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AbstractAbstract
[en] 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)
Primary Subject
Source
25 Jun 1980; 1 p; GB PATENT DOCUMENT 2035977/A/
Record Type
Patent
Country of publication
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Fuller, E.F.
Exxon Research and Engineering Co., Florham Park, NJ (USA)
Exxon Research and Engineering Co., Florham Park, NJ (USA)
AbstractAbstract
[en] Dissolved metal values are extracted from a liquid e.g. uranium from phosphoric acid by contacting the liquid with agglomerates for a time to load the agglomerate with the metal value, separating the loaded agglomerates from the liquid phase and stripping the metal value from the loaded agglomerate. The agglomerate may be made by combining finely divided solid particles with a binding liquid to form a paste, adding a suspending liquid to form a mixture, the suspending liquid and binding liquid being immiscible in each other and the solid particles being insoluble in the suspending liquid and shearing the mixture to form the agglomerate. (author)
Primary Subject
Source
6 Aug 1980; 11 p; GB PATENT DOCUMENT 2039268/A/
Record Type
Patent
Country of publication
ACTINIDE COMPLEXES, ACTINIDE COMPOUNDS, ACTINIDES, AROMATICS, COMPLEXES, ELEMENTS, ESTERS, HYDROCARBONS, HYDROGEN COMPOUNDS, INORGANIC ACIDS, METALS, ORGANIC COMPOUNDS, ORGANIC PHOSPHORUS COMPOUNDS, ORGANIC POLYMERS, OXYGEN COMPOUNDS, PHOSPHORIC ACID ESTERS, PHOSPHORUS COMPOUNDS, POLYMERS, POLYOLEFINS, POLYPHENYLS, POLYVINYLS, TERPHENYLS, URANIUM COMPLEXES, URANIUM COMPOUNDS
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AbstractAbstract
[en] Steps in the typical pre-conceptual engineering study are: problem identification, identify alternatives, evaluate alternatives, and recommend solution. Three examples of the use of pre-conceptual approach to save money are given; they include the water supply for fire protection in the 200-West Area of Hanford, emergency power for the Purex plant, and new filter for the Purex plant canyon exhaust
Primary Subject
Source
1981; 14 p; Civil engineering department seminar; Seattle, WA, USA; 1 Jun 1981; CONF-8106124--1; Available from NTIS., PC A02/MF A01
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
Publication YearPublication Year
Reference NumberReference Number
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INIS IssueINIS Issue
AbstractAbstract
[en] A method of fabricating fuel pellets includes apparatus for compacting, granulating, classifying, lubricating and pelletizing the materials is described. The components are arranged below one another, allowing gravity flow, but are offset relative to the preceding stage so that overhead access is provided without interference. (U.K.)
Primary Subject
Source
29 Apr 1981; 9 p; GB PATENT DOCUMENT 1588606/A/
Record Type
Patent
Country of publication
Publication YearPublication Year
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
Bae, K.S.; Chung, K.S.; Yang, H.B.; Kim, B.W.
Korea Advanced Energy Research Inst., Seoul (Republic of Korea)
Korea Advanced Energy Research Inst., Seoul (Republic of Korea)
AbstractAbstract
[en] Uranium was extracted with amines dissolved in diluents. Using alamine 336 as an extractant, dodecane as a diluent, and n-octanol as an additive for third phase inhibition, UO2++ extraction from H2SO4-Na2SO4 aqueous solutions was studied. When the extraction coefficient of uranium was plotted as a function of the initial Na2SO4 concentration at a constant sulfuric acid concentration, the typical initial spike type extraction curve was obtained in this case too. Thus, at low Na2SO4 concentrations, a reaction such as UO2SO4 + 2 (AlamineH)2SO4 reversible (AlamineH)4UO2(SO4)3...(1) is believed to take place, while at high Na2SO4 concentrations reaction (2) is thought to be occurring. UO2(SO4)34- + 2(R3NH)2SO4 reversible (R3NH)4UO2(SO4)3 = 2SO4-2...(2). The same result obtained with Na2SO4 as with H2SO4 is explained by the action of sulfate ions
Primary Subject
Source
1980; 52 p
Record Type
Report
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Publication YearPublication Year
Reference NumberReference Number
INIS VolumeINIS Volume
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AbstractAbstract
[en] A method for loading ion exchange resins is described. The process comprises contacting a weak acid cation exchange resin in the ammonium form with a uranyl fluoride salt solution
