treatment of both such ore types is as a rule carried out by specific technological processes. Consequently, low-grade uranium ores can be divided into two groups: (1) Ores with a low uranium content. To this category belong in our country uraniferous materials which originate as a by-product of technological processes used in processing other materials. This is primarily gangue and tailings of various physical or physico-chemical pretreatment operations to which the ore is subjected at the mining site. Mention should be made in this connection of mine waters, which represent a useful complementary source of uranium despite their low uranium content (of the order of milligrams per litre). (2) Ores whose economical treatment is problematic. To this category belong deposits of conventional ore types with a uranium content on the limit of profitable treatment. Also, those deposits containing atypical materials possessing such properties which impair the economy of their treatment. This includes ores with a considerable amount of components which are difficult to separate and which at the same time consume the leaching agents. Finally, it covers uranium-bearing materials in refractory forms which are difficult to dissolve and also some special materials, such as lignites, uranium-bearing shales, loams

  1. Chattanooga shale: uranium recovery by in situ processing

    International Nuclear Information System (INIS)

    Jackson, D.D.

    1977-01-01

    The increasing demand for uranium as reactor fuel requires the addition of sizable new domestic reserves. One of the largest potential sources of low-grade uranium ore is the Chattanooga shale--a formation in Tennessee and neighboring states that has not been mined conventionally because it is expensive and environmentally disadvantageous to do so. An in situ process, on the other hand, might be used to extract uranium from this formation without the attendant problems of conventional mining. We have suggested developing such a process, in which fracturing, retorting, and pressure leaching might be used to extract the uranium. The potential advantages of such a process are that capital investment would be reduced, handling and disposing of the ore would be avoided, and leaching reagents would be self-generated from air and water. If successful, the cost reductions from these factors could make the uranium produced competitive with that from other sources, and substantially increase domestic reserves. A technical program to evaluate the processing problems has been outlined and a conceptual model of the extraction process has been developed. Preliminary cost estimates have been made, although it is recognized that their validity depends on how successfully the various processing steps are carried out. In view of the preliminary nature of this survey (and our growing need for uranium), we have urged a more detailed study on the feasibility of in situ methods for extracting uranium from the Chattanooga shale

  2. Research and application for wastewater treatment technology in a southern uranium mine

    International Nuclear Information System (INIS)

    Tan Jianhua; Zhao Jinfang; Huang Yunbai; Deng Jianguo

    2014-01-01

    This paper analyzes the source and property of a southern uranium mine's drainage and the treatment technology is tested, and proposed by employing the process of '408 (Ⅱ) resin adsorption-NaCl + NaHCO 3 elution '. The results show that the treated drainage can meet the emission requirement of Regulations for radiation and environment protection in uranium mining and milling (GB23727-2009), with the uranium content being less than 0.3 mg/L -l . The econo-technical norms such as material consumption are improved as the new technology has been applied in practical production. (authors)

  3. Innovative Elution Processes for Recovering Uranium from Seawater

    International Nuclear Information System (INIS)

    Wai, Chien; Tian, Guoxin; Janke, Christopher

    2014-01-01

    Utilizing amidoxime-based polymer sorbents for extraction of uranium from seawater has attracted considerable interest in recent years. Uranium collected in the sorbent is recovered typically by elution with an acid. One drawback of acid elution is deterioration of the sorbent which is a significant factor that limits the economic competitiveness of the amidoxime-based sorbent systems for sequestering uranium from seawater. Developing innovative elution processes to improve efficiency and to minimize loss of sorbent capacity become essential in order to make this technology economically feasible for large-scale industrial applications. This project has evaluated several elution processes including acid elution, carbonate elution, and supercritical fluid elution for recovering uranium from amidoxime-based polymer sorbents. The elution efficiency, durability and sorbent regeneration for repeated uranium adsorption- desorption cycles in simulated seawater have been studied. Spectroscopic techniques are used to evaluate chemical nature of the sorbent before and after elution. A sodium carbonate-hydrogen peroxide elution process for effective removal of uranium from amidoxime-based sorbent is developed. The cause of this sodium carbonate and hydrogen peroxide synergistic leaching of uranium from amidoxime-based sorbent is attributed to the formation of an extremely stable uranyl peroxo-carbonato complex. The efficiency of uranium elution by the carbonate-hydrogen peroxide method is comparable to that of the hydrochloric acid elution but damage to the sorbent material is much less for the former. The carbonate- hydrogen peroxide elution also does not need any elaborate step to regenerate the sorbent as those required for hydrochloric acid leaching. Several CO2-soluble ligands have been tested for extraction of uranium from the sorbent in supercritical fluid carbon dioxide. A mixture of hexafluoroacetylacetone and tri-n-butylphosphate shows the best result but uranium

  4. Innovative Elution Processes for Recovering Uranium from Seawater

    Energy Technology Data Exchange (ETDEWEB)

    Wai, Chien [Univ. of Idaho, Moscow, ID (United States); Tian, Guoxin [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Janke, Christopher [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-05-29

    Utilizing amidoxime-based polymer sorbents for extraction of uranium from seawater has attracted considerable interest in recent years. Uranium collected in the sorbent is recovered typically by elution with an acid. One drawback of acid elution is deterioration of the sorbent which is a significant factor that limits the economic competitiveness of the amidoxime-based sorbent systems for sequestering uranium from seawater. Developing innovative elution processes to improve efficiency and to minimize loss of sorbent capacity become essential in order to make this technology economically feasible for large-scale industrial applications. This project has evaluated several elution processes including acid elution, carbonate elution, and supercritical fluid elution for recovering uranium from amidoxime-based polymer sorbents. The elution efficiency, durability and sorbent regeneration for repeated uranium adsorption- desorption cycles in simulated seawater have been studied. Spectroscopic techniques are used to evaluate chemical nature of the sorbent before and after elution. A sodium carbonate-hydrogen peroxide elution process for effective removal of uranium from amidoxime-based sorbent is developed. The cause of this sodium carbonate and hydrogen peroxide synergistic leaching of uranium from amidoxime-based sorbent is attributed to the formation of an extremely stable uranyl peroxo-carbonato complex. The efficiency of uranium elution by the carbonate-hydrogen peroxide method is comparable to that of the hydrochloric acid elution but damage to the sorbent material is much less for the former. The carbonate- hydrogen peroxide elution also does not need any elaborate step to regenerate the sorbent as those required for hydrochloric acid leaching. Several CO2-soluble ligands have been tested for extraction of uranium from the sorbent in supercritical fluid carbon dioxide. A mixture of hexafluoroacetylacetone and tri-n-butylphosphate shows the best result but uranium

  5. PROCESS FOR THE RECOVERY AND PURIFICATION OF URANIUM DEPOSITS

    Science.gov (United States)

    Carter, J.M.; Kamen, M.D.

    1958-10-14

    A process is presented for recovering uranium values from UCl/sub 4/ deposits formed on calutrons. Such deposits are removed from the calutron parts by an aqueous wash solution which then contains the uranium values in addition to the following impurities: Ni, Cu, Fe, and Cr. This impurity bearing wash solution is treated with an oxidizing agent, and the oxidized solution is then treated with ammonia in order to precipitate the uranium as ammonium diuranate. The metal impurities of iron and chromium, which form insoluble hydroxides, are precipitated along with the uranium values. The precipitate is separated from the solution, dissolved in acid, and the solution again treated with ammonia and ammonium carbonate, which results in the precipitation of the metal impurities as hydroxides while the uranium values remain in solution.

  6. Uranium mining and processing: their radiation impact into the environment

    International Nuclear Information System (INIS)

    Ostapczuk, Peter; Zoriy, Petro; Dederichs, Herbert; Lennartz, Reinhard

    2008-01-01

    Based on Thorium and Uranium determination in soil and plants samples collected in the region of Aktau, Kazakhstan the distribution pattern of environmental pollution by these elements was correlated with the radiation dose. The main radiation source was the waste deposit of the equipment used by the uranium processing (dose higher than 5 μSv/h). The mining area and also the transportation way from mine to the uranium factory has also an radiation impact which is difficult to estimate. Based on the data found by plants and soil samples all the area under study has a higher pollution level by Thorium and Uranium than the control area (about 0.1μSv/h). Due to observed strong wind blowing in different directions it is possible that the particle of uranium ore has been transported for long distance and polluted the plants and upper soil layer. The further investigations should get more information about this supposition. (author)

  7. Application of physical separation techniques in uranium resources processing

    International Nuclear Information System (INIS)

    Padmanabhan, N.P.H.; Sreenivas, T.

    2008-01-01

    The planned economic growth of our country and energy security considerations call for increasing the overall electricity generating capabilities with substantial increase in the zero-carbon and clean nuclear power component. Although India is endowed with vast resources of thorium, its utilization can commence only after the successful completion of the first two stages of nuclear power programme, which use natural uranium in the first stage and natural uranium plus plutonium in the second stage. For the successful operation of first stage, exploration and exploitation activities for uranium should be vigorously followed. This paper reviews the current status of physical beneficiation in processing of uranium ores and discusses its applicability to recover uranium from low grade and below-cut-off grade ores in Indian context. (author)

  8. Biogeochemical Processes Regulating the Mobility of Uranium in Sediments

    Energy Technology Data Exchange (ETDEWEB)

    Belli, Keaton M.; Taillefert, Martial

    2016-07-01

    This book chapters reviews the latest knowledge on the biogeochemical processes regulating the mobility of uranium in sediments. It contains both data from the literature and new data from the authors.

  9. Flotation process of lead-, copper-, uranium-, and rare earth minerals

    International Nuclear Information System (INIS)

    Broman, P.G.; Kihlstedt, P.G.; Du Rietz, C.

    1977-01-01

    This invention relates to a flotation process of oxide or sulfide ores containing lead-, copper-, uranium-, and rare earth minerals applicating a new collector. Flotation is in the presence of a tertiary amine

  10. Process for winning uranium from wet process phosphoric acid

    International Nuclear Information System (INIS)

    1980-01-01

    A process is described for winning uranium from wet process phosphoric acid by means of liquid-liquid extraction with organic phosphoric acid esters. The process is optimised by keeping the sulphate percentage in the phosphoric acid below 2% by weight, and preferably below 0.6% by weight, as compared to P 2 O 5 in the phosphoric acid. This is achieved by adding an excess of Ba and/or Ca carbonate or sulfide solution and filtering off the formed calcium and/or barium sulphate precipitates. Solid KClO 3 is then added to the filtrate to oxidise U 4+ to U 6+ . The normal extraction procedure using organic phosphoric esters as extraction liquid, can then be applied. (Th.P.)

  11. Alternative leaching processes for uranium ores

    International Nuclear Information System (INIS)

    Ring, R.J.

    1979-01-01

    Laboratory studies have been carried out to compare the extraction of uranium from Australian ores by conventional leaching in sulphuric acid with that obtained using hydrochloric acid and acidified ferric sulphate solutions. Leaching with hydrochloric acid achieved higher extractions of radium-226 but the extraction of uranium was reduced considerably. The use of acidified ferric sulphate solution reduced acid consumption by 20-40% without any detrimental effect on uranium extraction. The ferric ion, which is reduced during leaching, can be reoxidized and recycled after the addition of acid makeup. Hydrogen peroxide was found to be an effective oxidant in conventional sulphuric acid leaching. It is more expensive than alternative oxidants, but it is non-polluting, lesser quantities are required and acid consumption is reduced

  12. Process for enriching uranium from seawater

    International Nuclear Information System (INIS)

    Heitkamp, D.; Inden, P.

    1982-01-01

    In selective elutriation of uranium deposited on titanium oxide hydrate by carbonate solution, only uranium should be dissolved from the absorption material forming carbonate compounds, without the deposited ballast ions, above all of magnesium, calcium and sodium being elutriated. The uranium elutriation according to the invention is therefore carried out in the presence of these ballast ions in the same concentrations as those in seawater. The carbonate concentration can only be raised as far as the solubility product of the basic magnesium carbonate permits, so that magnesium remains in the solution, as well as carbonate, in the concentration present in seawater. One must accept the absence of calcium ions in the elutriation solution, as their solubility product with carbonate is considerably less than that for magnesium. (orig./PW) [de

  13. Uranium

    International Nuclear Information System (INIS)

    Hamdoun, N.A.

    2007-01-01

    The article includes a historical preface about uranium, discovery of portability of sequential fission of uranium, uranium existence, basic raw materials, secondary raw materials, uranium's physical and chemical properties, uranium extraction, nuclear fuel cycle, logistics and estimation of the amount of uranium reserves, producing countries of concentrated uranium oxides and percentage of the world's total production, civilian and military uses of uranium. The use of depleted uranium in the Gulf War, the Balkans and Iraq has caused political and environmental effects which are complex, raising problems and questions about the effects that nuclear compounds left on human health and environment.

  14. Uranium recovery from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    McCullough, J.F.; Phillips, J.F. Jr.; Tate, L.R.

    1979-01-01

    A method of recovering uranium from wet-process phosphoric acid is claimed where the acid is treated with a mixture of an ammonium salt or ammonia, a reducing agent, and then a miscible solvent. Solids are separated from the phosphoric acid liquid phase. The solid consists of a mixture of metal phosphates and uranium. It is washed free of adhering phosphoric acid with fresh miscible solvent. The solid is dried and dissolved in acid whereupon uranium is recovered from the solution. Miscible solvent and water are distilled away from the phosphoric acid. The distillate is rectified and water discarded. All miscible solvent is recovered for recycle. 5 claims

  15. Uranium tetrafluoride production via dioxide by wet process

    International Nuclear Information System (INIS)

    Aquino, A.R. de.

    1988-01-01

    The study for the wet way obtention of uranium tetrafluoride by the reaction of hydrofluoric acid and powder uranium dioxide, is presented. From the results obtained at laboratory scale a pilot plant was planned and erected. It is presently in operation for experimental data aquisition. Time of reaction, temperature, excess of reagents and the hydrofluoric acid / uranium dioxide ratio were the main parameters studied to obtain a product with the following characteristics: - density greater than 1 g/cm 3 , conversion rate greater than 96%, and water content equal to 0,2% that allows its application to heaxafluoride convertion or to magnesiothermic process. (author) [pt

  16. Oxidizing attack process of uranium ore by a carbonated liquor

    International Nuclear Information System (INIS)

    Maurel, Pierre; Nicolas, Francois.

    1981-01-01

    A continuous process for digesting a uraniferous ore by oxidation with a recycling aqueous liquor containing alkaline carbonates and bicarbonates in solution as well as uranium in a concentration close to its solubility limit at digestion temperature, and of recuperation of the precipitated uranium within the solid phase remaining after digestion. The digestion is carried out by spraying oxygen into the hot reactional medium in order not only to permit oxidation of the uranium and its solubilization but also to ensure that the sulphides of impurities and organic substances present in the ore are oxidized [fr

  17. Oxidation-extraction of uranium from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    Lawes, B.C.

    1985-01-01

    The invention involves an improvement to the reductive stripping process for recovering uranium values from wet-process phosphoric acid solution, where uranium in the solution is oxidized to uranium (VI) oxidation state and then extracted from the solution by contact with a water immiscible organic solvent, by adding sufficient oxidant, hydrogen peroxide, to obtain greater than 90 percent conversion of the uranium to the uranium (VI) oxidation state to the phosphoric acid solution and simultaneously extracting the uranium (VI)

  18. Process for the preparation of uranium dioxide

    International Nuclear Information System (INIS)

    Watt, G.W.; Baugh, D.W. Jr.

    1981-01-01

    A method for the preparation of actinide dioxides using actinide nitrate hexahydrates as starting materials is described. The actinide nitrate hexahydrate is reacted with sodium dithionite, and the product is heated in the absence of oxygen to obtain the dioxide. Preferably, the actinide is uranium, plutonium or neptunium. (LL)

  19. Thirty years of uranium ore processing in Spain

    International Nuclear Information System (INIS)

    Josa, J.M.

    1982-01-01

    Spanish background in the uranium ore processing includes ores from pegmatitic type deposits, vein deposits, sandstone, enrichments in metamorphic rocks, radioactive coals and non-conventional sources of uranium, such as wet phosphoric acid or copper liquors. Some tests have also done in order to recover uranium from very low grade paleozoic quartzites. We have also been involved in by-products recovery (copper) from uranium ores. The technologies that have been used are: physical concentration, combustion and roasting, conventional alkaline or acid methods, pressure, heap and bacteria leaching. Special attention was paid to recover uranium from the pregnant liquors and to develop suited equipment for it; solvent extraction and continuous ion exchange equipment was carefully studied. We have been involved in commercial size (500-3000 t/d) mills, but we have also developed transportable and reussable modular plants specially designed and suited to recover uranium from small and isolated deposits. In both cases the reduction of the environmental impact was taken in account. Spanish experience also includes nuclear purification aspects in order to get uranium nuclear compounds (ADU, UO 2 , UF 4 and UF 6 ). Wet (nitric-TBP) and dry (Fluid-bed) methods have been used. The best of these 30 years of experience in studies and in industrial practice, together with our new developments towards the future, could become in a good contribution for the medium size countries which are going to develop its own uranium industry. The way for these countries could be easier if they know what is valuable and what must be avoid in the uranium ore processing development. In this aim the whole paper was thought and written. (author)

  20. Water Treatment for Uranium at the U.S. Department of Energy's Legacy Management Sites

    International Nuclear Information System (INIS)

    Dayvault, J.; Bush, R.; Ribeiro, T.; Surovchak, S.; Powell, J.; Bartlett, T.; Carpenter, C.; Jacobson, C.; Miller, D.; Morrison, S.; Boylan, J.; Broberg, K.; Glassmeyer, C.; Hertel, W.

    2009-01-01

    year from the aquifer. The system requires minimal operation and maintenance; however, the reactive media requires occasional replacement. At a former uranium milling site in Shiprock, New Mexico, uranium-contaminated ground water is captured by pumping wells and subsurface collection drains. The captured water is conveyed to an 11-acre evaporation pond. The total flow rate of contaminated ground water to the evaporation pond is about 190 lpm. Influent uranium concentration is about 800 μg/L, and about 80 kg of uranium is removed from the subsurface annually. Because of the evaporation process, the ground-water resource is lost. Operation of the system is limited to occasional pump maintenance. A pump-and-treat system is used at the Fernald Preserve in Ohio to lower uranium concentrations to less than 30 μg/L prior to discharge to the Great Miami River. The treatment system uses six flow-through vessels, each containing 8.9 m 3 of anion-exchange resin. The treatment flow rate is currently about 5,678 lpm, and the system is removing about 54 kg of uranium per year. Some ground water is blended with treated water such that about 300 kg of uranium is removed from the aquifer per year. The treatment process requires continuous operation and maintenance. At a former uranium milling site near Tuba City, Arizona, uranium-contaminated ground water is pumped from extraction wells and treated by ion exchange followed by distillation. The average flow rate is about 340 lpm, and the influent uranium concentration is about 250 μg/L. About 40 kg of uranium is removed from the aquifer per year. The distillation treatment process is operated full time, with the treated water being injected back into the aquifer. A wide variety of water treatments are used by the LM Program to remove uranium from contaminated ground water. If uranium is the only contaminant, it can be removed by simple flow-through columns containing an ion exchanger (Dowex) or a reductant (ZVI). Ion exchange with

  1. Modeling of geochemical processes related to uranium mobilization in the groundwater of a uranium mine

    International Nuclear Information System (INIS)

    Gomez, P.; Garralon, A.; Buil, B.; Turrero, Ma.J.; Sanchez, L.; Cruz, B. de la

    2006-01-01

    This paper describes the processes leading to uranium distribution in the groundwater of five boreholes near a restored uranium mine (dug in granite), and the environmental impact of restoration work in the discharge area. The groundwater uranium content varied from < 1 μg/L in reduced water far from the area of influence of the uranium ore-containing dyke, to 104 μg/L in a borehole hydraulically connected to the mine. These values, however, fail to reflect a chemical equilibrium between the water and the pure mineral phases. A model for the mobilization of uranium in this groundwater is therefore proposed. This involves the percolation of oxidized waters through the fractured granite, leading to the oxidation of pyrite and arsenopyrite and the precipitation of iron oxyhydroxides. This in turn leads to the dissolution of the primary pitchblende and, subsequently, the release of U(VI) species to the groundwater. These U(VI) species are retained by iron hydroxides. Secondary uranium species are eventually formed as reducing conditions are re-established due to water-rock interactions

  2. Hydrometallurgic treatment of a mineral containing uranium, vanadium and phosphorus

    International Nuclear Information System (INIS)

    Echenique, Patricia; Fruchtenicht, Fernando; Gil, Daniel; Vigo, Daniel; Bouza, Angel; Vert, Gabriela; Becquart, Elena

    1987-01-01

    A preliminary study of a mineral has been made towards the hydrometallurgy separation of uranium, vanadium and phosphorus. After the ore dressing, work on sulfuric acid with oxidation leaching has been made, to get the uranium, vanadium and phosphorus in solution. For the separation and purification of these elements, two alternative solvent extraction methods have been tested. One of them has been the extraction of uranium and vanadium and a selective stripping of both elements. The second one has been the selective extraction of uranium and vanadium at different aqueous solutions pH. In both methods, the same reagent has been used: di(2-ethylhexyl) phosphoric acid, kerosene as diluent with two different synergistic agents: TOPO (tri-n-octyl phosphine oxide) and TBP (tri-n-butyl phosphate). Batch studies have been made to determine the equilibrium isotherms for uranium and vanadium. A continuous countercurrent simulation method has been used to get the best phase ratio and to test different stripping agents. For the first method, an important loss of uranium and vanadium at the feed solution conditioning for the extraction step has been observed. For the second method, a good recovery of uranium has been reached, but there has been losses of vanadium in pH adjustment. Nevertheless, among these processes, the last seems to work better in this mineral hydrometallurgy. (Author) [es

  3. Enriched uranium processing with 7-1/2% TBP

    International Nuclear Information System (INIS)

    Orth, D.A.; Martin, W.H.; Pickett, C.E.

    1983-01-01

    The 7-1/2% TBP flowsheet gives adequate recovery of uranium and neptunium or plutonium, with reduced waste volume as compared to the prior aluminum-salted 3-1/2% TBP flowsheet. Decontamination from fission products is sensitive to numerous variables, including aluminum nitrate concentration in the feed, impeller speeds, and prior treatment of the fuel solution in head end operations. The impeller speed in the 1A bank also influences uranium losses as well as the fission product decontamination. The magnitudes of these effects suggest that stage efficiency is poor with this flowsheet in this mixer settler unit. The existing continuous solvent washers give evidence of low washing efficiency that limits permissible feed activity and that may be related to low contact time between the solvent and the carbonate wash solution. The most general conclusion is that satisfactory operation can be obtained with all projected domestic and foreign fuels under consideration for processing, by suitable adjustment of operating conditions. Also, possible flowsheet and equipment changes are known that could improve operations with these fuels further. 7 references

  4. Occupational exposures to uranium: processes, hazards, and regulations

    International Nuclear Information System (INIS)

    Stoetzel, G.A.; Fisher, D.R.; McCormack, W.D.; Hoenes, G.R.; Marks, S.; Moore, R.H.; Quilici, D.G.; Breitenstein, B.D.

    1981-04-01

    The United States Uranium Registry (USUR) was formed in 1978 to investigate potential hazards from occupational exposure to uranium and to assess the need for special health-related studies of uranium workers. This report provides a summary of Registry work done to date. The history of the uranium industry is outlined first, and the current commercial uranium industry (mining, milling, conversion, enrichment, and fuel fabrication) is described. This description includes information on basic processes and areas of greatest potential radiological exposure. In addition, inactive commercial facilities and other uranium operations are discussed. Regulation of the commercial production industry for uranium fuel is reported, including the historic development of regulations and the current regulatory agencies and procedures for each phase of the industry. A review of radiological health practices in the industry - facility monitoring, exposure control, exposure evaluation, and record-keeping - is presented. A discussion of the nonradiological hazards of the industry is provided, and the final section describes the tissue program developed as part of the Registry

  5. Research on deeply purifying effluent from uranium mining and metallurgy to remove uranium by ion exchange. Pt.2: Elution uranium from lower loaded uranium resin by the intense fractionation process

    International Nuclear Information System (INIS)

    Zhang Jianguo; Chen Shaoqiang; Qi Jing

    2002-01-01

    Developing macroporous resin for purifying uranium effluent from uranium mining and metallurgy is presented. The Intense Fractionation Process is employed to elute uranium from lower loaded uranium resin by the eluent of sulfuric acid and ammonium sulfate. The result is indicated that the uranium concentration in the rich elutriant is greatly increased, and the rich liquor is only one bed column volume, uranium concentration in the elutriant is increased two times which concentration is 10.1 g/L. The eluent is saved about 50% compared with the conventional fixed bed elution operation. And also the acidity in the rich elutriant is of benefit to the later precipitation process in uranium recovery

  6. Analysis of Hazards Associated with a Process Involving Uranium Metal and Uranium Hydride Powders

    Energy Technology Data Exchange (ETDEWEB)

    Bullock, J.S.

    2000-05-01

    An analysis of the reaction chemistry and operational factors associated with processing uranium and uranium hydride powders is presented, focusing on a specific operation in the Development Division which was subjected to the Job Hazard Analysis (JHA) process. Primary emphasis is on the thermodynamic factors leading to pyrophoricity in common atmospheres. The discussion covers feed powders, cold-pressed and hot-pressed materials, and stray material resulting from the operations. The sensitivity of the various forms of material to pyrophoricity in common atmospheres is discussed. Operational recommendations for performing the work described are given.

  7. Waste water treatment of CO2+O2 in-situ leaching uranium

    International Nuclear Information System (INIS)

    Xu Lechang; Liu Naizhong; Du Zhiming; Wang Hongying

    2012-01-01

    An in-situ leaching uranium mine located in Northern China uses CO 2 +O 2 leaching process to leach uranium. The consumption of industrial reagent and water, and generation and discharge of waste water are minimized by comprehensive waste water treatment technology with process water recycle, reverse osmosis and natural evaporation. The process water of the mine that can be recycled and reused includes barren fluid, solution washing loaded resin, precipitating mother solution and filtered liquor of yellow cake. Solution regenerating barren resin is treated by reverse osmosis. Concentrated water from reverse osmosis and solution washing barren resin are naturally evaporated. (authors)

  8. Determination of uranium in the red blood cells of the workers in the chemical processing of uranium ore

    International Nuclear Information System (INIS)

    Nosek, J.; Simkova, M.; Kukula, F.; Musil, K.

    1975-04-01

    Neutron activation analysis was used in determining uranium in the venous blood erythrocytes of controls and of workers exposed to occupational hazards in a uranium chemical treatment plant. While 4.1 +- 2.6 ppb of uranium was found in dry matter of the erythrocytes in controls, 6.5 +- 2.1 ppb of uranium was ascertained in dry matter of the erythrocytes in occupationally exposed workers of a wet preparation plant, and 37.2 +- 20.2 ppb of uranium in the erythrocytes in workers of a dry cleaning plant. (author)

  9. Manual on laboratory testing for uranium ore processing

    International Nuclear Information System (INIS)

    1990-01-01

    Laboratory testing of uranium ores is an essential step in the economic evaluation of uranium occurrences and in the development of a project for the production of uranium concentrates. Although these tests represent only a small proportion of the total cost of a project, their proper planning, execution and interpretation are of crucial importance. The main purposes of this manual are to discuss the objectives of metallurgical laboratory ore testing, to show the specific role of these tests in the development of a project, and to provide practical instructions for performing the tests and for interpreting their results. Guidelines on the design of a metallurgical laboratory, on the equipment required to perform the tests and on laboratory safety are also given. This manual is part of a series of Technical Reports on uranium ore processing being prepared by the IAEA's Division of Nuclear Fuel Cycle and Waste Management. A report on the Significance of Mineralogy in the Development of Flowsheets for Processing Uranium Ores (Technical Reports Series No. 196, 1980) and an instruction manual on Methods for the Estimation of Uranium Ore Reserves (No. 255, 1985) have already been published. 17 refs, 40 figs, 17 tabs

  10. Symposium 'geology, mining and extractive processing of uranium, with special reference to Europe'

    International Nuclear Information System (INIS)

    Pietsch, H.B.

    1977-01-01

    This review of the symposium 'Geology, mining and extractive processing of uranium' gives a survey from the point of view of ore processing rather than exploration. A reason for the uranium consumption assumed is given, and uranium deposits and availability, methods of exploration, and interesting facts on uranium extraction from ores are gone into. (HK) [de

  11. Fixed capital investments for the uranium soils integrated demonstration soil treatment technologies

    Energy Technology Data Exchange (ETDEWEB)

    Douthat, D.M.; Armstrong, A.Q. [Oak Ridge National Lab., TN (United States); Stewart, R.N. [Univ. of Tennessee, Knoxville, TN (United States)

    1995-05-01

    The development of a nuclear industry in the United States required mining, milling, and fabricating a large variety of uranium products. One of these products was purified uranium metal which was used in the Savannah River and Hanford Site reactors. Most of this feed material was produced at the United States Department of Energy (DOE) facility formerly called the Feed Materials Production Center at Fernald, Ohio. During operation of this facility, soils became contaminated with uranium from a variety of sources. To address remediation and management of uranium-contaminated soils at sites owned by DOE, the Uranium Soils Integrated Demonstration (USID) Program was formed to evaluate and compare the versatility, efficiency, and economics of various technologies that may be combined into systems designed to characterize and remediate uranium contaminated soils. The USID Program has five major tasks in developing and demonstrating these technologies. Each must be able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from soil, (3) treat or dispose of resulting waste streams, (4) meet necessary state and federal regulations, and (5) meet performance assessment objectives. The role of the performance assessment objectives is to provide the information necessary to conduct evaluations of the technologies. These performance assessments provide the basis for selecting the optimum system for remediation of large areas contaminated with uranium. One of the performance assessment tasks is to address the economics of full-scale implementation of soil treatment technologies developed by the USID Program. The cost of treating contaminated soil is one of the criteria used in the decision-making process for selecting remedial alternatives.

  12. Operating and life-cycle costs for uranium-contaminated soil treatment technologies

    International Nuclear Information System (INIS)

    Douthat, D.M.; Armstrong, A.Q.

    1995-09-01

    The development of a nuclear industry in the US required mining, milling, and fabricating a large variety of uranium products. One of these products was purified uranium metal which was used in the Savannah River and Hanford Site reactors. Most of this feed material was produced at the US Department of Energy (DOE) facility formerly called the Feed Materials Production Center at Fernald, Ohio. During operation of this facility, soils became contaminated with uranium from a variety of sources. To avoid disposal of these soils in low-level radioactive waste burial sites, increasing emphasis has been placed on the remediating soils contaminated with uranium and other radionuclides. To address remediation and management of uranium-contaminated soils at sites owned by DOE, the DOE Office of Technology Development (OTD) evaluates and compares the versatility, efficiency, and economics of various technologies that may be combined into systems designed to characterize and remediate uranium-contaminated soils. Each technology must be able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from soil, (3) treat or dispose of resulting waste streams, (4) meet necessary state and federal regulations, and (5) meet performance assessment objectives. The role of the performance assessment objectives is to provide the information necessary to conduct evaluations of the technologies. These performance assessments provide the basis for selecting the optimum system for remediation of large areas contaminated with uranium. One of the performance assessment tasks is to address the economics of full-scale implementation of soil treatment technologies. The cost of treating contaminated soil is one of the criteria used in the decision-making process for selecting remedial alternatives

  13. Fixed capital investments for the uranium soils integrated demonstration soil treatment technologies

    International Nuclear Information System (INIS)

    Douthat, D.M.; Armstrong, A.Q.; Stewart, R.N.

    1995-05-01

    The development of a nuclear industry in the United States required mining, milling, and fabricating a large variety of uranium products. One of these products was purified uranium metal which was used in the Savannah River and Hanford Site reactors. Most of this feed material was produced at the United States Department of Energy (DOE) facility formerly called the Feed Materials Production Center at Fernald, Ohio. During operation of this facility, soils became contaminated with uranium from a variety of sources. To address remediation and management of uranium-contaminated soils at sites owned by DOE, the Uranium Soils Integrated Demonstration (USID) Program was formed to evaluate and compare the versatility, efficiency, and economics of various technologies that may be combined into systems designed to characterize and remediate uranium contaminated soils. The USID Program has five major tasks in developing and demonstrating these technologies. Each must be able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from soil, (3) treat or dispose of resulting waste streams, (4) meet necessary state and federal regulations, and (5) meet performance assessment objectives. The role of the performance assessment objectives is to provide the information necessary to conduct evaluations of the technologies. These performance assessments provide the basis for selecting the optimum system for remediation of large areas contaminated with uranium. One of the performance assessment tasks is to address the economics of full-scale implementation of soil treatment technologies developed by the USID Program. The cost of treating contaminated soil is one of the criteria used in the decision-making process for selecting remedial alternatives

  14. Method of processing plutonium and uranium solution

    International Nuclear Information System (INIS)

    Otsuka, Katsuyuki; Kondo, Isao; Suzuki, Toru.

    1989-01-01

    Solutions of plutonium nitrate solutions and uranyl nitrate recovered in the solvent extraction step in reprocessing plants and nuclear fuel production plants are applied with low temperature treatment by means of freeze-drying under vacuum into residues containing nitrates, which are denitrated under heating and calcined under reduction into powders. That is, since complicate processes of heating, concentration and dinitration conducted so far for the plutonium solution and uranyl solution are replaced with one step of freeze-drying under vacuum, the process can be simplified significantly. In addition, since the treatment is applied at low temperature, occurrence of corrosion for the material of evaporation, etc. can be prevented. Further, the number of operators can be saved by dividing the operations into recovery of solidification products, supply and sintering of the solutions and vacuum sublimation. Further, since nitrates processed at a low temperature are powderized by heating dinitration, the powderization step can be simplified. The specific surface area and the grain size distribution of the powder is made appropriate and it is possible to obtain oxide powders of physical property easily to be prepared into pellets. (N.H.)

  15. Study of rolled uranium annealing process

    International Nuclear Information System (INIS)

    Cabane, G.

    1954-06-01

    The dilatometric study of rolled uranium clearly shows not only the expansions or contractions induced by stress relief or diffusion of vacancies, but also the slope variations of the cooling curves, which are the best evidence of a texture change. Under the microscope, hard-rolled sheets appear as a mixture of two distinct structures; it is also possible by intermediate annealing to prepare homogeneous sheets of either structure, i.e. twinned or untwinned. All these sheets which have similar textures, undergo at first a primary recrystallization beginning at 320 deg C, then a texture change without any apparent crystal growth, at about 430 deg C. (author) [fr

  16. Uranium: the exploration process and recent developments

    International Nuclear Information System (INIS)

    Merwin, S.S.

    1977-01-01

    Mineral exploration is a combination of technical and nontechnical disciplines seasoned with competence, imagination, tenacity, and luck. The objectives and phases of mineral exploration are discussed. The roles of incentive, finance, staff, area, techniques, time, and luck are discussed briefly. Some of the recent developments in the uranium industry include exploitation of lower-grade deposits, vertical integration in the industry, involvement of governments, hardrock deposits, and technical innovations. The costs involved in a hypothetical exploration program are described. The time element is also considered. The odds of successful exploration is 0.5%, but persistence with a competent staff over a long period of time will improve the odds

  17. Test operation of the uranium ore processing pilot plant and uranium conversion plant

    International Nuclear Information System (INIS)

    Suh, I.S.; Lee, K.I.; Whang, S.T.; Kang, Y.H.; Lee, C.W.; Chu, J.O.; Lee, I.H.; Park, S.C.

    1983-01-01

    For the guarantee of acid leaching process of the Uranium Ore Processing Pilot Plnat, the KAERI team performed the test operation in coorperation with the COGEMA engineers. The result of the operation was successful achieving the uranium leaching efficiency of 95%. Completing the guarentee test, a continuous test operation was shifted to reconform the reproducibility of the result and check the functions of every units of the pilot plant feeding the low-grade domestic ore, the consistency of the facility was conformed that the uranium can easily be dissolved out form the ore between the temperature range of 60degC-70degC for two hours of leaching with sulfuric acid and could be obtained the leaching efficiency of 92% to 95%. The uranium recovery efficiencies for the processes of extraction and stripping were reached to 99% and 99.6% respectively. As an alternative process for the separation of solid from the ore pulp, four of the Counter Current Decanters were shifted replacing the Belt Filter and those were connected in a series, which were not been tested during the guarantee operation. It was found out that the washing efficiencies of the ore pulp in each tests for the decanters were proportionally increased according to the quantities of the washing water. As a result of the test, it was obtained that washing efficiencies were 95%, 85%, 83% for the water to ore ratio of 3:1, 2:1, 1.5:1 respectively. (Author)

  18. The uranium enrichment industry and the SILEX process

    International Nuclear Information System (INIS)

    Goldsworthy, M.

    1999-01-01

    Silex Systems Limited has been developing a new laser isotope separation process since 1992. The principle application of the SILEX Technology is Uranium Enrichment, the key step in the production of fuel for nuclear power plants. The Uranium Enrichment industry, today worth ∼ US$3.5 Billion p.a., is dominated by four major players, the largest being USEC with almost 40% of the market. In 1996, an agreement was signed between Silex and USEC to develop SILEX Technology for potential application to Uranium Enrichment. The SILEX process is a low cost, energy efficient scheme which may provide significant commercial advantage over current technology and competing laser processes. Silex is also investigating possible application to the enrichment of Silicon, Carbon and other materials. Significant markets may develop for such materials, particularly in the semiconductor industry

  19. Treatment of wastewater for removal of soluble uranium species at Cameco's Port Hope Conversion Facility

    International Nuclear Information System (INIS)

    Dumont, H.; Tairova, G.; Kwong, A.K.; Smith, B.D.

    2000-01-01

    Ion exchange (IX) resin processes have been used for many years in the uranium mining industry for the recovery of uranium from both acid and alkaline leach solutions. More recently, IX processes have been shown to be an effective approach to control the uranium levels in non-process waters, such as mine water, public drinking water supply and well water. Bench scale and mini-pilot plant tests were conducted at the Cameco's Port Hope Conversion Facility to demonstrate the economic and technical viability of an IX process as an uranium remediation treatment for trace amounts of uranium in non-process laundry water. In the mini-pilot plant study, waste laundry water containing between 10 mg U/L and 200 mg U/L was treated at a rate ranging from 120 L/h to 240 L/h, using a typical 'merry-go-round' fixed-bed ion exchange system with three ion exchange columns. Each column contained 14 L of strongly basic Purolite A300 resin type II. The results indicated that the breakthrough limit, set at 0.1 mg U/L was obtained after a minimum of 1,200 equivalent bed volumes, while saturation was obtained at 3,300 equivalent bed volumes. Recovery parameters are discussed along with feed and effluent stream quality and modifications to the upstream operation. (author)

  20. Biological processes for concentrating trace elements from uranium mine waters. Technical completion report

    International Nuclear Information System (INIS)

    Brierley, C.L.; Brierley, J.A.

    1981-12-01

    Waste water from uranium mines in the Ambrosia Lake district near Grants, New Mexico, USA, contains uranium, selenium, radium and molybdenum. The Kerr-McGee Corporation has a novel treatment process for waters from two mines to reduce the concentrations of the trace contaminants. Particulates are settled by ponding, and the waters are passed through an ion exchange resin to remove uranium; barium chloride is added to precipitate sulfate and radium from the mine waters. The mine waters are subsequently passed through three consecutive algae ponds prior to discharge. Water, sediment and biological samples were collected over a 4-year period and analyzed to assess the role of biological agents in removal of inorganic trace contaminants from the mine waters. Some of the conclusions derived from this study are: (1) The concentrations of soluble uranium, selenium and molybdenum were not diminished in the mine waters by passage through the series of impoundments which constituted the mine water treatment facility. Uranium concentrations were reduced but this was due to passage of the water through an ion exchange column. (2) The particulate concentrations of the mine water were reduced at least ten-fold by passage of the waters through the impoundments. (3) The sediments were anoxic and enriched in uranium, molybdenum and selenium. The deposition of particulates and the formation of insoluble compounds were proposed as mechanisms for sediment enrichment. (4) The predominant algae of the treatment ponds were the filamentous Spirogyra and Oscillatoria, and the benthic alga, Chara. (5) Adsorptive processes resulted in the accumulation of metals in the algae cells. (6) Stimulation of sulfate reduction by the bacteria resulted in retention of molybdenum, selenium, and uranium in sediments. 1 figure, 16 tables

  1. Technologies for the treatment of effluents from uranium mines, mills and tailings. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-06-01

    Effluent treatment is an important aspect of uranium mining and milling operations that continues through decommissioning and site rehabilitation. During the life of a mine, effluent treatment is an integral part of the operation with all effluent either being recycled to the mill or processed through a water treatment plant before being released into the environment. During decommissioning and rehabilitation, effluent treatment must continue either through a water treatment plant of by using passive treatment techniques. Because of the recent closing of several uranium mines or mining districts, particularly in eastern Europe, effluent treatment is becoming an ever increasing concern. Therefore the IAEA convened a technical committee meeting (TCM) so that experts from different countries could discuss information and knowledge on effluent treatment processes and methods. The papers presented at the meeting describe techniques for treatment of effluents from uranium production operations - both past and present. This publication contains ten papers presented at the meeting; each of the papers was indexed separately

  2. Process for uranium separation and preparation of UO4.2NH3.2HF

    International Nuclear Information System (INIS)

    Dokuzoguz, H.Z.

    1976-01-01

    A process for treating the aqueous effluents that are produced in converting gaseous UF 6 (uranium hexafluoride) into solid UO 2 (uranium dioxide) by way of an intermediate (NH 4 ) 4 UO 2 (CO 3 ) 3 (''AUC'' Compound) is disclosed. These effluents, which contain large amounts of NH 4 + , CO 3 2- , F - , and a small amount of U are mixed with H 2 SO 4 (sulfuric acid) in order to expel CO 2 (carbon dioxide) and thereby reduce the carbonate concentration. The uranium is precipitated through treatment with H 2 O 2 (hydrogen peroxide) and the fluoride is easily recovered in the form of CaF 2 (calcium fluoride) by contacting the process liquid with CaO (calcium oxide). The presence of SO 4 2- (sulfate) in the process liquid during CaO contacting seems to prevent the development of a difficult-to-filter colloid. The process also provides for NH 3 recovery and recycling. Liquids discharged from the process, moreover, are essentially free of environmental pollutants. The waste treatment products, i.e., CO 2 , NH 3 , and U are economically recovered and recycled back into the UF 6 → UO 2 conversion process. The process, moreover, recovers the uranium as a precipitate in the second stage. This precipitate is a new inorganic chemical compound UO 4 .2NH 3 .2HF [uranyl peroxide-2-ammonia-2-(hydrogen fluoride)

  3. Behaviour of organic matters in uranium ore processing

    International Nuclear Information System (INIS)

    Wu Sanmin

    1991-01-01

    The oxidation-reduction behaviour of organic matters in the course of oxidation roasting, acid leaching and alkali leaching, the regeneration of humic acid and the consumption of reagents are described. The mineralogical characteristics of the organic matter samples were studied. The results show that its organic matter rich in volatile carbon and with the shorter evolutionary process and lower association is easily oxidized with higher consumption of oxidant during its acid leaching; it is easily oxidized with forming humic acid during alkali leaching; and pretreating it by oxidation roasting is beneficial to the oxidation of uranium. On the contrary, the organic matter rich in fixed carbon, and with longer evolutionary process and higher association is difficultly oxidized with lower consumption of oxidant during its acid leaching; it is difficult to regenerate humic acid for it during alkali leaching; and the uranium can be easily reduced and the leaching performance of uranium can be lowered

  4. Processing hexavalent uranium gels and their properties

    International Nuclear Information System (INIS)

    Landspersky, H.; Benadik, A.; Spitzer, Z.

