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Browsing Scientific and Technical Reports by Author "Alfredson, PG"
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- ItemAnalysis of dilute natural uranium solutions by gamma-ray excited x-ray fluorescence(Australian Atomic Energy Commission, 1974-06) Ryan, RK; Ridley, JL; Alfredson, PGApplication of a gamma-ray excited X-ray fluorimeter to batch analysis of uranium solutions containing less than 10 g ℓ-1, and particularly of dilute impure solutions, is described. An outline is given of design details of the gamma-ray excited X-ray source and of the comparison of backscattered incident X-rays and fluorescent X-rays as a means of compensating for variations in the matrix solution. For pure solutions the lower limit of detection was 0.03 g ℓ-1, compared with 0.05 g ℓ-l for typical impure solutions. In the range 0.1 to 10 g ℓ-1, an accuracy of ± 5 per cent was obtained, while in the range 0.05 to 0.1 g ℓ-1, the accuracy was ± 30 per cent.
- ItemDevelopment of processes for pilot plant production of purified uranyl nitrate solutions(Australian Atomic Energy Commission, 1975-01) Alfredson, PG; Charlton, BG; Ryan, RK; Vilkaitis, VKNuclear purity uranyl nitrate solutions were produced from Rum Jungle yellow cake by dissolution in nitric acid and purification by solvent extraction with 20 vol.% tributyl phosphate in kerosene using pump -mix mixer-settler contactors. The design of the equipment, experimental studies and operating experience are described. Dissolution of yellow cake and recycled uranium oxide materials was readily carried out in a 100 ℓ dissolver to give solutions containing 300 gU ℓ -1 and 0.5 to 4 П nitric acid. Filtration of silica from this solution prior to solvent extraction was not necessary in this work for yellow cake containing 0.25 per cent silica. A low acid flowsheet for uranium purification was developed in which the nitric acid consumption was reduced by 76 per cent and the throughput of the mixer-settler units was increased by 67 per cent compared with the initial design flowsheet. Nine extraction and seven scrubbing stages were used with a feed solution containing 300 gU ℓ -1 and 1.0 П nitric acid and with a portion of the product recycled as scrub solution. The loaded organic phase was stripped in 16 stages with 0.05 П nitric acid heated to 60º C to give a 120 gU ℓ -1 product. The uranium concentration in the raffinate was < 0.04 g ℓ-1, corresponding to ~ 0.01 per cent of the feed.
- ItemInfluence of precipitation conditions on the properties of ammonium diuranate and uranium dioxide powders(Australian Atomic Energy Commission, 1971-05) Janov, J; Alfredson, PG; Vilkaitis, VKA comprehensive investigation of the factors affecting the properties of ADU precipitates in relation to the properties of the subsequent UO2 powders in pellet fabrication is reported and the importance of precipitation parameters is demonstrated. Variables investigated include continuous single-versus two-stage precipitation, pH, residence time, washing of ADU to remove nitrate, and calcination-reduction conditions. The most important variable was the pH at which precipitation occurred. In particular, this governed the size of agglomerate which determined the settling and filtering characteristics of the ADU slurry. In two-stage precipitation, the ADU properties were determined by the proportion of uranium precipitated at different pH values.
- ItemInvestigation of batch-tray calcination-reduction of ammonium diuranate to uranium dioxide(Australian Atomic Energy Commission, 1971-08) Alfredson, PG; Janov, JProcess cycles have been developed on the half-kilogram scale for the conversion of ADU to UO2 powder of specified surface area in the range 3-10 m2/g. The recommended cycle involves calcination of ADU in nitrogen during heating to the reduction temperature, followed by reaction with 30 volume per cent hydrogen in nitrogen and stabilisation at ambient temperature with 2 volume per cent oxygen in nitrogen. UO2 surface area increased from 3 to 10 m2/g as the reduction temperature changed from 700 to 500°C, but was not sensitive to the surface area of the precursor ADU or the hydrogen flow rate.