Original Title
Patent
Primary Subject
Source
30 Nov 1976; 6 p; US PATENT DOCUMENT 3,995,009/A/; PAT-APPL-613,704.
Record Type
Patent
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Reference NumberReference Number
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AbstractAbstract
[en] This document contains responses by the NRC Task Force to comments received on the report ''Environmental Survey of the Reprocessing and Waste Management Portions of the LWR Fuel Cycle'' (NUREG-0116). These responses are directed at all comments, inclding those received after the close of the comment period. Additional information on the environmental impacts of reprocessing and waste management which has either become available since the publication of NUREG-0116 or which adds requested clarification to the information in that document
Primary Subject
Secondary Subject
Source
Mar 1977; 425 p; Available from NTIS. $11.75
Record Type
Report
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Schilling, T.
Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung; Technische Hochschule Aachen (Germany, F.R.)
Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung; Technische Hochschule Aachen (Germany, F.R.)
AbstractAbstract
[en] In case of an enhanced utilisation of nuclear energy from fission reactors it is necessary to preserve the natural deposits of fissile material. One possibility for this preservation could be the breeding of synthetic fuel by the means of spallation neutrons. The configuration discussed here consists of an integral spallation target/breeding blanket pebble bed of hollow Thorium spheres, which is cooled alternatively with helium gas or liquid sodium or lithium. With helium cooling the neutron yield is 35 neutrons par initial proton and the breeding gain is 775 kg U-233/a. With sodium as coolant the n-yields is 25 n/p, and the fuel production is 520 kg U-233/a. If liquid lithium is used as coolant the neutron yield is 30 n/p with a production of 440 kg Uranium-233 per year. Supposing a fuel requirement of a high converting HTR of 90 kg U-233/GW-a, 5 to 9 1-GWsub(e)-reactors could be supplied depending on the coolant. (orig.)
[de]
Fuer den Fall einer verstaerkten Nutzung der Kernenergie aus Fission-Reaktoren sind Massnahmen zur Schonung der natuerlichen Spaltstoffreserven erforderlich. Eine Moeglichkeit hierzu ist das Erbrueten synthetischen Spaltstoffs mit Hilfe von Neutronen aus Spallationsprozessen. Das hier dargestellte Konzept besteht aus einer integrierten Spallationstarget/Brutblanket-Kugelhaufenanordnung mit einer Schuettung aus Thorium Hohlkugeln, welche mit gasfoermigem Helium oder alternativ mit fluessigem Natrium bzw. Lithium gekuehlt wird. Fuer den Fall der Heliumkuehlung ergibt sich eine Neutronenausbeute von 35 Neutronen pro Initialproton und eine Brutrate von 775 kg U-233/a, waehrend bei Natriumkuehlung die Neutronenausbeute 26 Neutronen pro Proton betraegt, womit 520 kg U-233/a erzeugt werden koennen. Bei der Verwendung von Lithium als Kuehlmedium betraegt die Neutronenausbeute 30 n/p, was zu einer Spaltstoffproduktion von 440 kg U-233/a fuehrt. Legt man fuer einen hochkonvertierenden Kugelhaufen-HTR einen Spaltstoffbedarf von 90 kg U-233/GWsub(e)a zugrunde, so koennen mit einer solchen Anlage je nach Kuehlmittel zwischen 5 und 9 1-GW3-Reaktoren versorgt werden. (orig.)Original Title
Neutronenphysikalische und thermohydraulische Untersuchungen zur Spaltstoffproduktion mit Spallationsneutronen
Primary Subject
Secondary Subject
Source
Mar 1983; 67 p; Diss.