    1980-01-01

    The properties of xerogels of ammonium polyuranate prepared by various drying procedures were studied. The individual drying procedures affect differently both the chemical structure of the material (its composition) and the physicochemical properties of the final product (specific surface area, porosity). In addition, the physicochemical properties of xerogels depend on the properties of the starting material, i.e., on the type of the initial gel. The physicochemical properties of xerogels, in particular their porosity, are in turn relevant for their subsequent high-temperature processing. The porous structure is essential for thermal treatment. The structure of xerogels obtained by distillation procedures is affected both by the conditions of azeotropic distillation and by the medium employed. By judicious selection of these two variables it is possible to prepare materials with different pore size distributions. (author)

  5. Optimization of dissolution process parameters for uranium ore concentrate powders

    Energy Technology Data Exchange (ETDEWEB)

    Misra, M.; Reddy, D.M.; Reddy, A.L.V.; Tiwari, S.K.; Venkataswamy, J.; Setty, D.S.; Sheela, S.; Saibaba, N. [Nuclear Fuel Complex, Hyderabad (India)

    2013-07-01

    Nuclear fuel complex processes Uranium Ore Concentrate (UOC) for producing uranium dioxide powder required for the fabrication of fuel assemblies for Pressurized Heavy Water Reactor (PHWR)s in India. UOC is dissolved in nitric acid and further purified by solvent extraction process for producing nuclear grade UO{sub 2} powder. Dissolution of UOC in nitric acid involves complex nitric oxide based reactions, since it is in the form of Uranium octa oxide (U{sub 3}O{sub 8}) or Uranium Dioxide (UO{sub 2}). The process kinetics of UOC dissolution is largely influenced by parameters like concentration and flow rate of nitric acid, temperature and air flow rate and found to have effect on recovery of nitric oxide as nitric acid. The plant scale dissolution of 2 MT batch in a single reactor is studied and observed excellent recovery of oxides of nitrogen (NO{sub x}) as nitric acid. The dissolution process is automated by PLC based Supervisory Control and Data Acquisition (SCADA) system for accurate control of process parameters and successfully dissolved around 200 Metric Tons of UOC. The paper covers complex chemistry involved in UOC dissolution process and also SCADA system. The solid and liquid reactions were studied along with multiple stoichiometry of nitrous oxide generated. (author)

  6. Purification process of uranium hexafluoride containing traces of plutonium fluoride and/or neptunium fluoride

    International Nuclear Information System (INIS)

    Aubert, J.; Bethuel, L.; Carles, M.

    1983-01-01

    In this process impure uranium hexafluoride is contacted with a metallic fluoride chosen in the group containing lead fluoride PbF 2 , uranium fluorides UFsub(4+x) (0 3 at a temperature such as plutonium and/or neptunium are reduced and pure uranium hexafluoride is recovered. Application is made to uranium hexafluoride purification in spent fuel reprocessing [fr

  7. Experience with water treatment and restoration technologies during and after uranium mining

    International Nuclear Information System (INIS)

    Benes, V.; Mitas, J.; Rihak, I.

    2002-01-01

    DIAMO, state owned enterprise, has a wide experience in uranium mining with the use of classical deep mining, acid in situ leaching and uranium ore processing. The sandstone deposits in Straz block have been exploited since 1968. Geological and hydrogeological conditions of the deposits and the short distance between the deep mine and ISL wellfields requires pumping huge amounts of fresh and/or acid mine water, their treatment and subsequent discharge into streams. DIAMO developed and applied several technologies for different types of wastewater treatment from the start of mining. Practically all of these technologies are used in the current phase of uranium deposit restoration after mining. It is possible to apply these technologies both in the production phase and during the restoration of underground water. In some cases, it is very desirable to combine two or several of them. (author)

  8. Conceptual process design for uranium recovery from sea water

    International Nuclear Information System (INIS)

    Suzuki, Motoyuki; Chihara, Kazuyuki; Fujimoto, Masahiko; Yagi, Hiroshi; Wada, Akihiko.

    1985-01-01

    Based on design of uranium recovery process from sea water, total cost for uranium production was estimated. Production scale of 1,000 ton-uranium per year was supposed, because of the big demand for uranium in the second age, i.e., fast breeder reactor age. The process is described as follows: Fluidized bed of hydrous titanium oxide (diameter is 0.1 mm, saturated adsorption capacity is 510 μg-U/g-Ad, adsorption capacity for ten days is 150 μg-U/g-Ad) is supposed, as an example, to be utilized as the primarily concentration unit. Fine adsorbent particles can be transferred as slurry in all of the steps of adsorption, washing, desorption, washing, regeneration. As an example, ammonium carbonate is applied to desorb the adsorbed uranium from titanium oxide. Then, stripping method is adopted for desorbent recovery. As for the secondary concentration, strong basic anion exchange method is supposed. The first step of process design is to determine the mass balance of each component through the whole process system by using the signal diagram. Then, the scale of each unit process, with which the mass balances are satisfied, is estimated by detailed chemical engineering calculation. Also, driving cost of each unit operation is estimated. As a result, minimum total cost of 160,000 yen/kg-U is obtained. Adsorption process cost is 80 to 90 % of the total cost. Capital cost and driving cost are fifty-fifty in the adsorption process cost. Pump driving cost forms a big part of the driving cost. Further concentrated study should be necessary on the adsorption process design. It might be important to make an effort on direct utilization of ocean current for saving the pump driving cost. (author)

  9. Non-polluting treatment of uranium effluents from the alkaline digestion of an uranium ore containing sulfur

    International Nuclear Information System (INIS)

    Berger, Bernard.

    1978-01-01

    New non-polluting process for treating uranium effluents from the alkaline digestion of an uranium ore containing sulphur, which makes it possible (a) to extract and obtain relatively pure uranium and (b) to process the digestion liquor freed from the uranium and containing in an aqueous solution a mixture of alkaline carbonate and/or bicarbonate and sodium sulphate, consisting in the selective extraction of the sodium sulphate present and the recycling of the liquor free of SO 4 = ions, containing in solution the sole carbonates and/or bicarbonates involved, towards the digestion of the ore [fr

  10. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    The article briefly discusses the Australian government policy and the attitude of political party factions towards the mining and exporting of the uranium resources in Australia. Australia has a third of the Western World's low-cost uranium resources

  11. Uranium

    International Nuclear Information System (INIS)

    Mackay, G.A.

    1978-01-01

    The author discusses the contribution made by various energy sources in the production of electricity. Estimates are made of the future nuclear contribution, the future demand for uranium and future sales of Australian uranium. Nuclear power growth in the United States, Japan and Western Europe is discussed. The present status of the six major Australian uranium deposits (Ranger, Jabiluka, Nabarlek, Koongarra, Yeelerrie and Beverley) is given. Australian legislation relevant to the uranium mining industry is also outlined

  12. Lung cancer among workers at a uranium processing plant

    International Nuclear Information System (INIS)

    Cookfair, D.L.; Beck, W.L.; Shy, C.; Lushbaugh, C.C.; Sowder, C.L.

    1983-01-01

    This study examined the risk of dying from lung cancer among white males who received radiation to the lung as a result of inhaling uranium dust or the dust of uranium compounds. Cases and controls were chosen from a cohort of workers employed in a uranium processing plant during World War II. Cumulative radiation lung dose among study population members ranged from 0 to 75 rads. Relative risk was found to increase with increasing level of exposure even after controlling for age and smoking status, but only for those who were over the age of 45 when first exposed. A statistically significant excess in risk was found for men in this age group with a cumulative lung dose of 20 rads of more. These data suggest that older age groups may be more susceptible to radiation-induced lung cancer than younger age groups

  13. Study of the dry processing of uranium ores; Etude des traitements de minerais d'uranium par voie seche

    Energy Technology Data Exchange (ETDEWEB)

    Guillet, H

    1959-02-01

    A description is given of direct fluorination of pre-concentrated uranium ores in order to obtain the hexafluoride. After normal sulfuric acid treatment of the ore to eliminate silica, the uranium is precipitated by a load of lime to obtain: either impure calcium uranate of medium grade, or containing around 10% of uranium. This concentrate is dried in an inert atmosphere and then treated with a current of elementary fluorine. The uranium hexafluoride formed is condensed at the outlet of the reaction vessel and may be used either for reduction to tetrafluoride and the subsequent manufacture of uranium metal or as the initial product in a diffusion plant. (author) [French] Il s'agit d'une description de fluoration directe de preconcentres de minerais d'uranium en vue d'obtention d'hexafluorure. Apres attaque sulfurique normale du minerai, afin d' eliminer la silice, l' uranium est precipite par un toit de chaux pour obtenir: ou uranate de chaux impur de titre moyen, ou uranium de la dizaine du pourcentage. Ce concentre seche en atmosphere inerte est soumis a un courant de fluor elementaire. L'hexafluorure d'uranium forme est condense a la sortie du reacteur et peut etre utilise soit apres reduction en tetrafluorure par l'elaboration d'uranium metal, soit comme produit de base dans le cadre d'une usine de diffusion. (auteur)

  14. Study of the dry processing of uranium ores; Etude des traitements de minerais d'uranium par voie seche

    Energy Technology Data Exchange (ETDEWEB)

    Guillet, H

    1959-02-01

    A description is given of direct fluorination of pre-concentrated uranium ores in order to obtain the hexafluoride. After normal sulfuric acid treatment of the ore to eliminate silica, the uranium is precipitated by a load of lime to obtain: either impure calcium uranate of medium grade, or containing around 10% of uranium. This concentrate is dried in an inert atmosphere and then treated with a current of elementary fluorine. The uranium hexafluoride formed is condensed at the outlet of the reaction vessel and may be used either for reduction to tetrafluoride and the subsequent manufacture of uranium metal or as the initial product in a diffusion plant. (author) [French] Il s'agit d'une description de fluoration directe de preconcentres de minerais d'uranium en vue d'obtention d'hexafluorure. Apres attaque sulfurique normale du minerai, afin d' eliminer la silice, l' uranium est precipite par un toit de chaux pour obtenir: ou uranate de chaux impur de titre moyen, ou uranium de la dizaine du pourcentage. Ce concentre seche en atmosphere inerte est soumis a un courant de fluor elementaire. L'hexafluorure d'uranium forme est condense a la sortie du reacteur et peut etre utilise soit apres reduction en tetrafluorure par l'elaboration d'uranium metal, soit comme produit de base dans le cadre d'une usine de diffusion. (auteur)

  15. Uranium,Radium and Iron Absorption from Liquid Waste Uranium Ore Processing by Zeolite

    International Nuclear Information System (INIS)

    Wismawati, T; Sorot sudiro, A; Herjati, T

    1998-01-01

    The aim of this work is to determine zeolites sorption capacity and the distribution coefficient of uranium, radium, and iron in zeolite-liquid waste system. Mineralogical composition of zeolite used in the experiment has been determine by examining the thin sections of zeolite grains under a microscope. Zeolite has ben activated by the dilute sulfuric acid or sodium hydroxide solution. The results show that the use of 0.25 N sodium hydroxide solution could be optimizing the zeolite for uranium and iron ions sorption and that of 0.1 N sulfuric acid solution is for radium sorption. The re-activation process has been carried out in three hours. Under such a condition, the sorption efficiency of zeolite to those ions have been known to be 45.85% for uranium, 96.63 % for iron and 87.80 % for radium. The distribution coefficients of uranium, radium and iron ion in zeolite-liquid waste system have been calculated 0.85, 7.02, and 28.65 ml/g respectively

  16. Rirang uranium ore processing: continuous solvent extraction of uranium from Rirang ore acid digestion solution

    International Nuclear Information System (INIS)

    Riza, F.; Nuri, H. L.; Waluya, S.; Subijanto, A.; Sarono, B.

    1998-01-01

    Separation of uranium from Rirang ore acid digestion solution by means of continuous solvent extraction using mixer-settlers has been studied and a mixture of 0.3 M D2EHPA and 0.075 M TOPO extracting agent and kerosene diluent is employed to recover and separate uranium from Th, RE, phosphate containing solution. The experiments have been conducted batch-wise and several parameters have been studied including the aqueous to organic phase ratio, A/O, the extraction and the stripping times, and the operation temperature. The optimum conditions for extraction have been found to be A/O = 2 ratio, five minute extraction time per stage at room temperature. The uranium recovery of 99.07% has been achieved at those conditions whilst U can be stripped from the organic phase by 85% H 3 PO 4 solution with an O/A = 1 for 5 minutes stripping time per stage, and in a there stage operation at room temperature yielding a 100% uranium recovery from the stripping process

  17. Uranium

    International Nuclear Information System (INIS)

    Toens, P.D.

    1981-03-01

    The geological setting of uranium resources in the world can be divided in two basic categories of resources and are defined as reasonably assured resources, estimated additional resources and speculative resources. Tables are given to illustrate these definitions. The increasing world production of uranium despite the cutback in the nuclear industry and the uranium requirements of the future concluded these lecture notes

  18. Continuous precipitation of uranium peroxide in process pilot plant

    International Nuclear Information System (INIS)

    Quinelato, A.L.

    1990-01-01

    An experimental study on uranium peroxide precipitation has been carried out with the objective to evaluate the influence of the main process parameters with a technological approach. The uraniferous solution used was obtained from the hydrometallurgical processing of an ore from Itataia - CE. Studies were developed in two distinct experimental stages. In the first stage, the precipitation was investigated by means of laboratory batch tests and, in the second stage, by means of continuous operation in a process pilot plant. (author)

  19. Case study: remediation of a former uranium mining/processing site in Hungary

    International Nuclear Information System (INIS)

    Csovari, M. et al.

    2004-01-01

    The Hungarian uranium mining activities near Pecs lasted from 1958 to 1997. Approximately 46 Mt of rock were mined, from which 18.8 Mt of upgraded ore were processed. Some ore had been exported prior to the construction of the processing plant at the site. Remediation of the former uranium-related industrial sites is being carried out by the Mecsek Ore Environment Ltd. and started in the 1990s. Today the former mines and their surroundings are rehabilitated, former heap piles and a number of smaller waste rock piles have been relocated to a more protected area (waste rock pile N 3). Ongoing core remediation activities are directed to the remediation of the tailings ponds, and also water treatment issues are most important. Three water treatment facilities are currently in operation: a mine water treatment system with the objective to remove uranium and gain a marketable by-product; a pump-and-treat system to restore the groundwater quality in the vicinity of the tailing ponds; a pilot-scale, experimental passive in-situ groundwater treatment system to avoid migration of uranium contaminated groundwater. Refs. 5 (author)

  20. Status Report from the United States of America [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, R H [United States Atomic Energy Commission, Washington, D.C. (United States)

    1967-06-15

    ores. However, there have been occasional circumstances in which the uranium in low-grade or marginal materials could be recovered at costs competitive with the costs of conventional ore mining and treatment. The procedures which are employed in treating these materials are given in the technical papers which were prepared for this meeting. The USAEC conducted a vigorous research programme in uranium ore processing technology from about 1947 through 1958. This work was carried out by many different types of organizations including USAEC laboratories, other government agencies, universities, and private companies. Research activities were also closely co-ordinated with the work of laboratories in several other countries.

  1. Process evaluations for uranium recovery from scrap material

    International Nuclear Information System (INIS)

    Westphal, B.R.; Benedict, R.W.

    1992-01-01

    The integral Fast Reactor (IFR) concept being developed by Argonne National Laboratory is based on pyrometallurgical processing of spent nuclear metallic fuel with subsequent fabrication into new reactor fuel by an injection casting sequence. During fabrication, a dilute scrap stream containing uranium alloy fines and broken quartz (Vycor) molds in produced. Waste characterization of this stream, developed by using present operating data and chemical analysis was used to evaluate different uranium recovery methods and possible process variations for the return of the recovered metal. Two methods, comminution with size separation and electrostatic separation, have been tested and can recover over 95% of the metal. Recycling the metal to either the electrochemical process or the injection casting was evaluated for the different economic and process impacts. The physical waste parameters and the important separation process variables are discussed with their effects on the viability of recycling the material. In this paper criteria used to establish the acceptable operating limits is discussed

  2. Process for the preparation of uranium dioxide

    International Nuclear Information System (INIS)

    Watt, G.W.; Baugh, D.W. Jr.

    1977-01-01

    An actinide dioxide, e.g., uranium dioxide, plutonium dioxide, neptunium dioxide, etc., is prepared by reacting the actinide nitrate hexahydrate with sodium dithionite as a first step; the reaction product from this first step is a novel composition of matter comprising the actinide sulfite tetrahydrate. The reaction product resulting from this first step is then converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) to a temperature of about 500 0 to about 950 0 C for about 15 to about 135 minutes. If the reaction product resulting from the first step is, prior to carrying out the second heating step, exposed to an oxygen-containing atmosphere such as air, the resultant product is a novel composition of matter comprising the actinide oxysulfite tetrahydrate which can also be readily converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) at a temperature of about 400 0 to about 900 0 C for about 30 to about 150 minutes. Further, the actinide oxysulfite tetrahydrate can be partially dehydrated at reduced pressures (and in the presence of a suitable dehydrating agent such as phosphorus pentoxide). The partially dehydrated product may be readily converted to the dioxide form by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) at a temperature of about 500 0 to about 900 0 C for about 30 to about 150 minutes. 16 claims

  3. Treatment of effluent containing uranium with magnetic zeolite

    International Nuclear Information System (INIS)

    Craesmeyer, Gabriel Ramos

    2013-01-01

    Within this work, a magnetic-zeolite composite was successfully synthesized using ferrous sulfate as raw material for the magnetic part of the composite, magnetite, and coal fly ash as raw material for the zeolitic phase. The synthesis of the zeolitic phase was made by alkali hydrothermal treatment and the magnetite nanoparticles were obtained through Fe 2+ precipitation on alkali medium. The synthetic process was repeated many times and showed good reproducibility comparing the zeolitic nanocomposite from different batches. The final product was characterized using infrared spectroscopy, powder X-ray diffraction, X-ray fluorescence, scanning electron microscopy with coupled EDS. Specific mass, specific surface area and other physicochemical proprieties. The main crystalline phases found in the final product were magnetite, zeolites types NaP1 and hydroxysodalite, quartz and mullite, those last two remaining from the raw materials. Uranium removal capacity of the magnetic zeolite composite was tested using batch techniques. The effects of contact time and initial concentration of the adsorbate over the adsorption process were evaluated. Equilibrium time was resolved and the following kinetics and diffusion models were evaluated: pseudo-first order kinetic model, pseudo-second order kinetic model and interparticle diffusion model. A contact time of 120 min turned out to be enough to reach equilibrium of the adsorption process. The rate of adsorption followed the pseudo-second order model and the intra particle diffusion did not turn out to be a speed determinant step. Two adsorption isotherms models, the Langmuir model and the Freundlich model, were also evaluated. The Langmuir model was the best fit for the obtained experimental data. Using the best fitted adsorption isotherm and kinetic model, the theoretical maximum adsorption capacity of uranium over the composite was determined for both models. The maximum removal capacity calculated was 20.7 mg.g -1 for the

  4. PHWR fuel fabrication with imported uranium - procedures and processes

    International Nuclear Information System (INIS)

    Rao, R.V.R.L.V.; Rameswara Rao, A.; Hemantha Rao, G.V.S.; Jayaraj, R.N.

    2010-01-01

    Following the 123 agreement and subsequent agreements with IAEA & NSG, Government of India has entered into bilateral agreements with different countries for nuclear trade. Department of Atomic Energy (DAE), Government of India, has entered into contract with few countries for supply of uranium material for use in the safeguarded PHWRs. Nuclear Fuel Complex (NFC), an industrial unit of DAE, established in the early seventies, is engaged in the production of Nuclear Fuel and Zircaloy items required for Nuclear Power Reactors operating in the country. NFC has placed one of its fuel fabrication facilities (NFC, Block-A, INE-) under safeguards. DAE has opted to procure uranium material in the form of ore concentrate and fuel pellets. Uranium ore concentrate was procured as per the ASTM specifications. Since no international standards are available for PHWR fuel pellets, Specifications have to be finalized based on the present fabrication and operating experience. The process steps have to be modified and fine tuned for handling the imported uranium material especially for ore concentrate. Different transportation methods are to be employed for transportation of uranium material to the facility. Cost of the uranium material imported and the recoveries at various stages of fuel fabrication have impact on the fuel pricing and in turn the unit energy costs. Similarly the operating procedures have to be modified for safeguards inspections by IAEA. NFC has successfully manufactured and supplied fuel bundles for the three 220 MWe safeguarded PHWRs. The paper describes various issues encountered while manufacturing fuel bundles with different types of nuclear material. (author)

  5. Technology of uranium recovery from wet-process phosphoric acid

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, Katsutoshi [Saga Univ. (Japan). Faculty of Science and Engineering; Nakashio, Fumiyuki

    1982-12-01

    Rock phosphate contains from 0.005 to 0.02 wt.% of uranium. Though the content is a mere 5 to 10 % of that in uranium ore, the total recovery of uranium is significant since it is used for fertilizer manufacture in a large quantity. Wet-process phosphoric acid is produced by the reaction of rock phosphate with sulfuric acid. The recovery of uranium from this phosphoric acid is mostly by solvent extraction at present. According to U/sup 4 +/ or UO/sub 2//sup 2 +/ as the form of its existence, the technique of solvent extraction differs. The following matters are described: processing of rock phosphate; recovery techniques including the extraction by OPPA-octyl pyrophosphoric acid for U/sup 4 +/, and by mixed DEHPA-Di-(2)-ethylhexyl phosphoric acid and TOPO-tryoctyl phosphine oxide for UO/sub 2//sup 2 +/, and by OPAP-octylphenyl acid phosphate for U/sup 4 +/; the recent progress of the technology as seen in patents.

  6. Processing Uranium-Bearing Materials Containing Coal and Loam

    Energy Technology Data Exchange (ETDEWEB)

    Civin, V; Prochazka, J [Research and Development Laboratory No. 3 of the Uranium Industry, Prague, Czechoslovakia (Czech Republic)

    1967-06-15

    Among the ores which are classified as low-grade in the CSSR are mixtures of coal and bentonitic loam of tertiary origin, containing approximately 0.1% U and with a moisture content at times well above 20-30%. The uranium is held mainly by the carbonaceous component. Conventional processing of these materials presents various difficulties which are not easily overcome. During leaching the pulp thickens and frequently becomes pasty, due to the presence of montmorillonites. Further complications arise from the high sorption capacity of the materials (again primarily due to montmorillonites) and poor sedimentation of the viscous pulps. In addition, the materials are highly refractory to the leaching agents. The paper presents experience gained in solving the problems of processing these ores. The following basic routes were explored: (1) separation of the carbonaceous and loamy components: The organic component appears to be the main activity carrier. Processing the concentrated material upon separation of the inactive or less active loam may not only remove the thixotropic behaviour but also substantially reduce the cost of the ore treatment; (2) 'liquifying' the pulps or preventing the thickening of the pulp by addition of suitable agents; (3) joint acid or carbonate processing of the materials in question with current ore types; (4) removal or suppression of thixotropic behaviour by thermal pretreatment of the material; and (5) application of the 'acid cure' method. The first method appears to be the most effective, but it presents considerable difficulties due to the extreme dispersion of the carbonaceous phase and further research is being carried out. Methods 2 and 3 proved to be unacceptable. Method 4, which includes roasting at 300-400{sup o}C, is now being operated on an industrial scale. The final method has also shown definite advantages for particular deposits of high montmorillonite content material. (author)

  7. Development of a pyro-partitioning process for long-lived radioactive nuclides. Process test for pretreatment of simulated high-level waste containing uranium

    International Nuclear Information System (INIS)

    Kurata, Masateru; Hijikata, Takatoshi; Kinoshita, Kensuke; Inoue, Tadashi

    2000-01-01

    A pyro-partitioning process developed at CRIEPI requires a pre-treatment process to convert high-level liquid waste to chloride. A combination process of denitration and chlorination has been developed for this purpose. Continuous process tests using simulated high-level waste were performed to certify the applicability of the process. Test results indicated a successful material balance sufficient for satisfying pyro-partitioning process criteria. In the present study, process tests using simulated high-level waste containing uranium were also carried out to prove that the pre-treatment process is feasible for uranium. The results indicated that uranium can be converted to chloride appropriate for the pyro-partitioning process. The material balance obtained from the tests is to be used to revise the process flow diagram. (author)

  8. Advances in treatment methods for uranium contaminated soil and water

    International Nuclear Information System (INIS)

    Navratil, J.D.

    2002-01-01

    Water and soil contaminated with actinides, such as uranium and plutonium, are an environmental concern at most U.S. Department of Energy sites, as well as other locations in the world. Remediation actions are on going at many sites, and plans for cleanup are underway at other locations. This paper will review work underway at Clemson University in the area of treatment and remediation of soil and water contaminated with actinide elements. (author)

  9. Remote Handling Devices for Disposition of Enriched Uranium Reactor Fuel Using Melt-Dilute Process

    International Nuclear Information System (INIS)

    Heckendorn, F.M.

    2001-01-01

    Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment

  10. The preparation of reports of a significant event at a uranium processing or uranium handling facility

    International Nuclear Information System (INIS)

    1988-08-01

    Licenses to operate uranium processing or uranium handling facilities require that certain events be reported to the Atomic Energy Control Board (AECB) and to other regulatory authorities. Reports of a significant event describe unusual events which had or could have had a significant impact on the safety of facility operations, the worker, the public or on the environment. The purpose of this guide is to suggest an acceptable method of reporting a significant event to the AECB and to describe the information that should be included. The reports of a significant event are made available to the public in accordance with the provisions of the Access to Information Act and the AECB's policy on public access to licensing information

  11. The conceptual flowsheet of effluent treatment during total gelation of uranium process for preparing ceramic UO2 particles of high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Quan Ying; Chen Xiaotong; Wang Yang; Liu Bing; Tang Yaping; Tang Chunhe

    2014-01-01

    Today, more and more people pay attention to the environmental protection and ecological environment. Along with the development of nuclear industry, many radioactive effluents may be discharged into environment, which can lead to the pollutions of water, atmosphere and soil. So radioactive effluents including low-activity and medium-level wastes solution treatments have been becoming one of significant subjects. High temperature gas-cooled reactor (HTR) is one of advanced nuclear reactors owing to its reliability, security and broad application in which the fabrication of spherical fuel element is a key technology. During the production of spherical fuel elements, the radioactive effluent treatment is necessary. Referring to the current treatment technologies and methods, the conceptual flowsheet of low-level radioactive effluent treatment during preparing spherical fuel elements was summarized which met the 'Zero Emission' demand. (authors)

  12. Recovery of uranium by a reverse osmosis process

    International Nuclear Information System (INIS)

    Cleary, J.G.; Stana, R.R.

    1980-01-01

    A method for concentrating and recovering uranium material from an aqueous solution, comprises passing a feed solution containing uranium through at least one reverse osmosis membrane system to concentrate the uranium, and then flushing the concentrated uranium solution with water in a reverse osmosis membrane system to further concentrate the uranium

  13. The study on process of recycling uranium in mixture of residue and liquid

    International Nuclear Information System (INIS)

    Zhang Jie; Shen Weiwei; Hao Jidong; Wu Jiangming

    2014-01-01

    The treat method of mixture of residue and liquid produced from HWR nuclear fuel chemical process using some kind of U_3O_8 powder was studied in this experiment. For recycling the uranium in mixture of residue and liquid, chemical dissolving method, washing and centrifuging method and dilute nitric acid leaching uranium method was contrasted in this test. The merit of dilute nitric acid leaching uranium method is simpler, more effective and higher uranium recycling ratio. Next, dilute nitric acid leaching uranium method was studied systematically. As a result, the main influence factors of uranium recycling ratio is dip sour degree and dip sour temperature. The influence law of factors to uranium recycling ratio and filtering effect was found out also. Along with increasing of dip sour degree and dip sour temperature, uranium recycling ratio increases and speed of filtrate increases also. At last, the process of batch treating mixture of residue and liquid was build and abundant uranium was recycled. (authors)

  14. Non-filtration method of processing of uranium ores

    International Nuclear Information System (INIS)

    Laskorin, B.N.; Vodolazov, L.I.; Tokarev, N.N.; Vyalkov, V.I.; Goldobina, V.A.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1977-01-01

    The development of the filterless sorption method has lead to working out the sorption leaching process and the process of extraction desorption, which has made possible to intensify the process of uranium ore working and to improve greatly the technical economic indexes by liquidating the complex method of multiple filtration and repulping of cakes. This method makes possible to involve more poor uranium raw materials and at the same time to extract valuable components: molybdenum, vanadium, copper, etc. Great industrial experience has been accumulating in sorption of dense pulp with the ratio of solid phase to liquid one equal to 1:1. This has lead to the increase of productivity of working plants by 1,5-3,0 times, the increase of uranium extraction by 5-10%, the increase of labour capacity of main workers by 2-3 times, and to the decrease of reagents expense, auxiliary materials, electric energy and vapour by several times. In fact the developed technology is continuous in all its steps with complete complex automatization of the process with the help of the most simple and available means of regulation and controlling. The process is equipped with high productivity apparatuses of great power with mechanic and pneumatic mixing for high density pulps, and with the columns KDS, KDZS, KNSPR and PIK for the regeneration of saturated sorbent in the counterflow regime. The exploitation of fine-granular hydrophilic ion-exchange resins in hydrophobized state is foreseen [ru

  15. Processing of low-grade uranium ores

    International Nuclear Information System (INIS)

    Michel, P.

    1975-01-01

    Four types of low grade ores are studied. Low grade ores which must be extracted because they are enclosed in a normal grade deposit. Heap leaching is the processing method which is largely used. It allows to obtain solutions or preconcentrates which may be delivered at the nearest plant. Normal grade ores contained in a low amplitude deposit which can be processed using leaching as far as the operation does not need any large expensive equipment. Medium grade ores in medium amplitude deposits to which a simplified conventional process can be applied using fast heap leaching. Low grade ores in large deposits. The processing possibilities leading to use in place leaching are explained. The operating conditions of the method are studied (leaching agent, preparation of the ore deposit to obtain a good tightness with regard to the hydrological system and to have a good contact between ore and reagent) [fr

  16. Processing of low grade uranium ores

    International Nuclear Information System (INIS)

    Michel, P.

    1978-10-01

    Four types of low-grade ores are studied: (1) Low-grade ores that must be extracted because they are enclosed in a normal-grade deposit. Heap leaching is the processing method which is largely used. (2) Normal-grade ores contained in low-amplitude deposits. They can be processed using in-place leaching as far as the operation does not need any large and expensive equipment. (3) Medium-grade ores in medium-amplitude deposits. A simplified conventional process can be applied using fast heap leaching. (4) Low-grade ores in large deposits. The report explains processing possibilities leading in most cases to the use of in-place leaching. The operating conditions of this method are laid out, especially the selection of the leaching agents and the preparation of the ore deposit

  17. 76 FR 60941 - Policy Regarding Submittal of Amendments for Processing of Equivalent Feed at Licensed Uranium...

    Science.gov (United States)

    2011-09-30

    ... Processing of Equivalent Feed at Licensed Uranium Recovery Facilities AGENCY: Nuclear Regulatory Commission... State-licensed uranium recovery site, either conventional, heap leach, or in situ recovery. DATES... Regarding Submittal of Amendments for Processing of Equivalent Feed at Licensed Uranium Recovery Facilities...

  18. Process of quantity determination of uranium by chromatography in liquid zone

    International Nuclear Information System (INIS)

    Muller, J.P.; Cojean, J.; Daubizit, M.

    1993-01-01

    The invention concerns a process of quantity determination of uranium by chromatography in liquid zone, usable to determine the quantity of uranium traces. Solutions to be treated can be aqueous or organic

  19. Treatment of waste water from uranium ore preparation

    International Nuclear Information System (INIS)

    Klicka, V.; Mitas, J.; Vacek, J.

    1976-01-01

    An improved closed-loop process is described for treating waste water resulting from chemical extraction of uranium from ore. The water is evaporated to form a concentrated solution and is then subjected to crystallization of the least soluble salt component thereof via further evaporation, or cooling or simultaneous cooling and a partial vacuum. The crystallized component is then separated from the mother liquor, whereupon the latter is fed back after removal of residual uranium therefrom to the extraction installation to replace the acids used therein. Additionally, the pure condensate produced during evaporation of the waste waters is employed as a replacement for the fresh water employed in processing of the ore. 6 claims, 2 figures

  20. Uranium enrichment by the gaseous diffusion process

    International Nuclear Information System (INIS)

    Petit, J.F.

    1977-01-01

    After a brief description of the process and technology (principle, stage constitution, cascade constitution, and description of a plant), the author gives the history of gaseous diffusion and describes the existing facilities. Among the different enrichment processes contemplated in the USA during and after the last world war, gaseous diffusion has been the only one to be developed industrially on a wide scale in the USA, then in the UK, in the USSR and in France. The large existing capacities in the USA provided the country with a good starting base for the development of a light-water nuclear power plant programme, the success of which led to a shortfall in production means. France and the USA, possessing the necessary know-how, have been able to undertake the realization of two industrial programmes: the CIP-CUP programme in the USA and the Eurodif project in France. Current plans still call for the construction of several plants by 1990. Can the gaseous diffusion process meet this challenge. Technically, there is no doubt about it. Economically, this process is mainly characterized by large energy consumption and the necessity to build large plants. This leads to a large investment, at least for the first plant. Other processes have been developed with a view to reducing both energy and capital needs. However, in spite of continuous studies and technological progress, no process has yet proved competitive. Large increments in capacities are still expected to come from gaseous diffusion, and several projects taking into account the improvements in flexibility, automatization, reliability and reduced investment, are analysed in the paper. Combining new facilities with existing plants has already proved to be of great interest. This situation explains why gaseous diffusion is being further investigated and new processes are being studied. (author)

  1. Treatment of Uranium-Contaminated Concrete for Reducing Secondary Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Han, G. S; Park, U. K; Kim, G. N.; Moon, J. K. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A volume reduction of the concrete waste by appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a decontamination process for uranium-contaminated (U-contaminated) concrete, and some experiments were performed to reduce the second radioactive waste. A decontamination process was developed to remove uranium from concrete waste. The yellow or brown colored surface of the wall brick with high concentration of uranium was removed by a chisel until the radioactivity of remaining block reached less than 1 Bq/g. The concrete waste coated with epoxy was directly burned by an oil flame, and the burned surface was then removed using the same method as the treatment of the brick. The selective mechanical removal of the concrete block reduced the amount of secondary radioactive waste. The concrete blocks without an epoxy were crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm were sequentially washed with a clear recycle solution and 1.0 M of nitric acid, their radioactivity reached below the limit value of uranium for self-disposal. For the concrete pieces smaller than 1 mm, a rotary washing machine and electrokinetic equipment were also used.

  2. Treatment of Uranium-Contaminated Concrete for Reducing Secondary Radioactive Waste

    International Nuclear Information System (INIS)

    Kim, Seung Soo; Han, G. S; Park, U. K; Kim, G. N.; Moon, J. K.

    2014-01-01

    A volume reduction of the concrete waste by appropriate treatment technologies will decrease the amount of waste to be disposed of and result in a reduction of the disposal cost and an enhancement of the efficiency of the disposal site. Our group has developed a decontamination process for uranium-contaminated (U-contaminated) concrete, and some experiments were performed to reduce the second radioactive waste. A decontamination process was developed to remove uranium from concrete waste. The yellow or brown colored surface of the wall brick with high concentration of uranium was removed by a chisel until the radioactivity of remaining block reached less than 1 Bq/g. The concrete waste coated with epoxy was directly burned by an oil flame, and the burned surface was then removed using the same method as the treatment of the brick. The selective mechanical removal of the concrete block reduced the amount of secondary radioactive waste. The concrete blocks without an epoxy were crushed to below 30 mm and sifted to 1 mm. When the concrete pieces larger than 1 mm were sequentially washed with a clear recycle solution and 1.0 M of nitric acid, their radioactivity reached below the limit value of uranium for self-disposal. For the concrete pieces smaller than 1 mm, a rotary washing machine and electrokinetic equipment were also used

  3. Study of immobilization of waste from treatment of acid waters of a uranium mining facility

    International Nuclear Information System (INIS)

    Goda, R.T.; Oliveira, A.P. de; Silva, N.C. da; Villegas, R.A.S.; Ferreira, A.M.

    2017-01-01

    This study aimed to produce scientific and technical knowledge aiming at the development of techniques to immobilize the waste generated in the treatment of acid waters in the UTM-INB Caldas uranium mining and processing facility using Portland cement. This residue (calcium diuranate - DUCA) contains uranium compounds and metal hydroxides in a matrix of calcium sulfate. It is observed that this material, in contact with the lake of acid waters of the mine's own pit, undergoes resolubilization and, therefore, changes the quality of the acidic water contained therein, changing the treatment parameters. For the study of immobilization of this residue, the mass of water contained in both the residue deposited in the pit of the mine and in the pulp resulting from the treatment of the acid waters was determined. In addition, different DUCA / CEMENT / WATER ratios were used for immobilization and subsequent mechanical strength and leaching tests. The results showed that in the immobilized samples with 50% cement mass condition, no uranium was detected in the leaching tests, and the mechanical strength at compression was 9.4 MPa, which indicates that more studies are needed, but indicate a good capacity to immobilize uranium in cement

  4. Uranium processing in South Africa from 1961 to 1981

    International Nuclear Information System (INIS)

    Boydell, D.W.; Viljoen, E.B.

    1982-01-01

    The production of uranium in South Africa reached a peak of 5,846 kt of U 3 O 8 in 1959, when 17 plants treated material from a total of 27 mines. By 1965 production had fallen to 2,669 kt of U 3 O 8 and only 6 plants remained in operation. A new record in production of 7,295 kt of U 3 O 8 was set in 1980. The revival in the industry during the intervening years was accompanied by improvements in all sections of the processing route employed to treat Witwatersrand ores. Ferric leaching, countercurrent decantation, belt filters, hopper clarification, solvent extraction, and continuous ion exchange have all found application in the new or modified plants that have been built. These developments are described, together with the novel process use by Palabora Mining Company for the recovery of uranium from uranothorianite concentrates as a byproduct from copper production

  5. Australian uranium exports: nuclear issues and the policy process

    International Nuclear Information System (INIS)

    Trood, R.B.

    1983-01-01

    The subject is discussed as follows: general introduction; formulation of uranium policy (the public debate; the Ranger Enquiry into all environmental aspects of a proposal by the AAEC and Ranger Uranium Mines to develop certain uranium deposits in the Northern Territory of Australia; the Government's decision); issues (non-proliferation and uranium safeguards policy; uranium enrichment in Australia; government involvement in uranium development; U development and environmental protection; U development and the Australian aborigines); conclusions. (U.K.)

  6. Releases of radioactivity from uranium mills and effluent treatment costs

    International Nuclear Information System (INIS)

    Witherspoon, J.P.; Sears, M.B.; Blanco, R.E.

    1977-01-01

    Airborne releases of radioactive materials from uranium milling to the environment consist of ore dust, yellowcake dust, tailings dust, and radon gas while the mill is active. After a mill has ceased operations, tailings may be stabilized to minimize or prevent airborne releases of radioactive particulates. However, radon gas will continue to be released in amounts inversely proportional to the degree of stabilization treatment (and expense). Liquid waste disposal is by evaporation and natural seepage to the ground beneath the tailings impoundment area. The release of radioactive materials (and potential radiation exposures) determines the majority of costs associated with minimizing the environmental impact of uranium milling. Radwaste treatments to reduce estimated radiation doses to individuals to 3 to 5% of those received with current milling practices are equivalent to $0.66 per pounds of U 3 O 8 and 0.032 mill per kWhr of electricity. This cost would cover a high efficiency reverse jet bag filter and high energy venturi scrubbers for dusts, neutralization of liquids, and an asphalt-lined tailings basin with a clay core dam to reduce seepage. In addition, this increased cost would cover stabilization of tailings, after mill closure, with a 1-in. asphalt membrane topped by 2 ft of earth and 0.5 ft of crushed rock to provide protection against future leaching and wind erosion. The cost of reducing the radiological hazards associated with uranium milling to this degree would contribute about 0.4% to the current total cost of nuclear power

  7. Bioleaching - an alternate uranium ore processing technology for India

    International Nuclear Information System (INIS)

    Abilash; Mehta, K.D.; Kumar, V.; Pandey, B.D.; Tamarakar, P.K.

    2010-01-01

    Meeting the feed supply of uranium fuel in the present and planned nuclear reactors calls for huge demand of uranium, which at the current rate of production, shows a mismatch. The processing methods at UCIL (DAE) needs to be modified/changed or re-looked into because of its very suitability in near future for low-index raw materials which are either unmined or stacked around if mined. There is practically no way to process tailings with still some values. Efforts were made to utilize such resources (low-index ore of Turamdih mines, containing 0.03% U 3 O 8 ) by NML in association with UCIL as a national endeavor. In this area, the R and D work showed the successful development of a bioleaching process from bench scale to lab scale columns and then finally to the India's first ever large scale column, from the view point of harnessing such a processing technology as an alternative for the uranium industry and nuclear sector in the country. The efforts culminated into the successful operation of large scale trials at the 2 ton level column uranium bioleaching that was carried out at the site of UCIL, Jaduguda yielding a maximum recovery of 69% in 60 days. This achievement is expected to pave the way for scaling up the activity to a 100T or even more heap bioleaching trials for realization of this technology, which needs to be carried out with the support of the nuclear sector in the country keeping in mind the national interest. (author)

  8. Application of nanofiltration to the treatment of uranium mill effluents

    International Nuclear Information System (INIS)

    Macnaughton, S.J.; McCulloch, J.K.; Marshall, K.; Ring, R.J.

    2002-01-01

    Nanofiltration is widely used in water treatment due to the lower energy requirements and higher yields than reverse osmosis. Separation characteristics are dependent on both the molecular size and charge of the dissolved species in the feed solution as well as membrane properties. In this investigation the potential of nanofiltration to remove dissolved species from uranium mill effluent has been studied. The background behind the application is discussed and the results of the first testwork programme are presented. An initial screening of seventeen commercially available membranes was completed and it was found that uranium rejections of greater than 75% were consistently achieved. Selected membranes also showed potential for the separation of radium, sulfate and manganese. (author)

  9. Uranium

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1981-01-01

    Events in the Canadian uranium industry during 1980 are reviewed. Mine and mill expansions and exploration activity are described, as well as changes in governmental policy. Although demand for uranium is weak at the moment, the industry feels optimistic about the future. (LL)

  10. Correction in the efficiency of uranium purification process by solvent extraction

    International Nuclear Information System (INIS)

    Franca Junior, J.M.

    1981-01-01

    An uranium solvent extraction, of high purification, with full advantage of absorbed uranium in the begining of process, is described. Including a pulsed column, called correction column, the efficiency of whole process is increased, dispensing the recycling of uranium losses from leaching column. With the correction column the uranium losses go in continuity, for reextraction column, increasing the efficiency of process. The purified uranium is removed in the reextraction column in aqueous phase. The correction process can be carried out with full efficiency using pulsed columns or chemical mixer-settlers. (M.C.K.) [pt

  11. Processing and Applications of Depleted Uranium Alloy Products

    Science.gov (United States)

    1976-09-01

    ammunition, weapons, gyrorotors, and ballast. Depleted uranium used in fly- wheel devices, nuclear fuel casks, and ammunition could consume a significant...from straight in the range of 0,002 to 0.060-inch TIR (total indicated runout ) with an average of 0.025-inch TIR.* Solution heat treatment of the as-cast...an envelope thickness of 0.050 inch to allow for runout and to clean up surface imperfections. The runout resulting from heat treatment was in the

  12. The extraction of uranium from wet process phosphoric acid using a liquid surfactant membrane system

    International Nuclear Information System (INIS)

    Dickens, N.; Davies, G.A.

    1984-01-01

    A liquid membrane extraction process is examined for the extraction of uranium from wet process phosphoric acid. Uranium is present in the acid in concentrations up to 100 ppm which in principle makes it ideal for treatment with a membrane process. The membrane system studied is based on extraction using DEHPA-TOPO reagents which are contained within the organic phase of a water in oil emulsion. Formulations of the emulsion membrane system have been studied, the limitations of acid temperature, P 2 O 5 concentration and solid dispersed impurities in the acid have been studied in laboratory batch experiments and in a continuous pilot plant unit capable of treating 5l of concentrated acid per minute. Data from the pilot plant work has been used to develop a flowsheet for a commercial unit based on this process. (author)

  13. Process for producing uranium carbide spheroids

    International Nuclear Information System (INIS)

    Shennan, J.V.; Ford, L.H.

    1976-01-01

    The invention deals with a method to produce UC spheroids which are filled into molded bodies of fire-proof material for fuel elements. The UC fuel particles are doubly coated: a first thin layer of pyrolytic carbon is coated at low temperature (1,200-1,400 0 C), a second layer of fire-proof material (e.g. SiC) is coated at a higher temperature (above 1,500 0 C) which holds back the fission products. The process is explained in more detail using an example. (GSCH) [de

  14. Improvements on heap leaching process for a refractory uranium ore and yellow cake precipitation process

    International Nuclear Information System (INIS)

    Feng Jianke

    2013-01-01

    Some problems such as formed harden matrix, ore heap compaction, poor permeability, and agglomeration of absorption resin occur during extracting uranium from a refractory uranium ore by heap leaching process. After some measures were taken, i.e. spraying a new ore heap by low concentration acid, two or more ore heaps in series leaching, turning ores in ore heap, the permeability was improved, acid consumption was reduced. Through precipitate circulation and aging, the yellow cake slurry in amorphous or microlite form was transformed to crystal precipitate, thus uranium content in yellow cake was improved, and water content in yellow cake was lowered with good economic benefits. (author)

  15. Separating uranium by laser: the atomic process

    Energy Technology Data Exchange (ETDEWEB)

    Destro, Marcelo G.; Damiao, Alvaro J.; Neri, Jose W.; Schwab, Carlos; Rodrigues, Nicolau A.S.; Riva, Rudimar [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados

    1996-07-01

    Among the countries around the world that utilizes nuclear energy, several ones are investing significantly in the development of laser techniques applied to isotope separation. In Brazil these studies are concentrated in one research institute, the IEAv (Institute for Advanced Studies), and aim at demonstrating the viability of this process using, as much as possible, resources available in the country. In this paper we briefly describe the laser methods for isotope separation, giving an overview of the present research and development status in this area. We also show some results obtained our laboratories. We focused this report on the atomic route for laser isotope separation, mainly in the areas of laser development and spectroscopy. (author)

  16. Separating uranium by laser: the atomic process

    International Nuclear Information System (INIS)

    Destro, Marcelo G.; Damiao, Alvaro J.; Neri, Jose W.; Schwab, Carlos; Rodrigues, Nicolau A.S.; Riva, Rudimar

    1996-01-01

    Among the countries around the world that utilizes nuclear energy, several ones are investing significantly in the development of laser techniques applied to isotope separation. In Brazil these studies are concentrated in one research institute, the IEAv (Institute for Advanced Studies), and aim at demonstrating the viability of this process using, as much as possible, resources available in the country. In this paper we briefly describe the laser methods for isotope separation, giving an overview of the present research and development status in this area. We also show some results obtained our laboratories. We focused this report on the atomic route for laser isotope separation, mainly in the areas of laser development and spectroscopy. (author)

  17. Non-filtration method of processing uranium ores

    International Nuclear Information System (INIS)

    Laskorin, B.N.; Vodolazov, L.I.; Tokarev, N.N.; Vyalkov, V.I.; Goldobina, V.A.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1977-01-01

    The development of the non-filtration sorption method has lead to procedures of the sorption leaching and the extraction desorption, which have made it possible to intensify the processing of uranium ores and to improve greatly the technical and economic indexes by eliminating the complex method of multiple filtration and re-pulping of cakes. This method makes it possible to involve more poor uranium raw materials, at the same time extracting valuable components such as molybdenum, vanadium, copper, etc. Considerable industrial experience has been acquired in the sorption of dense pulp with a solid-to-liquid phase ratio of 1:1. This has led to a plant production increase of 1.5-3.0 times, an increase of uranium extraction by 5-10%, a two- to- three-fold increase of labour capacity of the main workers, and to a several-fold decrease of reagents, auxiliary materials, electric energy and vapour. This non-filtration method is a continuous process in all its phases thanks to the use of high-yield and high-power equipment for high-density pulps. (author)

  18. Treatment of uranium-bearing wastewater by vacuum membrane distillation

    International Nuclear Information System (INIS)

    Duan Xiaolin; Li Qicheng; Chen Bingbing

    2006-01-01

    The removal of uranium from wastewater was carried out by vacuum membrane distillation (VMD) using microporous polypropylene membrane. The effects of feed temperature, mass concentration of U, flow rate and vacuum-side pressure on permeation flux and rejection were studied. The optimum experimental conditions are as follows: feed flow rate is 0.5 m/s, feed temperature is 55 degree C, vacuum-side pressure is 2.66 kPa. When the mass concentrations of U in the feed solution range from 1 mg/L to 9 mg/L, the membrane flux is 3.5 kg/(m 2 ·h) and the rejection rate is 99.1% under the optimum conditions. The water separated from uranium solution by vacuum membrane distillation could meet the state-controlled discharge standard 0.05 mg/L. The VMD as a novel technology will play an important role in the treatment of uranium-bearing wastewater. (authors)

  19. Process for in-situ leaching of uranium

    International Nuclear Information System (INIS)

    Espenscheid, W.F.; Yan, F.Y.

    1983-01-01

    The present invention relates to the recovery of uranium from subterranean ore deposits, and more particularly to an in-situ leaching operation employing an aqueous solution of sulfuric acid and carbon dioxide as the lixiviant. Uranium is solubilized in the lixiviant as it traverses the subterranean uranium deposit. The lixiviant is subsequently recovered and treated to remove the uranium

  20. 76 FR 63330 - Policy Regarding Submittal of Amendments for Processing of Equivalent Feed at Licensed Uranium...

    Science.gov (United States)

    2011-10-12

    ... Processing of Equivalent Feed at Licensed Uranium Recovery Facilities AGENCY: Nuclear Regulatory Commission... NRC and Agreement State-licensed uranium recovery site. This action is necessary to correct several... read ``(see Page A2 of SECY-99-011, ``Draft Rulemaking Plan: Domestic Licensing of Uranium and Thorium...

  1. Recent Developments in the Treatment of Uranium Ores from the Elliot Lake District

    Energy Technology Data Exchange (ETDEWEB)

    Downes, K W [Extraction Metallurgy Division, Department of Mines and Technical Surveys, Ottawa (Canada)

    1967-06-15

    A summary of the results obtained during investigations on the treatment of uranium ores from the Elliot Lake district in the laboratories of the Mines Branch, and of developments in operating procedures introduced in the uranium mills in the Elliot Lake district, is presented. Concentration of Elliot Lake ore on a pilot-plant scale by a combined gravity-flotation procedure yielded a 90% recovery of uranium at a ratio of concentration of 2.4 to 1.0. The mineralogical composition of the ore, the flow sheet used and the reagents employed are described. An approximate cost estimate indicates that, although the capacity of an existing uranium leaching plant would be doubled by introducing the procedure, the production cost per pound of U{sub 3}0{sub 8} would not be affected. Bacterial leaching of Elliot Lake ore on a laboratory scale yielded, under favourable conditions, extractions of 90 per cent in 5 weeks, and of 95 per cent in 15 weeks. The conditions that were found to influence the leaching results are outlined, and the effects of the leaching solutions are discussed. The purification of ion exchange eluates by liquid-liquid extraction, using tri-n-butyl phosphate, dibutyl butylphosphonate and tri-capryl amine in a continuous process, yielded solutions from which refined ammonium diuranate was precipitated using gaseous ammonia. The effectiveness of the three extractants is discussed, and the effects of the procedures employed on the production costs per pound of U{sub 3}O{sub 8} is estimated. Some improvements in operating procedures introduced in the Elliot Lake district uranium mills are briefly described, and their effects on the operations are indicated. Present methods of controlling radiological pollution of drainage waters by uranium mill tailings are outlined. (author)

  2. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

    1979-04-01

    Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.

  3. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    International Nuclear Information System (INIS)

    Hyder, M.L.; Perkins, W.C.; Thompson, M.C.; Burney, G.A.; Russell, E.R.; Holcomb, H.P.; Landon, L.F.

    1979-04-01

    Uranium fuels containing 235 U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction with dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of 238 Pu is high enough to make its recovery desirable. Most of the 238 Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, 239 Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse

  4. Set up of Uranium-Molybdenum powder production (HMD process)

    International Nuclear Information System (INIS)

    Lopez, Marisol; Pasqualini, Enrique E.; Gonzalez, Alfredo G.

    2003-01-01

    Powder metallurgy offers different alternatives for the production of Uranium-Molybdenum (UMo) alloy powder in sizes smaller than 150 microns. This powder is intended to be used as a dispersion fuel in an aluminum matrix for research, testing and radioisotopes production reactors (MTR). A particular process of massive hydriding the UMo alloy in gamma phase has been developed. This work describes the final adjustments of process variables to obtain UMo powder by hydriding-milling-de hydriding (HMD) and its capability for industrial scaling up. (author)

  5. Determination of uranium and thorium during chemical treatment of monazite

    International Nuclear Information System (INIS)

    El-Nadi, Y.A.; Daoud, J.A.; Aly, H.F.; Kregsamer, P.

    2000-01-01

    Total reflection x-ray fluorescence (TXRF) is a very useful technique for both qualitative and quantitative analysis because of its high detection power and its needed to small sample volumes (less than 100 μl are sufficient). In this work TXRF was used to determine the initial concentrations of the elements included in monazite sand and following up the chemical steps for treatment of monazite with special attention to uranium and thorium concentration as well as lanthanides. The results were compared to those obtained from EDXRF and ICP-MS techniques. (author)

  6. Metallurgical processing of the uranium-0.75 titanium alloy

    International Nuclear Information System (INIS)

    Jessen, N.C.

    1976-01-01

    Although the addition of titanium is an effective means of strengthening uranium, careful control of casting, homogenization, and heat treatment are necessary to optimize mechanical properties. Quenching of the alloy provides increased strength and elongation; however, subsequent low temperature aging will increase the strength even higher at the sacrifice of ductility. The properties of the alloy are quench rate sensitive and quenching produces high residual stresses in the alloy. The residual stresses can be reduced by mechanical deformation with only slight degradation of the mechanical properties. 15 figures

  7. Uranium hexaflouride freezer/sublimer process simulator/trainer

    International Nuclear Information System (INIS)

    Carnal, C.L.; Belcher, J.D.; Tapp, P.A.; Ruppel, F.R.; Wells, J.C.

    1991-01-01

    This paper describes a software and hardware simulation of a freezer/sublimer unit used in gaseous diffusion processing of uranium hexafluoride (UF 6 ). The objective of the project was to build a plant simulator that reads control signals and produces plant signals to mimic the behavior of an actual plant. The model is based on physical principles and process data. Advanced Continuous Simulation Language (ACSL) was used to develop the model. Once the simulation was validated with actual plant process data, the ACSL model was translated into Advanced Communication and Control Oriented Language (ACCOL). A Bristol Babcock Distributed Process Controller (DPC) Model 3330 was the hardware platform used to host the ACCOL model and process the real world signals. The DPC will be used as a surrogate plant to debug control system hardware/software and to train operators to use the new distributed control system without disturbing the process. 2 refs., 4 figs

  8. In situ leaching process for recording uranium values

    International Nuclear Information System (INIS)

    McKnight, W.M.; Timmins, T.H.; Sherry, H.S.

    1977-01-01

    A method of recovering uranium values from a subterranean deposit comprising: injecting an alkaline carbonate lixiviant into said deposit; flowing said alkaline carbonate lixiviant through said deposit to dissolve said uranium values into said lixiviant; producing said lixiviant and said dissolved uranium values from said deposit; flowing said lixiviant and said dissolved uranium values through an adsorption material to adsorp said uranium values from said lixiviant; eluting said adsorption material with an eluant of ammonium carbonate to desorb said uranium values from said adsorption material into said eluate in a concentration greater than in said lixiviant; heating said eluate and said desorbed uranium values to vaporize off ammonia and carbon dioxide therefrom, thereby causing uranium values to crystallize from the eluate; and recovering said solid uranium values

  9. Treatment of back flow fluids from shale gas exploration with recovery of uranium

    International Nuclear Information System (INIS)

    Gajda, D.; Zakrzewska-Koltuniewicz, G.; Abramowska, A.; Kiegiel, K.; Niescior-Borowinska, P.; Miskiewicz, A.; Olszewska, W.; Kulisa, K.; Samszynski, Z.; Drzewicz, P.; Konieczynska, M.

    2015-01-01

    Shale gas exploitation is the cause of many social protests. According to the protesters gas extraction technology threatens the environment: it consumes huge amounts of water, creates danger of poisoning drinking water, the formation of toxic wastewater, air contamination, noise, etc. Hydro-fracturing fluids could also leach radioactive isotopes e.g. uranium from the rock. The upper content of the main elements found in examined back flow fluids in Poland are the following: chlorine: 100.00 Kg/m 3 , sodium: 40.00 kg/m 3 , potassium: 0.90 kg/m 3 , lithium: 0.15 kg/m 3 , magnesium: 2.00 kg/m 3 , calcium: 20.00 kg/m 3 , strontium: 0.80 kg/m 3 and cesium: 0.06 kg/m 3 while the upper content of trace elements are the following: uranium: 3.5 g/m 3 , lanthanum: 12.4 g/m 3 , vanadium: 1.3 g/m 3 , yttrium: 1.3 g/m 3 , molybdenum: 2.0 g/m 3 and manganese: 9.7 g/m 3 . The recovery of uranium, and other valuable metals, from back flow fluids will reduce an environmental impact of hydro-fracturing process. This poster details the treatment of back flow fluids in Poland allowing rare earth elements and uranium recovery

  10. Modelling of uranium/plutonium splitting in purex process

    International Nuclear Information System (INIS)

    Boullis, B.; Baron, P.

    1987-06-01

    A mathematical model simulating the highly complex uranium/plutonium splitting operation in PUREX process has been achieved by the french ''Commissariat a l'Energie Atomique''. The development of such a model, which includes transfer and redox reactions kinetics for all the species involved, required an important experimental work in the field of basis chemical data acquisition. The model has been successfully validated by comparison of its results with those of specific trials achieved (at laboratory scale), and with the available results of the french reprocessing units operation. It has then been used for the design of french new plants splitting operations

  11. Optimization of desalting process with centrifugation for condensation process of uranium from sea water

    International Nuclear Information System (INIS)

    Yamamoto, Tatsuya; Takase, Hisao; Fukuoka, Fumio

    1984-01-01

    Optimization of desalting of the slurry on the condensation process by the deposited slurry method for the recovery of uranium from sea water was studied. We have already published that the uranium rich deposit containing seven ppm uranium could be made on the sea bottom by the deposited slurry method. Uranium can be transferred to the anion exchange resin from titanic acid in the slurry. But in this case Cl - ions obstruct the adsorption of uranium on the anion exchange resin, so the slurry must be desalted before RIP method. It is considered that the cost of desalting of the slurry stage would be a large portion of the capital cost for the recovery of uranium from sea water. The cost of water required is comparable to the cost of energy so that the objective function consists of the cost of energy and the quantity of water. The consumption of energy and water required for desalting of the slurry with the multi-stage centrifugation were oprimized based on dynamic programming. (author)

  12. Characterization of depleted uranium oxides fabricated using different processing methods

    International Nuclear Information System (INIS)

    Hastings, E.P.; Lewis, C.; FitzPatrick, J.; Rademacher, D.; Tandon, L.

    2008-01-01

    Identifying both physical and chemical characteristics of Special Nuclear Material (SNM) production processes is the corner stone of nuclear forensics. Typically, processing markers are based on measuring an interdicted sample's bulk chemical properties, such as the elemental or isotopic composition, or focusing on the chemical and physical morphology of only a few particles. Therefore, it is imperative that known SNM processes be fully characterized from bulk to trace level for each particle size range. This report outlines a series of particle size measurements and fractionation techniques that can be applied to a bulk SNM powders, categorizing both chemical and physical properties in discrete particle size fractions. This will be demonstrated by characterizing the process signatures of a series of different depleted uranium oxides prepared at increasing firing temperatures (350-1100 deg C). Results will demonstrate how each oxides' material density, particle size distribution, and morphology varies. (author)

  13. Non-polluting treatment of alkaline uranium effluents contaning SO42- ions

    International Nuclear Information System (INIS)

    Berger, Bernard.

    1978-01-01

    New non-polluting process for treating uranium effluents from the alkaline digestion of an uranium ore containing sulfur, which makes it possible, on the one hand, to extract uranium and SO 4 2- contained in these effluents allowing the recycling of the sole alkaline carbonates and/or bicarbonates involved, towards the digestion of the ore and on the other hand the separation of the mixture uranium and SO 4 2- ions extracted simultaneously to obtain relatively pure uranium in oxide form [fr

  14. Uranium

    International Nuclear Information System (INIS)

    Perkin, D.J.

    1982-01-01

    Developments in the Australian uranium industry during 1980 are reviewed. Mine production increased markedly to 1841 t U 3 O 8 because of output from the new concentrator at Nabarlek and 1131 t of U 3 O 8 were exported at a nominal value of $37.19/lb. Several new contracts were signed for the sale of yellowcake from Ranger and Nabarlek Mines. Other developments include the decision by the joint venturers in the Olympic Dam Project to sink an exploration shaft and the release of an environmental impact statement for the Honeymoon deposit. Uranium exploration expenditure increased in 1980 and additions were made to Australia's demonstrated economic uranium resources. A world review is included

  15. Uranium

    International Nuclear Information System (INIS)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-01-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U 3 O 8 ; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables

  16. Preliminary study for treatment methodology establishment of liquid waste containing uranium in refining facility lagoon

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Jik; Lee, Kune Woo; Won, Hui Jun; Ahn, Byung Gil; Shim, Joon Bo

    1999-12-01

    The preliminary study which establishes the treatment methodology of the sludge waste containing uranium in the conversion facility lagoon was performed. The property of lagoon liquid waste such as the initial water content, the density including radiochemical analysis results were obtained using the samples taken from the lagoon. The objective of this study is to provide some basically needed materials for selection of the most proper lagoon waste treatment methodology by reviewing the effective processes and methods for minimizing the secondary waste resulting from the treatment and disposition of large amount of radioactive liquid waste according to the facility closing. The lagoon waste can be classified into two sorts, such as supernatant and precipitate. The supernatants contain uranium less than 5 ppm and their water content are about 35 percent. Therefore, supernatants are solutions composed of mainly salt components. However, the precipitates have lots of uranium compound contained in the coagulation matrix, and are formed as two kinds of crystalline structures. The most proper method minimizing the secondary waste would be direct drying and solidification of the supernatants and precipitates after separation of them by filtering. (author)

  17. Study on the treatment of waste waster containing uranium by organic modified vermiculite

    International Nuclear Information System (INIS)

    Liu Wenjuan; Zeng Yanhong

    2012-01-01

    The adsorption capability of uranium on organic modified Vermiculite was studied. The influence factors of the amount of adsorbent, initial pH, initial concentration of uranium and adsorption time have been investigated too. Through the orthogonal test, the primary factors of impacting the adsorption treatment can be obtained. Finally, the preliminary research and analysis on the principle adsorption of organic modified vermiculite test of uranium have been conducted. The results show that: Modifying Vermiculite by CTMAB makes Vermiculite adsorption capacity stronger when treating solution containing uranium. Combined flocculants with vermiculite to treat with low concentration of uranium solution has synergy, significantly enhancing its adsorption capacity. The impact factors of organic modified vermiculite's adsorption of uranium are adsorbent dosage, pH, initial concentration of uranium solution and adsorption time. The best adsorption pH is between 5∼6.5. (authors)

  18. Managing the heritage of east-German uranium mining and uranium processing

    International Nuclear Information System (INIS)

    Hagen, M.

    1997-01-01

    The corporate aim of the WISMUT GmbH, in accordance with the current statutory regulations of the Federal Republic of Germany, is the decommissioning of its installations as well as the reclamation and revegetation of a landscape and an environment on which decades of uninhibited extraction and processing of uranium ore have left their imprint. Expenditure for this major ecological project of international scale is put at 13 billion marks. These funds are provided by the Federal government in the course of an envisaged period of 10 to 15 years. They enable WISMUT to buy the best know-how to be obtained in Germany and abroad for the decommissioning and reclamation works. (orig./RHM) [de

  19. Uranium Rirang ore processing: extraction of uranium from Rirang ore digestion solution with tributyl phosphate

    International Nuclear Information System (INIS)

    Arief, E. R.; Zahardi; Susilaningtyas

    1998-01-01

    Uranium is extracted from Rirang ore acid digestion solution containing rare earths. A mixture of tributyl phosphate solvent and kerosene diluent is employed. Several parameters of solvent extraction have been studied included aqueous to organic phase ratio, H 2 O 2 reductor concentration and Tbp concentration in the solvent mixture, as well as the aqueous to organic phase ratio in the stripping process. The optimum conditions for the extraction step include the use of 25% H 2 O 2 (v/v), one to one aqueous to organic ratio, and 40% Tbp in kerosene. The extraction recovery for U, RE, Th, and PO 4 3 - are 99%, 4%, 70%, and 30%, respectively. The stripping step optimum conditions include the use of one to five organic to aqueous phase ratio 0.24 N HNO 3 . and the stripping recovery for U, RE, Th, and PO 4 3 - are 84%, 80%, 72%, and 83%, respectively

  20. Uranium separation from phosphates and the fuel cycle process

    International Nuclear Information System (INIS)

    Lavi, J.

    1978-01-01

    A short introduction on the recycle of uranium and plutonium is presented. The uranium world market at present, the prices during the last few years, the actual requirements and those for the years 1978-1983 are given. In a special paragraph the present resources of uranium in Israel as well as the extraction possibilities are discussed. (B.G.)

  1. Uranium enrichment in Europe by the gas centrifuge process

    International Nuclear Information System (INIS)

    Severin, D.J.E.

    1975-01-01

    To begin with, this lesson gives an outline of the expected energy demand of the Western World and the concentration of the European companies participating in uranium enrichment by the gas centrifuge method. Next, a) the principles of the gas centrifuge method are outlined, b) its advantages over other industrial processes are stressed, and c) the characteristic data of complete plants are given. The existing German, Dutch, and British pilot plants are mentioned as examples for the perfected state of the process. The Capenhurst (UK) and Almedo (NL) demonstration plants, each with a capacity of 200 t SW/a, will have been extended to 2 x 1.000 t SW/a by 1982. Finally, economic data of the gas centrifuge process are given. The term 'separative work' is explained in an annex. (GG) [de

  2. Extraction of uranium with emulsion membrane process use tributylphosphate extractant

    International Nuclear Information System (INIS)

    Basuki, K.T.; Sudibyo, R.; Bambang EHB; Muhadi, A.W.

    1996-01-01

    To increase the effectiveness of extraction process with so for to occur, it was tried the extraction with a couple of extraction and stripping process. This couple process was called liquid membrane emulsion. As membrane was used mix surfactant (Span-80), tributylphosphate in kerosene, natrium carbonate, while as a feeder was uranium solution with 500 concentration ppm in 0.5 - 3 M nitrate acid. In this experiment the variable investigated were % surfactant (1 - 5 %), rotary speed for membrane making (2,500 - 10.000 rpm). The optimal condition result of experiment were 5 % surfactant, 3 M nitrate acid, rotary speed 10.000 rpm and (Kd eksU ) 57 %, and (Kd strippU ) 87 %, Kd eksU at liquid-liquid extraction is 44 %. (author)

  3. PROCESS FOR RECOVERY OF URANIUM VALUES FROM IMPURE SOLUTIONS THEREOF

    Science.gov (United States)

    Kilner, S.B.

    1959-11-01

    A process is presented for the recovery of uraninm values from impure solutions which are obtained, for example, by washing residual uranium salt or uranium metal deposits from stainless steel surfaces using an aqueous or certain acidic aqueous solutions. The solutions include uranyl and oxidized iron, chromium, nickel, and copper ions and may contain manganese, zinc, and silver ions. In accordance with one procedure. the uranyl ions are reduced to the uranous state, and the impurity ions are complexed with cyanide under acidic conditions. The solution is then treated with ammonium hydroxide or alkali metal hydroxide to precipitate uranous hydroxide away from the complexed impurity ions in the solution. Alternatively, an excess of alkali metal cyanide is added to the reduced solution until the solution becomes sufficiently alkaline for the uranons hydroxide to precipitate. An essential feature in operating the process is in maintaining the pH of the solution sufficiently acid during the complexing operation to prevent the precipitation of the impurity metal hydroxides.

  4. TEM and SEM observation of uranium induced renal necrosis and the result of chelates treatment on rats

    International Nuclear Information System (INIS)

    Sun Shiquan; Li Baoxing; Lai Chixiang; You Zhanyun

    1987-01-01

    The TEM (transmission electron microscope) and SEM (scanning electron microscope) observation of uranium induced renal necrosis and the result of chelates treatment on rats are reported. Ultrastructural changes in kidney related with the impairment of intracellular fluid transportation can be found after acute uranium intoxication in rats, such as: condensation and swelling of mitochondria, matrix edema, dilatation of intercellular space, disappearance of basal folds, thickening of basal web, intensification of basal lamina of the proximal convoluted tubule epithelium cells, and foot processes swelling, diminishing of endothelium fenestrae of the renal glomerulus. Heavy metal chelates DTPA and H-73-10 treatment may result in intracellular fluid accumulation and condensed grannule formation in lysosome. Treatment with these chelates in the critical stage of uranium intoxication may accelerate the necrosis instead of diminishing. This may be related to the augment of the load of lysosome and intracellular system of fluid transportation

  5. About the elaboration of pure uranium dicarbide

    International Nuclear Information System (INIS)

    Besson, J.; Blum, P.; Guinet, Ph.; Spitz, J.

    1963-01-01

    In order to develop methods for the elaboration of as pure as possible uranium dicarbide, the authors report the study of different elaboration processes based on the reaction between uranium and carbon, or between uranium and hydrocarbon, or between uranium oxide and carbon. They finally choose a method which comprises an arc-induced fusion of a mixture of uranium dioxide and carbon. The fusion process is described. The influence of thermal treatments is discussed as well as the graphite electrode carburization

  6. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    Recent decisions by the Australian Government will ensure a significant expansion of the uranium industry. Development at Roxby Downs may proceed and Ranger may fulfil two new contracts but the decision specifies that apart from Roxby Downs, no new mines should be approved. The ACTU maintains an anti-uranium policy but reaction to the decision from the trade union movement has been muted. The Australian Science and Technology Council (ASTEC) has been asked by the Government to conduct an inquiry into a number of issues relating to Australia's role in the nuclear fuel cycle. The inquiry will examine in particular Australia's nuclear safeguards arrangements and the adequacy of existing waste management technology. In two additional decisions the Government has dissociated itself from a study into the feasibility of establishing an enrichment operation and has abolished the Uranium Advisory Council. Although Australian reserves account for 20% of the total in the Western World, Australia accounts for a relatively minor proportion of the world's uranium production

  7. Uranium

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The French Government has decided to freeze a substantial part of its nuclear power programme. Work has been halted on 18 reactors. This power programme is discussed, as well as the effect it has on the supply of uranium by South Africa

  8. Measures for waste water management from recovery processing of Zhushanxia uranium deposit

    International Nuclear Information System (INIS)

    Liu Yaochi; Xu Lechang

    2000-01-01

    Measures for waste water management from recovery processing of Zhushanxia uranium deposit of Wengyuan Mine is analyzed, which include improving process flow, recycling process water used in uranium mill as much as possible and choosing a suitable disposing system. All these can decrease the amount of waste water, and also reduce costs of disposing waste water and harm to environment

  9. Operating experience in processing of differently sourced deeply depleted uranium oxide and production of deeply depleted uranium metal ingots

    International Nuclear Information System (INIS)

    Manna, S.; Ladola, Y.S.; Sharma, S.; Chowdhury, S.; Satpati, S.K.; Roy, S.B.

    2009-01-01

    Uranium Metal Plant (UMP) of BARC had first time experience on production of three Depleted Uranium Metal (DUM) ingots of 76kg, 152kg and 163kg during March 1991. These ingots were produced by processing depleted uranyl nitrate solution produced at Plutonium Plant (PP), Trombay. In recent past Uranium Metal Plant (UMP), Uranium Extraction Division (UED), has been assigned to produce tonnage quantity of Deeply DUM (DDUM) from its oxide obtained from PP, PREFRE and RMP, BARC. This is required for shielding the high radioactive source of BHABHATRON Tele-cobalt machine, which is used for cancer therapy. The experience obtained in processing of various DDU oxides is being utilized for design of large scale DDU-metal plant under XIth plan project. The physico- chemical characteristics like morphology, density, flowability, reactivity, particle size distribution, which are having direct effect on reactivity of the powders of the DDU oxide powder, were studied and the shop-floor operational experience in processing of different oxide powder were obtained and recorded. During campaign trials utmost care was taken to standardized all operating conditions using the same equipment which are in use for natural uranium materials processing including safety aspects both with respect to radiological safety and industrial safety. Necessary attention and close monitoring were specially arranged and maintained for the safety aspects during the trial period. In-house developed pneumatic transport system was used for powder transfer and suitable dust arresting system was used for reduction of powder carry over

  10. Uranium extraction process in a sulfuric medium by means of liquid emulsified membranes

    International Nuclear Information System (INIS)

    Monteillet, A.

    1985-02-01

    Uranium ore processing, after leaching by sulfuric acid, by liquid-liquid extraction is a rather heavy process, not suitable for small deposits. Extraction by emulsions was suggested. In this process the leachate is contacted with an oil in water type emulsion, a liquid organic membrane is formed by the continuous phase. Uranium complexes diffuse through the liquid membrane towards the dispersed aqueous phase of the emulsion (stripping solution). Uranium is recovered by breaking the emulsion. Are successively studied: development of stable emulsions, influence of emulsion composition on uranium transfer kinetics, transfer mechanisms through the membrane and modelling of kinetics data obtained in the experimental study [fr

  11. Liquid membranes and process for uranium recovery therewith

    International Nuclear Information System (INIS)

    Frankenfeld, J.W.; Li, N.N.T.; Bruncati, R.L.

    1981-01-01

    A liquid membrane system consisting of water-in-oil type emulsions dispersed in water, which is capable of extracting uranium-containing ions from an aqueous feed solution containing uranium ions at a temperature in the range of 25 0 C to 80 0 C, is described. The emulsion comprises an aqueous interior phase surrounded by a surfactant-containing exterior phase. The exterior phase is immiscible with the interior phase and comprises a transfer agent capable of transporting selectively the desired uranium-containing ions and a solvent for the transfer agent. The interior phase comprises a reactant capable of removing uranium-containing ions from the transfer agent and capable of changing the valency of the uranium in uranium-containing ions to a second valency state and converting the uranium-containing ions into a nonpermeable form. (U.K.)

  12. Process for recovering uranium and other base metals

    International Nuclear Information System (INIS)

    Jan, R. J-J.

    1979-01-01

    Uranium and other base metals are leached from their ores with aqueous solutions containing bicarbonate ions that have been generated or reconstituted by converting other non-bicarbonate anions into bicarbonate ions. The conversion is most conveniently effected by contacting solutions containing SO 4 - and Cl - ions with a basic anion exchange resin so that the SO 4 - and Cl - ions are converted into or exhanged for HCO 3 - ions. CO 2 may be dissolved in the solution so it is present during the exhange. The resin is preferably in bicarbonate form prior to contact and CO 2 partial pressure is adjusted so that the resin is not fouled by depositing metal precipitates. In-situ uranium mining is conducted by circulating such solutions through the ore deposit. Oxidizing agents are included in the injected lixiviant. The leaching strength of the circulating bicarbonate lixiviant is maintained by converting the anions generated during leaching or above-ground recovery processes into HCO 3 - ions. The resin may conveniently be eluted and reformed intermittently

  13. Process for recovering uranium and other base metals

    International Nuclear Information System (INIS)

    Jan, R.J.

    1981-01-01

    Uranium and other base metals are leached from their ores with aqueous solutions containing bicarbonate ions that have been generated or reconstituted by converting other non-bicarbonate anions into bicarbonate ions. The conversion is most conveniently effected by contacting solutions containing SO 4 -- and C1 - ions with a basic anion exchange resin so that the SO 4 -- and Cl - ions are converted into or exchanged for HCO 3 - ions. CO 2 may be dissolved in the solution so it is present during the exchange. The resin is preferably in bicarbonate form prior to contact and CO 2 partial pressure is adjusted so that the resin is not fouled by depositing metal precipitates. In-situ uranium mining is conducted by circulating such solutions through the ore deposit. Oxidizing agents are included in the injected lixiviant. The leaching strength of the circulating bicarbonate lixiviant is maintained by converting the anions generated during leaching or above-ground recovery processes into HCO 3 - ions. The resin may conveniently be eluted and performed intermittently. (author)

  14. Disposal of residue from uranium ore processing in France

    International Nuclear Information System (INIS)

    Crochon, Ph.

    2011-01-01

    Between 1949 and 2001, French mines produced 76, 000 metric tons of uranium and 50 million metric tons of ore, processing residues are stored at 17 sites (in ponds enclosed by dykes or in former open-cast mines) subject to ICPE (classified facility for environment protection) regulation. These disposal sites cover surface areas of between one and several tens of hectares and several thousands to several millions of metric tons of waste are stored at them. When uranium mining stopped in France, these sites were redeveloped, with caps placed over the residue to provide mechanical and radiological protection. All these sites are still monitored by AREVA. In the last fifteen years, these sites have been the subject of a number of studies, especially regarding the long-term evolution and impact of the residue. These studies are now being pursued within the framework of the national plan for the management of nuclear materials and waste (PNGMDR). A regulatory and institutional framework regarding long-term management of these disposal sites needs to be defined. (author)

  15. Processing of stored uranium tetrafluoride for productive use

    International Nuclear Information System (INIS)

    Whinnery, W.N. III

    1987-01-01

    Waste uranium tetrafluoride (UF4) was created from converting uranium hexafluoride (UF6) to UF4 for generation of hydrogen fluoride. This resulted in more tails cylinders being made available in the early days of the Paducah Gaseous Diffusion Plant. A need arose for the UF4; however, a large portion of the material was stored outside in 55-gallon drums where the material became caked and very hard. Chemical operations crushed, ground, and screened a large portion of the waste UF4 from 1981-1987. Over 111,935,000 pounds of the material has been processed and put into productive use at Westinghouse Materials Company of Ohio or at Department of Defense facilities. This long-term effort saved the disposal cost of the material which is estimated at $9,327,900. In addition, the work was for an outside contract which lowered the operating cost of the Chemical Operations Department by $4,477,400. Disposal options for the material still present in the current inventory are outlined

  16. Process for recovering uranium and other base metals

    International Nuclear Information System (INIS)

    Jan, R.J.

    1984-01-01

    Uranium and other base metals are leached from their ores with aqueous solutions containing bicarbonate ions that have been generated or reconstituted by converting other non-bicarbonate anions into bicarbonate ions. The conversion is most conveniently effected by contacting solutions containing SO 4 2- and Cl - ions with a basic anion exchange resin so that the SO 4 2- and Cl - ions are converted into or exchanged for HCO 3 - ions. CO 2 may be dissolved in the solution so it is present during the exchange. The resin is preferably in bicarbonate form prior to contact and CO 2 partial pressure is adjusted so that the resin is not fouled by depositing metal precipitates. In-situ uranium mining is conducted by circulating such solutions through the ore deposit. Oxidizing agents are included in the injected lixiviant. The leaching strength of the circulating bicarbonate lixiviant is maintained by converting the anions generated during leaching or above-ground recovery processes into HCO 3 - ions. The resin may conveniently be eluted and reformed intermittently

  17. Selectivity of NF membrane for treatment of liquid waste containing uranium

    International Nuclear Information System (INIS)

    Oliveira, Elizabeth E.M.; Barbosa, Celina C.R.; Afonso, Julio C.

    2013-01-01

    The performance of two nanofiltration membranes were investigated for treatment of liquid waste containing uranium through two conditions permeation: permeation test and concentration test of the waste. In the permeation test solution permeated returned to the feed tank after collected samples each 3 hours. In the test of concentration the permeated was collected continuously until 90% reduction of the feed volume. The liquid waste ('carbonated water') was obtained during conversion of UF 6 to UO 2 in the cycle of nuclear fuel. This waste contains uranium concentration on average 7.0 mg L -1 , and not be eliminated to the environmental. The waste was permeated using a cross-flow membrane cell in the pressure of the 1.5 MPa. The selectivity of the membranes for separation of uranium was between 83% and 90% for both tests. In the concentration tests the waste was concentrated around for 5 times. The surface layer of the membranes was evaluated before and after the tests by infrared spectroscopy (ATR-FTIR), field emission microscopy (FESEM) and atomic force spectroscopy (AFM). The membrane separation process is a technique feasible to and very satisfactory for treatment the liquid waste. (author)

  18. UDAD, Radiation Exposure to Man at Uranium Processing Plant

    International Nuclear Information System (INIS)

    Momeni, M.H.; Yuan, Y.; Zielen, A.J.

    1983-01-01

    1 - Description of problem or function: The Uranium Dispersion and Dosimetry (UDAD) program provides estimates of potential radiation exposure to individuals and to the general population in the vicinity of a uranium processing facility such as a uranium mine or mill. Only transport through the air is considered. Exposure results from inhalation, external irradiation from airborne and ground- deposited activity, and ingestion of foodstuffs. Individual dose commitments, population dose commitments, and environmental dose commitments are computed. The program was developed for application to uranium mining and milling; however, it may be applied to dispersion of any other pollutant. 2 - Method of solution: The removal of radioactive particles from a contaminated area such as uranium tailings by wind action is estimated from theoretical and empirical wind-erosion equations according to the wind speed, particle size distribution, surface roughness, and other parameters. Atmospheric concentrations of radioactivity from specific sources are calculated by means of a dispersion-deposition-resuspension model. Source depletion as a result of deposition, fallout of the heavier particulates, and radioactive decay and ingrowth of radon daughters are included in a sector-averaged, Gaussian plume dispersion model. The average air concentration at any given receptor location is assumed to be constant during each annual release period, but to increase from year to year because of resuspension. Surface contamination is estimated by including buildup from deposition, ingrowth of radio- active daughters, and removal by radioactive decay, weathering, and other environmental processes. Deposition velocity is estimated on the basis of particle size, density, and physical and chemical environmental conditions which influence the behavior of the smaller particles. Calculation of the inhalation dose to an individual is based on the ICRP Task Group Lung Model (TGLM). Estimates of the dose to

  19. Uranium and Thorium in zircon sands processed in Northeastern Brazil

    International Nuclear Information System (INIS)

    Hazin, Clovis A.; Farias, Emerson E. G. de

    2008-01-01

    Zircon the main mineral of zirconium is a silicate mineral product (ZrSiO 4 ) obtained from beach sand deposits, along with other minerals such as kyanite, ilmenite, and rutile. All zircons contain some radioactive impurities due to the presence of uranium, thorium and their respective decay products in the crystalline structure of zircon, as well as potassium-40. Uranium and thorium substitute Zr 4+ in the mineral through an internal process called isomorphous replacement of zirconium. For this study, samples were collected both from a mineral sand processing plant located in the coastal region of Northeastern brazil and from the beach sands used in the process. The aim of this study was to assess the 238 U, 232 Th and 40 K contents in the beach sands and in the mineral products extracted from the sands in that facility, with special emphasis on zircon. Measurements were performed through gamma spectrometry, by using a high-purity germanium detector (HPGe) coupled to a multichannel analyzer. Activity concentration for 238 U and 232 Th in zircon sands ranged from 5462±143 to 19286±46 Bq kg -1 and from 1016±7 to 7162±38 Bq kg -1 , respectively. For 40 K, on the other hand, activity concentration values ranged from 81±14 to 681±26 Bq Kg -1 . The results of the measurements carried out for raw sand samples showed activity concentrations between 2.7±0.6 and 7.9±0.9 Bq kg -1 and 6.5±0.4 and 9.4±0.6 Bq kg -1 for 238 U and 23T h respectively, and from 48.8±3.1 to 76.1±2.4 Bq kg -1 for 40 K. Activity concentrations of 238 U and 232 Th in kyanite, ilmenite and rutile samples were also determined. (author)

  20. Obtention of uranium tetrafluoride from effluents generated in the hexafluoride conversion process

    International Nuclear Information System (INIS)

    Silva Neto, J.B.; Urano de Carvalho, E.F.; Durazzo, M.; Riella, H.G.

    2009-01-01

    Full text: The uranium silicide (U3Si2) fuel is produced from uranium hexafluoride (UF6) as the primary raw material. The uranium tetrafluoride (UF4) and metallic uranium are the two subsequent steps. There are two conventional routes for UF4 production: the first one reduces the uranium from the UF6 hydrolysis solution by adding stannous chloride (SnCl2). The second one is based on the hydrofluorination of solid uranium dioxide (UO2) produced from the ammonium uranyl carbonate (AUC). This work introduces a third route, a dry way route which utilizes the recovering of uranium from liquid effluents generated in the uranium hexafluoride reconversion process adopted at IPEN/CNEN-SP. Working in the liquid phase, this route comprises the recovery of ammonium fluoride by NH4HF2 precipitation. The crystallized bifluoride is added to the solid UO2 to get UF4, which returns to the metallic uranium production process and, finally, to the U3Si2 powder production. The UF4 produced by this new route was chemically and physically characterized and will be able to be used as raw material for metallic uranium production by magnesiothermic reduction. (author)

  1. Contribution to the study of the new international philosophy of the radiological safety in the natural uranium chemical treatment

    International Nuclear Information System (INIS)

    Moraes da Silva, T. de

    1990-01-01

    The objective of this work is to adapt the Radiological Safety System in the facilities concerned to the chemical treatment of the uranium concentrated (yellow-cake) until conversion in uranium hexafluoride in the pilot plant of IPEN-CNEN/SP, to the new international philosophy adopted by ICRP and IAEA. The new philosophy changes fully the Radiological Protection concepts of preceding philosophy, changes, also, the concept of the workplace and individual monitoring as well as the classification of the working areas. In this paper we show the monitoring program, in each phase of the natural uranium treatment chemical process in conversion facility for external irradiation, surface contamination and air contamination. The results were analysed according with the new philosophy and used to reclassify the workplace. It was introduced the condition work concept taking account the time spent by the worker in that workplace. (author)

  2. Study contribution to the new international philosophy of the radiological safety system on chemical processing of the natural uranium

    International Nuclear Information System (INIS)

    Silva, T.M. da.

    1988-01-01

    The objective of the work is to adapt the radiological Safety System in the facilities concerned to the chemical treatment of the uranium concentrated (yellow-cake) until conversion in uranium hexafluoride in the pilot plant of IPEN-CNEN/SP, to the new international philosophy adopted by the International Commission Radiological on Protection ICPR publication 22(1973), 26(1977), 30(1978) and the International Atomic Energy Agency IAEA publication 9(1982). The new philosophy changes fully the Radiological Protection concepts of preceding philosophy, changes, also, the concept of the work place and individual monitoring as well as the classification of the working areas. These new concepts are applied in each phase of the natural uranium treatment chemical process in conversion facility. (author)

  3. Process for recovering uranium using an alkyl pyrophosphoric acid and alkaline stripping solution

    International Nuclear Information System (INIS)

    Worthington, R.E.; Magdics, A.

    1987-01-01

    A process is described for stripping uranium for a pregnant organic extractant comprising an alkyl pyrophosphoric acid dissolved in a substantially water-immiscible organic diluent. The organic extractant contains tetravalent uranium and an alcohol or phenol modifier in a quantity sufficient to retain substantially all the unhydrolyzed alkyl pyrophosphoric acid in solution in the diluent during stripping. The process comprises adding an oxidizing agent to the organic extractant and thereby oxidizing the tetravalent uranium to the +6 state in the organic extractant, and contacting the organic extractant containing the uranium in the +6 state with a stripping solution comprising an aqueous solution of an alkali metal or ammonium carbonate or hydroxide thereby stripping uranium from the organic extractant into the stripping solution. The resulting barren organic extractant containing substantially all of the unhydrolyzed alkyl pyrophosphoric acid dissolved in the diluent is separated from the stripping solution containing the stripped uranium, the barren extractant being suitable for recycle

  4. Indian uranium scenario and a new process technology for alkaline leaching

    International Nuclear Information System (INIS)

    Suri, A.K.; Ghosh, S.K.; Padmanabhan, N.P.H.