Record Type
Report
Literature Type
Thesis/Dissertation
Report Number
Country of publication
ACTINIDE NUCLEI, ACTINIDES, ALKALI METALS, ALPHA DECAY RADIOISOTOPES, ELEMENTS, EVEN-ODD NUCLEI, FLUIDS, GAS COOLED REACTORS, GRAPHITE MODERATED REACTORS, HEAVY NUCLEI, HOMOGENEOUS REACTORS, ISOTOPES, LIQUIDS, MATERIALS, METALS, NUCLEAR FUEL CONVERSION, NUCLEAR REACTIONS, NUCLEI, RADIOISOTOPES, REACTORS, SOLID HOMOGENEOUS REACTORS, URANIUM ISOTOPES, YEARS LIVING RADIOISOTOPES
Publication YearPublication Year
Reference NumberReference Number
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INIS IssueINIS Issue
de Marteau, J.M.; Sorale, S.
Rockwell International Corp., Golden, CO (USA). Rocky Flats Plant
Rockwell International Corp., Golden, CO (USA). Rocky Flats Plant
AbstractAbstract
[en] A process is claimed for recovering uranium from a uranium-containing acid leach solution, consisting of extraction of the uranium leach solution with an organic extractant solution containing a pyrophosphoric dialkyl ester, a diluent and a solubilizing agent for uranium-ammonium pyrophosphoric dialkyl ester double salt in at least a proportion of 0.5 mole per mole of the above pyrophosphoric dialkyl ester
Primary Subject
Source
Jul 1983; 8 p; Available from NTIS, PC A02/MF A01; 1 as DE83015028
Record Type
Report
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Lavrinovich, Yu.G.; Kirillovich, A.P.; Vorobe , M.P.; Pimonov, Yu.I.; Gazizov, R.K.
Nauchno-Issledovatel'skij Inst. Atomnykh Reaktorov, Dimitrovgrad (USSR)
Nauchno-Issledovatel'skij Inst. Atomnykh Reaktorov, Dimitrovgrad (USSR)
AbstractAbstract
[en] Results of investigation into certain physicochemical and radiation properties of fluoride wastes resulted from experimental processing of irradiated uranium oxide and uranium-plutonium fuel, as well as behaviour of highly active wastes in the process of their 6-year controlled storage, are presented. Yield of radionuclides into gaseous phase, its composition, leaching rate, thermal and radiation stability, heat conductivity, heat release, compatibility of wastes with structural materials, are determined. An increase in the waste storage time is shown to increase their thermal and radiation stability. Recommendations for the waste storage are given
Original Title
Radiatsionnye, teplofizicheskie i fiziko-khimicheskie svojstva otkhodov, poluchennykh pri opytnoj pererabotke obluchennogo topliva reaktora BOR-60 gazoftoridnym metodom
Primary Subject
Source
1982; 27 p; 11 refs.; 4 figs.; 7 tabs.
Record Type
Report
Report Number
Country of publication
ACTINIDE COMPOUNDS, CHALCOGENIDES, ENERGY SOURCES, EXTRACTIVE METALLURGY, FUELS, MANAGEMENT, MATERIALS, METALLURGY, NUCLEAR FACILITIES, NUCLEAR FUELS, OXIDES, OXYGEN COMPOUNDS, PLUTONIUM COMPOUNDS, PYROMETALLURGY, RADIOACTIVE MATERIALS, REACTOR MATERIALS, REPROCESSING, SEPARATION PROCESSES, STORAGE, TRANSURANIUM COMPOUNDS, URANIUM COMPOUNDS, WASTE MANAGEMENT, WASTE PROCESSING, WASTE STORAGE, WASTES
Publication YearPublication Year
Reference NumberReference Number
INIS VolumeINIS Volume
INIS IssueINIS Issue
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