    2008-01-01

    The growing demand of uranium for the nuclear power reactors in the country necessitates maximal utilization of the indigenously available uranium resources. In addition to the single operating uranium mine and the mill at Jaduguda, new mines need to be opened to meet the requirements. However, for the exploitation of the various uranium deposits no single elixir process technology is available and needs to necessarily be developed based on the uranium and gangue mineralogy. One such challenge was development of techno-economic process for exploitation of a reasonably vast deposit at Tummalapalle, Andhra Pradesh. The ore characteristics are much different from that of Jaduguda ore and required alkaline pressure leaching technique to bring the uranium values from the ore into solution. Based on the laboratory and pilot plant studies a working flow sheet has been developed and this paper describes the challenges and how they were tackled. (author)

  5. Surface preparation process of a uranium titanium alloy, in particular for chemical nickel plating

    International Nuclear Information System (INIS)

    Henri, A.; Lefevre, D.; Massicot, P.

    1987-01-01

    In this process the uranium alloy surface is attacked with a solution of lithium chloride and hydrochloric acid. Dissolved uranium can be recovered from the solution by an ion exchange resin. Treated alloy can be nickel plated by a chemical process [fr

  6. Processes for extracting radium from uranium mill tailings

    International Nuclear Information System (INIS)

    Nirdosh, I.; Baird, M.H.; Muthuswami, S.V.

    1987-01-01

    This patent describes a process for the extraction of radium from uranium mill tailings solids including the steps of contacting the tailings with a liquid leaching agent, leaching the tailings therewith and subsequently separating the leachate liquid and the leached solids. The improvement described here is wherein the leaching agent comprises: (a) a complexing agent in an amount of from 2 to 10 times the stoichiometric amount needed to complex the metal ions to be removed thereby from the tailings; and (b) a reducing agent reducing the hydrolysable ions of the metal ions to be removed to their lower oxidation states, the reduction agent being present in an amount from 2 to 10 times the stoichiometric amount needed for reducing the hydrolysable metals present in the tailings

  7. Predictor of regulation of uranium dioxide powder pressing process

    International Nuclear Information System (INIS)

    Motta, Eduardo Souza; Araujo, Victor Hugo Leal de; Bernardelli, Sergio Henrique

    2007-01-01

    One of the most important steps of the uranium dioxide pellets fabrication used in the nuclear fuel elements is the green pellets pressing. The target density of the pellets after the sintering process determines the density of the green pellet. To meet the same sintered target density the green density may vary according to the powder characteristics. These variations implies in changing the regulation of the press for different powder's patches. The regulation done empirically imply in productivity loss and necessity of reprocessing the pellets pressed during the press regulation and also depends on the operator experience. At this work, was developed an artificial neural network feed forward back propagation to predict the press regulation, depending on the powder characteristics and the green pellet's target density. The results obtained at INB - Industrias Nucleares do Brasil S. A. during the fabrication of the fifth recharge of Angra II nuclear power plant are presented. (author)

  8. Process for preparing sintered uranium dioxide nuclear fuel

    International Nuclear Information System (INIS)

    Carter, R.E.

    1975-01-01

    Uranium dioxide is prepared for use as fuel in nuclear reactors by sintering it to the desired density at a temperature less than 1300 0 C in a chemically controlled gas atmosphere comprised of at least two gases which in equilibrium provide an oxygen partial pressure sufficient to maintain the uranium dioxide composition at an oxygen/uranium ratio of at least 2.005 at the sintering temperature. 7 Claims, No Drawings

  9. Rapid determination of fluoride in uranyl nitrate solution obtained in conversion process of uranium tetrafluoride

    International Nuclear Information System (INIS)

    Levin, R.; Feldman, R.; Sahar, E.

    1976-01-01

    In uranium production the conversion of impure uranium tetrafluoride by sodium hydroxide was chosen as a current process. A rapid method for determination of fluoride in uranyl-nitrate solution was developed. The method includes precipitation of uranium as diuranate, separation by centrifugation, and subsequent determination of fluoride in supernate by titration with thorium nitrate. Fluoride can be measured over the range 0.15-2.5 gr/gr U, with accuracy of +-5%, within 15 minutes. (author)

  10. The use of slightly alloyed uranium as fuel: its influence on the dissolution and other stages of treatment

    International Nuclear Information System (INIS)

    Faugeras, P.; Leroy, P.; Lheureux, C.

    1959-01-01

    This report deals chiefly with the treatment of binary alloys (UAI, UMo, UZr, UCr, USi) with a low concentration of the additional element (≤2 per cent). The investigation was pursued with a view to the continued utilisation, with a minimum of modification, of the existing plants for treatment of non-alloyed irradiated uranium. In the first part, the usual process for the treatment of irradiated uranium by solvent extraction is briefly recalled. The second part is devoted to a study of the selective dissolution of the canning around certain of these alloys. The third part gives the behaviour of these different alloys at various phases of the usual treatment: a) dissolution; b) extractions; c) final treatment of fission products; d) final purification of plutonium. To conclude, possible alloys are classed as a function of their repercussions on the normal treatment. (author) [fr

  11. Chlorine/chloride based processes for uranium ores

    International Nuclear Information System (INIS)

    1980-11-01

    The CE Lummus Minerals Division was commissioned by The Department of Supply and Services to develop order-of-magnitude capital and operating cost estimates for chlorine/chloride-based processes for uranium ores. The processes are designed to remove substantially all radioactive consituents from the ores to render the waste products harmless. Two processes were selected, one for a typical low grade ore (2 lb. U 3 O 8 /ton ore) and one for a high grade ore (50 lbs U 3 O 8 /ton). For the low grade ore a hydrochloric acid leaching process was chosen. For high grade ore, a more complex process, including gaseous chlorination, was selected. Capital cost estimates were compiled from information obtained from vendors for the specified equipment. Building cost estimates and the piping, electrical and instrumentation costs were developed from the plant layout. Utility diagrams and mass balances were used for estimating utilities and consumables. Detailed descriptions of the bases for capital and operating cost estimates are given

  12. Uranium mining, processing and nuclear energy - opportunities for Australia?

    International Nuclear Information System (INIS)

    2006-12-01

    On 6 June 2006, the Prime Minister announced the appointment of a taskforce to undertake an objective, scientific and comprehensive review of uranium mining, value-added processing and the contribution of nuclear energy in Australia in the longer term. This is known as the Review of Uranium Mining Processing and Nuclear Energy in Australia, referred to in this report as the Review. The Prime Minister asked the Review to report by the end of 2006. A draft report was released for public comment on 21 November 2006 and was also reviewed by an expert panel chaired by the Chief Scientist (see Appendix F). The Review is grateful for comments provided on the draft report by members of the public. The report has been modified in the light of those comments. In response to its initial call for public comment in August 2006 the Review received over 230 submissions from interested parties. It also conducted a wide range of consultations with organisations and individuals in Australia and overseas, and commissioned specialist studies on various aspects of the nuclear industry. Participating in the nuclear fuel cycle is a difficult issue for many Australians and can elicit strong views. This report is intended to provide a factual base and an analytical framework to encourage informed community discussion. Australia's demand for electricity will more than double before 2050. Over this period, more than two-thirds of existing electricity generation will need to be substantially upgraded or replaced and new capacity added. The additional capacity will need to be near-zero greenhouse gas emitting technology if Australia is just to keep greenhouse gas emissions at today's levels. Many countries confront similar circumstances and have therefore considered the use of nuclear power for some of the following reasons: the relative cost competitiveness of nuclear power versus the alternatives; security of supply and independence from fossil fuel energy imports; diversity of domestic

  13. Status Report from Yugoslavia [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Bunji, B [Institute for Technology of Nuclear and Other Raw Materials, Belgrade, Yugoslavia (Serbia)

    1967-06-15

    Full text: The greater part of our activities is connected with the problem of extracting uranium from low-grade ores. In this paper, a brief review of the most important recent developments will be presented. In this connection, it may be useful to determine the definition of low-grade ores. This term can be applied to ore from which the uranium content cannot be extracted under normal economic conditions. Thus this term can be applied to uranium-bearing material with a uranium content of no more than 0. 05%. But, in general, it could be said that there is a very large range of uranium content where uranium extraction may not be economic for such different reasons as; (a) the size or other facts in connection with the orebodies themselves; (b) refractory ore; or (c) other local conditions. During research on the treatment of low-grade ore from the deposit at Gabrovnica (Stara Planina, Yugoslavia) it became apparent that an alkaline leaching process would have to be carried out. The treatment of this granitic type of ore causes no particular difficulties. The required temperature is about 90{sup o}C. The retention time in the leaching stage is from 4 to 12 hours. Sodium carbonate consumption is not higher than 15 kg/t of ore. Pachuca-type leaching shows satisfactory maintenance and processing costs. At Kalna uranium precipitation by means of hydrogen pressure reduction has been developed, and is being developed and investigated in full-scale operation. Details of the process were published in Geneva in 1963. On the basis of the experience gained from full-scale operation, many refinements and cost-saving changes have been made. A normal steel wire screen used as a catalyst carrier shows a very good improvement over free-moving UO{sub 2} as catalyst. In large-scale operation (200 t/d), after the precipitation of uranium the barren solution content is about 1 g U/m{sup 3}. The content of the pregnant solution is of the order of 300-600 g/m{sup 3}. Recycling the

  14. New insights into uranium (VI) sol-gel processing

    International Nuclear Information System (INIS)

    King, C.M.; Thompson, M.C.; Buchanan, B.R.; King, R.B.; Garber, A.R.

    1990-01-01

    Nuclear Magnetic Resonance (NMR) investigations on the Oak Ridge National Laboratory process for sol-gel synthesis of microspherical nuclear fuel (UO 2 ), has been extremely useful in sorting out the chemical mechanism in the sol-gel steps. 13 C, 15 N, and 1 H NMR studies on the HMTA gelation agent (Hexamethylene tetramine, C 6 H 12 N 4 ) has revealed near quantitative stability of this adamantane-like compound in the sol-gel process, contrary to its historical role as an ammonia source for gelation from the worldwide technical literature. 17 O NMR of uranyl (UO 2 ++ ) hydrolysis fragments produced in colloidal sols has revealed the selective formation of a uranyl trimer, [(UO 2 ) 3 (μ 3 -O)(μ 2 -OH) 3 ] + , induced by basic hydrolysis with the HMTA gelation agent. Spectroscopic results will be presented to illustrate that trimer condensation occurs during sol-gel processing leading to layered polyanionic hydrous uranium oxides in which HMTAH + is occluded as an ''intercalation'' cation. Subsequent sol-gel processing of microspheres by ammonia washing results in in-situ exchange and formation of a layered hydrous ammonium uranate with a proposed structural formula of (NH 4 ) 2 [(UO 2 ) 8 O 4 (OH) 10 ] · 8H 2 O. This compound is the precursor to sintered UO 2 ceramic fuel. 23 refs., 10 figs

  15. Denitrification of acid wastes from uranium purification processes

    International Nuclear Information System (INIS)

    Clark, F.E.; Francis, C.W.; Francke, H.C.; Strohecker, J.W.

    1975-11-01

    Laboratory and pilot-plant investigations have shown the technical feasibility of removing nitrates from neutralized acid wastes from uranium purification processes by biological denitrification, a dissimilatory process in which the nitrate ion is reduced to nitrogen gas by specific bacteria. The process requires anaerobic conditions and an organic carbon source, as well as other life-sustaining constituents. These denitrification studies produced process design information on a columnar denitrification plant and on continuous-flow, stirred-bed reactors. Denitrification, using packed columns, was found to be desirable for soluble salts, such as those of sodium and ammonium; denitrification, using stirred reactors, was found to be desirable for mixtures containing insoluble salts, such as those of calcium and aluminum. Packed columns were found to have denitrification rates ranging up to 122 grams of nitrate per day per cubic decimeter of column volume; stirred-bed reactors have been shown to have reaction rates near 10 grams of nitrate per day per cubic decimeter of reactor volume. The continuous-flow, stirred-bed reactors were selected for scaleup studies because of the solids-removal problems associated with packed columns when operating on feeds containing high concentrations of insoluble salts or ions which form insoluble salts with the products of the denitrification reaction

  16. Uranium conversion

    International Nuclear Information System (INIS)

    Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

    2006-03-01

    FOI, has performed a study on uranium conversion processes that are of importance in the production of different uranium compounds in the nuclear industry. The same conversion processes are of interest both when production of nuclear fuel and production of fissile material for nuclear weapons are considered. Countries that have nuclear weapons ambitions, with the intention to produce highly enriched uranium for weapons purposes, need some degree of uranium conversion capability depending on the uranium feed material available. This report describes the processes that are needed from uranium mining and milling to the different conversion processes for converting uranium ore concentrate to uranium hexafluoride. Uranium hexafluoride is the uranium compound used in most enrichment facilities. The processes needed to produce uranium dioxide for use in nuclear fuel and the processes needed to convert different uranium compounds to uranium metal - the form of uranium that is used in a nuclear weapon - are also presented. The production of uranium ore concentrate from uranium ore is included since uranium ore concentrate is the feed material required for a uranium conversion facility. Both the chemistry and principles or the different uranium conversion processes and the equipment needed in the processes are described. Since most of the equipment that is used in a uranium conversion facility is similar to that used in conventional chemical industry, it is difficult to determine if certain equipment is considered for uranium conversion or not. However, the chemical conversion processes where UF 6 and UF 4 are present require equipment that is made of corrosion resistant material

  17. The application of illite supported nanoscale zero valent iron for the treatment of uranium contaminated groundwater.

    Science.gov (United States)

    Jing, C; Landsberger, S; Li, Y L

    2017-09-01

    In this study, nanoscale zero valent iron I-NZVI was investigated as a remediation strategy for uranium contaminated groundwater from the former Cimarron Fuel Fabrication Site in Oklahoma, USA. The 1 L batch-treatment system was applied in the study. The result shows that 99.9% of uranium in groundwater was removed by I-NZVI within 2 h. Uranium concentration in the groundwater stayed around 27 μg/L, and there was no sign of uranium release into groundwater after seven days of reaction time. Meanwhile the release of iron was significantly decreased compared to NZVI which can reduce the treatment impact on the water environment. To study the influence of background pH of the treatment system on removal efficiency of uranium, the groundwater was adjusted from pH 2-10 before the addition of I-NZVI. The pH of the groundwater was from 2.1 to 10.7 after treatment. The removal efficiency of uranium achieved a maximum in neutral pH of groundwater. The desorption of uranium on the residual solid phase after treatment was investigated in order to discuss the stability of uranium on residual solids. After 2 h of leaching, 0.07% of the total uranium on residual solid phase was leached out in a HNO 3 leaching solution with a pH of 4.03. The concentration of uranium in the acid leachate was under 3.2 μg/L which is below the EPA's maximum contaminant level of 30 μg/L. Otherwise, the concentration of uranium was negligible in distilled water leaching solution (pH = 6.44) and NaOH leaching solution (pH = 8.52). A desorption study shows that an acceptable amount of uranium on the residuals can be released into water system under strong acid conditions in short terms. For long term disposal management of the residual solids, the leachate needs to be monitored and treated before discharge into a hazardous landfill or the water system. For the first time, I-NZVI was applied for the treatment of uranium contaminated groundwater. These results provide proof that I-NZVI has

  18. Treatment alternatives of liquid radioactive waste containing uranium in phosphoric acid

    International Nuclear Information System (INIS)

    Bustamante Escobedo, Mauricio

    2003-01-01

    The UGDR, receives annually 100 [l] of liquid radioactive waste containing, highly acid (pH=0) uranium in phosphoric acid from the Laboratory of Chemical Analysis. This waste must be chemically and radiologically decontaminated before it can be discharged in accordance with local environmental standards. Chemical precipitation and evaporation test were carried out to define the operating conditions for the radiological decontamination of this radioactive waste and to obtain a solid waste that can be conditioned in a cement matrix. The evaporation process generates excellent rates of volume reduction, over 80%, but generates a pulp that is hard handle when submitted to a drying process. Chemical precipitation generates good results for decontaminating these solutions and reducing volume (above 50%) to obtain a uranium free effluent. The treatment with calcium carbonate generated an effluent with a low concentration of polluting agents. A preliminary test was carried out condition these solids in a cement matrix, using ratios of 0.45 waste/cement and 2 of water/cement. The mix prepared with waste from the sodium hydroxide treatment had low mechanical resistance resulting from the saline incrustations. The waste from the calcium carbonate treatment was very porous due to the water evaporation from the highly exothermic reaction between the waste and the cement. The mix of the calcium carbonate generated waste and the cement matrix needs to be optimized, since it generates favorable conditions for adhering with the cement matrix (au)

  19. Secondary wastes and treatment of effluents from leaching of uranium from soils

    International Nuclear Information System (INIS)

    Ally, M.R.; Wilson, J.H.; Francis, C.W.

    1993-01-01

    The Department of Energy's Feed Materials and Production Center at Fernald, Ohio has over two million cubic meters of soil contaminated with Uranium which must be cleaned. Soil characterization studies show that Uranium is unevenly distributed between the clay, sand and silt fractions. This paper examines the option of using leaching agents to remove Uranium from the soil and the treatment of secondary wastes. Results of the effects of various leachants in removing Uranium and the complications of co-leaching minerals/organic matter that are important for maintaining soil integrity and structure shall be discussed. Candidate leachants must remove the Uranium level below 35pCi/g of soil and produce a secondary waste that is amenable to on-site treatment at reasonable cost

  20. Soil treatment to remove uranium and related mixed radioactive contaminants. Final report September 1992--October 1995

    International Nuclear Information System (INIS)

    1996-07-01

    A research and development project to remove uranium and related radioactive contaminants from soil by an ultrasonically-aided chemical leaching process began in 1993. The project objective was to develop and design, on the basis of bench-scale and pilot-scale experimental studies, a cost-effective soil decontamination process to produce a treated soil containing less than 35 pCi/g. The project, to cover a period of about thirty months, was designed to include bench-scale and pilot-scale studies to remove primarily uranium from the Incinerator Area soil, at Fernald, Ohio, as well as strontium-90, cobalt-60 and cesium-137 from a Chalk River soil, at the Chalk River Laboratories, Ontario. The project goal was to develop, design and cost estimate, on the basis of bench-scale and pilot-scale ex-situ soil treatment studies, a process to remove radionuclides form the soils to a residual level of 35 pCi/g of soil or less, and to provide a dischargeable water effluent as a result of soil leaching and a concentrate that can be recovered for reuse or solidified as a waste for disposal. In addition, a supplementary goal was to test the effectiveness of in-situ soil treatment through a field study using the Chalk River soil

  1. Applications of Ecological Engineering Remedies for Uranium Processing Sites, USA

    Energy Technology Data Exchange (ETDEWEB)

    Waugh, William [Navarro Research and Engineering

    2016-05-23

    The U.S. Department of Energy (USDOE) is responsible for remediation of environmental contamination and long-term stewardship of sites associated with the legacy of nuclear weapons production during the Cold War in the United States. Protection of human health and the environment will be required for hundreds or even thousands of years at many legacy sites. USDOE continually evaluates and applies advances in science and technology to improve the effectiveness and sustainability of surface and groundwater remedies (USDOE 2011). This paper is a synopsis of ecological engineering applications that USDOE is evaluating to assess the effectiveness of remedies at former uranium processing sites in the southwestern United States. Ecological engineering remedies are predicated on the concept that natural ecological processes at legacy sites, once understood, can be beneficially enhanced or manipulated. Advances in tools for characterizing key processes and for monitoring remedy performance are demonstrating potential. We present test cases for four ecological engineering remedies that may be candidates for international applications.

  2. Chemical Separation of Fission Products in Uranium Metal Ingots from Electrolytic Reduction Process

    International Nuclear Information System (INIS)

    Lee, Chang-Heon; Kim, Min-Jae; Choi, Kwang-Soon; Jee, Kwang-Yong; Kim, Won-Ho

    2006-01-01

    Chemical characterization of various process materials is required for the optimization of the electrolytic reduction process in which uranium dioxide, a matrix of spent PWR fuels, is electrolytically reduced to uranium metal in a medium of LiCl-Li 2 O molten at 650 .deg. C. In the uranium metal ingots of interest in this study, residual process materials and corrosion products as well as fission products are involved to some extent, which further adds difficulties to the determination of trace fission products. Besides it, direct inductively coupled plasma atomic emission spectrometric (ICP-AES) analysis of uranium bearing materials such as the uranium metal ingots is not possible because a severe spectral interference is found in the intensely complex atomic emission spectra of uranium. Thus an adequate separation procedure for the fission products should be employed prior to their determinations. In present study ion exchange and extraction chromatographic methods were adopted for selective separation of the fission products from residual process materials, corrosion products and uranium matrix. The sorption behaviour of anion and tri-nbutylphosphate (TBP) extraction chromatographic resins for the metals in acidic solutions simulated for the uranium metal ingot solutions was investigated. Then the validity of the separation procedure for its reliability and applicability was evaluated by measuring recoveries of the metals added

  3. 75 FR 71677 - Reimbursement for Costs of Remedial Action at Active Uranium and Thorium Processing Sites

    Science.gov (United States)

    2010-11-24

    ... DEPARTMENT OF ENERGY Reimbursement for Costs of Remedial Action at Active Uranium and Thorium... in FY 2011 from eligible active uranium and thorium processing site licensees for reimbursement under... approximately $24.3 million of Recovery Act funds available for reimbursement in FY 2011, as well as the $10...

  4. Processing of Sierra Albarrana uranium ores; Tratamiento de los minerales de uranio de Sierra Albarrana

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez Jodra, L; Perez Luina, A; Perarnau, M

    1960-07-01

    Uranium recovery by hydrometallurgy from brannerite, found in Hornachuelos (Cordoba) is described. It has been studied the acid and alkaline leaching and salt roasting, proving as more satisfactory the acid leaching. Besides the uranium solubilization by acid leaching, is described the further process to obtain pure uranyl nitrate. (Author)

  5. Uranium Metal to Oxide Conversion by Air Oxidation –Process Development

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, A

    2001-12-31

    Published technical information for the process of metal-to-oxide conversion of uranium components has been reviewed and summarized for the purpose of supporting critical decisions for new processes and facilities for the Y-12 National Security Complex. The science of uranium oxidation under low, intermediate, and high temperature conditions is reviewed. A process and system concept is outlined and process parameters identified for uranium oxide production rates. Recommendations for additional investigations to support a conceptual design of a new facility are outlined.

  6. Laboratory studies on leaching of low grade uranium ores and treatment of low level liquid waste generated by leaching experiments

    International Nuclear Information System (INIS)

    Palabrica, O.T.; Antonino, E.J.; Caluag, L.A.; Villamater, D.

    1980-07-01

    Acid leaching experiments of preconcentrated uranium ore were carried out at a pulp density of 50% solids, using sulfuric acid with sodium chlorate as oxidant. The different leaching parameters considered in this work were temperature, oxidant level and leaching time. In the experimental procedure, the concentration of oxidant and the temperature were varied to determine how they affect the leaching process. Experimental results are illustrated in tabulated form for better interpretation. Uranium analyses were done by fluorimetric and delayed-neutron activation analysis. An anion exchange method using Dowex 1 x 8, 200-400 mesh (Cl - ) was used in treating the low-level liquid waste generated by leaching experiments. The purpose of this treatment was to minimize radioactive contamination in the waste materials and also to recover some of the uranium left in the liquid waste. (author)

  7. Decontamination process development for gravels contaminated with uranium

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gye Nam; Park, Uk Ryang; Kim, Seung Su; Kim, Won Suk; Moon, Jei Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    It is impossible to scrub gravels in a washing tank, because gravels sinks to the bottom of the washing tank. In addition, when electrokinetic decontamination technology is applied to gravels larger than 10 cm, the removal efficiency of uranium from the gravels is reduced, because electro-osmotic flux at the surface of the gravel in electrokinetic cell reduces owing to a reduction of the particle surface area attributable to large-sized gravel. The volume ratio of gravel larger than10 cm in total volume of the soil in KAERI was about 20%. Therefore, it is necessary to study the decontamination process of gravels contaminated with radionuclides. The optimum number of washings for contaminated gravels is considered to be two. In addition, the removal efficiency of contaminated gravel was not related to its weight. For an electrokinetic-electrodialytic decontamination period of 5 days, 10 days, 15 days, and 20 days, {sup 238}U in gravel was removed by about 42%, 64%, 74%, and 80%, respectively. The more the decontamination time elapsed, the greater the reduction of the removal efficiency ratio of {sup 238}U. The decontamination process for gravels was generated on the basis of the results of washing and electrokinetic electrodialtic experiments.

  8. The separation nozzle process for uranium isotope enrichment

    International Nuclear Information System (INIS)

    Becker, E.W.

    1977-01-01

    In the separation nozzle process, uranium isotope separation is brought about by the mass dependence of the centrifugal forces in a curved flow of a UF 6 /H 2 -mixture. Due to the large excess in hydrogen the high ration of UF 6 flow velocity to thermal velocity required for an effective isotope separation is obtained at relatively low expansion ratios and, accordingly, with relatively low gas-dynamic losses. As the optimum Reynolds number of the curved jet is comparatively low and a high absolute pressure is essential for economic reasons, the characteristic dimensions of the nozzle systems are made as small as possible. For commercial application in the near future systems involving mechanical jet deflection were developed. However, promising results were also obtained with separation nozzle systems generating a streamline curvature by the interaction of opposed jets. Most of the development work has been done at the Nuclear Research Center of Karlsruhe. Since 1970 the German company STEAG has been involved in the commercial implementation of the process. Two industrial-scale separative stages were tested successfully. This work constitutes the basis of planning of a separation nozzle demonstration plant to be built in Brazil

  9. Management of uranium mining and processing wastes at Turamdih project

    International Nuclear Information System (INIS)

    Puri, R.C.; Verma, R.P.

    1991-01-01

    Based on environmental impact assessment, comprehensive plan for management of wastes has been drawn up. No solid waste from the mine is being disposed off outside the project area. The quantity of waste generated after processing of ore is large because of low content of uranium in the ore. A big tailings pond has been planned in specially selected suitable valley near the plant. No liquid effluents are to be discharged into general surrounding environment. Mine water is to be fed to the process plant. Effluents from tailings pond will be collected in a storage cum evaporation pond. All water from different zones of the project shall be collected in zonal ponds and then pumped to tailings effluent storage pond. All the ponds will be provided with requisite impervious liners. The effluents of the storage pond will be treated for removal of radium and manganese and discharged into monitoring pond. Large surface areas for various ponds are envisaged to take advantage of evaporation with aim for zero discharge. To reduce impact from gaseous emissions, high efficiency dust suppression and extraction systems shall be provided. High stacks have been incorporated for DG set, boiler plants, sulphuric acid plant and dust extraction systems for crushing and grinding section and the quality of discharges will be very much within the prescribed limits. The paper describes the management plan in detail. (author)

  10. Analytical control of reducing agents on uranium/plutonium partitioning at purex process

    International Nuclear Information System (INIS)

    Araujo, Izilda da Cruz de

    1995-01-01

    Spectrophotometric methods for uranium (IV), hydrazine (N 2 H 4 ) and its decomposition product hydrazoic acid(HN 3 ), and hydroxylamine (NH 2 OH) determinations were developed aiming their applications for the process control of CELESTE I installation at IPEN/CNEN-SP. These compounds are normally present in the U/Pu partitioning phase of the spent nuclear treatment via PUREX process. The direct spectrophotometry was used for uranium (IV) analysis in nitric acid-hydrazine solutions based on the absorption measurement at 648 nm. The azomethine compound formed by reaction of hydrazine and p-dimethylamine benzaldehyde with maximum absorption at 457 nm was the basis for the specific analytical method for hydrazine determination. The hydrazoic acid analysis was performed indirectly by its conversion into ferric azide complex with maximum absorption at 465 nm. The hydroxylamine detection was accomplished based on its selective oxidation to nitrous acid which is easily analyzed by the reaction with Griess reagent. The resulted azocompound gas a maximum absorption at 520 nm. The sensibility of 1,4x10 -6 M for U(IV) with 0,8% of precision, 1,6x10 -6 M for hydrazine with 0,8% of precision, 2,3x10 -6 M hydrazoic acid with 0,9% of precision and 2,5x10 -6 M for hydroxylamine with 0,8% of precision were achieved. The interference studies have shown that each reducing agent can be determined in the presence of each other without any interference. Uranium(VI) and plutonium have also shown no interference in these analysis. The established methods were adapted to run inside glove-boxes by using an optical fiber colorimetry and applied to process control of the CELESTE I installation. The results pointed out that the methods are reliable and safety in order to provide just-in-time information about process conditions. (author)

  11. Chlorination of antimony and its volatilization treatment of waste antimony-uranium composite oxide catalyst

    International Nuclear Information System (INIS)

    Sawada, K.; Enokida, Y.

    2011-01-01

    For the waste antimony-uranium composite oxide catalyst, the chlorination of antimony and its volatilization treatment were proposed, and evaluated using hydrogen chloride gas at 873-1173 K. During the treatment, the weight loss of the composite oxide sample, which resulted from the volatilization of antimony, was confirmed. An X-ray diffraction analysis showed that uranium oxide, U 3 O 8 , was formed during the reaction. After the treatment at 1173 K for 1 h, almost all the uranium contained in the waste catalyst was dissolved by a 3 M nitric acid solution at 353 K within 10 min, although that of the non-treated catalyst was less than 0.1%. It was found that the chlorination and volatilization treatment was effective to separate antimony from the composite oxide catalyst and change uranium into its removable form. (orig.)

  12. Improvements in process technology for uranium metal production

    International Nuclear Information System (INIS)

    Meghal, A.M.; Singh, H.; Koppiker, K.S.

    1991-01-01

    The research reactors in Trombay use uranium metal as a fuel. The plant to produce nuclear grade uranium metal ingots has been in operation at Trombay since 1959. Recently, the capacity of the plant has been expanded to meet the additional demand of the uranium metal. The operation of the expanded plant, has brought to the surface various shortcomings. This paper identifies various problems and describes the measures to be taken to upgrade the technology. Some comments are made on the necessity for development of technology for future requirement. (author). 6 refs., 1 fig

  13. Occupational dermatoses in the uranium mining and processing industry

    Energy Technology Data Exchange (ETDEWEB)

    Sevcova, M [Zavodni Ustav Narodniho Zdravi Uranoveho Prumyslu, Pribram (Czechoslovakia)

    1978-04-01

    Experience gained so far by the Department of Dermatovenerology in the uranium industry discloses that the incidence of occupational dermatoses is relatively low in this industry. It represents about 1% of all newly ascertained skin diseases per year. Allergic contact eczemas after having been in contact with rubber products, chiefly rubber boots, predominate. Under the working conditions in mining and preparing uranium ore, ionizing radiation cannot induce non-stochastic effects of the type of radiation dermatitis on the skin. A higher incidence was, however, ascertained in uranium miners of basaliomas, which agrees with the estimation of the dose of external alpha radiation in the basal epidermis layer.

  14. New techniques for the treatment of uranium ores

    International Nuclear Information System (INIS)

    Renaud, J.; Boutonnet, G.

    1977-01-01

    The growth in nuclear power programmes since 1970 has led to an increasing demand for uranium, and tenders have been invited from all parts of the world for the construction of new treatment plants. What types of plant could be suggested. The diversity of ores and sites, even more stringent safety requirements, greater care for the environment and economic facts called for numerous, if not basic, reviews of the conventional techniques. Two examples illustrate this point. In the case of a plant to treat a refractory ore situated in a desert area with limited water resources, Pechiney Ugine Kuhlmann studied and applied a new technique of leaching by sulphuric acid pulping, which gives a considerable saving of sulphuric acid and water in comparison with conventional leaching techniques. In dealing with a problem which arose at a plant situated in a mountainous region of touristic interest, where a tailings settling tank could not be installed, Pechiney Ugine Kuhlmann studied and developed techniques involving the use of band filters for solid-liquid separation and pulp washing. Apart from lowering investment costs by about 15% in comparison with the techniques used so far, this technique produces the tailings in solid form so that they do not require a settling tank for storage. (author)

  15. The effect of dispersed materials on baro-membrane treatment of uranium-containing waters

    International Nuclear Information System (INIS)

    Kryvoruchko, Antonina P.; Atamanenkoa, Irina D.

    2007-01-01

    The paper investigated a treatment process of uranium-containing waters in a membrane reactor while using natural mineral kizelgur and synthetic sorbent SKN-1K with subsequent ultra- and nano-filtration separation of the mixture. The retention coefficient of U(VI) by membrane UPM-20 under conditions of quasi-stationary equilibrium reached the levels of 0.87-0.89 and 0.89-0.91, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. In the case of membrane OPMN-P and natural mineral kizelgur the retention coefficient of U(VI) was 0.990-0.991 and 0.993-0.996, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. Data regarding the state of water in membranes formed from natural mineral or synthetic sorbent on the surface of substrate membranes UPM-20 and OPMN-P made it possible to conclude that dispersed materials of different chemical nature affect the process of baro-membrane treatment of uranium-containing waters. (authors)

  16. Significance of mineralogy in the development of flowsheets for processing uranium ores

    International Nuclear Information System (INIS)

    1980-01-01

    This report has been prepared from material developed at and subsequent to a consultants' meeting held in Vienna in January 1978. The main purpose of the meeting was to prepare a document in the form of a guide for planning and developing treatment flowsheets for uranium ore processing. It was apparent that ore mineralogy, analysed, described and interpreted in ways most meaningful to the metallurgist, is the most essential information required for forming the basis of such planning. This topic, here termed metallurgical mineralogy, is therefore a major theme of this publication. In preparing the report the Agency has borne in mind the important need to impart the experience and knowledge gained in the more developed countries to those who are in the early stages of exploiting their uranium resources. The contents may be criticized as lacking, in some respects, the requisite depth and detail of treatment. The Agency and the consultants are conscious of the need to expand the information in a number of ways. However, the report is presented in its present form in the belief that, as the first attempt to correlate, on a world-wide basis, ore type with processing, it will be considered as a useful basis for future development of these themes

  17. Situation of radioactive wastes and their prevention and treatment measures in China's uranium mining and metallurgy

    International Nuclear Information System (INIS)

    Li Renjie.

    1988-01-01

    The sorts of radioactive wastes produced in uranium mining and metallurgy and their hazards are discribed in this paper. The characteristics of the radioactive wastes are discussed. The measurements and results are introduced for treatment and disposal of the radioactive wastes. The way to deal with prevention and treatment of radioactive wastes is presented in the stages of engineering design, construction, production and decommission of uranium mines and plants

  18. Treatment technology of low concentration uranium-bearing wastewater and its research progress

    International Nuclear Information System (INIS)

    Wei Guangzhi; Xu Lechang

    2007-01-01

    With growth of the discharged uranium-bearing wastewater capacity, a low cost and effective treatment technology is required to avoid transferring and diffusion of the radioactive nuclides. On the basis of analyses of the source and characteristics of the low-concentration uranium-bearing wastewater, the conventional treatment technologies, such as, flocculating settling, ion exchange, concentration, adsorption, and some innovatory technologies, such as, membrane, microorganism, phytoremediation and zero-valent iron technology are introduced. (authors)

  19. Development of practical decontamination process for the removal of uranium from gravel.

    Science.gov (United States)

    Kim, Ilgook; Kim, Gye-Nam; Kim, Seung-Soo; Choi, Jong-Won

    2018-01-01

    In this study, a practical decontamination process was developed to remove uranium from gravel using a soil washing method. The effects of critical parameters including particle size, H 2 SO 4 concentration, temperature, and reaction time on uranium removal were evaluated. The optimal condition for two-stage washing of gravel was found to be particle size of 1-2 mm, 1.0 M H 2 SO 4 , temperature of 60°C, and reaction time of 3 h, which satisfied the required uranium concentration for self-disposal. Furthermore, most of the extracted uranium was removed from the waste solution by precipitation, implying that the treated solution can be reused as washing solution. These results clearly demonstrated that our proposed process can be indeed a practical technique to decontaminate uranium-polluted gravel.

  20. Reverse osmosis treatment in CO_2 + O_2 to the application of the in-situ leaching of uranium

    International Nuclear Information System (INIS)

    Ruan Zhilong; Li Xilong; Yang Shaowu

    2014-01-01

    Advantages and disadvantages of various groundwater management methods, combined with CO_2 + O_2 characteristics of in situ leaching uranium mining process, use reverse osmosis wastewater treatment technology, has carried on the laboratory test, field condition test and industrial test. Obtained by indoor experiment and field conditions for Cl"- ion concentration variation characteristics; Reverse osmosis treatment effect of wastewater is verified by industrial test, obtained the technical parameters and consumption data, as well as the leaching liquid and adsorption tail liquid pH, SO_4"2"-; Cl"- in the plasma concentration monitoring, and further prove that the reverse osmosis treatment technology is suitable for in-situ leaching of uranium in CO_2 + O_2 in wastewater treatment. (authors)

  1. Hypertension and hematologic parameters in a community near a uranium processing facility

    International Nuclear Information System (INIS)

    Wagner, Sara E.; Burch, James B.; Bottai, Matteo; Pinney, Susan M.; Puett, Robin; Porter, Dwayne; Vena, John E.; Hebert, James R.

    2010-01-01

    Background: Environmental uranium exposure originating as a byproduct of uranium processing can impact human health. The Fernald Feed Materials Production Center functioned as a uranium processing facility from 1951 to 1989, and potential health effects among residents living near this plant were investigated via the Fernald Medical Monitoring Program (FMMP). Methods: Data from 8216 adult FMMP participants were used to test the hypothesis that elevated uranium exposure was associated with indicators of hypertension or changes in hematologic parameters at entry into the program. A cumulative uranium exposure estimate, developed by FMMP investigators, was used to classify exposure. Systolic and diastolic blood pressure and physician diagnoses were used to assess hypertension; and red blood cells, platelets, and white blood cell differential counts were used to characterize hematology. The relationship between uranium exposure and hypertension or hematologic parameters was evaluated using generalized linear models and quantile regression for continuous outcomes, and logistic regression or ordinal logistic regression for categorical outcomes, after adjustment for potential confounding factors. Results: Of 8216 adult FMMP participants 4187 (51%) had low cumulative uranium exposure, 1273 (15%) had moderate exposure, and 2756 (34%) were in the high (>0.50 Sievert) cumulative lifetime uranium exposure category. Participants with elevated uranium exposure had decreased white blood cell and lymphocyte counts and increased eosinophil counts. Female participants with higher uranium exposures had elevated systolic blood pressure compared to women with lower exposures. However, no exposure-related changes were observed in diastolic blood pressure or hypertension diagnoses among female or male participants. Conclusions: Results from this investigation suggest that residents in the vicinity of the Fernald plant with elevated exposure to uranium primarily via inhalation exhibited

  2. Hypertension and hematologic parameters in a community near a uranium processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, Sara E., E-mail: swagner@uga.edu [College of Public Health, Department of Epidemiology and Biostatistics, Paul D. Coverdell Center for Biomedical and Health Sciences, University of Georgia, 500 D.W. Brooks Drive, Athens, GA 30602-7396 (United States); Burch, James B. [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); South Carolina Statewide Cancer Prevention and Control Program, Columbia, SC (United States); WJB Dorn Veteran' s Affairs Medical Center, Columbia, SC (United States); Bottai, Matteo [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); Pinney, Susan M. [College of Medicine, Department of Environmental Health, University of Cincinnati, Cincinnati, OH (United States); Puett, Robin [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); South Carolina Statewide Cancer Prevention and Control Program, Columbia, SC (United States); Arnold School of Public Health, Department of Environmental Health Sciences, University of South Carolina, Columbia, SC (United States); Porter, Dwayne [Arnold School of Public Health, Department of Environmental Health Sciences, University of South Carolina, Columbia, SC (United States); Vena, John E. [College of Public Health, Department of Epidemiology and Biostatistics, Paul D. Coverdell Center for Biomedical and Health Sciences, University of Georgia, 500 D.W. Brooks Drive, Athens, GA 30602-7396 (United States); Hebert, James R. [Arnold School of Public Health, Department of Epidemiology and Biostatistics, University of South Carolina, Columbia, SC (United States); South Carolina Statewide Cancer Prevention and Control Program, Columbia, SC (United States)

    2010-11-15

    Background: Environmental uranium exposure originating as a byproduct of uranium processing can impact human health. The Fernald Feed Materials Production Center functioned as a uranium processing facility from 1951 to 1989, and potential health effects among residents living near this plant were investigated via the Fernald Medical Monitoring Program (FMMP). Methods: Data from 8216 adult FMMP participants were used to test the hypothesis that elevated uranium exposure was associated with indicators of hypertension or changes in hematologic parameters at entry into the program. A cumulative uranium exposure estimate, developed by FMMP investigators, was used to classify exposure. Systolic and diastolic blood pressure and physician diagnoses were used to assess hypertension; and red blood cells, platelets, and white blood cell differential counts were used to characterize hematology. The relationship between uranium exposure and hypertension or hematologic parameters was evaluated using generalized linear models and quantile regression for continuous outcomes, and logistic regression or ordinal logistic regression for categorical outcomes, after adjustment for potential confounding factors. Results: Of 8216 adult FMMP participants 4187 (51%) had low cumulative uranium exposure, 1273 (15%) had moderate exposure, and 2756 (34%) were in the high (>0.50 Sievert) cumulative lifetime uranium exposure category. Participants with elevated uranium exposure had decreased white blood cell and lymphocyte counts and increased eosinophil counts. Female participants with higher uranium exposures had elevated systolic blood pressure compared to women with lower exposures. However, no exposure-related changes were observed in diastolic blood pressure or hypertension diagnoses among female or male participants. Conclusions: Results from this investigation suggest that residents in the vicinity of the Fernald plant with elevated exposure to uranium primarily via inhalation exhibited

  3. Two cases of physical treatment of uranium ore; Deux cas de traitement physique de minerai d'uranium

    Energy Technology Data Exchange (ETDEWEB)

    Ginocchio, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Until now, all uranium deposits exploited in France are vein type and present a very big variety of mineralization and structure. The process of concentration of these ores requires a study for each of them and, in a lot of cases, only the chemical attack can solve the problem. However, the flotation succeeded to results permitting a very interesting enrichment with lower investments expenses and cost prices. Enrichment by flotation would be foreseeable for poor layers and weak tonnage, permitting to absorb facilities on tonnages three times less important than the acidic treatment, or, to equality of cost price, to treat ores having contents of 2,5 to 4 times weaker. (M.B.) [French] Jusqu'ici, tous les gisements uraniferes exploites en France sont du type filonien et presentent une tres grande variete de mineralisation et de structure. Le procede de concentration de ces minerais necessite une etude pour chacun d'eux et, dans bien des cas, seule l'attaque chimique peut resoudre le probleme. Toutefois, la flottation a abouti a des resultats permettant un enrichissement tres interessant avec les depenses d'investissements et des prix de revient beaucoup plus bas. Un enrichissement par flottation serait envisageable pour des gisements pauvres et de faible tonnage, permettant d'amortir les installations sur des tonnages trois fois moins importants que le traitement acide, ou encore, a egalite de prix de revient, de traiter des minerais ayant des teneurs de 2,5 a 4 fois plus faibles. (M.B.)

  4. Separation and purification of uranium product from thorium in thorex process by precipitation technique

    International Nuclear Information System (INIS)

    Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Mukherjee, A.; Dhumwad, R.K.

    1989-01-01

    A sequential precipitation technique is reported for the separation of uranium and thorium present in the uranium product stream of a single cycle 5 per cent TBP Thorex Process. It involves the precipitation of thorium as oxalate in 1M HNO 3 medium at 60-70degC and after filtration, precipitation of uranium as ammonium diuranate at 80-90degC from the oxalate supernatant. This technique has several advantages over the ion-exchange process normally used for treating these products. In order to meet the varying feed conditions, this method has been tested for feeds containing 10 g/1 uranium and 1-50 g/1 thorium in 1-6M HNO 3 . Various parameters like feed acidities, uranium and thorium concentrations, excess oxalic acid concentrations in the oxalate supernatant, precipitation temperatures, precipitate wash volumes etc. have been optimised to obtain more than 99 per cent recovery of thorium and uranium as their oxides with less than 50 ppm uranium losses to ammonium diuranate filtrate. The distribution patterns of different fission products and stainless steel corrosion products during various steps of this procedure have also been studied. For simulating the actual Thorex plant scale operation, experiments have been conducted with 25g and 100g lots of uranium per batch. (author). 6 tabs., 8 figs., 22 refs

  5. Development of a process analyzer for trace uranium

    International Nuclear Information System (INIS)

    Hiller, J.M.

    1990-01-01

    A process analyzer, based on time-resolved laser-induced luminescence, is being developed for the Department of Energy's Oak Ridge Y-12 Plant for the ultra-trace determination of uranium. The present instrument has a detection limit of 1 μg/L; the final instrument will have a detection limit near 1 ng/L for continuous environmental monitoring. Time-resolved luminescence decay is used to enhance sensitivity, reduce interferences, and eliminate the need for standard addition. The basic analyzer sequence is: a pulse generator triggers the laser; the laser beam strikes a photodiode which initiates data acquisition and synchronizes the timing, nearly simultaneously, laser light strikes the sample; intensity data are collected under control of the gated photon counter; and the cycle repeats as necessary. Typically, data are collected in 10 μs intervals over 700 μs (several luminescence half-lives). The final instrument will also collect and prepare samples, calibrate itself, reduce the raw data, and transmit reduced data to the control station(s)

  6. Comparative assessment of licensing processes of uranium mines in Brazil

    International Nuclear Information System (INIS)

    Silva, K.M.; Menezes, R.M.; Mezrahi, A.

    2002-01-01

    Commercial operation of uranium mining and milling started in Brazil, at the Pocos de Caldas Unit, State of Minas Gerais, in 1982. The Pocos de Caldas Unit was licensed by the Brazilian Regulatory Body (CNEN) and its is now in the decommissioning process. In 2000, a new mining and milling installation, the Caetite Unit, located in State of Bahia, started operation. This paper will discuss how Brazilian Nuclear Energy Commission is licensing the Caetite Unit based on the lessons learned from the Pocos de Caldas Unit. The objective is to draw attention to the importance of the safety assessment for a new unit, specially considering that some wrong decisions were taken for the Pocos de Caldas unit. These decisions lead to less effective long term solutions to protect the environment. Notwithstanding the differences between the two units, it is of great value to use the acquired experience to avoid or minimize the short, medium and long term impacts to the environment and population in the new operation. (author)

  7. Mining and processing of uranium deposits in Salamanca, Spain

    International Nuclear Information System (INIS)

    Gomez Jaen, J.P.; Otero, J.; Serrano, J.R.; Membrillera, J.R.; Josa, J.M.

    1977-01-01

    In July, 1974, Empresa Nacional del Uranio, S.A. (ENUSA), took the decision to mine uranium in the province of Salamanca, based on geological and processing studies carried out by the Junta de Energia Nuclear (JEN). The milling plant was designed by JEN and assembled by ENUSA, and operations were begun on 22 May, 1975. The orebody, FE-1, is composed of slate of Cambrain age and the fissures are filled by primary minerals. Secondary minerals are impregnated in the zone affected by the hydrostatic level. The orebody is of the stockwork type in which carbonaceous matter has acted as a reducing agent. The average grade of the ore is 0.09% U 3 O 8 at a cutoff grade of 0.02% U 3 O 8 : the deposit is therefore among the lowest-grade deposits that are currently mined. Annual production is 1 200 000 t of rock, of which 200 000 t is ore-bearing. The milling plant uses a static heap-leaching method, followed by solvent extraction (tertiary amines) and precipitation by ammonia. Joint studies by JEN and ENUSA have led to the introduction of modifications that have increased the production capacity from 75 to 112 t U 3 O 8 per annum with no significant alteration in the initial planned investment. The total recovery after processing is 75% of the U 3 O 8 contained in the ore. Approximately 100 people are employed in the overall operation. ENUSA has decided to expand operations in Salamanca with the construction of a new milling plant (technological aid by JEN), which will be capable of processing 825 000 t of ore per year, with an annual production of 500 t U 3 O 8 . The new plant is expected to begin operations in 1979. (author)

  8. Studies on uranium recovery from inlet stream of Effluent Treatment Plant by novel 'In-House' sorbent

    International Nuclear Information System (INIS)

    Sangita Pal; Tewari, P.K.; Suchismita Mishra; Pandit, G.G.; Puranik, V.D.; Satpati, S.K.

    2011-01-01

    'In-House' resin Polyacrylhydroxamic acid (PHOA) has been synthesized and utilized targeting ground water remediation; recovery of uranium from low concentration aqueous solution e.g., mining activities related water, flooding of excavated or deplumed areas, nuclear plant washed effluent and process generated effluents in nuclear plant during front-end as well as back-end treatment. In the present study, treatment of field effluent containing heavy metals and radio-nuclides from contaminated mining sites reflected preference for uranium with respect to manganese. The specific complexation between the extractant and metal ion especially uranium provides high distribution co-efficient (K d ) for uranium (K d,U = 1,450 mL/g from inlet of Effluent Treatment Plant (ETP) and K d,U = 74,950 mL/g for synthetic solution) compared to high level impurity (1,000 times higher concentration) of manganese (K d,Mn = 111 mL/g from inlet of ETP and K d,Mn = 10,588 mL/g for synthetic solution). The 'In-House' resin showed significant extractability (70-95% elution efficiency) and indicates a possibility of selective removal/recovery of the valuable metal ions even from secondary sources. As a specialty, resin can be regenerated and reused. (author)

  9. The development of the production process for the thorium/uranium dicarbide fuel kernels for the first charge of the Dragon Reactor

    International Nuclear Information System (INIS)

    Burnett, R.C.; Hankart, L.J.; Horsley, G.W.

    1965-05-01

    The development of methods of producing spheroidal sintered porous kernels of hyperstoichiometric thorium/uranium dicarbide solid solution from thorium/uranium monocarbide/carbon and thoria/urania/carbon powder mixes is described. The work has involved study of (i) Methods of preparing green kernels from UC/Th/C powder mixes using the rotary sieve technique. (ii) Methods of producing green kernels from UO2/Th02/C powder mixes using the planetary mill technique. (iii) The conversion by appropriate heat treatment of green kernels produced by both routes to sintered porous kernels of thorium/uranium carbide. (iv) The efficiency of the processes. (author)

  10. Treatment of liquid effluents from uranium analytical method 'DAVIES & GRAY' by electrodialysis and electrodialysis reactive tests

    International Nuclear Information System (INIS)

    Zuniga Alvear, Karina Andrea

    2014-01-01

    This work describes the process which produces liquid waste coming from the chemical analysis laboratory of the Chilean Nuclear Energy Commission (CCHEN), from the analytical technique called 'Davies and Gray' and their further treatment, using electro dialysis (ED) and reactive electro dialysis (RED), in order to achieve lower uranium contents in solution. The contamination in water is a big problem, since there are many places in the world where is limited. For these reasons new treatments must be done, and the ion-selective membrane has opened a new path for these processes. The radioactive liquid waste have lots of other restrictions in their final disposal, which difficult even more the water recovery, because the law has very strict security margins with respect to these ones. In the case of liquid waste containing uranium, the concern increases, because being the uranium a radioactive element has it has to be lowered at its maximum, or eliminated directly, in order to avoid any kind of contamination. There exist national regulations and international recommendations. They have stipulated the correct management and disposal for radioactive waste. These can come from any uranium production process. In any of these, the liquid waste contains certain uranium content, which after the end of the process; the discarded waste must go through a conditioning and cleaning process for its afterward liberation or recycling. In this study, it was tested the electro dialysis as a radioactive waste treatment, only uranium containing waste coming from the chemical analysis laboratory in CCHEN. The electro dialysis process has a direct competition with other separation process, such as distillation, ionic exchange, and reverse osmosis, among others. The classic electro dialysis has been developed during the 50's, and until today, there has been different version, as inverse, reactive, reversible. The unidirectional and reactive electro dialysis will be the

  11. Uranium oxide catalysts: environmental applications for treatment of chlorinated organic waste from nuclear industry.

    Science.gov (United States)

    Lazareva, Svetlana; Ismagilov, Zinfer; Kuznetsov, Vadim; Shikina, Nadezhda; Kerzhentsev, Mikhail

    2018-02-05

    Huge amounts of nuclear waste, including depleted uranium, significantly contribute to the adverse environmental situation throughout the world. An approach to the effective use of uranium oxides in catalysts for the deep oxidation of chlorine-containing hydrocarbons is suggested. Investigation of the catalytic activity of the synthesized supported uranium oxide catalysts doped with Cr, Mn and Co transition metals in the chlorobenzene oxidation showed that these catalysts are comparable with conventional commercial ones. Physicochemical properties of the catalysts were studied by X-ray diffraction, temperature-programmed reduction with hydrogen (H 2 -TPR), and Fourier transform infrared spectroscopy. The higher activity of Mn- and Co-containing uranium oxide catalysts in the H 2 -TPR and oxidation of chlorobenzene in comparison with non-uranium catalysts may be related to the formation of a new disperse phase represented by uranates. The study of chlorobenzene adsorption revealed that the surface oxygen is involved in the catalytic process.

  12. Aeromagnetic data processing and application in the evaluation of uranium resource potential in China

    International Nuclear Information System (INIS)

    Wang Yuanzhi; Zhang Junwei; Feng Chunyuan

    2012-01-01

    The article introduces the main methods to deduce geological structures with aeromagnetic data, and summarizes the prediction elements of aeromagnetic characteristics for granite, volcanic, carbonaceous-siliceous-argillaceous rock and sandstone type uranium deposits. By analysing the relationship of aeromagnetic deduced geological structures and uranium mineralization, the prediction model of combined factors was summarized for each type uranium deposit. A case study in Taoshan-Zhuguang mineralization belt shows that the fault, plutons and volcanic structures deduced from areomagnetic information can judge the favorable mineralization environment and ore control structure. Therefore, the process and application of aeromagnetic data can play an important role in the evaluation of uranium resource potential and uranium exploration. (authors)

  13. Candidate processes for diluting the 235U isotope in weapons-capable highly enriched uranium

    International Nuclear Information System (INIS)

    Snider, J.D.

    1996-02-01

    The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile 235 U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile 235 U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel

  14. Internationally Standardized Reporting (Checklist) on the Sustainable Development Performance of Uranium Mining and Processing Sites

    International Nuclear Information System (INIS)

    Harris, Frank

    2014-01-01

    The Internationally Standardized Reporting Checklist on the Sustainable Development Performance of Uranium Mining and Processing Sites: • A mutual and beneficial work between a core group of uranium miners and nuclear utilities; • An approach based on an long term experience, international policies and sustainable development principles; • A process to optimize the reporting mechanism, tools and efforts; • 11 sections focused on the main sustainable development subject matters known at an operational and headquarter level. The WNA will make available the sustainable development checklist for member utilities and uranium suppliers. Utilities and suppliers are encouraged to use the checklist for sustainable development verification.

  15. Uranium exploration

    International Nuclear Information System (INIS)

    De Voto, R.H.

    1984-01-01

    This paper is a review of the methodology and technology that are currently being used in varying degrees in uranium exploration activities worldwide. Since uranium is ubiquitous and occurs in trace amounts (0.2 to 5 ppm) in virtually all rocks of the crust of the earth, exploration for uranium is essentially the search of geologic environments in which geologic processes have produced unusual concentrations of uranium. Since the level of concentration of uranium of economic interest is dependent on the present and future price of uranium, it is appropriate here to review briefly the economic realities of uranium-fueled power generation. (author)

  16. Converting the Caetité Mill Process to Enhance Uranium Recovery and Expand Production

    Energy Technology Data Exchange (ETDEWEB)

    Gomiero, L. A.; Scassiotti Filho, W.; Veras, A., E-mail: gomiero@inb.gov.br [Indústrias Nucleares do Brasil S/A — INB, Caetité, BA (Brazil); Cunha, J. W. [Instituto de Engenharia Nuclear-IEN/CNEN, Rio de Janeiro, RJ (Brazil); Morais, C. A. [Centro do Desenvolvimento da Tec. Nuclear-CDTN/CNEN, Belo Horizonte, MG (Brazil)

    2014-05-15

    The Caetité uranium mill was commissioned in 2000 to produce about 340 t U per year from an uranium ore averaging 0.29% U{sub 3}O{sub 8}. This production is sufficient to supply the two operating nuclear power plants in the country. As the Brazilian government has recently confirmed its plan to start building another ones from 2009, the uranium production will have to expand its capacity in the next two years. This paper describes the changes in the milling process that are being evaluated in order to not only increase the production but also the uranium recovery, to fulfil the increasing local demand. The heap leaching process will be changed to conventional tank agitated leaching of ground ore slurry in sulphuric acid medium. Batch and pilot plant essays have shown that the uranium recovery can increase from the 77% historical average to about 93%. As the use of sodium chloride as the stripping agent has presented detrimental effects in the extraction and stripping process, two alternatives are being evaluated for the uranium recovery from the PLS: (a) uranium peroxide precipitation at controlled pH from a PLS that was firstly neutralized and filtered. Batch essays have shown good results with a final calcined precipitate averaging 99% U{sub 3}O{sub 8}. Conversely the results obtained at the first pilot plant essay has shown that the precipitation conditions of the continuous process calls for further evaluation. The pilot plant is being improved and another essay will be carried out. (b) uranium extraction with a tertiary amine followed by stripping with concentrated sulphuric acid solution. Efforts are being made to recover the excess sulphuric acid from the pregnant stripping solution to enhance the economic viability of the process and to avoid the formation of a large quantity of gypsum in the pre-neutralization step before the uranium peroxide precipitation. (author)

  17. Determination of uranium and plutonium in metal conversion products from electrolytic reduction process

    International Nuclear Information System (INIS)

    Lee, Chang Heon; Suh, Moo Yul; Joe, Kih Soo; Sohn, Se Chul; Jee, Kwang Young; Kim, Won Ho

    2005-01-01

    Chemical characterization of process materials is required for the optimization of an electrolytic reduction process in which uranium dioxide, a matrix of spent PWR fuels, is electrolytically reduced to uranium metal in a medium of LiCl-Li 2 O molten at 650 .deg. C. A study on the determination of fissile materials in the uranium metal products containing corrosion products, fission products and residual process materials has been performed by controlled-potential coulometric titration which is well known in the field of nuclear science and technology. Interference of Fe, Ni, Cr and Mg (corrosion products), Nd (fission product) and LiCl molten salt (residual process material) on the determination of uranium and plutonium, and the necessity of plutonium separation prior to the titration are discussed in detail. Under the analytical condition established already, their recovery yields are evaluated along with analytical reliability

  18. Main means for reducing the production costs in process of leaching uranium

    International Nuclear Information System (INIS)

    Jiang Lang

    2000-01-01

    The production costs in process of leaching uranium have been reduced by controlling mixture ratio of crudes, milling particle size, liquid/solid mass ratio of leaching pulp, potential and residue acidity, and improving power equipment

  19. Bioleaching of uranium in batch stirred tank reactor: Process optimization using Box–Behnken design

    International Nuclear Information System (INIS)

    Eisapour, M.; Keshtkar, A.; Moosavian, M.A.; Rashidi, A.

    2013-01-01

    Highlights: ► High amount of uranium recovery achieved using Acidithiobacillus ferrooxidans. ► ANOVA shows individual variables and their squares are statistically significant. ► The model can accurately predict the behavior of uranium recovery. ► The model shows that pulp density has the greatest effect on uranium recovery. - Abstract: To design industrial reactors, it is important to identify and optimize the effective parameters of the process. Therefore, in this study, a three-level Box–Behnken factorial design was employed combining with a response surface methodology to optimize pulp density, agitation speed and aeration rate in uranium bioleaching in a stirred tank reactor using a pure native culture of Acidithiobacillus ferrooxidans. A mathematical model was then developed by applying the least squares method using the software Minitab Version 16.1.0. The second order model represents the uranium recovery as a function of pulp density, agitation speed and aeration rate. An analysis of variance was carried out to investigate the effects of individual variables and their combined interactive effects on uranium recovery. The results showed that the linear and quadratic terms of variables were statistically significant whilst the interaction terms were statistically insignificant. The model estimated that a maximum uranium extraction (99.99%) could be obtained when the pulp density, agitation speed and aeration rate were set at optimized values of 5.8% w/v, 510 rpm and 250 l/h, respectively. A confirmatory test at the optimum conditions resulted in a uranium recovery of 95%, indicating a marginal error of 4.99%. Furthermore, control tests were performed to demonstrate the effect of A. ferrooxidans in uranium bioleaching process and showed that the addition of this microorganism greatly increases the uranium recovery

  20. Tailings treatment techniques for uranium mill waste: a review of existing information

    International Nuclear Information System (INIS)

    Sherwood, D.R.; Serne, R.J.

    1983-07-01

    Of primary concern at uranium mill sites in the United States is the potential of ground-water contamination from mill wastes that are disposed in tailings impoundments. Although many systems have been used to control seepage from tailings impoundments, most of these systems are limited in their ability to handle an excess of tailings solution. Three general amelioration methods were identified: neutralization, fixation and specific constituent removal. During neutralization, a reagent is added to the tailings solution to neutralize the acidity and raise the pH to reduce the solubility of various pH sensitive contaminants. Fixation processes add materials such as lime, cement or asphalt to the waste to produce a physically stable composition that resists leaching of hazardous constituents. Specific constituent removal encompasses varying techniques, such as alternate ore leaching processes, effluent treatment with sorption, or ion exchange agents or selected precipitation that reduce specific constituent concentrations in tailings solution. Neutralization processes appear to be best suited for treating uranium mill tailings because they can, at a reasonable cost, limit the solution concentration of many contaminants. The effectiveness of the process depends on the reagent used as well as the waste being treated. Of the six reagents studied (lime, limestone, caustic soda, soda ash, combined limestone/lime and combined alumina/lime/soda), a combined treatment of limestone and lime seems best, especially for tailings containing ferric iron as the limestone economically buffers the solution acidity while the lime takes the pH to 8.0, an optimum level for heavy metal removal. For those tailings containing ferrous iron, lime alone works best. The costs for the lime/limestone or lime processes range from $0.20 to $1.00 per 1000 gal of treated water, excluding capital equipment costs

  1. Bomb reduction of uranium tetrafluoride. Part II: Influence of the addition elements in the reduction process

    International Nuclear Information System (INIS)

    Anca Abati, R.; Lopez Rodriguez, M.

    1962-01-01

    This work shows the influence of uranium oxide and uranyl fluoride in the reduction of uranium with Ca and Mg. These additions are more harmful when using smaller bombs. The uranyl fluoride has influence in the reduction process; the curves yield-concentration shows two regions depending upon the salt concentration. The behaviour of this addition in these regions can be explained following the different decompositions that can take place during the reduction process. (Author) 9 refs

  2. Oxidation experiment of metal uranium waste for the treatment of depleted uranium waste

    International Nuclear Information System (INIS)

    Kang, K. H.; Kwac, K. I.; Kim, K. J.

    2001-01-01

    A study was conducted on the oxidation behavior of U-Ti chips(Depleted Uranium, DU chips) using an XRD and a thermogravimetric analyzer in the temperature range from 250 to 500 .deg. C in air. At the temperature lower than 400 .deg. C, DU chips were converted to UO 2 , U 3 O 7 and U 3 O 8 whereas at the temperature higher than 400 .deg. C, DU chips were completely converted to U 3 O 8 , the most stable form of uranium oxide. The activation energy for the oxidation of U-Ti chips is found, 44.9 kJ/mol and the oxidation rate in terms of weight gain (%) can be expressed as ; dW/dt=8.4 x 10 2 e(-44.9 kJ/mol /RT) wt %/min (250≤T(deg. C)≤500) where W=weight gain (%), t=time and T=temperature

  3. A new methodology using mathematical treatment in uranium recovery of slags from U-metal production

    International Nuclear Information System (INIS)

    Ferreto, Helio Fernando Rodrigues; Araujo, Berta Floh de

    1999-01-01

    U 3 Si 2 fuel was developed by the Fuel Cycle Department of IPEN/CNEN - SP in order to provide high density fuel elements for the IEA-R1m swimming pool reactor. Uranium containing magnesium fluoride slags are produced during the reduction of U F 4 to metallic uranium, the first step of U 3 Si 2 production. Since enriched uranium is used and taking in account process economics and environmental impacts, the recovery of uranium from the slags is highly recommended. This work deals with the uranium recovery from magnesium fluoride slag via nitric acid leaching process using a new methodology for the study. A statistical procedure for process optimization was applied using a fractional factorial design at two levels and four variables represented as 2 4-1 . Variance analysis followed by multiple regression was used, setting up a first order polygonal model, as follow: y 92,409 +3,825 x 1 - 0,875 x 3 + 1,65 x 4 - 0,95 x 3 x 4 Standard error 1,04572. This equation represents the variables and the most suitable interactions in the uranium recovery process. By using this equation, one can obtain in advance and without making experiments the values from the process variables for a giving process yield. (author)

  4. EFFECT OF CURRENT, TIME, FEED AND CATHODE TYPE ON ELECTROPLATING PROCESS OF URANIUM SOLUTION

    Directory of Open Access Journals (Sweden)

    Sigit Sigit

    2017-02-01

    Full Text Available ABSTRACT   EFFECT OF CURRENT, TIME, FEED AND CATHODE TYPE ON ELECTROPLATING PROCESS OF URANIUM SOLUTION. Electroplating process of uranyl nitrate and effluent process has been carried out in order to collect uranium contained therein using electrode Pt / Pt and Pt / SS at various currents and times. Material used for electrode were Pt (platinum and SS (Stainlees Steel. Feed solution of 250 mL was entered into a beaker glass equipped with Pt anode - Pt cathode or Pt anode - SS cathode, then fogged direct current from DC power supply with specific current and time so that precipitation of uranium sticking to the cathode. After the processes completed, the cathode was removed and weighed to determine weight of precipitates, while the solution was analyzed to determine the uranium concentration decreasing after and before electroplating process. The experiments showed that a relatively good time to acquire uranium deposits at the cathode was 1 hour by current 7 ampere, uranyl nitrate as feed, and Pt (platinum as cathode. In these conditions, uranium deposits attached to the cathode amounted to 74.96% of the original weight of uranium oxide in the feed or 206.5 mg weight. The use of Pt cathode for  uranyl nitrate, SS and Pt cathode for effluent process feed gave uranium specific weight at the cathode of 12.99 mg/cm2, 2.4 mg/cm2 and 5.37 mg/cm2 respectively for current 7 ampere and electroplating time 1 hour. Keywords: Electroplating, uranyl nitrate, effluent process, Pt/Pt electrode, Pt/SS electrode

  5. Study on extraction of uranium from clayey sandstone with floatation-leaching process

    International Nuclear Information System (INIS)

    Meng Guangshou; Zhao Manchang; Wu Peisheng; Song Wenlan; Li Wenxia.

    1985-01-01

    An improved floatation-leaching process is proposed to extract uranium from some clayey sandstone type of ore. By two-step flotation, the ground feed ore can be divided into three urani-ferous sections, i.e., the sulfidic concentrate carrying organic matter, the carbonate concentrate, and the tailings. The sulfidic concentrate is mixed with the tailings and then treated by acid-leaching with the result that 93% uranium extraction can be attained. The excess free acid of the leached slurry is further neutralized with the carbonate concentrate instead of lime commonly used. As a result, approximately 60% uranium extraction can be attained. As a whole, by the flotation-leaching process the acid consumption can be reduced from 200 kg/t down to < 80 kg/t and the uranium extraction can be raised from 85% to 90% as compared with the conventional acid-leaching process

  6. Providing radiation safety for the environment and people at uranium ore mining and primary processing operations and treatment of radioactive wastes in the Navoi mining and metallurgy combinat, Uzbekistan

    International Nuclear Information System (INIS)

    Kuchersky, N.I.

    1997-01-01

    The rise of the uranium industry in the Republic of Uzbekistan is closely connected with the discovery of a number of significant uranium deposits of sheet sandstone type in the Kyzylkum desert area, between the Amu-Darya and Syr-Darya rivers. Based on these deposits, in 1958 the construction of Navoi Mining and Metallurgical Combinat (NMMC) commenced. In 1965 the Hydrometallurgical Plant No. 1 (HMP-1), located in the industrial zone near the Navoi town, started producing the uranium protoxide-oxide (yellow cake). The structure of the NMMC uranium production operations includes HMP-1 and three mining facilities. Conventional open-pit and underground uranium mining were shut down here in 1994 and at present all the uranium is extracted by in situ leaching (ISL). The radiation and hygiene sanitary monitoring aimed at collecting information on the radiation conditions at the working places and in the environment, on the current and expected irradiation doses taken by the personnel and various population groups inhabiting the area involved in the activities of the existing, liquidated or temporarily closed NMMC facilities constitutes an integral part of the radiation safety providing system. No radioactive contamination of any environmental objects has been detected outside the sanitary zones and production sites. The radionuclides content in the atmosphere, on the ground surface, in the underground water was found to be at the background level. Current annual average exposure doses of the limited critical population groups were detected to be essentially lower than the current international standards and were about 1 mSv per year

  7. Non-cyanide process for flotation of a uranium-bearing lead-zinc polymetallic sulphide ore

    International Nuclear Information System (INIS)

    Li Qingxin

    1988-01-01

    The characteristics of the minerals of a urnium-bearing lead-zinc ore are described in this paper, And the experimentsl results of non-cyanide flotation process are given. The tests show that the selective flotation process of lead and zinc followed by uranium treatment is feasible in technology and reasonable in economics. When the run-of-mine contains 2.86%Pb, 2.47%Zn and 0,019%U, the lead concentrate containing 65.13%Pb, and 4.51%Zn, the zinc concentrate containing 52.00%Zn and 1.22%Pb, and the uranium concentrate containing 0.028%U can be obtained with the recoveries of 94.87%Pb, 87.61%Zn and 66.13%U respectively. The influence of sodium sulphite on flotaion process, the effect of sodium sulphite and the flotation mechanism of dibutyldithiophosphate ammonium are also discussed

  8. Study on technology for radioactive waste treatment and management from uranium production

    International Nuclear Information System (INIS)

    Vu Hung Trieu; Vu Thanh Quang; Nguyen Duc Thanh; Trinh Giang Huong; Tran Van Hoa; Hoang Minh Chau; Ngo Van Tuyen; Nguyen Hoang Lan; Vuong Huu Anh

    2007-01-01

    There is some solid and liquid radioactive waste created during producing Uranium that needs being treated and managed to keep our environment safe. This radioactive waste contains Uranium (U-238), Thorium (Th-232), Radium (Ra-226) and some heavy metals and mainly is low radioactive waste. Our project has researched and built up appropriate technology for treating and managing the radioactive waste. After researching and experimenting, we have built up four technology processes as follows: Technology for separating Radium from liquid waste; Technology for treating and managing solid waste containing Ra; Technology for separating Thorium from liquid waste after recovering radium; Technology for stabilizing solid waste from Uranium production. (author)

  9. Selected bibliography for the extraction of uranium from seawater: chemical process and plant design feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Binney, S.E.; Polkinghorne, S.T.; Jante, R.R.; Rodman, M.R.; Chen, A.C.T.; Gordon, L.I.

    1979-02-01

    A selected annotated bibliography of 521 references was prepared as a part of a feasibility study of the extraction of uranium from seawater. For the most part, these references are related to the chemical processes whereby the uranium is removed from the seawater. A companion docment contains a similar bibliography of 471 references related to oceanographic and uranium extraction plant siting considerations, although some of the references are in common. The bibliography was prepared by computer retrieval from Chemical Abstracts, Nuclear Science Abstracts, Energy Data Base, NTIS, and Oceanic Abstracts. References are listed by author, country of author, and selected keywords.

  10. Recovery of uranium from wet process by the chloridic leaching of phosphate rocks

    International Nuclear Information System (INIS)

    Santana, A.O.; Paula, H.C.B.; Dantas, C.C.

    1984-01-01

    Uranium was recovered from chloridic leach liquor of phosphate rocks by solvent extraction on a laboratory scale. The extractor system is a mixture of di-(2-ethylhexyl) phosphoric acid (D 2 EHPA) and tributyl-phosphate (TBP) in a varsol diluent. The uranium concentration is 150 ppm in the rocks and 12 ppm in the leach liquor. The phosphate rocks are leached on a semi-industrial scale for dicalcium phosphate production. The recovery process comprises the following steps: extraction, reextraction, iron removal and uranium precipitation. (orig./EF)

  11. Selected bibliography for the extraction of uranium from seawater: chemical process and plant design feasibility study

    International Nuclear Information System (INIS)

    Binney, S.E.; Polkinghorne, S.T.; Jante, R.R.; Rodman, M.R.; Chen, A.C.T.; Gordon, L.I.

    1979-02-01

    A selected annotated bibliography of 521 references was prepared as a part of a feasibility study of the extraction of uranium from seawater. For the most part, these references are related to the chemical processes whereby the uranium is removed from the seawater. A companion docment contains a similar bibliography of 471 references related to oceanographic and uranium extraction plant siting considerations, although some of the references are in common. The bibliography was prepared by computer retrieval from Chemical Abstracts, Nuclear Science Abstracts, Energy Data Base, NTIS, and Oceanic Abstracts. References are listed by author, country of author, and selected keywords

  12. Recovery of uranium from wet process by the chloridic leaching of phosphate rocks

    Energy Technology Data Exchange (ETDEWEB)

    Santana, A O; Paula, H C.B.; Dantas, C C

    1984-03-01

    Uranium was recovered from chloridic leach liquor of phosphate rocks by solvent extraction on a laboratory scale. The extractor system is a mixture of di-(2-ethylhexyl) phosphoric acid (D/sub 2/EHPA) and tributyl-phosphate (TBP) in a varsol diluent. The uranium concentration is 150 ppm in the rocks and 12 ppm in the leach liquor. The phosphate rocks are leached on a semi-industrial scale for dicalcium phosphate production. The recovery process comprises the following steps: extraction, reextraction, iron removal and uranium precipitation.

  13. Disposal of wastes from uranium conversion and enrichment processes

    International Nuclear Information System (INIS)

    Costello, J.M.

    1981-11-01

    This paper reviews the general principles and objectives in radioactive waste management, and shows how these are applied in options for management and disposal of wastes from uranium upgrading operations. Some estimates of radiological dose commitments and health effects from LWR nuclear power and its fuel cycle have been made for US conditions

  14. Process for extracting uranium from phosphoric acid solutions

    International Nuclear Information System (INIS)

    1977-01-01

    The description is given of a method for extracting uranium from phosphoric acid solutions whereby the previously oxided acid is treated with an organic solvent constituted by a mixture of dialkylphosphoric acid and trialkylphosphine oxide in solution in a non-reactive inert solvent so as to obtain de-uraniated phosphoric acid and an organic extract constituted by the solvent containing most of the uranium. The uranium is then separated from the extract as uranyl ammonium tricarbonate by reaction with ammonia and ammonium carbonate and the extract de-uraniated at the extraction stage is recycled. The extract is treated in a re-extraction apparatus comprising not less than two stages. The extract to be treated is injected at the top of the first stage. At the bottom of the first stage, ammonia is introduced counter current as gas or as an aqueous solution whilst controlling the pH of the first stage so as to keep it to 8.0 or 8.5 and at the bottom of the last stage an ammonium carbonate aqueous solution is injected in a quantity representing 50 to 80% of the stoichiometric quantity required to neutralize the dialkylphosphoric acid contained in the solvent and transform the uranium into uranyl ammonium tricarbonate [fr

  15. Process for iron separation from an organic solution containing uranium

    International Nuclear Information System (INIS)

    Textoris, A.; Lyaudet, G.; Bathelier, A.

    1987-01-01

    Iron is separated from an organic solution of U and Fe in a phosphine oxide and an acid organic phosphorus compound by reaction on oxalic acid or a mixture of sulfuric and phosphoric acid or phosphoric acid. Uranium stays in the initial organic solution and iron is transferred to the aqueous phase [fr

  16. Risks associated with mining and processing of uranium

    International Nuclear Information System (INIS)

    Archer, V.E.

    1976-01-01

    Mortality from all causes was determined for groups of white and Indian underground uranium miners, and for a small group of uranium mill workers. Analysis was by a life table method. A significant excess of respiratory cancer was found among both white and Indian miners. A significant excess of nonmalignant respiratory disease was found among whites. It was attributed primarily to diffuse parenchymal damage by radiation; it approaches respiratory cancer in importance as a cause of death among whites. A significant excess of malignant disease of the lymphatic and hematopoietic tissue was found among uranium mill workers. This was attributed primarily to irradiation of lymph nodes by thorium-230. Exposure-response curves for nonsmoking uranium miners are linear for both respiratory cancer and ''other respiratory disease.'' Cigarette smoking elevates and distorts that curve. Light cigarette smokers appear to be most vulnerable to lung parenchymal damage by the radiation. The predominant histological type of cancer among nonsmokers, white smokers and Indians is small cell undifferentiated. Accidental deaths are high among inexperienced miners, and even among experienced miners it is about 3 times what is expected

  17. Uranium recovery from AVLIS slag

    International Nuclear Information System (INIS)

    D'Agostino, A.E.; Mycroft, J.R.; Oliver, A.J.; Schneider, P.G.; Richardson, K.L.

    2000-01-01

    Uranium metal for the Atomic Vapor Laser Isotope Separation (AVLIS) project was to have been produced by the magnesiothermic reduction of uranium tetrafluoride. The other product from this reaction is a magnesium fluoride slag, which contains fine and entrained natural uranium as metal and oxide. Recovery of the uranium through conventional mill leaching would not give a magnesium residue free of uranium but to achieve more complete uranium recovery requires the destruction of the magnesium fluoride matrix and liberation of the entrapped uranium. Alternate methods of carrying out such treatments and the potential for recovery of other valuable byproducts were examined. Based on the process flowsheets, a number of economic assessments were performed, conclusions were drawn and the preferred processing alternatives were identified. (author)

  18. Recovery of uranium in the production of concentrated phosphoric acid by a hemihydrate process

    International Nuclear Information System (INIS)

    Nakajima, S.; Miyamoto, M.

    1983-01-01

    Nissan Chemical Industries as manufacturers of phosphoric acid have studied the recovery of uranium, based on a concentrated phosphoric acid production process. The process consists of two stages, a hemihydrate stage with a formation of hemihydrate and a filtration section, followed by a dihydrate stage with hydration and a filtration section. In the hemihydrate stage, phosphate is treated with a mixture of phosphoric acid and sulphuric acid to produce phosphoric acid and hydrous calcium sulphate; the product is recovered in the filtration section and its concentration is 40-50% P 2 O 3 . In the dihydrate stage, the hemihydrate is transformed by re-dissolution and hydration, producing hydrous calcium sulphate, i.e. gypsum. This process therefore comprises two parts, each with different acid concentrations. As the extraction of uranium is easier in the case of a low concentration of phosphoric acid, the process consists of the recovery of uranium starting from the filtrate of the hydration section. The tests have shown that the yield of recovery of uranium was of the order of 80% disregarding the handling losses and no disadvantageous effect has been found in the combination of the process of uranium extraction with the process of concentrated phosphoric acid production. Compared with the classical process where uranium is recovered from acid with 30% P 2 O 5 , the process of producing high-concentration phosphoric acid such as the Nissan process, in which the uranium recovery is effected from acid with 15% P 2 O 5 from the hydration section, presents many advantages [fr

  19. A study on oxidation treatment of uranium metal chip under controlling atmosphere for safe storage

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Ji, Chul Goo; Bae, Sang Oh; Woo, Yoon Myeoung; Kim, Jong Goo; Ha, Yeong Keong

    2011-01-01

    The U metal chips generated in developing nuclear fuel and a gamma radioisotope shield have been stored under immersion of water in KAERI. When the water of the storing vessels vaporizes or drains due to unexpected leaking, the U metal chips are able to open to air. A new oxidation treatment process was raised for a long time safe storage with concepts of drying under vacuum, evaporating the containing water and organic material with elevating temperature, and oxidizing the uranium metal chips at an appropriate high temperature under conditions of controlling the feeding rate of oxygen gas. In order to optimize the oxidation process the uranium metal chips were completely dried at higher temperature than 300 .deg. C and tested for oxidation at various temperatures, which are 300 .deg. C, 400 .deg. C, and 500 .deg. C. When the oxidation temperature was 400 .deg. C, the oxidized sample for 7 hours showed a temperature rise of 60 .deg. C in the self-ignition test. But the oxidized sample for 14 hours revealed a slight temperature rise of 7 .deg. C representing a stable behavior in the self-ignition test. When the temperature was 500 .deg. C, the shorter oxidation for 7 hours appeared to be enough because the self-ignition test represented no temperature rise. By using several chemical analyses such as carbon content determination, X-ray deflection (XRD), Infrared spectra (IR) and Thermal gravimetric analysis (TGA) on the oxidation treated samples, the results of self-ignition test of new oxidation treatment process for U metal chip were interpreted and supported

  20. Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

    Energy Technology Data Exchange (ETDEWEB)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2008-07-08

    Uraninite (UO2) and metaschoepite (UO3·2H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21°C and 50°C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.004±0.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21°C than the particles prepared at 50°C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  1. Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

    International Nuclear Information System (INIS)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2008-01-01

    Uraninite (UO2) and metaschoepite (UO3-2H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21 C and 50 C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.004 ± 0.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21 C than the particles prepared at 50 C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  2. Process for separately recovering uranium, transuranium elements, and fission products of uranium from atomic reactor fuel

    International Nuclear Information System (INIS)

    Balal, A.L.; Metscher, K.; Muehlig, B.; Reichmuth, C.; Schwarz, B.; Zimen, K.E.

    1976-01-01

    Spent reactor fuel elements are dissolved in dilute nitric acid. After addition of acetic acid as a complexing agent, the nitric acid is partly decomposed and the mixture subjected to electrolysis while a carrier liquid, which may be dilute acetic acid or a dilute mixture of acetic acid and nitric acid is caused to flow in the electric field between the electrodes either against the direction of ion migration or transversely thereto. The ions of uranium, plutonium, and other transuranium elements, and of fission products accumulate in discrete portions of the electrolyte and are separately withdrawn as at least three fractions after one or more stages of electrolysis

  3. Reducing uranium and thorium level in Zircon: effect of heat treatment on rate of leaching

    International Nuclear Information System (INIS)

    Meor Yusoff Meor Sulaiman

    2002-01-01

    Considerable amount of uranium and thorium are found in Malaysian zircon and the level is much higher than the minimum value adopted by many importing countries. Selective leaching had been applied as an important technique to reduce these elements. An initial study was carried out using hydrochloric acid leaching system but the result was not favourable. The rate of uranium and thorium leached can be further improved by introducing a heat pretreatment process prior to leaching (Author)

  4. Status report from USSR [Processing of Low-Grade Uranium Ores]; Doklad o sostoyanii voprosa v SSSR

    Energy Technology Data Exchange (ETDEWEB)

    Zefirov, A P [Gosudarstvennyj Komitet Po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moskva, Union of Soviet Socialist Republics (Russian Federation)

    1967-06-15

    The uranium industry for processing poor uranium ores in the USSR was established in recent years. As a result of research work institutions and enterprises in the development of this industry was provided by rapid technological advances that allowed dramatically increased productivity, reduced consumption of reagents, simplified process flow diagrams, and reduced production costs. At present, the basis for uranium industry, including and poor uranium ore deposits in the USSR are with different content valuable components (uranium, phosphorus, molybdenum, rare earth elements, thorium, iron, .. .)

  5. Study of rolled uranium annealing process; Etude du recuit de l'uranium lamine

    Energy Technology Data Exchange (ETDEWEB)

    Cabane, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1954-06-15

    The dilatometric study of rolled uranium clearly shows not only the expansions or contractions induced by stress relief or diffusion of vacancies, but also the slope variations of the cooling curves, which are the best evidence of a texture change. Under the microscope, hard-rolled sheets appear as a mixture of two distinct structures; it is also possible by intermediate annealing to prepare homogeneous sheets of either structure, i.e. twinned or untwinned. All these sheets which have similar textures, undergo at first a primary recrystallization beginning at 320 deg C, then a texture change without any apparent crystal growth, at about 430 deg C. (author) [French] L'anisotropie de l'uranium {alpha} se manifeste fortement dans les coefficients de dilatation. Aussi la dilatometrie permet lle de reperer facilement les expansions ou contractions dues a des relachements de tensions ou a des disparitions de lacunes, ainsi que les variations de pente des courbes de refroidissement, qui constituent la plus importante manifestation d'un changement de texture. Au microscope, les toles fortement ecrouies apparaissent generalement formees d'un melange de deux structures differentes; on a pu aussi preparer des toles de structure homogene: les unes formees de cristaux macles, les autres apparemment depourvues de macles. Ces toles, qui ont toutes a peu pres la meme texture, subissent d'abord une recristallisation primaire a partir de 320 deg C, puis a 430 deg C environ, un changement de texture sans grossissement apparent de cristaux. (auteur)

  6. Migration of uranium process wastes from the uranium-233--thorium-232 cycle

    International Nuclear Information System (INIS)

    Fried, S.; Sabau, C.; Hines, J.; Friedman, A.

    1978-03-01

    With the advent of fuel loadings of 233 U in the Shippingport Reactor, it has become important to understand the migratory behavior of uranium. The purpose of this study is the determination of the parameters influencing the migration of U(VI), the most likely chemical form of uranium to be mobilized from a repository. Samples of rhyolite tuff were used to measure the absorption coefficients of solutions of U(VI) in ground waters. In addition, columns of tuff were used to measure the elution behavior of U(VI) at various conditions of pH, U(VI) concentration, and flow saturation. These results indicate that there are several elution peaks with values of K/sub d/ between 35 and 120. This behavior is not the same as that of Pu(VI) on tuff; and the experimental results to date have not revealed the reason for this difference. Values of K/sub d/ in this range imply that geological containment would be difficult in strata of this type. It may be possible to find more retentive strata than tuff. Rocks containing reducing components are the most likely candidates and further investigation is urgently needed if the 233 U-Th cycle is to be widely used

  7. Study of the impact of environmental bacteria ob uranium speciation in order to engage bioremediation process

    International Nuclear Information System (INIS)

    Untereiner, G.

    2008-11-01

    Uranium is both a radiological and a chemical toxic. Its concentration in the environment is low except when human activities have caused pollution. Uranium is a heavy reactive element, and thus it is easily complexed with soil component like minerals or organic molecules. These different complexes can be more or less bioavailable for microorganisms and plants, and then get in the human food chain. The knowledge and the understanding of transfer mechanisms and also the fate of toxic elements in the biosphere are a key issue to estimate health and ecological hazards. The knowledge of the speciation is very important for bioremediation processes. Here, we focused on the microorganisms effects onto uranium speciation in environment. Bacteria can accumulate and/or transform uranium depending on the initial form of the element. Thus, its bioavailability could be changed. The species used in this work are Cupriavidus metallidurans CH34, which is an environmental bacteria with a high resistance to heavy metal, Deinococcus radiodurans R1, which is known for his radiological resistance, and Rhodopseudomonas palustris, which is a purple photo-trophic bacteria capable of degrading aromatic compounds. Two forms of uranium were used with these bacteria, a mineral one, uranyl carbonate, and an organic one, uranyl citrate. In a first step, the growth media were modified in order to stabilize uranium complexes thanks to a simulation program. Then, the capacity of the bacteria to accumulate or transform uranium was studied. We saw a difference between minimal inhibition concentrations of these two speciation which is due to a difference between phosphate bioavailability. No accumulation was observed with environmental pH but uranium precipitation was observed with acidic pH (pH 1). Uranium speciation seemed to be well controlled in the growth media and the precipitates were uranyl phosphate. (author)

  8. Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks

    International Nuclear Information System (INIS)

    Mosteller, R.D.; Little, R.C.

    1998-01-01

    Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment

  9. Evaluation of selected neutralizing agents for the treatment of uranium tailings leachates. Laboratory progress report

    International Nuclear Information System (INIS)

    Sherwood, D.R.; Serne, R.J.

    1983-02-01

    Laboratory experiments were conducted to evaluate the performance of selected neutralizing agents for the treatment of uranium tailings solutions. Highly acidic tailings solutions (pH 3 ) reagent grade; Calcium hydroxide [Ca(OH) 2 ] reagent grade; Magnesium oxide (MgO) reagent grade; Sodium carbonate (Na 2 CO 3 ) reagent grade; and Sodium hydroxide (NaOH) reagent grade. Evaluation of the effectiveness for the treatment of uranium tailings solutions for the selected neutralizing agents under controlled laboratory conditions was based on three criteria. The criteria are: (1) treated effluent water quality, (2) neutralized sludge handling and hydraulic properties, and (3) reagent costs and acid neutralizing efficiency. On the basis of these limited laboratory results calcium hydroxide or its dehydrated form CaO (lime) appears to be the most effective option for treatment of uranium tailings solutions

  10. Canadian uranium mines and mills evolution of regulatory expectations and requirements for effluent treatment

    International Nuclear Information System (INIS)

    LeClair, J.; Ashley, F.

    2006-01-01

    The regulation of uranium mining in Canada has changed over time as our understanding and concern for impacts on both human and non-human biota has evolved. Since the mid-1970s and early 1980s, new uranium mine and mill developments have been the subject of environmental assessments to assess and determine the significance of environmental effects throughout the project life cycle including the post-decommissioning phase. Water treatment systems have subsequently been improved to limit potential effects by reducing the concentration of radiological and non-radiological contaminants in the effluent discharge and the total loadings to the environment. This paper examines current regulatory requirements and expectations and how these impact uranium mining/milling practices. It also reviews current water management and effluent treatment practices and performance. Finally, it examines the issues and challenges for existing effluent treatment systems and identifies factors to be considered in optimizing current facilities and future facility designs. (author)

  11. Environmental monitoring data review of a uranium ore processing facility in Argentina

    International Nuclear Information System (INIS)

    Bonetto, J.

    2014-01-01

    An uranium ore processing facility in the province of Mendoza (Argentina) that has produced uranium concentrate from 1954 to 1986 is currently undergoing the last steps of environmental restoration. The operator has been performing post-closure environmental monitoring since 1986, while the Nuclear Regulatory Authority (ARN) has been carrying out its own independent radiological environmental monitoring for verification purposes since its creation, in 1995. A detailed revision of ARN´s monitoring plan for uranium mining and milling facilities has been undergoing since 2013, starting with this particular site. Results obtained from long-time sampling locations (some of them currently unused) have been analyzed and potentially new sampling points have been studied and proposed. In this paper, some statistical analysis and comparison of sampling-points’ datasets are presented (specifically uranium and radium concentration in groundwater, surface water and sediments) with conclusions pertaining to their keeping or discarding as sampling points in future monitoring plans. (author)

  12. Uranium recovery from the concentrated phosphoric acid prepared by the hemi-hydrate process

    Energy Technology Data Exchange (ETDEWEB)

    Fouad, E A; Mahdy, M A; Bakr, M Y [Nuclear materials authority, Cairo, (Egypt); Zatout, A A [Faculty of engineering, Alex. university, Alex, (Egypt)

    1995-10-01

    It has been proved that the uranium dissolution from El-sebaiya phosphate ore was possible by using 10 Kg of K Cl O{sub 4}/ ton rock during the preparation of high strength phosphoric acid using the hemi hydrate process. In the present work, effective extraction of uranium (about 90%) from the high strength phosphoric acid using a new synergistic solvent mixture of 0.75 M D 2 EHPA/0.1 M TOHPO had been a success. Stripping of uranium from the organic phase was possible by 10 M phosphoric acid while the direct precipitation of uranium concentrate from the later was feasible by using N H{sub 4} F in presence of acetone. 8 figs.

  13. Uranium recovery from the concentrated phosphoric acid prepared by the hemi-hydrate process

    International Nuclear Information System (INIS)

    Fouad, E.A.; Mahdy, M.A.; Bakr, M.Y.; Zatout, A.A.

    1995-01-01

    It has been proved that the uranium dissolution from El-sebaiya phosphate ore was possible by using 10 Kg of K Cl O 4 / ton rock during the preparation of high strength phosphoric acid using the hemi hydrate process. In the present work, effective extraction of uranium (about 90%) from the high strength phosphoric acid using a new synergistic solvent mixture of 0.75 M D 2 EHPA/0.1 M TOHPO had been a success. Stripping of uranium from the organic phase was possible by 10 M phosphoric acid while the direct precipitation of uranium concentrate from the later was feasible by using N H 4 F in presence of acetone. 8 figs

  14. Status Report from Sweden [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, A [AB Atomenergi, Stockholm (Sweden)

    1967-06-15

    The Ministry of Education was authorized in November 1945 to appoint a commission to study the organization of nuclear energy research. In April 1947 this commission, the Swedish Atomic Energy Commission, proposed the formation of a semi-state-owned company to be a central body for applied research work and development in the nuclear energy field in Sweden. In November 1947 the Atomic Energy Company (AB Atomenergi) had its statutory meeting. The State owns 4/7 of the share capital and the remaining 3/7 is owned by 71 private and municipal share-holders. Except for a part of the stock capital, all investments and running costs of the company have been financed by the Government. The company is in practice answerable to the Department of Commerce which has an advisory body, the Atomic Energy Board. AB Atomenergi is responsible for Government-financed research on the industrial applications of nuclear energy, the milling of uranium ores and refining of uranium. The total number of employees is at present about 1400, 800 of which work at the company's research establishment Studsvik about 120 km south of Stockholm. As early as 1945 the Research Institute of the Swedish National Defence started work in the field of uranium processing. Similar work was also started quite early by the Boliden Mining Company, the Swedish Shale Oil Company and Wargons AB. After the establishment of AB Atomenergi, all work in the uranium processing field was transferred to this company. In fact one of the main reasons for the formation of AB Atomenergi was the need for Swedish uranium production as there was no possibility of importing uranium at that time. As a result of research and development in uranium processing a pilot plant at Kvarntorp near Orebro in central Sweden started milling a low-grade uranium ore (shale) in 1953. The capacity of this plant was 5-10 tons of uranium a year. A uranium mill at Ranstad in south-west Sweden, near Skovde, with a capacity of 120 tons of uranium a

  15. Aquifer restoration at in-situ leach uranium mines: evidence for natural restoration processes

    International Nuclear Information System (INIS)

    Deutsch, W.J.; Serne, R.J.; Bell, N.E.; Martin, W.J.

    1983-04-01

    Pacific Northwest Laboratory conducted experiments with aquifer sediments and leaching solution (lixiviant) from an in-situ leach uranium mine. The data from these laboratory experiments and information on the normal distribution of elements associated with roll-front uranium deposits provide evidence that natural processes can enhance restoration of aquifers affected by leach mining. Our experiments show that the concentration of uranium (U) in solution can decrease at least an order of magnitude (from 50 to less than 5 ppM U) due to reactions between the lixiviant and sediment, and that a uranium solid, possibly amorphous uranium dioxide, (UO 2 ), can limit the concentration of uranium in a solution in contact with reduced sediment. The concentrations of As, Se, and Mo in an oxidizing lixiviant should also decrease as a result of redox and precipitation reactions between the solution and sediment. The lixiviant concentrations of major anions (chloride and sulfate) other than carbonate were not affected by short-term (less than one week) contact with the aquifer sediments. This is also true of the total dissolved solids level of the solution. Consequently, we recommend that these solution parameters be used as indicators of an excursion of leaching solution from the leach field. Our experiments have shown that natural aquifer processes can affect the solution concentration of certain constituents. This effect should be considered when guidelines for aquifer restoration are established

  16. Ore leaching processing for yellow cake production and assay of their uranium content by radiometric analysis

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Rahman, Mohamed A.E. [Nuclear Engineering Department, Military Technical College, Kobry El-Kobbah, Cairo (Egypt); El-Mongy, Sayed A. [Nuclear and Radiological Regularity Authority (ENRRA), Nasr City, Cairo (Egypt)

    2018-01-17

    In this study, Ore granite samples were collected from Gattar site for leashing of yellow cake. The process involves heap leaching of uranium through four main steps; size reduction, leaching, uranium purification, and finally precipitation and filtration. The separation process has been given in details and as flow chart. Gamma spectrometry based on HpGe detector and energy dispersive X-ray (EDX) were used to assay uranium content and activity before and after separation. The uranium weight percentage value as measured by EDX were found to be 40.5 and 67.5 % before and after purification respectively. The results of the calculations based on gamma measurements show high uranium activity and the uranium activity ratios values are 0.045 ± 4.9, 0.043 ± 4.7, and 0.046 ± 2.3 %, before purification, whereas these values were found to be 0.050 ± 3.3, 0.049 ± 3.3, and 0.050 ± 2.7 %, after purification, respectively. The results are discussed in details in the paper. (copyright 2018 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  17. Measurement of the oxidation-extraction of uranium from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    Lawes, B.C.

    1985-01-01

    The present invention relates to processes for the recovery of uranium from wet-process phosphoric acid and more particularly to the oxidation-extraction steps in the DEPA-TOPO process for such recovery. A more efficient use of oxidant is obtained by monitoring the redox potential during the extraction step

  18. Uranium extraction from gold-uranium ores

    Energy Technology Data Exchange (ETDEWEB)

    Laskorin, B.N.; Golynko, Z.Sh.

    1981-01-01

    The process of uranium extraction from gold-uranium ores in the South Africa is considered. Flowsheets of reprocessing gold-uranium conglomerates, pile processing and uranium extraction from the ores are presented. Continuous counter flow ion-exchange process of uranium extraction using strong-active or weak-active resins is noted to be the most perspective and economical one. The ion-exchange uranium separation with the succeeding extraction is also the perspective one.

  19. Reaction of Antimony-Uranium Composite Oxide in the Chlorination Treatment of Waste Catalyst - 13521

    International Nuclear Information System (INIS)

    Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi

    2013-01-01

    The effect of oxygen gas concentration on the chlorination treatment of antimony-uranium composite oxide catalyst waste was investigated by adding different concentrations of oxygen at 0-6 vol% to its chlorination agent of 0.6 or 6 vol% hydrogen chloride gas at 1173 K. The addition of oxygen tended to prevent the chlorination of antimony in the oxide. When 6 vol% hydrogen chloride gas was used, the addition of oxygen up to 0.1 vol% could convert the uranium contained in the catalyst to U 3 O 8 without any significant decrease in the reaction rate compared to that of the treatment without oxygen. (authors)

  20. The recycle of depleted uranium waste products by a hydrometallurgical process

    International Nuclear Information System (INIS)

    Nachtrab, William T.; Schlier, David S.; Pollock, Eugene N.; Shinopulos, George

    1992-01-01

    Nuclear Metals, Inc. has developed a process for recycling uranium scrap materials into high quality metal. The process involves the dissolution of scrap metal in an aqueous solution of 2.4 N HCI and 0.16 N HBF 4 , followed by precipitation of UF 4 through the addition of HF. The precipitated green salt is Filtered, washed, dried, and heat treated after which it is suitable for reduction to metal. The product and the process are referred to as Hydromet, since it is a hydrometallurgical approach to producing green salt. Conventionally, green salt is produced by a pyrometallurgical technique. The steps of the process are described and results presented for derbies produced using Hydromet green salt. With proper process selection and appropriate heat treatment, green salt produced by Hydromet is fully equivalent to pyrometallurgical green salt. Hydromet green salt can be reduced to metal using the identical process used for pyromet green salt. Good quality, well-formed derbies can be readily produced. (author)

  1. The study on microb and organic metallogenetic process of the interlayer oxidized zone uranium deposit. A case study of the Shihongtan uranium deposit in Turpan-Hami basin

    International Nuclear Information System (INIS)

    Qiao Haiming; Shang Gaofeng

    2010-01-01

    Microbial and organic process internationally leads the field in the study of metallogenetic process presently. Focusing on Shi Hongtan uranium deposit, a typical interlayer oxidized zone sandstone-type deposit, this paper analyzes the geochemical characteristics of microb and organic matter in the deposit, and explores the interaction of microb and organic matter. It considers that the anaerobic bacterium actively takes part in the formation of the interlayer oxidized zone, as well as the mobilization and migration of uranium. In the redox (oxidation-reduction) transition zone, sulphate-reducing bacteria reduced sulphate to stink damp, lowing Eh and acidifying pH in the groundwater, which leads to reducing and absorbing of uranium, by using light hydrocarbon which is the product of the biochemical process of organism and the soluble organic matter as the source of carbon. The interaction of microb and organic matter controls the metallogenetic process of uranium in the deposit. (authors)

  2. Environmental assessment of remedial action at the Naturita uranium processing site near Naturita, Colorado: Revision 5

    Energy Technology Data Exchange (ETDEWEB)

    1994-10-01

    Title 1 of the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978, Public Law (PL) 95-604, authorized the US Department of Energy (DOE) to perform remedial action at the inactive Naturita, Colorado, uranium processing site to reduce the potential health effects from the radioactive materials at the site and at vicinity properties associated with the site. Title 2 of the UMTRCA authorized the US Nuclear Regulatory Commission (NRC) or agreement state to regulate the operation and eventual reclamation of active uranium processing sites. The uranium mill tailings at the site were removed and reprocessed from 1977 to 1979. The contaminated areas include the former tailings area, the mill yard, the former ore storage area, and adjacent areas that were contaminated by uranium processing activities and wind and water erosion. The Naturita remedial action would result in the loss of 133 acres (ac) of contaminated soils at the processing site. If supplemental standards are approved by the NRC and the state of Colorado, approximately 112 ac of steeply sloped contaminated soils adjacent to the processing site would not be cleaned up. Cleanup of this contamination would have adverse environmental consequences and would be potentially hazardous to remedial action workers.

  3. The relationship of JNC and JCO in the uranium processing plant criticality accident

    International Nuclear Information System (INIS)

    Kanamori, Masashi; Yanagibashi, Katsumi; Okamoto, Naritoshi

    2002-12-01

    On September 30th 1999, the criticality accident occurred at JCO's uranium conversion building in Tokai. The accident occurred during reconversion from U 3 O 8 to uranium nitrate solution (UNH) with uranium enriched 18.8% and about 60 kgU. JCO contacted with JNC to supply UNH that is fuel material for the experimental fast breeder reactor 'JOYO'. JNC has contracted with JCO that had started nuclear fuel material processing business following a definite policy of Japanese government and developed SUMITOMO ADU PROCESS'. JNC made the first contract with JCO in 1985 and has made a contact every year. There had never been a problem in their products. JNC inspected products based on contract. JNC discharge our duty as customer inspecting products based on contract. As for safety control, JCO had taken licensing safety review and had been permitted to be 'a processing facility'. Therefore JNC understood that JCO produced following this license. 'The Uranium Processing Plant Criticality Accident Investigation' showed that JCO had been taking a different method from the permit and violating the license. However JNC had never been explained about that and JCO's operation procedures had never described about that. Therefore the Criticality Accident couldn't be avoided. This report describes the relationship of JNC and JCO in the uranium reconversion contract for JOYO, atomic development policy of Japanese government, process to the order and the contents of contract. (author)

  4. Environmental assessment of remedial action at the Naturita uranium processing site near Naturita, Colorado: Revision 5

    International Nuclear Information System (INIS)

    1994-10-01

    Title 1 of the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978, Public Law (PL) 95-604, authorized the US Department of Energy (DOE) to perform remedial action at the inactive Naturita, Colorado, uranium processing site to reduce the potential health effects from the radioactive materials at the site and at vicinity properties associated with the site. Title 2 of the UMTRCA authorized the US Nuclear Regulatory Commission (NRC) or agreement state to regulate the operation and eventual reclamation of active uranium processing sites. The uranium mill tailings at the site were removed and reprocessed from 1977 to 1979. The contaminated areas include the former tailings area, the mill yard, the former ore storage area, and adjacent areas that were contaminated by uranium processing activities and wind and water erosion. The Naturita remedial action would result in the loss of 133 acres (ac) of contaminated soils at the processing site. If supplemental standards are approved by the NRC and the state of Colorado, approximately 112 ac of steeply sloped contaminated soils adjacent to the processing site would not be cleaned up. Cleanup of this contamination would have adverse environmental consequences and would be potentially hazardous to remedial action workers

  5. Treatment of liquid effluent from uranium mines and mills. Report of a co-ordinated research project 1996-2000

    International Nuclear Information System (INIS)

    2004-10-01

    Treatment and control of liquid effluents produced during uranium mining and milling operations is an integral part of environmental project management. Research has continued to add to the large body of science that has been built up around the treatment of radioactive and non-radioactive effluents to minimize their long-term environmental impact. The objective of the meetings on which this publication is based was to exchange information on active effluent treatment technologies that have application during operations and passive treatment techniques such as constructed wetlands and use of micro-organisms that are applicable during project reclamation and long-term care and maintenance. Papers describe effluent treatment case histories from active uranium mining and processing operations as well as effluent treatment research on both active and passive systems that have potential application under a wide range of operating and post-operational conditions including new information on high-density sludge from effluent neutralization (Australia), aerated manganese hydroxide for removal of radium (China), nanofiltration and macropore resins to treat mine water (Australia and China), in situ microbial treatment and permeable reactive walls for treatment of contaminated groundwater (Germany), construction of wetlands to treat mine water runoff (Australia and Germany), biogenic granules to remove 226 Ra from mill effluent (India), self-remediation of acidic in situ leach aquifers (Kazakhstan) and sorption characteristics of soil for self-remediation of contaminated groundwater (Hungary). These and other topics presented in this publication will be of interest to technical personnel who deal with day-to-day practical aspects of liquid effluent control and treatment at uranium production facilities worldwide

  6. Development and technical implementation of the separation nozzle process for enrichment of uranium 235

    International Nuclear Information System (INIS)

    Syllus Martins Pinto, C.; Voelcker, H.; Becker, E.W.

    1977-12-01

    The separation nozzle process for the enrichment of uranium-235 has been developed at the Karlsruhe Nuclear Research Center as an alternative to the gaseous diffusion and centrifuge process. The separation of uranium isotopes is achieved by the deflection of a jet of uranium hexafluoride mixed with hydrogen. Since 1970, the German company of STEAG, has been involved in the technological development and commercial implementation of the nozzle process. In 1975, the Brazilian company of NUCLEBRAS, and the German company of Interatom, joined the effort. The primary objective of the common activity is the construction of a separation nozzle demonstration plant with an annual capacity of about 200 000 SWU and the development of components of a commercial plant. The paper covers the most important steps in the development and the technical implementation of the process. (orig.) [de

  7. Uranium mining and milling

    International Nuclear Information System (INIS)

    Floeter, W.

    1976-01-01

    In this report uranium mining and milling are reviewed. The fuel cycle, different types of uranium geological deposits, blending of ores, open cast and underground mining, the mining cost and radiation protection in mines are treated in the first part of this report. In the second part, the milling of uranium ores is treated, including process technology, acid and alkaline leaching, process design for physical and chemical treatment of the ores, and the cost. Each chapter is clarified by added figures, diagrams, tables, and flowsheets. (HK) [de

  8. Analysis of lagoon sludge characteristics for choice of treatment process

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. H.; Hwang, D. S.; Choi, Y. D.; Lee, K. I.; Hwang, S. T.; Jung, K. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    The Korea Atomic Energy Research Institute has launched a decommissioning program of uranium conversion plant. One of the important tasks in the decommissioning program is the treatment of the sludge, which was generated during operation and stored in the two ponds of the lagoon. The treatment requires the volume reduction of lagoon sludges for the low cost of the program and the conversion of the chemical forms, including uranium, for the acceptance at the final disposal site. The physical properties, such as densities, were measured and chemical compositions and radiological properties were analyzed. The denitration was a candidate process which would satisfy the requirements for sludge treatment, and the characteristics of thermal decomposition and dissolution with water were analyzed. The main compounds of the sludge were ammonium and sodium nitrate from conversion plant and calcium nitrate, calcium carbonate from Ca precipitation and impurities of the yellow cake. The content of uranium, thorium and Ra-226 was high in pond-1 and low in pond-2 because those were removed during Ca precipitation. On the base of the characteristics of the sludge and available technologies, reviewed in this study and being developed in Korea Atomic Energy Research Institute, two processes were proposed and evaluated in points of the expected technological difficulties. And the cost for treatment of sludges are estimated for both processes. 79 refs., 44 figs., 37 tabs. (Author)

  9. Analysis of Uranium and Thorium in Radioactive Wastes from Nuclear Fuel Cycle Process

    International Nuclear Information System (INIS)

    Gunandjar

    2008-01-01

    The assessment of analysis method for uranium and thorium in radioactive wastes generated from nuclear fuel cycle process have been carried out. The uranium and thorium analysis methods in the assessment are consist of Titrimetry, UV-VIS Spectrophotometry, Fluorimetry, HPLC, Polarography, Emission Spectrograph, XRF, AAS, Alpha Spectrometry and Mass Spectrometry methods. From the assessment can be concluded that the analysis methods of uranium and thorium content in radioactive waste for low concentration level using UV-VIS Spectrometry is better than Titrimetry method. While for very low concentration level in part per billion (ppb) can be used by Neutron Activation Analysis (NAA), Alpha Spectrometry and Mass Spectrometry. Laser Fluorimetry is the best method of uranium analysis for very low concentration level. Alpha Spectrometry and ICP-MS (Inductively Coupled Plasma Mass Spectrometry) methods for isotopic analysis are favourable in the precision and accuracy aspects. Comparison of the ICP-MS and Alpha Spectrometry methods shows that the both of methods have capability to determining of uranium and thorium isotopes content in the waste samples with results comparable very well, but the time of its analysis using ICP-MS method is faster than the Alpha Spectrometry, and also the cost of analysis for ICP-MS method is cheaper. NAA method can also be used to analyze the uranium and thorium isotopes, but this method needs the reactor facility and also the time of its analysis is very long. (author)

  10. Gravity data processing and research in potential evaluation of uranium resource in China

    International Nuclear Information System (INIS)

    Liu Hu; Zhao Dan; Ke Dan; Li Bihong; Han Shaoyang

    2012-01-01

    Through data processing, anomaly extraction, geologic structure deduction from gravity in 39 uranium metallogenic zones and 29 prediction areas, the predicting factors such as tectonic units, faults, scope and depth of rocks, scope of basins and strata structure were provided for the evaluation of uranium resources potential. Gravity field features of uranium metallogenic environment were summarized for hydrothermal type uranium deposits (granite, volcanic and carbonate-siliceous-argillaceous type) as regional gravity transition from high to the low field or the region near the low field, and the key metallogenic factors as granite rocks and volcanic basins in the low gravity field. It was found that Large-scale sandstone type uranium mineralization basins are located in the high regional gravity field, provenance areas are in the low field, and the edge and inner uplift areas usually located in the high field of the residual gravity. Faults related to different type uranium mineralization occur as the gradient zones, boundaries, a string of bead anomalies and striped gravity anomalies in the gravity field. (authors)

  11. Uranium hexafluoride: Safe handling, processing, and transporting: Conference proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Strunk, W.D.; Thornton, S.G. (eds.)

    1988-01-01

    This conference seeks to provide a forum for the exchange of information and ideas of the safety aspects and technical issue related to the handling of uranium hexafluoride. By allowing operators, engineers, scientists, managers, educators, and others to meet and share experiences of mutual concern, the conference is also intended to provide the participants with a more complete knowledge of technical and operational issues. The topics for the papers in the proceedings are widely varied and include the results of chemical, metallurgical, mechanical, thermal, and analytical investigations, as well as the developed philosophies of operational, managerial, and regulatory guidelines. Papers have been entered individually into EDB and ERA. (LTN)

  12. Uranium hexafluoride: Safe handling, processing, and transporting: Conference proceedings

    International Nuclear Information System (INIS)

    Strunk, W.D.; Thornton, S.G.

    1988-01-01

    This conference seeks to provide a forum for the exchange of information and ideas of the safety aspects and technical issue related to the handling of uranium hexafluoride. By allowing operators, engineers, scientists, managers, educators, and others to meet and share experiences of mutual concern, the conference is also intended to provide the participants with a more complete knowledge of technical and operational issues. The topics for the papers in the proceedings are widely varied and include the results of chemical, metallurgical, mechanical, thermal, and analytical investigations, as well as the developed philosophies of operational, managerial, and regulatory guidelines. Papers have been entered individually into EDB and ERA

  13. Development of an on-line analyzer for organic phase uranium concentration in extraction process

    International Nuclear Information System (INIS)

    Dong Yanwu; Song Yufen; Zhu Yaokun; Cong Peiyuan; Cui Songru

    1998-10-01

    The working principle, constitution, performance of an on-line analyzer and the development characteristic of immersion sonde, data processing system and examination standard are reported. The performance of this instrument is reliable. For identical sample, the signal fluctuation in continuous monitoring for four months is less than +-1%. According to required measurement range by choosing appropriate length of sample cell the precision of measurement is better than 1% at uranium concentration 100 g/L. The detection limit is (50 +- 10) mg/L. The uranium concentration in process stream can be automatically displayed and printed out in real time and 4∼20 mA current signal being proportional to the uranium concentration can be presented. So the continuous control and computer management for the extraction process can be achieved

  14. The Canadian Nuclear Safety Commission regulatory process for decommissioning a uranium mining facility

    International Nuclear Information System (INIS)

    Scissons, K.; Schryer, D.M.; Goulden, W.; Natomagan, C.

    2002-01-01

    The Canadian Nuclear Safety Commission (CNSC) regulates uranium mining in Canada. The CNSC regulatory process requires that a licence applicant plan for and commit to future decommissioning before irrevocable decisions are made, and throughout the life of a uranium mine. These requirements include conceptual decommissioning plans and the provision of financial assurances to ensure the availability of funds for decommissioning activities. When an application for decommissioning is submitted to the CNSC, an environmental assessment is required prior to initiating the licensing process. A case study is presented for COGEMA Resources Inc. (COGEMA), who is entering the decommissioning phase with the CNSC for the Cluff Lake uranium mine. As part of the licensing process, CNSC multidisciplinary staff assesses the decommissioning plan, associated costs, and the environmental assessment. When the CNSC is satisfied that all of its requirements are met, a decommissioning licence may be issued. (author)

  15. Surveying and assessing the hazards associated with the processing of uranium

    International Nuclear Information System (INIS)

    Kruger, J.

    1980-01-01

    The control of uranium during the milling process has not received extensive attention. The results of several surveys of surface contamination, airborne contamination and external radiation made at South African processing facilities are presented and compared with derived norms for permissible exposure to uranium dust. The routine urine sampling results are used as an indicator of personnel exposures. Results of sampling identify the main sources of airborne activity and indicate the contribution of general surface contamination levels to airborne levels. The use of surface contamination levels together with frequent air sampling for assessing the environmental conditions is illustrated. It is concluded that infrequent grab air sampling alone is not adequate for assessing the hazards during uranium processing. Detailed surveys are required and proper area and personnel access control are indicated. (H.K.)

  16. Treatment of reactive process wastewater with high-level ammonia by blow-off method

    International Nuclear Information System (INIS)

    Chen Xiaotong; Quan Ying; Wang Yang; Fu Genna; Liu Bing; Tang Yaping

    2012-01-01

    The ceramic UO 2 kernels for nuclear fuel elements of high temperature gas cooled reactors were prepared through sol-gel process with uranyl nitrate, which produces process wastewater containing high-level ammonia and uranium. The blow-off method on a bench scale was investigated to remove ammonia from reactive wastewater. Under the optimized operating conditions, the ammonia can be removed by more than 95%, with little reactive uranium distilled. The effects of pH, heating temperature and stripping time were studied. Static tests with ion-exchange resin indicate that ammonia removal treatment increases uranium accumulation in anion exchange resin. (authors)

  17. Fate of soluble uranium in the I{sub 2}/KI leaching process for mercury removal

    Energy Technology Data Exchange (ETDEWEB)

    Bostick, W.D.; Davis, W.H.; Jarabek, R.J. [East Tennessee Technology Park, Oak Ridge, TN (United States). Materials and Chemistry Lab.

    1997-09-01

    General Electric Corporation has developed an extraction and recovery system for mercury, based upon the use of iodine (oxidant) and iodide ion (complexing agent). This system has been proposed for application to select mercury-contaminated mixed waste (i.e., waste containing radionuclides as well as other hazardous constituents), which have been generated by historic activities in support of US Department of Energy (DOE) missions. This system is compared to a system utilizing hypochlorite and chloride ions for removal of mercury and uranium from a sample of authentic mixed waste sludge. Relative to the hypochlorite (bleach) system, the iodine system mobilized more mercury and less uranium from the sludge. An engineering flowsheet has been developed to treat spent iodine-containing extraction medium, allowing the system to be recycled. The fate of soluble uranium in this series of treatment unit operations was monitored by tracing isotopically-enriched uranyl ion into simulated spent extraction medium. Treatment with use of elemental iron is shown to remove > 85% of the traced uranium while concurrently reducing excess iodine to the iodide ion. The next unit operation, adjustment of the solution pH to a value near 12 by the addition of lime slurry to form a metal-laden sludge phase (an operation referred to as lime-softening), removed an additional 57% of soluble uranium activity, for an over-all removal efficiency of {approximately} 96%. However, the precipitated solids did not settle well, and some iodide reagent is held up in the wet filtercake.

  18. Fate of soluble uranium in the I2/KI leaching process for mercury removal

    International Nuclear Information System (INIS)

    Bostick, W.D.; Davis, W.H.; Jarabek, R.J.

    1997-09-01

    General Electric Corporation has developed an extraction and recovery system for mercury, based upon the use of iodine (oxidant) and iodide ion (complexing agent). This system has been proposed for application to select mercury-contaminated mixed waste (i.e., waste containing radionuclides as well as other hazardous constituents), which have been generated by historic activities in support of US Department of Energy (DOE) missions. This system is compared to a system utilizing hypochlorite and chloride ions for removal of mercury and uranium from a sample of authentic mixed waste sludge. Relative to the hypochlorite (bleach) system, the iodine system mobilized more mercury and less uranium from the sludge. An engineering flowsheet has been developed to treat spent iodine-containing extraction medium, allowing the system to be recycled. The fate of soluble uranium in this series of treatment unit operations was monitored by tracing isotopically-enriched uranyl ion into simulated spent extraction medium. Treatment with use of elemental iron is shown to remove > 85% of the traced uranium while concurrently reducing excess iodine to the iodide ion. The next unit operation, adjustment of the solution pH to a value near 12 by the addition of lime slurry to form a metal-laden sludge phase (an operation referred to as lime-softening), removed an additional 57% of soluble uranium activity, for an over-all removal efficiency of ∼ 96%. However, the precipitated solids did not settle well, and some iodide reagent is held up in the wet filtercake

  19. Synthesis of uranium and thorium dioxides by Complex Sol-Gel Processes (CSGP). Synthesis of uranium oxides by Complex Sol-Gel Processes (CSGP)

    International Nuclear Information System (INIS)

    Deptula, A.; Brykala, M.; Lada, W.; Olczak, T.; Wawszczak, D.; Chmielewski, A.G.; Modolo, G.; Daniels, H.

    2010-01-01

    In the Institute of Nuclear Chemistry and Technology (INCT), a new method of synthesis of uranium and thorium dioxides by original variant of sol-gel method - Complex Sol-Gel Process (CSGP), has been elaborated. The main modification step is the formation of nitrate-ascorbate sols from components alkalized by aqueous ammonia. Those sols were gelled into: - irregularly agglomerates by evaporation of water; - medium sized microspheres (diameter <150) by IChTJ variant of sol-gel processes by water extraction from drops of emulsion sols in 2-ethylhexanol-1 by this solvent. Uranium dioxide was obtained by a reduction of gels with hydrogen at temperatures >700 deg. C, while thorium dioxide by a simple calcination in the air atmosphere. (authors)

  20. Recent trends in research and development work on the processing of uranium ore in South Africa

    International Nuclear Information System (INIS)

    James, H.E.

    1976-07-01

    The rapid increases in the price of gold and uranium in recent years have coincided with an unprecedented increase in working costs at South African gold mines. A re-examination of the existing flowsheets for the recovery of uranium, gold, and pyrite from Witwatersrand ores, in the light of these economic trends, has resulted in the identification of a number of profitable areas for research and development. The main topics under investigation in South Africa in the processing of uranium ore are the use of physical methods of concentration such as flotation, gravity concentration, and wet high-intensity magnetic separation; the wider adoption of the 'reverse leach', in which prior acid leaching for uranium improves the subsequent extraction of gold; the use of higher leaching temperatures and higher concentrations of ferric ion in the leach to increase the percentage of uranium extracted, including the production of ferric ion from recycled solutions; the application of pressure leaching to the recovery of uranium from low-grade ores and concentrates; the development of a continuous ion-exchange contactor capable of handling dilute slurries, so that simpler and cheaper techniques of solid-liquid separation can be used instead of the expensive filtration and clarification steps, and the improvement of instrumentation for the control of additions of sulphuric acid and manganese dioxide to the leach. A brief description is given of the essential features of the new or improved processing techniques under development that hold promise of full-scale application at existing or future uranium plants [af

  1. Recent trends in research and development work on the processing of uranium ore in South Africa

    International Nuclear Information System (INIS)

    James, H.E.

    1976-01-01

    The rapid increases in the price of gold and uranium in recent years have coincided with an unprecedented increase in working costs at South African gold mines. A re-examination of the existing flowsheets for the recovery of uranium, gold and pyrite from Witwatersrand ores, in the light of these economic trends, has resulted in the identification of a number of profitable areas for research and development. The main topics under investigation in South Africa in the processing of uranium ore are the use of physical methods of concentration such as flotation, gravity concentration and wet high-intensity magnetic separation; the wider adoption of the 'reverse leach', in which prior acid leaching for uranium improves the subsequent extraction of gold; the use of higher leaching temperatures and higher concentrations of ferric ion in the leach to increase the percentage of uranium extracted, including the production of ferric ion from recycled solutions; the application of pressure leaching to the recovery of uranium from low-grade ores and concentrates; the development of a continuous ion-exchange contactor capable of handling dilute slurries, so that simpler and cheaper techniques of solid/liquid separation can be used instead of the expensive filtration and clarification steps, and the improvement of instrumentation for the control of additions of sulphuric acid and manganese dioxide to the leach. A brief description is given of the essential features of the new or improved processing techniques under development that hold promise of full-scale application at existing or future uranium plants

  2. A new process for the fractionation of uranium; Un nuevo procedimiento para el fraccionamiento de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Costas, E.; Baselga, B.; Tarin, F.

    2015-07-01

    We propose a new biological process for uranium isotopic fractionation based on Chlamydomonas cf. fonticola (microalgae) isolated from a pond extremely contaminated by uranium (. 25 ppm) from the ENUSA mine in Saelices (Salamanca, Spain) and genetically improved. The metabolic activity of this genetically improved ChlSPGI strain allows recover 115 mg of U per gram of micoralgal biomass in a short time (because this strain complete their cell cycle in . 24 hours). During this process ChlSPGI microalgae selectively captures {sup 2}35U conducting an isotopic enrichment of {sup 2}35U ({sup 2}35U δ = + 3,983%). (Author)

  3. Iron behaviour in the process of stratum-infiltration uranium ore formation

    International Nuclear Information System (INIS)

    Shmariovich, E.M.; Golubev, V.S.

    1980-01-01

    Investigated has been the behaviour of iron in the process of stratum infiltration uranium mineralization. Iron is partially avacuated from the forward part of the stratum oxidation zone during the development of infiltration uranium mineralization in pyritiferous rocks. This phenomenon is characterized quantitatively and described on the basis of equations of physical chemistry and dynamics of geochemical processes. Local regions of epigenetic ferruginization caused by opposite diffusion of iron and its precipitation in oxygenous conditions often occur at the sections of sharp moderation of limonitization zone advance. Formation of similar ferruginous margins takes place in a very short geological period (less than thousand years)

  4. Semitechnical studies of uranium recovery from wet process phosphoric acid by liquid-liquid-extraction method

    International Nuclear Information System (INIS)

    Poczynajlo, A.; Wlodarski, R.; Giers, M.

    1987-01-01

    A semitechnical installation for uranium recovery from wet process phosphoric acid has been built. The installation is based on technological process comprising 2 extraction cycles, the first with a mixture of mono- and dinonylphenylphosphoric acids (NPPA) and the second with a synergic mixture of di-/2-ethylhexyl/-phosphoric acid (D2EHPA) and trioctylphosphine oxide (TOPO). The installation was set going and the studies on the concentration distributions of uranium and other components of phosphoric acid have been performed for all technological circuits. 23 refs., 15 figs., 3 tabs. (author)

  5. ELECTRODEPOSITION OF NICKEL ON URANIUM

    Science.gov (United States)

    Gray, A.G.

    1958-08-26

    A method is described for preparing uranium objects prior to nickel electroplating. The process consiats in treating the surface of the uranium with molten ferric chloride hexahydrate, at a slightiy elevated temperature. This treatment etches the metal surface providing a structure suitable for the application of adherent electrodeposits and at the same time plates the surface with a thin protective film of iron.

  6. Dissolution of metallic uranium and its alloys. Part 1. Review of analytical and process-scale metallic uranium dissolution

    International Nuclear Information System (INIS)

    Laue, C.A.; Gates-Anderson, D.; Fitch, T.E.

    2004-01-01

    This review focuses on dissolution/reaction systems capable of treating uranium metal waste to remove its pyrophoric properties. The primary emphasis is the review of literature describing analytical and production-scale dissolution methods applied to either uranium metal or uranium alloys. A brief summary of uranium's corrosion behavior is included since the corrosion resistance of metals and alloys affects their dissolution behavior. Based on this review, dissolution systems were recommended for subsequent screening studies designed to identify the best system to treat depleted uranium metal wastes at Lawrence Livermore National Laboratory (LLNL). (author)

  7. Effects of barium chlorine treatment of uranium ore on 222Rn emanation and 226Ra leachability from mill tailings

    International Nuclear Information System (INIS)

    Ibrahim, S.A.; Church, S.L.; Whicker, F.W.

    1985-01-01

    The purpose of this laboratory study was to investigate the effectiveness of barium chloride treatment of uranium ore on 222 Rn emanation from mill tailings, 226 Ra level in waste-water, and the leachability of radium from tailings. It has been shown that barium sulfate is an excellent carrier for radium and that barium sulfate crystals have high retention capacity for radon gas produced by radium trapped within the lattice. Ground uranium ore from a mine in Wyoming was mixed with water to form a 1:1 ratio before barium and potassium chlorides were added at concentrations of 0, 10, 25, 50, and 100 mg per liter of slurry. The ore was then subjected to a simulated mill process using sulfuric acid leaching. The liquid representing tailings pond water was separated and analyzed for 226 Ra and the solid fraction, representing mill tailings, was tested for radon emanation and the leachability of radium by deionized water. This study suggests that barium treatment of uranium ore prior to sulfuric acid leaching could be effective in reducing radon emanation from tailings and also in reducing the 226 Ra concentration of waste-water. Leachability of radium from treated tailings was markedly reduced

  8. Preparation, heat treatment, and mechanical properties of the uranium-5 weight percent chromium eutectic alloy

    International Nuclear Information System (INIS)

    Townsend, A.B.

    1980-10-01

    The eutectic alloy of uranium-5 wt % chromium (U-5Cr) was prepared from high-purity materials and cast into 1-in.-thick ingots. This material was given several simple heat treatments, the mechanical properties of these heat-treated samples were determined; and the microstructure was examined. Some data on the melting point and transformation temperatures were obtained

  9. Study On The Choice Of Leaching System For Thanh My, Quang Nam Province Uranium Ores Treatment

    International Nuclear Information System (INIS)

    Than Van Lien; Nguyen Dinh Van; Tran The Dinh

    2011-01-01

    In order to implement the plan of peaceful uses of atomic energy, the Radioactive and Rare Earth Geology Division have been carried out the uranium ores exploitation project in Thanh My area of Quang Nam province since 2010. The treatment uranium ores samples is one of works of this project. In order to preparing for uranium ores samples treatment, the Institute for Technology of Radioactive and Rare Elements have been studied and have chosen the heap leaching method for Thanh My uranium ore treatment. The ore, which contained less than 0.07% U, was crushed to -1 cm before being placed in the heap. The acid consumption for this heap leach operation was approximately ranged 40 kg - 45 kg of H 2 SO 4 per tonne of ore, and oxidant 4 kg of MnO 2 per tonne of ore. The entire treatment cycle required 20-25 days, the recovery exceeded 80%, the leached tails contained less than 0.01% U. The experimental results were comparable with those obtained in the field scale heap leaching in the world. (author)

  10. Bio sorption of uranium with Sargassum filipendula: use in treatment of effluents of laboratories

    International Nuclear Information System (INIS)

    Rodrigues Silva, J.I.; Melo Ferreira, A.C.; Costa, A.C.A. da

    2008-01-01

    International and National Standards establish methodologies for the management of radioactive waste in order to comply with radiological protection principles. Thus, it is necessary to find alternatives with both low cost and effective results. This work studied the use of brown algae Sargassum filipendula in its ability to remove uranium in waste generated in the environmental analysis laboratories of the Institute of Radiation Protection and Dosimetry. At first, the kinetics of bio sorption was studied. This experiment was conducted on a batch at concentrations of 1 mg/L and 100 mg/L. Mathematical models of the first and second order were used to fit the experimental results. In evaluating the maximum removal capacity by marine biomass on a batch, different uranium concentrations were analyzed, and iso terms of bio sorption were plotted. The experimental results have been adjusted by Langmuir and Freundlich models. The Freundlich model presented the best correlation coefficients (0.99 and 0.94) for studies with an hour and three hours of contact, respectively. In order to determine the best conditions for removal of uranium using the Sargassum filipendula, it was necessary to hold experiments in a continuous flow. A study on the critical height of bed depth was carried out by filling a column with different masses of seaweed. It was obtained a lower outlet concentration o f uranium (0.07 mg/L) in 40 cm bed depth. This best height of bed was applied to the waste treatment of SEANA laboratories. It was monitored the increase of retention in biomass for a known quantity of uranium. The results showed an excellent uptake of uranium (1.25 mg U/g of biomass) even in the presence of other metals and reagents. Decontamination of the effluent for uranium reached values below those set by CONAMA for water classes I and II, making it possible to reuse the water. (author)

  11. Optimization of the recycling process of precipitation barren solution in a uranium mine

    International Nuclear Information System (INIS)

    Long Qing; Yu Suqin; Zhao Wucheng; Han Wei; Zhang Hui; Chen Shuangxi

    2014-01-01

    Alkaline leaching process was adopted to recover uranium from ores in a uranium mine, and high concentration uranium solution, which would be later used in precipitation, was obtained after ion-exchange and elution steps. The eluting agent consisted of NaCl and NaHCO 3 . Though precipitation barren solution contained as high as 80 g/L Na 2 CO 3 , it still can not be recycled due to presence of high Cl - concentration So, both elution and precipitation processes were optimized in order to control the Cl - concentration in the precipitation barren solution to the recyclable concentration range. Because the precipitation barren solution can be recycled by optimization, the agent consumption was lowered and the discharge of waste water was reduced. (authors)

  12. Remedial action standards for inactive uranium processing sites (40 cfr 192). Draft environmental impact statement

    International Nuclear Information System (INIS)

    1980-12-01

    The Environmental Protection Agency is proposing standards for disposing of uranium mill tailings from inactive processing sites and for cleaning up contaminated open land and buildings. These standards were developed pursuant to the Uranium Mill Tailings Radiation Control Act of 1978 (Public Law 95-604). This Act requires EPA to promulgate standards to protect the environment and public health and safety from radioactive and nonradioactive hazards posed by uranium mill tailings at designated inactive processing sites. The Draft Environmental Impact Statement examines health, technical, cost, and other factors relevant to determining standards. The proposed standards for disposal of the tailings piles cover radon emissions from the tailings to the air, protection of surface and ground water from radioactive and nonradioactive contaminants, and the length of time the disposal system should provide a reasonable expectation of meeting these standards. The proposed cleanup standards limit indoor radon decay product concentrations and gamma radiation levels and the residual radium concentration of contaminated land after cleanup

  13. No fluorinated compounds in the uranium conversion process: risk analysis and proposition of pictograms

    International Nuclear Information System (INIS)

    Jeronimo, Adroaldo Clovis; Oliveira, Wagner dos Santos

    2012-01-01

    The plants comprising the chemical conversion of uranium, which are part of the nuclear fuel cycle, present some risks, among others, because are associated with the non-fluorinated compounds handled in these processes. This study is the analysis of the risks associated with these compounds, i e, the non-fluorinated reactants and products, handled in different chemical processing plants, which include the production of uranium hexafluoride, while emphasizing the responsibilities and actions that fit to the chemical engineer with regard to minimizing risks during the various stages. The work is based on the experience gained during the development and mastery of the technology of production of uranium hexafluoride, the IPEN/ CNEN-SP, during the '80s, with the support of COPESP -Navy of Brazil. (author)

  14. Chemical process for recovery of uranium values contained in phosphoric mineral lixivia

    International Nuclear Information System (INIS)

    Conceicao, E.L.H. da; Awwal, M.A.; Coelho, S. V.

    1980-01-01

    A recovery process of uranium values from phosporic mineral lixivia for obtaining uranio oxide concentrate adjusted to specifications of purity for its commercialization the process consists of the adjustment of electromotive force of lixiviem to suitable values for uranium extraction, extraction with organic solvent containing phosphoric acid ester and oxidant reextraction from this solvent with phosphoric acid solution, suggesting a new solvent extraction containing synergetic mixture of di-2-ethyl hexyl phosphoric acid and tri-octyl phosphine, leaching this solvent with water and re-extraction/precipitation with ammonium carbonate solution, resulting in the formation of uranyl tricarbonate and ammonium, that by drying and calcination gives the uranium oxide with purity degree for commercialization. (M.C.K.) [pt

  15. Development of an improved two-cycle process for recovering uranium from wet-process phosphoric acid

    International Nuclear Information System (INIS)

    Chen, H.M.; Chen, H.J.; Tsai, Y.M.; Lee, T.W.; Ting, G.

    1987-01-01

    An improved two-cycle separation process for the recovery of uranium from wet-process phosphoric acid by extraction with bis(2-ethylhexyl)phosphoric acid (D2EHPA) plus dibutyl butylphosphonate (DBBP) in kerosene has been developed and demonstrated successfully in bench-scale, continuous mixer-settler tests. The sulfuric acid and water scrubbing steps for the recycled extraction in the second cycle solve the problems of the contamination and dilution of the phosphoric acid by the ammonium ion and water and also avoid the formation of undesirable phosphatic precipitates during the subsequent extraction of uranium by recycled organic extractant

  16. Processing device for gaseous waste containing uranium hexafluoride

    International Nuclear Information System (INIS)

    Hirosawa, Jun-ichi.

    1985-01-01

    Purpose: To enable to detect the inactivation of chemical traps thereby reduce the amount of adsorbents. Constitution: Two chemical traps are disposed in series and γ-detector for detecting γ-rays generated from U-235 in hexafluoride is disposed to the outer surface of a pipeway connecting these two chemical traps. Further, chemical traps are adapted to be swtichable between the first stage and the second stage thereof by the ON-OFF operation of a valve. Then, by determining γ-rays from U-235 at the pipeway downstream from the gas exit of the chemical traps, the counted value for the γ-rays is substantially at the background level so long as the chemical trap has an adsorbing performance for uranium hexafluoride. Then, since the γ-ray counted value is increased at the step upon inactivation of the chemical trap, the inactivation of the trap can be detected. (Yoshino, Y.)

  17. Rirang Uranium Ore Processing System Design: Agitated Digester

    International Nuclear Information System (INIS)

    Erni, R.A.; Susilaningtyas

    1996-01-01

    A closed tank digester equipped with a pitched blades turbine agitator has been designed to facilities Rirang uranium ore dissolution using concentrated sulphuric acid at high temperature. The digester was designed to accommodate the digestion of 6 kg of-65 mesh ore at 200 o C, acid resistant material (SS-3 16). It has the dimension of 33 cm high, 22 cm diameter, and elliptical bottom and height of 4 cm. Moreover, the dimension of the 4 blades agitator is as follows: 8 cm long, 1,6 cm blades width. The distance between the blades and digester required 0, 007 Hp for a 500 rpm agitation speed and + 24. 103 kcal energy equipment for heating. Digestion experiment using the agitated digester yielded data that are in good agreement with laboratory scale experiment

  18. Uranium recovery from wet-process phosphoric acid with octylphenyl acid phosphate. Progress report

    International Nuclear Information System (INIS)

    Arnold, W.D.; McKamey, D.R.; Baes, C.F.

    1980-01-01

    Studies were continued of a process for recovering uranium from wet-process phosphoric acid with octylphenyl acid phosphate (OPAP), a mixture of mono- and dioctylphenyl phosphoric acids. The mixture contained at least nine impurities, the principal one being octyl phenol, and also material that readily hydrolyzed to octyl phenol and orthophosphoric acid. The combination of mono- and dioctylphenyl phosphoric acids was the principal uranium extractant, but some of the impurities also extracted uranium. Hydrolysis of the extractant had little effect on uranium extraction, as did the presence of moderate concentrations of octyl phenol and trioctylphenyl phosphate. Diluent choice among refined kerosenes, naphthenic mixtures, and paraffinic hydrocarbons also had little effect on uranium extraction, but extraction was much lower when an aromatic diluent was used. Purified OPAP fractions were sparingly soluble in aliphatic hydrocarbon diluents. The solubility was increased by the presence of impurities such as octyl phenol, and by the addition of water or an acidic solution to the extractant-diluent mixture. In continuous stability tests, extractant loss by distribution to the aqueous phase was much less to wet-process phosphoric acid than to reagent grade acid. Uranium recovery from wet-process acid decreased steadily because of the combined effects of extractant poisoning and precipitation of the extractant as a complex with ferric iron. Unaccountable losses of organic phase volume occurred in the continuous tests. While attempts to recover the lost organic phase were unsuccessful, the test results indicate it was not lost by entrainment or dissolution in the phosphoric acid solutions. 21 figures, 8 tables

  19. A review of experiment data processing method for uranium mining and metallurgy in BRICEM

    International Nuclear Information System (INIS)

    Ye Guoqiang; Lu Kehong; Wang Congying

    1997-01-01

    The authors investigates the methods of experiment data processing in Beijing Research Institute of Chemical Engineering and Metallurgy (BRICEM). It turns out that error analysis method is used to process experiment data, single-factor transformation and orthogonal test design method are adopted for arranging test, and regression analysis and mathematical process simulation are applied to process mathematical model for uranium mining and metallurgy. The methods above-mentioned lay a foundation for the utilization of mathematical statistics in our subject

  20. Treatment of Lagoon sludge waste generated from Uranium Conversion Plant

    International Nuclear Information System (INIS)

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

    2003-01-01

    This study investigated the dissolution property of nitrate salts in the desalination process by water and the drying property of residual solid after separating nitrates in a series of processes for the sludge treatment. Desalination was carried out with the adding ratio of water and drying property was analyzed by TG/DTA, FTIR, and XRD. Nitrate salts involved in the sludge were separated over 97 % at the water adding ratio of 2.5. But a small quantity of calcium and sodium nitrate remained in the residue. These were decomposed over 600 deg. C while calcium carbonate, which was a main compound of residual solid, was decomposed into calcium oxide over 750 deg. C. The residual solid has to be decomposed over 800 deg. C to converse uranyl nitrate of six values into the stable U 3 O 8 of four values. As a result of removing the nitrates at the adding ratio of 2.5 and drying the residue over 900 deg. C, volume of the sludge waste decreased over 80 %. (authors)

  1. Biomineral processing of high apatite containing low-grade indian uranium ore

    Energy Technology Data Exchange (ETDEWEB)

    Abhilash; Mehta, K.D.; Pandey, B.D., E-mail: biometnml@gmail.com [National Metallurgical Laboratory (CSIR), Jamshedpur (India); Ray, L. [Jadavpur Univ., FTBE Dept., Kolkata (India); Tamrakar, P.K. [Uranium Corp. of India Limited, CR& D Dept., Jaduguda (India)

    2010-07-01

    Microbial species isolated from source mine water, primarily an enriched culture of Acidithiobacillus ferrooxidans was employed for bio-leaching of uranium from a low-grade apatite rich uranium ore of Narwapahar Mines, India while varying pH, pulp density (PD), particle size, etc. The ore (0.047% U{sub 3}O{sub 8}), though of Singhbhum area (richest deposit of uranium ores in India), due to presence of some refractory minerals and high apatite (5%) causes a maximum 78% recovery through conventional processing. Bioleaching experiments were carried out by varying pH at 35{sup o}C using 20%(w/v) PD and <76μm size particles resulting in 83.5% and 78% uranium bio-recovery at 1.7 and 2.0 pH in 40 days as against maximum recovery of 46% and 41% metal in control experiments respectively. Finer size (<45μm) ore fractions exhibited higher uranium dissolution (96%) in 40 days at 10% (w/v) pulp density (PD), 1.7 pH and 35{sup o}C. On increasing the pulp density from 10% to 20% under the same conditions, the biorecovery of uranium fell down from 96% to 82%. The higher uranium dissolution during bioleaching at 1.7 pH with the fine size particles (<45μm) can be correlated with increase in redox potential from 598 mV to 708 mV and the corresponding variation of Fe(III) ion concentration in 40 days. (author)

  2. Biomineral processing of high apatite containing low-grade indian uranium ore

    International Nuclear Information System (INIS)

    Abhilash; Mehta, K.D.; Pandey, B.D.; Ray, L.; Tamrakar, P.K.

    2010-01-01

    Microbial species isolated from source mine water, primarily an enriched culture of Acidithiobacillus ferrooxidans was employed for bio-leaching of uranium from a low-grade apatite rich uranium ore of Narwapahar Mines, India while varying pH, pulp density (PD), particle size, etc. The ore (0.047% U_3O_8), though of Singhbhum area (richest deposit of uranium ores in India), due to presence of some refractory minerals and high apatite (5%) causes a maximum 78% recovery through conventional processing. Bioleaching experiments were carried out by varying pH at 35"oC using 20%(w/v) PD and <76μm size particles resulting in 83.5% and 78% uranium bio-recovery at 1.7 and 2.0 pH in 40 days as against maximum recovery of 46% and 41% metal in control experiments respectively. Finer size (<45μm) ore fractions exhibited higher uranium dissolution (96%) in 40 days at 10% (w/v) pulp density (PD), 1.7 pH and 35"oC. On increasing the pulp density from 10% to 20% under the same conditions, the biorecovery of uranium fell down from 96% to 82%. The higher uranium dissolution during bioleaching at 1.7 pH with the fine size particles (<45μm) can be correlated with increase in redox potential from 598 mV to 708 mV and the corresponding variation of Fe(III) ion concentration in 40 days. (author)

  3. Identification of chemical processes influencing constituent mobility during in-situ uranium leaching

    International Nuclear Information System (INIS)

    Sherwood, D.R.; Hostetler, C.J.; Deutsch, W.J.

    1984-07-01

    In-situ leaching of uranium has become a widely accepted method for production of uranium concentrate from ore zones that are too small, too deep, and/or too low in grade to be mined by conventional techniques. One major environmental concern that exists with in-situ leaching of uranium is the possible adverse effects mining might have on regional ground water quality. The leaching solution (lixiviant), which extracts uranium from the ore zone, might also mobilize other potential contaminants (As, Se, Mo, and SO 4 ) associated with uranium ore. Column experiments were performed to investigate the geochemical interactions between a lixiviant and a uranium ore during in-situ leaching and to identify chemical processes that might influence contaminant mobility. The analytical composition data for selected column effluents were used with the MINTEQ code to develop a computerized geochemical model of the system. MINTEQ was used to calculate saturation indices for solid phases based on the composition of the solution. A potential constraint on uranium leaching efficiency appears to be the solubility control of schoepite. Gypsum and powellite solubilities may limit the mobilities of sulfate and molybdenum, respectively. In contrast, the mobilities of arsenic and selenium were not limited by solubility constraints, but were influenced by other chemical interaction between the solution and sediment, perhaps adsorption. Bulk chemical and mineralogical analyses were performed on both the original and leached ores. Using these analyses together with the column effluent data, mass balance calculations were performed on five constituents based on solution chemical analysis and bulk chemical and γ-spectroscopy analysis for the sediment. 6 references, 10 figures, 10 tables

  4. Optimization of lime treatment processes

    International Nuclear Information System (INIS)

    Zinck, J. M.; Aube, B. C.

    2000-01-01

    Lime neutralization technology used in the treatment of acid mine drainage and other acidic effluents is discussed. Theoretical studies and laboratory experiments designed to optimize the technology of lime neutralization processes and to improve the cost efficiency of the treatment process are described. Effluent quality, slaking temperature, aeration, solid-liquid separation, sludge production and geochemical stability have been studied experimentally and on site. Results show that through minor modification of the treatment process, costs, sludge volume generated, and metal released to the environment can be significantly reduced. 17 refs., 4 figs

  5. Methods of removing uranium from drinking water. 1. A literature survey. 2. Present municipal water treatment and potential removal methods

    International Nuclear Information System (INIS)

    Drury, J.S.; Michelson, D.; Ensminger, J.T.; Lee, S.Y.; White, S.K.

    1982-12-01

    Literature was searched for methods of removing uranium from drinking water. U.S. manufacturers and users of water-treatment equipment and products were also contacted regarding methods of removing uranium from potable water. Based on the results of these surveys, it was recommended that untreated, partially treated, and finished water samples from municipal water-treatment facilities be analyzed to determine the extent of removal of uranium by presently used procedures, and that additional laboratory studies be performed to determine what changes are needed to maximize the effectiveness of treatments that are already in use in existing water-treatment plants

  6. Selection of water treatment processes special study

    International Nuclear Information System (INIS)

    1991-11-01

    Characterization of the level and extent of groundwater contamination in the vicinity of Title I mill sites began during the surface remedial action stage (Phase 1) of the Uranium Mill Tailings Remedial Action (UMTRA) Project. Some of the contamination in the aquifer(s) at the abandoned sites is attributable to milling activities during the years the mills were in operation. The restoration of contaminated aquifers is to be undertaken in Phase II of the UMTRA Project. To begin implementation of Phase II, DOE requested that groundwater restoration methods and technologies be investigated by the Technical Assistance Contractor (TAC). and that the results of the TAC investigations be documented in special study reports. Many active and passive methods are available to clean up contaminated groundwater. Passive groundwater treatment includes natural flushing, geochemical barriers, and gradient manipulation by stream diversion or slurry walls. Active groundwater.cleanup techniques include gradient manipulation by well extraction or injection. in-situ biological or chemical reclamation, and extraction and treatment. Although some or all of the methods listed above may play a role in the groundwater cleanup phase of the UMTRA Project, the extraction and treatment (pump and treat) option is the only restoration alternative discussed in this report. Hence, all sections of this report relate either directly or indirectly to the technical discipline of process engineering

  7. Status Report from Canada [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Thunaes, A [Eldorado Mining and Refining Ltd., Ottawa (Canada)

    1967-06-15

    The Canadian production of uranium increased in a spectacular manner during the period 1955-1959 from 1000 to 15 500 tons U{sub 3}O{sub 8} per year. Since 1959 the production has declined to the 1966 level of 3900 tons U{sub 3}O{sub 8} per year; stretch-out of contracts and government stockpiling programmes has made the decline gradual, and is maintaining the current rate of production until 1970. Nineteen mills were in operation during the period of peak production but only three are operating today. Ten mills were shut down and dismantled because of exhaustion of ore bodies or because the operation was uneconomical; six mills are maintained in stand-by condition. The total daily capacity of mills in operation or standing by is about 28 000 tons ore, but some of these mills would not be reopened unless an appreciable increase in uranium price occurs. The tide of uranium demand is about ready to turn and prospecting for uranium is very active this year, particularly in the Elliot Lake and Beaverlodge areas. The estimates for uranium demand in 1975-1980 are such that new ore will have to be found and developed, and new treatment plants must be built. The new ore that is found will likely be of lower grade or more expensive to mine than most of the current proven reserves in Canada and the most efficient methods of treatment will be needed to avoid excessive increases in production costs. This seems an opportune time to review Canadian milling of uranium ore, the improvements that have been made and development work towards further improvements.

  8. Measurement system analysis (MSA) of the isotopic ratio for uranium isotope enrichment process control

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Josue C. de; Barbosa, Rodrigo A.; Carnaval, Joao Paulo R., E-mail: josue@inb.gov.br, E-mail: rodrigobarbosa@inb.gov.br, E-mail: joaocarnaval@inb.gov.br [Industrias Nucleares do Brasil (INB), Rezende, RJ (Brazil)

    2013-07-01

    Currently, one of the stages in nuclear fuel cycle development is the process of uranium isotope enrichment, which will provide the amount of low enriched uranium for the nuclear fuel production to supply 100% Angra 1 and 20% Angra 2 demands. Determination of isotopic ration n({sup 235}U)/n({sup 238}U) in uranium hexafluoride (UF{sub 6} - used as process gas) is essential in order to control of enrichment process of isotopic separation by gaseous centrifugation cascades. The uranium hexafluoride process is performed by gas continuous feeding in separation unit which uses the centrifuge force principle, establishing a density gradient in a gas containing components of different molecular weights. The elemental separation effect occurs in a single ultracentrifuge that results in a partial separation of the feed in two fractions: an enriched on (product) and another depleted (waste) in the desired isotope ({sup 235}UF{sub 6}). Industrias Nucleares do Brasil (INB) has used quadrupole mass spectrometry (QMS) by electron impact (EI) to perform isotopic ratio n({sup 235}U)/n({sup 238}U) analysis in the process. The decision of adjustments and change te input variables are based on the results presented in these analysis. A study of stability, bias and linearity determination has been performed in order to evaluate the applied method, variations and systematic errors in the measurement system. The software used to analyze the techniques above was the Minitab 15. (author)

  9. Scale-Up Information for Gas-Phase Ammonia Treatment of Uranium in the Vadose Zone at the Hanford Site Central Plateau

    Energy Technology Data Exchange (ETDEWEB)

    Truex, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Szecsody, James E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhong, Lirong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thomle, Jonathan N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Johnson, Timothy C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-09-01

    Uranium is present in the vadose zone at the Hanford Central Plateau and is of concern for protection of groundwater. The Deep Vadose Zone Treatability Test Plan for the Hanford Central Plateau identified gas-phase treatment and geochemical manipulation as potentially effective treatment approaches for uranium and technetium in the Hanford Central Plateau vadose zone. Based on laboratory evaluation, use of ammonia vapor was selected as the most promising uranium treatment candidate for further development and field testing. While laboratory tests have shown that ammonia treatment effectively reduces the mobility of uranium, additional information is needed to enable deployment of this technology for remediation. Of importance for field applications are aspects of the technology associated with effective distribution of ammonia to a targeted treatment zone, understanding the fate of injected ammonia and its impact on subsurface conditions, and identifying effective monitoring approaches. In addition, information is needed to select equipment and operational parameters for a field design. As part of development efforts for the ammonia technology for remediation of vadose zone uranium contamination, field scale-up issues were identified and have been addressed through a series of laboratory and modeling efforts. This report presents a conceptual description for field application of the ammonia treatment process, engineering calculations to support treatment design, ammonia transport information, field application monitoring approaches, and a discussion of processes affecting the fate of ammonia in the subsurface. The report compiles this information from previous publications and from recent research and development activities. The intent of this report is to provide technical information about these scale-up elements to support the design and operation of a field test for the ammonia treatment technology.

  10. Solid state processing of massive uranium mononitride, using uranium and uranium higher nitride powders as starting materials (1962); Preparation a l'etat solide de mononitrure d'uranium massif a partir de poudres d'uranium et de nitrures superieurs d'uranium (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Molinari, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1960-12-15

    The mechanism and the optimum conditions for preparing uranium mononitride have been studied. The results have been used for hot pressing (250 kg/cm{sup 2}, 1000 deg. C, under vacuum) a mixture of powders of uranium and uranium higher nitrides. The products obtained have been identified by X-ray measurements and may be - at will and depending upon the stoichiometry - either UN, or a cermet a U{sub {alpha}}-UN. As revealed by the curved shape of grain boundaries, the sinters obtained here do not easily evolve towards physico-chemical equilibrium when submitted to heat treatment. This behaviour is quite different from the one observed with uranium monocarbide prepared by a similar method. This fact may be ascribed to the insolubility in the matrix UN of particles of UO{sub 2} being present as impurities. The density, hardness and thermal conductivity of these products are higher than those measured on uranium nitride or cermets U-UN obtained by other methods. (author) [French] Apres une etude prealable du mecanisme et des conditions optimales de nitruration de l'uranium, on a montre qu'il est possible de preparer par frittage sous charge (250 kg/cm{sup 2}, 1000 deg. C sous vide) d'un melange de poudres d'uranium et de nitrures superieurs d'uranium, un produit qui a ete identifie par diffraction de rayons X. On peut ainsi obtenir a volonte, soit le monocarbure UN, soit un cermet U{sub {alpha}}-UN dans le cas de compositions sous-stoechiometriques. Au contraire du monocarbure d'uranium prepare dans des conditions analogues, les produits obtenus ici, soumis a un traitement thermique, n'evoluent pas facilement vers un etat d'equilibre physico-chimique caracterise par l'existence de joints de grains rectilignes. On attribue ce phenomene a l'insolubilite de l'impurete UO{sub 2} dans UN. La densite, la durete, la conductibilite thermique de ces produits se revelent superieures a celles des nitrures d'uranium ou des cermets U-UN obtenus par les autres methodes. (auteur)

  11. Pilot plant studies on the treatment of El Atshan Uranium Ores, Eastern Desert, Egypt.

    Energy Technology Data Exchange (ETDEWEB)

    Abd Elghany, M S; Mahdy, M A [Nuclear materials authority, El-Maadi, Cairo, (Egypt); Abd El-Monem, A M; El-Hazek, A T [Faculty of engineering, Cairo university, Cairo, (Egypt)

    1995-10-01

    The present work deals with studying the different processes leading to the preparation of commercial uranium concentrate (yellow cake) from El Atshan granitic ore material (0.077%U) after acid leading of the latter, the two common extraction techniques of uranium from the obtained sulphate leach liquor; namely, anion exchange rein and solvent extraction have been studied. The studied leaching and extraction conditions-realized on the lab scale-were applied to inches pilot plant unit (capacity 150 kg ore). An average leaching leaching efficiency exceeding 88% has been achieved. Using anion exchange resin, it has been possible to prepare a uranium peroxide concentrate assaying a uranium content of about 67% U{sub 3} O{sub 8}. Only trace amount of Ca, Fe, Po{sub 4}, Cr and Pb have been detected. On the other hand, sodium uranate, as a uranium precipitate was prepared from the strip solution of the loaded solvent (di-2-ethyl) phosphoric acid concerned with the evaluation of a new optimized technique for the principle of chloramine-T method used for insulin iodination for the modified procedure can be carried out under normal condition of room temperature, employed longer reaction times and omitted the addition of inorganic reducing salts maintaining efficient iodination and avoiding denaturation to obtain labels of exceedingly high specific activity and small quantities of insulin for in vitro usage in the investigation of human erythrocytes 125 I-inulin binding capacity in normal and in some disease status. 9 figs., 2 tabs.

  12. Pilot plant studies on the treatment of El Atshan Uranium Ores, Eastern Desert, Egypt

    International Nuclear Information System (INIS)

    Abd Elghany, M.S.; Mahdy, M.A.; Abd El-Monem, A.M.; El-Hazek, A.T.

    1995-01-01

    The present work deals with studying the different processes leading to the preparation of commercial uranium concentrate (yellow cake) from El Atshan granitic ore material (0.077%U) after acid leading of the latter, the two common extraction techniques of uranium from the obtained sulphate leach liquor; namely, anion exchange rein and solvent extraction have been studied. The studied leaching and extraction conditions-realized on the lab scale-were applied to inches pilot plant unit (capacity 150 kg ore). An average leaching leaching efficiency exceeding 88% has been achieved. Using anion exchange resin, it has been possible to prepare a uranium peroxide concentrate assaying a uranium content of about 67% U 3 O 8 . Only trace amount of Ca, Fe, Po 4 , Cr and Pb have been detected. On the other hand, sodium uranate, as a uranium precipitate was prepared from the strip solution of the loaded solvent (di-2-ethyl phosphoric acid concerned with the evaluation of a new optimized technique for the principle of chloramine-T method used for insulin iodination for the modified procedure can be carried out under normal condition of room temperature, employed longer reaction times and omitted the addition of inorganic reducing salts maintaining efficient iodination and avoiding denaturation to obtain labels of exceedingly high specific activity and small quantities of insulin for in vitro usage in the investigation of human erythrocytes 125 I-inulin binding capacity in normal and in some disease status. 9 figs., 2 tabs

  13. Treatment of Uranium and Plutonium solutions generated in Atalante by R and D activities

    International Nuclear Information System (INIS)

    Lagrave, H.; Beretti, C.; Bros, P.

    2008-01-01

    The Atalante complex operated by the 'Commissariat a l'Energie Atomique' (Cea) consolidates research programs on actinide chemistry, processing for recycling spent fuel, and fabrication of actinide targets for innovative concepts in future nuclear systems. In order to produce mixed oxide powder containing uranium, plutonium and minor actinides and to deal with increasing flows in the facility, a new shielded line will be built and is expected to be operational by 2012. Its main functions will be to receive, concentrate and store solutions, purify them, ensure co-conversion of actinides and conversion of excess uranium. (authors)

  14. Treatment of Uranium and Plutonium solutions generated in Atalante by R and D activities

    Energy Technology Data Exchange (ETDEWEB)

    Lagrave, H.; Beretti, C.; Bros, P. [CEA Rhone Valley Research Center, BP 17171, 30207 Bagnols-sur-Ceze Cedex (France)

    2008-07-01

    The Atalante complex operated by the 'Commissariat a l'Energie Atomique' (Cea) consolidates research programs on actinide chemistry, processing for recycling spent fuel, and fabrication of actinide targets for innovative concepts in future nuclear systems. In order to produce mixed oxide powder containing uranium, plutonium and minor actinides and to deal with increasing flows in the facility, a new shielded line will be built and is expected to be operational by 2012. Its main functions will be to receive, concentrate and store solutions, purify them, ensure co-conversion of actinides and conversion of excess uranium. (authors)

  15. The regulatory process for uranium mines in Canada -general overview and radiation health and safety in uranium mine-mill facilities

    International Nuclear Information System (INIS)

    Dory, A.B.

    1982-01-01

    This presentation is divided into two main sections. In the first, the author explores the issues of radiation and tailings disposal, and then examines the Canadian nuclear regulatory process from the point of view of jurisdiction, objectives, philosophy and mechanics. The compliance inspection program is outlined, and the author discussed the relationships between the AECB and other regulatory agencies, the public and uranium mine-mill workers. The section concludes with an examination of the stance of the medical profession on nuclear issues. In part two, the radiological hazards for uranium miners are examined: radon daughters, gamma radiation, thoron daughters and uranium dust. The author touches on new regulations being drafted, the assessment of past exposures in mine atmospheres, and the regulatory approach at the surface exploration stage. The presentation concludes with the author's brief observations on the findings of other uranium mining inquiries and on future requirements in the industry's interests

  16. Status report from South Africa [Processing of Low-Grade Uranium Ores

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, R E [Atomic Energy Board, Pretoria (South Africa)

    1967-06-15

    Most of the research work on the processing of uranium ores in South Africa is being conducted by the Extraction Metallurgy Division of the S.A. Atomic Energy Board. Nevertheless, a considerable amount of applied research has been done by the different mining groups concerned with the operation of uranium plants, and also by the Transvaal and Orange Free State Chamber of Mines research laboratories. There is, however, very close collaboration between the various research groups and the Atomic Energy Board and the main research described is conducted on a collaborative basis.

  17. A complete remediation process for a uranium-contaminated site and application to other sites

    International Nuclear Information System (INIS)

    Mason, C.F.V.; Lu, N.; Kitten, H.D.; Williams, M.; Turney, W.R.J.R.

    1998-01-01

    During the summer of 1996 the authors were able to test, at the pilot scale, the concept of leaching uranium (U) from contaminated soils. The results of this pilot scale operation showed that the system they previously had developed at the laboratory scale is applicable at the pilot scale. The paper discusses these results, together with laboratory scale results using soil from the Fernald Environmental Management Project (FEMP), Ohio. These FEMP results show how, with suitable adaptations, the process is widely applicable to other sites. The purpose of this paper is to describe results that demonstrate remediation of uranium-contaminated soils may be accomplished through a leach scheme using sodium bicarbonate

  18. A complete remediation process for a uranium-contaminated site and application to other sites

    Energy Technology Data Exchange (ETDEWEB)

    Mason, C.F.V.; Lu, N.; Kitten, H.D.; Williams, M.; Turney, W.R.J.R.

    1998-12-31

    During the summer of 1996 the authors were able to test, at the pilot scale, the concept of leaching uranium (U) from contaminated soils. The results of this pilot scale operation showed that the system they previously had developed at the laboratory scale is applicable at the pilot scale. The paper discusses these results, together with laboratory scale results using soil from the Fernald Environmental Management Project (FEMP), Ohio. These FEMP results show how, with suitable adaptations, the process is widely applicable to other sites. The purpose of this paper is to describe results that demonstrate remediation of uranium-contaminated soils may be accomplished through a leach scheme using sodium bicarbonate.

  19. Management and Handling of Rejected Fuel of MTR Type and Process Effluents Contained Uranium at FEPI

    International Nuclear Information System (INIS)

    Ghaib Widodo; Bambang Herutomo

    2007-01-01

    Research Reactor Fuel Element Production Installation (FEPI) - Serpong has performed management and handling of all kinds of rejected fuel material during production (solids, liquids, and gases) and process effluents contained uranium. The methods that has been implemented are precipitation, absorption, evaporation, electrolysis, and electrodialysis. By these methods will finally be obtained forms of product which can be used directly as fuel material feed and solid/liquid radioactive waste that fulfil the requirements (uranium contents < 50 ppm) to be send to Radioactive Waste Management Installation. (author)

  20. An Overview of Process Monitoring Related to the Production of Uranium Ore Concentrate

    Energy Technology Data Exchange (ETDEWEB)

    McGinnis, Brent [Innovative Solutions Unlimited, LLC

    2014-04-01

    Uranium ore concentrate (UOC) in various chemical forms, is a high-value commodity in the commercial nuclear market, is a potential target for illicit acquisition, by both State and non-State actors. With the global expansion of uranium production capacity, control of UOC is emerging as a potentially weak link in the nuclear supply chain. Its protection, control and management thus pose a key challenge for the international community, including States, regulatory authorities and industry. This report evaluates current process monitoring practice and makes recommendations for utilization of existing or new techniques for managing the inventory and tracking this material.

  1. Recovery treatment for the non fissioned uranium in the production of Mo-99

    International Nuclear Information System (INIS)

    Rodriguez S, A.; Acosta C, A.L.; Lopez M, B.E.

    1991-09-01

    An effective modification of the chemical processes has been obtained to dissolve at the uranium-IV and to extract it as uranyl triperoxidate that facilitates its manipulation and final conversion to uranyl nitrate like a concentrate of high purity. (Author)

  2. Potentiometric determination of uranium in simulated Purex Process solutions by acidiometry

    International Nuclear Information System (INIS)

    Cohen, V.H.; Matsuda, H.T.; Araujo, B.F. de; Araujo, J.A. de

    1983-01-01

    A potentiometric methods for sequential free acidity and uranium determination in simulated Purex Process solutions is described. An oxalate solution or a mixture of fluoride-oxalate pellets were used as complexing agent for free titration. Following this first equivalent point, uranium is determined-by indirect titration of H + liberated in the peruanate reaction. Some elements present in the standard fuel elements with a burn-up of 33.000 Mwd/t, neutron flux of 3,2 x 10 13 n.cm -2 .s -1 and cooling time of two years were considered as interfering elements in uranium analyses. As a substitute of Pu-IV, Th(NO 3 ) 4 solution was used. The method can be applied to aqueous and organic (TBP/diluent) solutions with 2% precision and 2% accuracy. (Autor) [pt

  3. Potentiometric determination of uranium in simulated Purex Process solutions by acidiometry

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, V H; Matsuda, H T; Araujo, B.F. de; Araujo, J.A. de

    1984-01-01

    A potentiometric methods for sequential free acidity and uranium determination in simulated Purex Process solutions is described. An oxalate solution or a mixture of fluoride-oxalate pellets were used as complexing agent for free titration. Following this first equivalent point, uranium is determined-by indirect titration of H/sup +/ liberated in the peruanate reaction. Some elements present in the standard fuel elements with a burn-up of 33.000 Mwd/t, neutron flux of 3,2 x 10/sup 13/n.cm/sup -2/.s/sup -1/ and cooling time of two years were considered as interfering elements in uranium analyses. As a substitute of Pu-IV, Th(NO/sub 3/)/sub 4/ solution was used. The method can be applied to aqueous and organic (TBP/diluent) solutions with 2% precision and 2% accuracy. (Autor).

  4. Uranium and thorium concentration process during partial fusion and crystallization of granitic magma

    International Nuclear Information System (INIS)

    Cuney, M.

    1982-01-01

    Two major processes, frequently difficult to distinguish, lead to uranium and thorium enrichment in igneous rocks and more particularly in granitoids; these are partial melting and fractional crystallization. Mont-Laurier uranothoriferous pegmatoids, Bancroft and Roessing deposits are examples of radioelement concentrations resulting mostly of low grade of melting on essentially metasedimentary formations deposited on a continental margin or intracratonic. Fractional crystallization follows generally partial melting even in migmatitic areas. Conditions prevailing during magma crystallization and in particular oxygen fugacity led either to the formation of uranium preconcentrations in granitoids, or to its partition in the fluid phase expelled from the magma. No important economic uranium deposit appears to be mostly related to fractional crystallization of large plutonic bodies

  5. Uranium and thorium recovery from a sub-product of monazite industrial processing

    International Nuclear Information System (INIS)

    Gomiero, L.A.; Ribeiro, J.S.; Scassiotti Filho, W.

    1994-01-01

    In the monazite alkaline leaching industrial process for the production of rare earth elements, a by-product is formed, which has a high concentration of thorium and a lower but significant one of uranium. A procedure for recovery of the thorium and uranium contents in this by-product is presented. The first step of this procedure is the leaching with sulfuric acid, followed by uranium extraction from the acid liquor with a tertiary amine, stripping with a Na Cl solutions and precipitation as ammonium diuranate with N H 4 O H. In order to obtain thorium concentrates with higher purity, it is performed by means of the extraction of thorium from the acid liquor, with a primary amine, stripping by a Na Cl solution and precipitation as thorium hydroxide or oxalate. (author)

  6. Technologies for processing low-grade uranium ores and their relevance to the Indian situation

    International Nuclear Information System (INIS)

    Murthy, T.K.S.

    1991-01-01

    The technology for uranium ore processing is well established. Various estimates have shown that on a global basis uranium resources are adequate to meet the forseeable demand. The Indian resources are estimated to be about 60,000 t U. The grade of the ores is low and the individual deposits are small. The nature of the deposits, precarious resources position and relatively small capacity of the mines do not permit the country to take advantage of large throughputs in the mill to achieve substantial cost reduction. However by resorting to as high a scale of milling as the mines would permit, by reducing the loss of solubilised uranium after leaching and by undertaking production of nuclear grade final product at the mill site, significant though not a major, economic benefit can be derived. (author). 2 figs., 3 tabs

  7. PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

    Science.gov (United States)

    Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

    1962-08-14

    A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

  8. The CIX uranium process: Blyvoors leads the way with full conversion

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The Atomic Energy Board has developed a promising technique - the continuous ion exchange(CIX) process - for the recovery of uranium. The Blyvooruitzicht Gold Mining Company, which accommodated the highly successful demonstration plant is now spending R10 500 000 on extentions and conversions to full CCD/CIX. This article outlines the system and its advantages

  9. Biomembrane oxidizing tank used in the process of bacterial heap leaching of uranium ore

    International Nuclear Information System (INIS)

    Meng Yunsheng; Fan Baotuan; Liu Jian; Zheng Ying; Liu Chao

    2004-01-01

    The construction characteristic of biomembrane oxidizing tank and specialty of packing material used in the process of bacterial heap leaching of uranium ore are introduced in this paper. Method for designing biomembrane oxidizing tank, layout principle of aeration system and measurements on running management are summarized

  10. Solution of environmental protection problems and complex utilization of raw materials during mining and processing of uranium ores

    International Nuclear Information System (INIS)

    Litvinenko, V.G.; Savva, P.P.

    1993-01-01

    Consideration is given to the complex of measures taken in Priargunsky industrial mine-chemical association and directed to environment protection, complex utilization of raw materials during mining and processing of uranium ores. These measures include: 1) reduction of toxic chemical agent effluents into atmosphere due to introduction of new methods and gas cleaning systems; 2) rational use of water resources owing to application of circulating water supply systems, waste waters treatment and effective control of the state of water consumption by industrial enterprises; 3) utilization of gangue and industrial solid wastes

  11. Processing used nuclear fuel with nanoscale control of uranium and ultrafiltration

    Energy Technology Data Exchange (ETDEWEB)

    Wylie, Ernest M.; Peruski, Kathryn M.; Prizio, Sarah E. [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, Notre Dame, IN 46556 (United States); Bridges, Andrea N.A.; Rudisill, Tracy S.; Hobbs, David T. [Savannah River National Laboratory, Aiken, SC 29808 (United States); Phillip, William A. [Department of Chemical and Biomolecular Engineering, University of Notre Dame, Notre Dame, IN 46556 (United States); Burns, Peter C., E-mail: pburns@nd.edu [Department of Civil and Environmental Engineering and Earth Sciences, University of Notre Dame, Notre Dame, IN 46556 (United States); Department of Chemistry and Biochemistry, University of Notre Dame, Notre Dame, IN 46556 (United States)

    2016-05-15

    Current separation and purification technologies utilized in the nuclear fuel cycle rely primarily on liquid–liquid extraction and ion-exchange processes. Here, we report a laboratory-scale aqueous process that demonstrates nanoscale control for the recovery of uranium from simulated used nuclear fuel (SIMFUEL). The selective, hydrogen peroxide induced oxidative dissolution of SIMFUEL material results in the rapid assembly of persistent uranyl peroxide nanocluster species that can be separated and recovered at moderate to high yield from other process-soluble constituents using sequestration-assisted ultrafiltration. Implementation of size-selective physical processes like filtration could results in an overall simplification of nuclear fuel cycle technology, improving the environmental consequences of nuclear energy and reducing costs of processing. - Highlights: • Nanoscale control in irradiated fuel reprocessing. • Ultrafiltration to recover uranyl cage clusters. • Alternative to solvent extraction for uranium purification.

  12. Extraction of uranium from seawater: chemical process and plant design feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, M.H.; Frame, J.M.; Dudey, N.D.; Kiel, G.R.; Mesec, V.; Woodfield, F.W.; Binney, S.E.; Jante, M.R.; Anderson, R.C.; Clark, G.T.

    1979-02-01

    A major assessment was made of the uranium resources in seawater. Several concepts for moving seawater to recover the uranium were investigated, including pumping the seawater and using natural ocean currents or tides directly. The optimal site chosen was on the southeastern Puerto Rico coast, with the south U.S. Atlantic coast as an alternate. The various processes for extracting uranium from seawater were reviewed, with the adsorption process being the most promising at the present time. Of the possible adsorbents, hydrous titanium oxide was found to have the best properties. A uranium extraction plant was conceptually designed. Of the possible methods for contacting the seawater with the adsorbent, a continuous fluidized bed concept was chosen as most practical for a pumped system. A plant recovering 500 tonnes of U/sub 3/O/sub 8/ per year requires 5900 cubic meters per second of seawater to be pumped through the adsorbent beds for a 70% overall recovery efficiency. Total cost of the plant was estimated to be about $6.2 billion. A computer model for the process was used for parametric sensitivity studies and economic projections. Several design case variations were developed. Other topics addressed were the impact of co-product recovery, environmental considerations, etc.

  13. Salt separation of uranium deposits generated from electrorefining in pyro process

    International Nuclear Information System (INIS)

    Kwon, S. W.; Park, K. M.; Jeong, J. H.; Lee, H. S.; Kim, J. G.

    2012-01-01

    Electrorefining is a key step in a pyro processing. Electrorefining process is generally composed of two recovery steps- deposit of uranium onto a solid cathode(electrorefining) and then the recovery of the remaining uranium and TRU(TransUranic) elements simultaneously by a liquid cadmium cathode(electrowinning). The uranium ingot is prepared from the deposits after the salt separation. In this study, the sequential operation of the liquid salt separation? distillation of the residual salt was attempted for the achievement of high throughput performance in the salt separation. The effects of deposit size and packing density were also investigated with steel chips, steel chips, and uranium dendrites. The apparent evaporation rate decreased with the increasing packing density or the increasing size of deposits due to the hindrance of the vapor transport by the deposits. It was found that the packing density and the geometry of deposit crucible are important design parameters for the salt separation system. Base on the results of the study, an engineering scale salt distiller was developed and installed in the argon cell. The salt distiller is a batch-type, and the process capacity to about 50 kg U-deposits/day. The design of the salt distiller is based on the remote operation by Master Slave Manipulator (MSM) and a hoist. The salt distiller is composed of two large blocks of the distillation tower and the crucible loading system for the transportation to maintenance room via the Large Transfer Lock (LTL)

  14. Extraction of uranium from seawater: chemical process and plant design feasibility study

    International Nuclear Information System (INIS)

    Campbell, M.H.; Frame, J.M.; Dudey, N.D.; Kiel, G.R.; Mesec, V.; Woodfield, F.W.; Binney, S.E.; Jante, M.R.; Anderson, R.C.; Clark, G.T.

    1979-02-01

    A major assessment was made of the uranium resources in seawater. Several concepts for moving seawater to recover the uranium were investigated, including pumping the seawater and using natural ocean currents or tides directly. The optimal site chosen was on the southeastern Puerto Rico coast, with the south U.S. Atlantic coast as an alternate. The various processes for extracting uranium from seawater were reviewed, with the adsorption process being the most promising at the present time. Of the possible adsorbents, hydrous titanium oxide was found to have the best properties. A uranium extraction plant was conceptually designed. Of the possible methods for contacting the seawater with the adsorbent, a continuous fluidized bed concept was chosen as most practical for a pumped system. A plant recovering 500 tonnes of U 3 O 8 per year requires 5900 cubic meters per second of seawater to be pumped through the adsorbent beds for a 70% overall recovery efficiency. Total cost of the plant was estimated to be about $6.2 billion. A computer model for the process was used for parametric sensitivity studies and economic projections. Several design case variations were developed. Other topics addressed were the impact of co-product recovery, environmental considerations, etc

  15. Radiation protection of workers in mining and processing of uranium ore

    International Nuclear Information System (INIS)

    Khan, A.H.; Sahoo, S.K; Puranik, V.D.

    2003-01-01

    Low grade of uranium ore mined from three underground mines is processed in a mill at Jaduguda in eastern India to recover uranium concentrate in the form of yellow cake. Radiation protection of workers is given due importance at all stages of these operations. Dedicated Health Physics Units and Environmental Survey Laboratories established at the site regularly carry out in-plant and environmental surveillance to keep radiation exposure of workers and the members of public within the limits prescribed by the regulatory body. The limits set by the national regulatory body based on the international standards recommended by the ICRP and the IAEA are followed. In the uranium mines, external gamma radiation, radon and airborne activity due to radioactive dust are monitored. Similarly, in the uranium ore processing mill, gamma radiation and airborne radioactivity due to long-lived α-emitters are monitored. Personal dosimeters are also issued to workers. The total radiation exposure of workers from external and internal sources is evaluated from the area and personal monitoring data. It has been observed that the average radiation dose to workers has been below 10 mSvy -1 and all exposures are well below 20 mSvy -1 at all stages of operations. Adequate ventilation is provided during mining and ore processing operations to keep the concentrations of airborne radioactivity well below the derived limits. Workers use personal protective appliances, where necessary, as a supplementary means of control. The monitoring methodologies, results and control measures are presented in the paper. (author)

  16. Salt separation of uranium deposits generated from electrorefining in pyro process

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, S. W.; Park, K. M.; Jeong, J. H.; Lee, H. S.; Kim, J. G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Electrorefining is a key step in a pyro processing. Electrorefining process is generally composed of two recovery steps- deposit of uranium onto a solid cathode(electrorefining) and then the recovery of the remaining uranium and TRU(TransUranic) elements simultaneously by a liquid cadmium cathode(electrowinning). The uranium ingot is prepared from the deposits after the salt separation. In this study, the sequential operation of the liquid salt separation? distillation of the residual salt was attempted for the achievement of high throughput performance in the salt separation. The effects of deposit size and packing density were also investigated with steel chips, steel chips, and uranium dendrites. The apparent evaporation rate decreased with the increasing packing density or the increasing size of deposits due to the hindrance of the vapor transport by the deposits. It was found that the packing density and the geometry of deposit crucible are important design parameters for the salt separation system. Base on the results of the study, an engineering scale salt distiller was developed and installed in the argon cell. The salt distiller is a batch-type, and the process capacity to about 50 kg U-deposits/day. The design of the salt distiller is based on the remote operation by Master Slave Manipulator (MSM) and a hoist. The salt distiller is composed of two large blocks of the distillation tower and the crucible loading system for the transportation to maintenance room via the Large Transfer Lock (LTL)

  17. Brazilian uranium exploration program

    International Nuclear Information System (INIS)

    Marques, J.P.M.

    1981-01-01

    General information on Brazilian Uranium Exploration Program, are presented. The mineralization processes of uranium depoits are described and the economic power of Brazil uranium reserves is evaluated. (M.C.K.) [pt

  18. Acid-curing and ferric-trickle leaching effluent used in closed circuit uranium extractive process

    International Nuclear Information System (INIS)

    Jin Suoqing; Xiang Qinfang; Guo Jianzheng; Lu Guizhu; Su Yanru

    1998-01-01

    The new uranium ore process consists of crushing ore, mixing crushed ore with strong acid in rotating drums and curing the mixture in piles, trickle-leaching the ore beds with ferric solution, extracting uranium from pregnant solution with tertiary amine, precipitating product and disposing residue tailings. All the process effluent is used in closed circuit. There will be no process water to be discharged in the flowsheet except the tailings carrying off 15% water because during leaching moisture content of the ore rises to 15%. Tailings produced by the process are moist and friable, and can be disposed of on a pile or returned to the mine. Main technical parameters of the process: (a) water consumption is 0.2∼0.3 m 3 /t ore, electric power consumption is 20∼30 kW·h/t ore; (b) ore crushing up to -5∼-7 mm, leaching period is 12∼45 d, U content of residue is 0.01%∼0.02%, producing pregnant solution is 0.3∼0.5 m 3 /t ore, which is 1/5∼1/8 that of conventional agitation leaching process; (c) organic agent consumption is 1/5∼1/8 that of the conventional agitation process. All the research results above are tested by the pilot-plant test and industrial test. The new process has been applied to recovery of uranium in the mine located at northeast of China

  19. Project C-018H, 242-A Evaporator/PUREX Plant Process Condensate Treatment Facility, functional design criteria. Revision 3

    International Nuclear Information System (INIS)

    Sullivan, N.

    1995-01-01

    This document provides the Functional Design Criteria (FDC) for Project C-018H, the 242-A Evaporator and Plutonium-Uranium Extraction (PUREX) Plant Condensate Treatment Facility (Also referred to as the 200 Area Effluent Treatment Facility [ETF]). The project will provide the facilities to treat and dispose of the 242-A Evaporator process condensate (PC), the Plutonium-Uranium Extraction (PUREX) Plant process condensate (PDD), and the PUREX Plant ammonia scrubber distillate (ASD)

  20. Extraction of uranium from coarse ore and acid-curing and ferric sulphate-trickle leaching process

    International Nuclear Information System (INIS)

    Jin Suoqing

    1994-01-01

    On the basis of analysis of the problems in the technology of the traditional uranium hydrometallurgy and the limitations of thin layer leaching process (TLL), a new leaching system-acid-curing and ferric sulphate-trickle leaching (AFL) process (NGJ in Chinese) has developed for extraction of uranium from the coarse ore. The ferric sulphate solution was used for trickling the acid-cured uranium ore and the residual leaching reaction incomplete in TLL process can be improved in this process. And the AFL process has a wide applicability to China's uranium ores, being in competition with the traditional agitation leaching process for treating coarse ores. The uranium ore processing technology based on the AFL process will become one of the new basic technologies of uranium hydrometallurgy. A series of difficulties will be basically overcome associated with fine grinding because of its elimination in the presented process. Moreover, the situation of the present uranium hydrometallurgy can be also changed owing to without technological effluent discharge

  1. Study on microstructure change of Uranium nitride coated U-7wt%Mo powder by heat treatment

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Woo Hyoung; Park, Jae Soon; Lee, Hae In; Kim, Woo Jeong; Yang, Jae Ho; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Uranium-molybdenum alloy particle dispersion fuel in an aluminum matrix with a high uranium density has been developed for a high performance research reactor in the RERTR program. In order to retard the fuel-matrix interaction in U-Mo/Al dispersion fuel in which the U-Mo fuel particles were dispersed in Al matrix, nitride layer coated U-Mo fuel particle has been designed and techniques to fabricate nitride-layer coated U-7wt%Mo particles have been developed in our lab. In this study, uranium nitride coated U-Mo particle has heat treatment for several times and degree. And we suggested for interaction layer remedy in U-Mo dispersion fuel. We investigate effect of heat treatment interaction layer evolution on uranium nitride coated U-Mo powder. The EDS and XRD analysis to investigate the phase evolution in uranium nitride coated layer is also a part of the present work

  2. Treatment of effluent containing uranium with magnetic zeolite; Tratamento de efluente contendo uranio com zeolita magnetica

    Energy Technology Data Exchange (ETDEWEB)

    Craesmeyer, Gabriel Ramos

    2013-07-01

    Within this work, a magnetic-zeolite composite was successfully synthesized using ferrous sulfate as raw material for the magnetic part of the composite, magnetite, and coal fly ash as raw material for the zeolitic phase. The synthesis of the zeolitic phase was made by alkali hydrothermal treatment and the magnetite nanoparticles were obtained through Fe{sup 2+} precipitation on alkali medium. The synthetic process was repeated many times and showed good reproducibility comparing the zeolitic nanocomposite from different batches. The final product was characterized using infrared spectroscopy, powder X-ray diffraction, X-ray fluorescence, scanning electron microscopy with coupled EDS. Specific mass, specific surface area and other physicochemical proprieties. The main crystalline phases found in the final product were magnetite, zeolites types NaP1 and hydroxysodalite, quartz and mullite, those last two remaining from the raw materials. Uranium removal capacity of the magnetic zeolite composite was tested using batch techniques. The effects of contact time and initial concentration of the adsorbate over the adsorption process were evaluated. Equilibrium time was resolved and the following kinetics and diffusion models were evaluated: pseudo-first order kinetic model, pseudo-second order kinetic model and interparticle diffusion model. A contact time of 120 min turned out to be enough to reach equilibrium of the adsorption process. The rate of adsorption followed the pseudo-second order model and the intra particle diffusion did not turn out to be a speed determinant step. Two adsorption isotherms models, the Langmuir model and the Freundlich model, were also evaluated. The Langmuir model was the best fit for the obtained experimental data. Using the best fitted adsorption isotherm and kinetic model, the theoretical maximum adsorption capacity of uranium over the composite was determined for both models. The maximum removal capacity calculated was 20.7 mg.g{sup -1

  3. Foreign research reactor uranium supply program: The Y-12 national security complex process

    International Nuclear Information System (INIS)

    Nelson, T.; Eddy, B.G.

    2010-01-01

    The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community

  4. Uranium Task Force final report

    International Nuclear Information System (INIS)

    1991-03-01

    Site-specific data on the management of uranium of 17 facilities have been assembled and analyzed to develop a comprehensive report on uranium processes, treatment, storage, and disposal on a Department of Energy-wide basis. By integrating a variety of waste generation sources, treatment processes, storage facilities, and disposal options, this waste management system study aims to effectively characterize and evaluate the performance and effectiveness of the total Department of Energy system for the management of uranium, as well as the individual sites. 7 refs., 7 figs., 2 tabs

  5. Preparation of Na4UO2(CO3)3 in presence of Ce-141. II, Treatment of uranium decontamination

    International Nuclear Information System (INIS)

    Lopez M, B.E.; Rodriguez S, A.

    1992-02-01

    It was settled down that the coexistence of chemical species structurally different of cerium, is a consequence of the preparation time; whose practical application, for the purification of the uranium, it can constitute the technological aspect but important in the ion exchange process, to separate the Ce-141 from the uranium. (Author)

  6. Method for the recovery of uranium from phosphoric acid, originating from the wet-process of uraniferous phosphate ores

    International Nuclear Information System (INIS)

    Pyrih, R.Z.; Rickard, R.S.; Carrington, O.F.

    1978-01-01

    Improvement in the process for recoverying uranium from wet-process phosphoric acid solution derived from the acidulation of uraniferous phosphate ores by the use of two ion exchange circuits is described. (Auth.)

  7. Research and development prospects for the atomic uranium laser isotope separation process. Research report 442

    International Nuclear Information System (INIS)

    Janes, G.S.; Forsen, H.K.; Levy, R.H.

    1977-06-01

    Research and development activities are being conducted on many aspects of the atomic uranium laser isotope separation process. Extensive laser spectroscopy studies have been made in order to identify attractive multi-step selective ionization schemes. Using low density (10 10 atoms/cm 3 ) apparatus, the excited state spectra of atomic uranium have been investigated via multiple step laser excitation and photoionization studies using two, three and four pulsed lasers. Observation of the spectra was accomplished by observing the yield of 235 U and 238 U ions as a function of the wavelength, intensities and delays of the various lasers. These data yielded information on the photoexcitation and photoionizatin cross sections, and on the location, J values, lifetimes, isotope shifts and hyperfine structure of the various atomic levels of uranium. Experiments on selective ionization of uranium vapor by multiple step laser excitation followed by ion extraction at 10 13 atoms/cm 3 density have produced 6% enriched 235 U. These indicate that this process is well adapted to produce light water reactor fuel but less suitable for highly enriched material. Application has been made for license for a 1979 experimental facility to provide data for a mid-1980 commercial plant

  8. Laboratory evaluation of the hydrogen sulfide gas treatment approach for remediation of chromate-, uranium(VI)-, and nitrate-contaminated soils

    International Nuclear Information System (INIS)

    Thornton, E.C.; Baechler, M.A.; Beck, M.A.; Amonette, J.E.

    1994-08-01

    Bench-scale soil treatment tests were conducted as part of an effort to develop and implement an in situ chemical treatment approach to the remediation of metal and radionuclide contaminated soils through the use of reactive gases. In general, > 90% immobilization of chromium and > 50% immobilization of uranium was achieved. Leach test results indicate that the treatment process is irreversible for chromium but partially reversible for uranium indicates that immobilization for this contaminant is more readily achieved in organic rich soils. This observation is ascribed to the reducing nature of organic matter. Additional tests were also conducted with soils contaminated to the 5,000 ppm level with nitrate. Nitrate was not found to interfere significantly with treatment of the contaminants. Nitrite was observed in the leachate samples obtained from tests with an organic-rich soil containing clay, however. Leachate chemistries suggested that no other significantly hazardous byproducts were generated by the treatment process and that soil alteration effects were minimal. Test results also suggest that treatment effectiveness is somewhat lower in very dry soils but still able to immobilize chromium and uranium to an acceptable degree. Results of these testing activities indicate that the concentration of hydrogen sulfide in the gas mixture is not a limited factor in treatment as long as a sufficient volume of the mixture is delivered to the soil to achieve a mole ratio of hydrogen sulfide to contaminant of at least 10

  9. Reaction of Antimony-Uranium Composite Oxide in the Chlorination Treatment of Waste Catalyst - 13521

    Energy Technology Data Exchange (ETDEWEB)

    Sawada, Kayo [EcoTopia Science Institute (Japan); Hirabayashi, Daisuke; Enokida, Youichi [Department of Materials, Physics and Energy Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan)

    2013-07-01

    The effect of oxygen gas concentration on the chlorination treatment of antimony-uranium composite oxide catalyst waste was investigated by adding different concentrations of oxygen at 0-6 vol% to its chlorination agent of 0.6 or 6 vol% hydrogen chloride gas at 1173 K. The addition of oxygen tended to prevent the chlorination of antimony in the oxide. When 6 vol% hydrogen chloride gas was used, the addition of oxygen up to 0.1 vol% could convert the uranium contained in the catalyst to U{sub 3}O{sub 8} without any significant decrease in the reaction rate compared to that of the treatment without oxygen. (authors)

  10. A study of aeration treatment of uranium-contained wastewater by saccharomyces cerevisiae-activated sludge

    International Nuclear Information System (INIS)

    Xia Liangshu; Chen Zhongqing

    2006-01-01

    Experiments of the aeration treatment of uranium-contained wastewater by saccharomyces cerevisiae-activated sludge were carried out. The experimental results indicate that, saccharomyces cerevisiae (S.C) can accumulate UO 2 2+ effectively from aqueous solution: the removal ratio of 100 mg·L -1 UO 2 2+ is 78.2% when S.C dosage is 10 g·L -1 , while with 8 g·L -1 activated sludge (A.S.) added in the solution the ratio has increased to 96.3%; then, 5-10 min effluent settling is clarified as a result of sludge flocculation; the optimum conditions of biosorption of U from wastewater by S.C.-A.S. are at pH 5, A.S concentration=8 g·L -1 , added dry weight of S.C.=10 g·L -1 , granularity of S.C=100-120 mesh; the quantity of U increases with the enhanced initial concentration of UO 2 2+ in the process of biosorption by S.C.-A.S., but the removal ratio decreases. The uptake of U could be described by the Freundlich and the Langmuir adsorption isotherms, which demonstrated that the adsorption was regarded as a physical adsorption. (authors)

  11. Recent technical changes in the treatment plants of Sismo for uranium ores

    International Nuclear Information System (INIS)

    Clappier, L.; Michel, P.

    1983-01-01

    The mills of the Ste. Industrielle des Minerais de l'Ouest (Simo) have undergone various reconstructions, consisting of the replacement of worn equipment, an extension of the capacity (from 600000 to 1100000 tonnes per year) and the introduction of new techniques which are described in the present article: endless belt filters; the grinding operations in a wet medium lead to a pulp which contains 1.11 m 3 of water per tonne of ore; the insulation of one endless-belt filter from the washing sector after leaching has permitted an increase of the solid concentration of the pulps; the recovery of uraniferous solutions includes generally a stage of concentration and of purification of the uranium in solution by ion exchange on a resin or liquid solvent substrate; instead of using them in parallel, the ''Eluex'' process utilizes them in series; the extraction by solvent is achieved with mixers and decanters and more recently with the ''pulsed column'' since 1981; the treatment of scum which hampers the liquid/liquid extraction has been planned in several ways; the latest solution adopted concerns the filtration under pressure; the drying of the concentrate has been effected for several years at ''l'Ecarpiere'' by atomization at the drying space on the belt [fr

  12. Dry uranium tetrafluoride process preparation using the uranium hexafluoride reconversion process effluents; Processo alternativo para obtencao de tetrafluoreto de uranio a partir de efluentes fluoretados da etapa de reconversao de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Silva Neto, Joao Batista da

    2008-07-01

    It is a well known fact that the use of uranium tetrafluoride allows flexibility in the production of uranium suicide and uranium oxide fuel. To its obtention there are two conventional routes, the one which reduces uranium from the UF{sub 6} hydrolysis solution with stannous chloride, and the hydro fluorination of a solid uranium dioxide. In this work we are introducing a third and a dry way route, mainly utilized to the recovery of uranium from the liquid effluents generated in the uranium hexafluoride reconversion process, at IPEN/CNEN-SP. Working in the liquid phase, this route comprises the recuperation of ammonium fluoride by NH{sub 4}HF{sub 2} precipitation. Working with the solid residues, the crystallized bifluoride is added to the solid UO{sub 2}, which comes from the U mini plates recovery, also to its conversion in a solid state reaction, to obtain UF{sub 4}. That returns to the process of metallic uranium production unity to the U{sub 3}Si{sub 2} obtention. This fuel is considered in IPEN CNEN/SP as the high density fuel phase for IEA-R1m reactor, which will replace the former low density U{sub 3}Si{sub 2}-Al fuel. (author)

  13. Impact of MCNP Unresolved Resonance Probability-Table Treatment on Uranium and Plutonium Benchmarks

    International Nuclear Information System (INIS)

    Mosteller, R.D.; Little, R.C.

    1999-01-01

    A probability-table treatment recently has been incorporated into an intermediate version of the MCNP Monte Carlo code named MCNP4XS. This paper presents MCNP4XS results for a variety of uranium and plutonium criticality benchmarks, calculated with and without the probability-table treatment. It is shown that the probability-table treatment can produce small but significant reactivity changes for plutonium and 233 U systems with intermediate spectra. More importantly, it can produce substantial reactivity increases for systems with large amounts of 238 U and intermediate spectra

  14. Uranium enrichment measurement by X- and γ-ray spectrometry with the 'URADOS' process

    International Nuclear Information System (INIS)

    Morel, Jean; Etcheverry, Michel; Riazuelo, Gilles

    1998-01-01

    The methods used for the uranium enrichment measurement require in general prior instrument calibration with several standards. Thus, it is possible to avoid the constraints involved in calibration by considering the complex spectral region called XK α . This spectral region is sufficiently limited so that the variation of the detector efficiency response is small enough to facilitate a self-calibration. Processing this region is critical and requires taking into account 3 elemental images, one corresponding to 235 U, one to 238 U and one to the X-ray fluorescence induced in the sample by radiation above 100 keV. A process called 'URADOS' based on this principle has been developed. Six uranium oxide standards with different enrichments and infinite thicknesses were counted several times to test this process; other samples, some highly enriched, were also used. The results obtained are compared to the declared values. From these measurements, it has been possible to improve the photon emission probability values

  15. Closedown programme for the uranium ore and processing plant at Eleshnitsa, Bulgaria

    International Nuclear Information System (INIS)

    Christiansson, R.; Hedman, K.; Oosterbaan, A.; Popov, I.

    1998-01-01

    The tailings pond of the uranium ore and resin processing plant at Eleshnitsa, Bulgaria has been subject to a rehabilitation programme under a phare contract. The tailings cover 34 ha behind a 70 m high dam. The volume of waste is 12 million m 3 . Furhermore, there are considerable areas with contaminated soils within the industrial area. In order to be able to prepare detailed design and full tender documents for the actual contractual works a number of additional studies has been performed. The studies include geophysical, geodetical and geotechnical surveys, piezometer installations, a sub-regional sampling survey of surface and groundwater and a laboratory scale water treatment test. From these studies final design data and parameters have been obtained with respect to long term dam-stability, composition of cover and shape of contouring of the tailings pond area, structures for diversion of upstream surface waters and size and type of waste water treatment plant. Based on this information are contract Dossiers prepared The main part of this paper deals with the rehabilitation concept to be used for the contractual works. A description of present-day conditions, starting point for the rehabilitation concept, is also presented. Specific emphasis is put on the long-term stability of the dam and covering of the tailings by a soil membrane, on the design of the waste water treatment plant and on the monitoring programmes to be put into place. Besides above mentioned more technical and environmental aspects of the closedown programme a preliminary planning as well as cost-estimates for the different contractual works are presented. (orig.)

  16. Czechoslovak uranium

    International Nuclear Information System (INIS)

    Pluskal, O.

    1992-01-01

    Data and knowledge related to the prospecting, mining, processing and export of uranium ores in Czechoslovakia are presented. In the years between 1945 and January 1, 1991, 98,461.1 t of uranium were extracted. In the period 1965-1990 the uranium industry was subsidized from the state budget to a total of 38.5 billion CSK. The subsidies were put into extraction, investments and geologic prospecting; the latter was at first, ie. till 1960 financed by the former USSR, later on the two parties shared costs on a 1:1 basis. Since 1981 the prospecting has been entirely financed from the Czechoslovak state budget. On Czechoslovak territory uranium has been extracted from deposits which may be classified as vein-type deposits, deposits in uranium-bearing sandstones and deposits connected with weathering processes. The future of mining, however, is almost exclusively being connected with deposits in uranium-bearing sandstones. A brief description and characteristic is given of all uranium deposits on Czechoslovak territory, and the organization of uranium mining in Czechoslovakia is described as is the approach used in the world to evaluate uranium deposits; uranium prices and actual resources are also given. (Z.S.) 3 figs

  17. Potential radiological impacts of recovery of uranium from wet-process phosphoric acid. Final report to the Environmental Protection Agency

    International Nuclear Information System (INIS)

    Davis, W. Jr.; Haywood, F.F.; Danek, J.L.; Moore, R.E.; Wagner, E.B.; Rupp, E.M.

    1979-01-01

    A study was made to determine the radiological impacts associated with recovery of uranium from wet-process (WP) phosphoric acid in central Florida. Removal of U and other radionuclides from phosphoric acid prevents their distribution on farm lands and urban gardens and grasses via fertilizers; this results in a positive impact (decreased dose commitment) on the associated populations. This study considers the potential negative impacts of current and project recovery processes in a site-specific manner using detailed state-of-the-art methodologies. Positive impacts are treated in a generic sense using U.S. average values for important variables such as average and maximum fertilizer application rates and quantities of radionuclides in fertilizer. Three model plants to recover U from WP phosphoric acid were selected and source terms for release of radionuclides are developed for all three and for two treatment methods for airborne particulates. Costs for radwaste treatment were developed. Field measurements were conducted at the only commercial uranium recovery plant in operation. Radiological doses to the population surrounding release points during plant operation were estimated

  18. Machining of uranium and uranium alloys

    International Nuclear Information System (INIS)

    Morris, T.O.

    1981-01-01

    Uranium and uranium alloys can be readily machined by conventional methods in the standard machine shop when proper safety and operating techniques are used. Material properties that affect machining processes and recommended machining parameters are discussed. Safety procedures and precautions necessary in machining uranium and uranium alloys are also covered. 30 figures

  19. Optimization and validation of a chemical process for uranium, mercury and cesium leaching from cemented radioactive wastes

    International Nuclear Information System (INIS)

    Reynier, N.; Riveros, P.; Lastra, R.; Laviolette, C.; Bouzoubaa, N.; Chapman, M.

    2015-01-01

    Atomic Energy of Canada Limited (AECL) is developing a treatment and long-term management strategy for a legacy cemented radioactive waste that contains uranium, mercury and fission products. Extracting the uranium would be advantageous for decreasing the waste classification and reducing the cost of long-term management. Consequently, there are safety and economic and environmental incentives for the extraction of uranium, mercury and cesium before subjecting the cemented waste to a stabilization process. The mineralogical analysis of the surrogate cemented waste (SCW) indicated that uranium forms calcium uranate, CaUO 4 , occurring as layers of several millimeters or as grains of 20 μm. Hg is found mostly as large (∼50 μm) and small grains (5-8 μm) of HgO. The chemical leachability of three key elements (U, Hg, and Cs) from a SCW was studied with several leaching materials. The results showed that the most promising approach to leach and recover U, Hg, and Cs is the direct leaching of the SCW with H 2 SO 4 in strong saline media. Operating parameters such as particle size, temperature, pulp density, leaching time, acid and salt concentrations, number of leaching/rinsing step, etc. were optimized to improve key elements solubilization. Sulfuric leaching in saline media of a SCW (U5) containing 1182 ppm of U, 1598 ppm of Hg, and 7.9 ppm of Cs in the optimized conditions allows key elements recovery of 98.5 ± 0.4%, 96.6 ± 0.1%, and 93.8 ± 1.1% of U, Hg, and Cs respectively. This solubilization process was then applied in triplicate to seven other SCW prepared with different cement, liquid ratio and at different aging time and temperature. Concentrated sulfuric acid is added to the slurry until the pH is about 2, which causes the complete degradation of cement and the formation of CaSO 4 . At this pH, the acid consumption is moderate and the formation of amorphous silica gel is avoided. Sulfuric acid is particularly useful because it produces a leachate that

  20. Initial cathode processing experiences and results for the treatment of spent fuel

    International Nuclear Information System (INIS)

    Westphal, B.R.; Laug, D.V.; Brunsvold, A.R.; Roach, P.D.

    1996-01-01

    As part of the spent fuel treatment demonstration at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a salt electrolyte, primarily consisting of a eutectic mixture of lithium and potassium chlorides, from uranium is achieved by a batch operation termed ''cathode processing.'' Cathode processing is performed in a retort furnace which enables the production of a stable uranium product that can be isotopically diluted and stored. To date, experiments have been performed with two distillation units; one for prototypical testing and the other for actual spent fuel treatment operations. The results and experiences from these initial experiments with both units will be discussed as well as problems encountered and their resolution

  1. Stake holder involvement in the Canadian review process for uranium production projects in Northern Saskatchewan

    International Nuclear Information System (INIS)

    Underhill, D.

    2004-01-01

    This report describes the Canadian environmental review process for uranium production projects as a case study for the purpose of understanding the nature and value of stakeholder involvement in the management of radiological hazards. While the Canadian review process potentially applies to any development, this case study focuses on the assessment of the uranium projects of northern Saskatchewan conducted during the 1990's. It describes the environmental assessment (EA) conducted in the 1990's for six new uranium facilities (including mines and mills and related tailings disposal sites) planned in northern Saskatchewan. Both the Canadian federal and the Saskatchewan provincial government have extensive environmental review processes that must under law be complete before any major industrial development judged to have potential environmental impacts is undertaken within their respective territories. However, even in those instances where no clear potential environmental impacts are evident, Canadian law mandates 'if public concern about the proposal is such that a public review is desirable, the initiating department shall refer the proposal to the Minister for review by a Panel'. (Wh95) As a stakeholder under law, in both Canada and Saskatchewan, the public plays an important role in the environmental review process. To encourage participation and assist the public in its review the two governments may provide funding (as done in this review) to assist qualified individuals or groups to participant in the review process. The first section of this case study sets the scene. It describes the Saskatchewan uranium mining story, focusing on how the importance of the public stakeholder evolved to become a major component, under law, in the EA process for new uranium mines. This increase in stakeholder involvement opportunities coincided with heightened public concern for the socio-economic impacts of the projects. In the late 1980's both governments were advised by

  2. A non-pedological hypothesis for the processes of uranium mineralization in calcrete

    International Nuclear Information System (INIS)

    Briot, P.; Fuchs, Y.

    1984-01-01

    The non-pedological hypothesis presented for the origin of the uraniferous calcrete deposits in Western Australia is based on the premise that alluvial and calcareous lacustrine sediments were initially formed during earlier wet periods, evidence for which has been found in the fossil records. These were followed by subsequent epigenetic alteration accompanied by the precipitation of uranium mineralization during drier semi-arid periods. Typical examples of the processes involved were found in the Yeelirrie uranium deposit. During the latter semi-arid period, the limited surface flow which consisted of periodic flash flood conditions probably contributed marginally to the recharge of the groundwater, and consequently, semi-stagnant groundwater conditions evolved, particularly where the hydraulic gradient was extremely small, for example, for the Yeelirrie channel it is approximately 0.001. In addition, ponding of water behind a natural barrier caused the groundwater to evolve along the following geochemical sequence: mild alkalinity, weak oxidizing conditions, and oversaturation in dissolved elements. These hydrological and hydrogeochemical conditions induced the epigenetic alteration of the palustral/lacustrine limestone, bringing about dolomite neogenesis and the precipitation of carnotite. The source of the uranium in the calcretes and the groundwater of the Yeelirrie channel is considered to be the weathered outcrops of the breakaways along its margins. The genetic hypothesis proposed in this paper, although somewhat different from those described previously and elsewhere in this volume, could be applied to the other uranium-bearing calcretes in Mauritania, Namibia, and Somalia

  3. Guidebook on the development of projects for uranium mining and ore processing

    International Nuclear Information System (INIS)

    1991-04-01

    Bringing a uranium operation into production involves a sequence of interrelated steps. These are outlined in the simplified diagram of Fig. 1. The challenge is to determine how the various steps of the development sequence should function and whether the costs are sufficiently low to return a positive benefit to the owner. This Guidebook has been prepared to aid in the planning, development and implementation of feasible uranium projects. It is one in a series of publications by the IAEA. This guidebook is essentially the executive summary of the other publications. It is an overview of the systematic approach to project development. It might be viewed as the ''road map'' of a project. A list of other publications in this series is provided in the Bibliography. Each chapter of the Guidebook addresses a critical aspect of project development. Chapters follow a general sequence, but none should be considered in isolation. Each Chapter presents an overview of the requirements for reaching decisions necessary to advance a project. References are provided to more definitive information and to documents which will be required by technical personnel on a project. Such detailed publications include IAEA books such as ''An Instruction Manual on Methods for Estimation of Uranium Ore Reserves'', and the ''Significance of Mineralogy in the Development of Flow Sheets for Processing Uranium Ores''. This Guidebook does not detail how to do project development but rather what must be done to insure that all critical elements of a project are considered. Refs, figs and tabs

  4. Process evaluation for treatment of aluminium bearing declad waste

    International Nuclear Information System (INIS)

    Banerjee, D.; Rao, Manjula A.; Srinivas, C.; Wattal, P.K.

    2012-01-01

    Declad waste generated by the process of chemical decladding of Al-cladded uranium metal fuel is characterized by highly alkaline, high Al bearing intermediate level waste. It was found that the process developed and adopted in India for plant scale treatment of alkaline intermediate level waste (ILW) is unsuitable for treatment of declad waste. This is mainly due to its exotic characteristics, notably substantial amounts of aluminium in the declad waste. As part of development of treatment scheme for this waste, 137 Cs removal by RFPR has been demonstrated earlier and the present paper reports the results of further processing of the Cs-lean effluent. The waste simulated with respect to the major chemical constituents of stored Al-bearing alkaline ILW after 137 Cs and 90 Sr removal by ion exchange, is used in this study

  5. Demonstrations of video processing of image data for uranium resource assessments

    International Nuclear Information System (INIS)

    Marrs, R.W.; King, J.K.

    1978-01-01

    Video processing of LANDSAT imagery was performed for nine areas in the western United States to demonstrate the applicability of such analyses for regional uranium resource assessment. The results of these tests, in areas of diverse geology, topography, and vegetation, were mixed. The best success was achieved in arid areas because vegetation cover is extremely limiting in any analysis dealing primarily with rocks and soils. Surface alteration patterns of large areal extent, involving transformation or redistribution of iron oxides, and reflectance contrasts were the only type of alteration consistently detected by video processing of LANDSAT imagery. Alteration often provided the only direct indication of mineralization. Other exploration guides, such as lithologic changes, can often be detected, even in heavily vegetated regions. Structural interpretation of the imagery proved far more successful than spectral analyses as an indicator of regions of possible uranium enrichment

  6. Finding of No Significant Impact, proposed remediation of the Maybell Uranium Mill Processing Site, Maybell, Colorado

    International Nuclear Information System (INIS)

    1995-01-01

    The U.S. Department of Energy (DOE) has prepared an environmental assessment (EA) (DOE/EA-0347) on the proposed surface remediation of the Maybell uranium mill processing site in Moffat County, Colorado. The mill site contains radioactively contaminated materials from processing uranium ore that would be stabilized in place at the existing tailings pile location. Based on the analysis in the EA, DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, Public Law 91-190 (42 U.S.C. section 4321 et seq.), as amended. Therefore, preparation of an environmental impact statement is not required and DOE is issuing this Finding of No Significant Impact (FONSI)

  7. Processing of Indian monazite for the recovery of thorium and uranium values

    International Nuclear Information System (INIS)

    Mukherjee, T.K.

    2004-01-01

    The mineral monazite, a phosphate of rare earths and thorium with significant quantity of uranium is one of the six heavy minerals present in the beach sands of specific coastal areas of India. Indian Rare Earths Ltd is mining and processing monazite at its Rare Earths Division for the last many decades with an aim of building up enough stock of thorium concentrate for its future use in the three stage nuclear power programme of the country. The present paper briefly describes the monazite resource position of he country, the past and present modified processing schemes and the future programme commensurate with the requirement of the country for quality thorium and uranium bearing nuclear materials

  8. Fluorinated compounds in the uranium conversion process: risk analysis and proposition of pictograms

    International Nuclear Information System (INIS)

    Jeronimo, Adroaldo Clovis; Oliveira, Wagner dos Santos

    2012-01-01

    In the process of uranium hexafluoride production there are risks that must be taken into account since the time of completing the project chemist, in its conceptual stage, until to the stage of detailed design and are associated with the handling of chemicals, especially fluoride hydrogen and fluorine. This paper aims to address issues related to the prevention of risks related to industrial safety and health and the environment, considering the different stages of the uranium conversion. Take into account the safety warnings of the plant and, accordingly, make the proposition of pictograms adequate to alert operators of care to be taken during the proposition of pictograms adequate to alert operators of care to be taken during the conduct of these chemical processes. (author)

  9. Research and economic evaluation on uranium enrichment by gaseous diffusion process in Japan

    International Nuclear Information System (INIS)

    Aochi, T.; Takahashi, S.

    1977-01-01

    Research and development works on uranium enrichment by gaseous diffusion process were carried out by JAERI, IPCR and industries since 1965. There are two important keys to reduce the uranium separation cost. One is the characteristics of the barrier and the other is financing and/or political planning. The technics to prepare the barrier with pore diameter of 40A have been developed with polytetrafluoroethylene, alumina and nickel. The experiment on corrosion behavior of PTFE barriers has shown better characteristics than the others. In the field of engineering research, the adiabatic efficiency of axial compressor for UF 6 was resulted to as high as 90% by long term operation tests. Based on these experimental data, techno-economic evaluation on a uranium enrichment plant was carried out with regard to the optimization of separation efficiency, numbers of step and operating conditions of the plant. Sensitivity in the separation cost were calculated as a function of pore diameter, uranium hexafluoride cost, plant capacity, electric power cost, and the plant annual expenditure. A financing plan must be such as to achieve 1. maximization of debt in a percentage of total capitalization 2. off-take contracts to utilities as security for financing 3. minimization of risks to equity and achievable cost of capital. Therefore the cash flow analysis and the schedule for construction and operation are very important for a economical feasibility of a uranium enrichment plant. To minimize the risk, not only economical but also political environment are important. The governmental supports and international agreements will be necessary

  10. Idaho Chemical Processing Plant and Plutonium-Uranium Extraction Plant phaseout/deactivation study

    International Nuclear Information System (INIS)

    Patterson, M.W.; Thompson, R.J.

    1994-01-01

    The decision to cease all US Department of Energy (DOE) reprocessing of nuclear fuels was made on April 28, 1992. This study provides insight into and a comparison of the management, technical, compliance, and safety strategies for deactivating the Idaho Chemical Processing Plant (ICPP) at Westinghouse Idaho Nuclear Company (WINCO) and the Westinghouse Hanford Company (WHC) Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this study is to ensure that lessons-learned and future plans are coordinated between the two facilities

  11. Extraction of uranium from wet process phosphoric acid in centrifugal and mixer-settler batteries

    International Nuclear Information System (INIS)

    Poczynajlo, A.; Giers, M.

    1986-01-01

    Five stage countercurrent batteries were comparatively applied for the extraction of uranium from wet phosphoric acid (Chemical Works, Police) in semitechnical scale. As an extractant phase the 0.16 M equimolar solution of mono- and dinonylphenyolphosphoric acids in kerosene was used. The optimum hydrodynamic and extraction conditions for the batteries were found. Process efficiencies of the apparatus were also determined. 5 refs., 5 figs., 2 tabs. (author)

  12. The acid aging as alternative process for uranium recovery from silicated ores

    International Nuclear Information System (INIS)

    Cipriani, M.; Della Testa, A.

    1984-01-01

    The influence of different variables on the extraction uranium efficiency and on the silicate solubility by means of acid aging is studied. The variables studied in bench scale were: acid/ore, oxidizing/ore and liquid/solid relationships; reaction time; temperature and recovery time. The results are discussed and compared with the ones of continuous operation of a semi-pilot plant. A flowsheet of the industrial process application is presented. (M.A.C.) [pt

  13. Pretreatment of phosphoric acid for uranium recovery by the wet phosphoric acid process

    International Nuclear Information System (INIS)

    Chern, S.L.P.; Chen, Y.C.L.; Chang, S.S.H.; Kuo, T.S.; Ting, G.C.M.

    1980-01-01

    The proposal deals with reprocessing of phosphoric acid arising from uranium separation according to the wet phosphoric acid process and being intended for recycling. In detail, the sludge will be removed by means of an inclined separating device containing corrugated plates, then the organic impurities are washed out with kerosene in suitable facilities, and the crude phase remaining in the settling tank will be separated from the kerosene in a separating centrifuge. The method has only got low cost of installation. (UWI) [de

  14. The impact of Canada's environmental review process on new uranium mine developments

    International Nuclear Information System (INIS)

    Whillans, R.T.

    1997-01-01

    Canada introduced and environmental assessment process in the mid 1970s. It was designed to ensure that the environmental consequences of all project proposals with federal government involvement were assessed for potential adverse effects early in the planning stage. In 1984, a Guidelines Order was approved to clarify the rules, responsibilities and procedures of the environmental Assessment and Review Process (EARP) that had evolved informally under earlier Cabinet directives. In 1989/1990, the Federal Court of Appeal effectively converted the Guidelines Order into a legal requirement for rigorous application. The Supreme Court of Canada upheld the constitutionally of the EARP Guidelines Order in 1992. Canada became the world's leading producer and exporter of uranium during the late 1980s. Since then, the Canadian public has become sensitized to numerous issues concerning environmental degradation, from the Chernobyl accident to ozone depletion. In 1991, during this period of increasing awareness, the Atomic Energy Control Board, the federal nuclear regulator, referred six new Saskatchewan uranium mining projects for environmental review, pursuant to the EARP Guidelines Order. The public review process provided an extremely valuable focus on aspects of these developments that needed to be addressed by proponents and regulators. It has helped to demonstrate that new uranium mining projects are being developed in a responsible manner, after full consideration has been given to the potential impacts and public concerns associated with these facilities. 4 figs, 1 tab

  15. Assessment of surface contamination level in an operating uranium ore processing facility of Jaduguda, India

    International Nuclear Information System (INIS)

    Meena, J.S.; Patnaik, R.L.; Jha, V.N.; Sahoo, S.K.; Ravi, P.M.; Tripathi, R.M.

    2014-01-01

    Radiological concern of the occupational workers and the area is given priority over other safety issue in confirmation with the stipulated guideline of national regulatory agency (AERB/FEFCF/SG-2, 2007). The key concern from the radiological hazard evaluation point of view is air activity, external gamma level and surface contamination. Present investigations was carried out to ascertain the surface contamination level of uranium ore processing facility at Jaduguda, Jharkhand. For a low grade uranium ore processing industry surface contamination is a major concern in product precipitation and recovery section. In view of this, the ore processing plant can broadly be classified into three areas i.e. ion exchange area, precipitation and product recovery section and other areas. The monitoring results incorporate the level of surface contamination of the plant during the last five years. The geometric mean activity of surface contamination level was 31.1, 34.5 and 9.8 Bq dm -2 in ion exchange, product precipitation and recovery and other areas with GSD of 2, 2.5 and 1.9. In most of the cases the surface contamination level was well within the recommended limit of 100 Bq dm -2 for M class uranium compound. Occasional cases of surface contamination levels exceeding the recommended limit were addressed and areas were decontaminated. Based on the study, modification in the design feature of the surface of the finished product section was also suggested so that the decontamination procedure can be more effectively implemented

  16. Development of dissolution process for metal foil target containing low enriched uranium

    International Nuclear Information System (INIS)

    Srinivasan, B.; Hutter, J.C.; Johnson, G.K.; Vandegrift, G.F.

    1994-01-01

    About six times more low enriched uranium (LEU) metal is needed to produce the same quantity of 99 Mo as from a high enriched uranium (HEU) oxide target, under similar conditions of neutron irradiation. In view of this, the post-irradiation processing procedures of the LEU target are likely to be different from the Cintichem process procedures now in use for the HEU target. The authors have begun a systematic study to develop modified procedures for LEU target dissolution and 99 Mo separation. The dissolution studies include determination of the dissolution rate, chemical state of uranium in the solution, and the heat evolved in the dissolution reaction. From these results the authors conclude that a mixture of nitric and sulfuric acid is a suitable dissolver solution, albeit at higher concentration of nitric acid than in use for the HEU targets. Also, the dissolver vessel now in use for HEU targets is inadequate for the LEU target, since higher temperature and higher pressure will be encountered in the dissolution of LEU targets. The desire is to keep the modifications to the Cintichem process to a minimum, so that the switch from HEU to LEU can be achieved easily

  17. Demographic studies of Sherpalle area, the proposed site for Uranium Processing Plant in Nalgondo district, Andhra Pradesh

    International Nuclear Information System (INIS)

    Padmaja, S.; Pavanaguru, R.; Venugopal Reddy, K.; Yadagiri, G.; Chougaonkar, M.P.

    2013-01-01

    Availability of nuclear fuel, in the wake of over stress on other power resources, for continuous production of nuclear energy is a crucial and essential factor. Uranium Corporation of India Ltd. (UCIL) is undertaking mining and processing of uranium ore on large scale and it is expanding its operation in the Nalgonda district of AP, which is endowed with huge uranium deposits. To initiate the continuous operation of mining processes, it is essential and prime requisite to generate baseline demographic data which can be compared to both past and future date to identify changes that may result due to mining operations

  18. Role of thermal analysis in uranium oxide fuel fabrication process

    International Nuclear Information System (INIS)

    Balaji Rao, Y.; Yadav, R.B.

    2006-01-01

    The present paper discusses the application of thermal analysis, particularly, differential thermal analysis (Dta) at various stages of fuel fabrication process. The useful role of Dta in knowing the decomposition pattern and calcination temperature of Adu along with de-nitration temperature is explained. The decomposition pattern depends upon the type of drying process adopted for wet ADU cake (ADU C). Also, the paper highlights the utility of DTA in determining the APS and SSA of UO 2+x and U 3 O 8 powders as an alternate technique. Further, the temperature difference (ΔT max ) between the two exothermic peaks obtained in UO 2+x powder oxidation is related to sintered density of UO 2 pellets. (author)

  19. Data processing in management of Dolni Rozinka uranium mines

    International Nuclear Information System (INIS)

    Benes, B.

    1987-01-01

    In 1985, a qualitative inovation was introduced of data processing by the commissioning of the EC 1026 computer with a terminal network and a remote data communication system. The design jobs which are being gradually implemented are mainly oriented to the creating of an automated information system for operative control of mining production, data preparation in mining plants, and to the personnel, wages, material consumptions, etc. areas. (J.B.)

  20. Application of alkaline leaching to the extraction of uranium from shale of the Vosges

    International Nuclear Information System (INIS)

    Mouret, P.; Pottier, P.; Le Bris, J.

    1958-01-01

    Description of chemical treatment of Vosges shales to obtain uranium by alkaline leaching. Mineralogy aspects of ore, physical and chemical conditions of leaching, solid/liquid separation, uranium recovery by either ion exchange process or electrolytic precipitation. (author) [fr