- ItemLiquid wastes from mining and milling of uranium ores - a laboratory study of treatment methods(Australian Atomic Energy Commission, 1976-10) Ryan, RK; Alfredson, PGMethods of reducing the concentration of contaminants in mine water and in the acidic raffinate from uranium milling operations have been studied. Lime, limestone, caustic soda and lime-soda ash mixtures were compared as reagents for neutralising raffinates and for removing amines and heavy metals including radium from solution. All methods of neutralisation reduced contaminant levels significantly. Two-stage neutralisation using limestone in the first stage to pH 4, followed by second stage lime treatment appears to be an economically attractive approach. This method usually gave the lowest residual radium concentration provided the solids from the first stage were not removed before adding lime. Radium can be further removed from neutralised raffinates or from mine water conditioned with sulphate by the addition of barium chloride to co-precipitate the sulphates of barium and radium. The concentration of radium was readily reduced to less than 3 pCi £-1 by adding 10 mg Ba £-1 raffinate. For mine waters conditioned to 0.01 M in sulphate, barium additions of 20 mg £-l were required to attain the same radium concentrations. Adsorption on barytes was also effective in removing radium from conditioned mine water and neutralised raffinates.
- ItemPapers presented to the AAEC symposium on uranium processing, Lucas Heights, 20-21 July, 1972(Australian Atomic Energy Commission, 1972-09) Miles, GL; Harltey, FR; Butler, RD; Henley, KJ; Cooper, RS; Kelly, A; Goldney, LH; Canning, RG; Gooden, JEA; Baillie, MG; Thomas, JA; Hardy, CJ; Alfredson, PG; Costello, JM; Silver, JM; Richmond, MReview of nuclear fuel cycles and world trends; Conventional processes to produce yellow cake; Carbonate leaching of uranium ores, a review; The application of mineralogy to uranium ore processing; Extraction investigations with some Australian uranium ores; Planned changes in the Mary Kathleen treatment plant for future operations; Possible trends and methods for the production of high purity products; Review of methods and technology for the reproduction of high purity products; Review of methods and technology for the production of uranium hexafluoride; Capital and production costs for the production of uranium hexafluoride; The market for Australian conversion services.
- ItemPerformance of a thermosiphon evaporator for concentration of uranyl nitrate solutions(Australian Atomic Energy Commission, 1973-05) Levins, DM; Alfredson, PGThe performance of a thermosiphon evaporator, with a heat transfer area of 0.74 m2, for the concentration of uranyl nitrate solutions up to 1000 gU ℓ-1 is evaluated. The effects of steam pressure, uranium concentration and liquid submergence on the overall heat transfer coefficient are reported and a method of designing thermosiphon evaporators handling concentrated uranyl nitrate solutions is proposed. Pressure drop characteristics of two packings used in the steam stripping and de-entrainment sections of the evaporator are also given. Overall heat transfer coefficients in the range 1.4 - 2.5 kW m-1 K-1 were obtained at uranyl nitrate concentrations up to 1000 gU ℓ-l. Heat transfer to uranyl nitrate solutions below a concentration of 100 gU ℓ-1 was satisfactorily predicted by employing conventional heat transfer correlations using physical property data for water. Results for more concentrated solutions were successfully correlated using an empirical correction factor which accounts for the change in physical properties of the boiling solution with concentration.
- ItemPilot plant development of processes for the production of ammonium diuranate(Australian Atomic Energy Commission, 1975-01) Janov, J; Alfredson, PG; Vilkaitis, VKNuclear grade ammonium diuranate (ADU) and UO powders were produced on a pilot plant scale by the continuous single-stage precipitation of ADU with ammonium hydroxide, dewatering with a rotary drum vacuum filter or a solid bowl centrifuge, and batch-tray drying and calcination-reduction to UO powder. Precipitation at 50ºC and pH values in the range 7.2 to 8.0 produced ADU materials which could be converted to UO powder by calcination and reduction at temperatures of 600 to 730ºC, and fabricated into sintered pellets with densities of 10.37 to 10.77 g cm-3. The lower the pH of precipitation the lower was the reduction temperature required to achieve a specified pellet density. Precipitation with ammonium hydroxide at 80ºC and with ammonia gas at 50ºC offered no advantages over precipitation with ammonium hydroxide at 50ºC. The UO2 powders and sintered pellets produced from ADU powders precipitated by the three methods were similar. Precipitation at pH 7.5 and 50ºC is recommended since a reasonably filterable precipitate can be produced reproducibly without a need for stringent control, and considerable flexibility is available in the subsequent production of a sinterable UO2 powder. Dewatering of ADU slurries was carried out more efficiently using a solid bowl centrifuge rather than a rotary drum vacuum filter. Clearer discharge liquids were produced at a higher rate of throughput in the solid bowl centrifuge.
- ItemPilot plant development of processes for the production of nuclear grade uranium oxide.(Australian Atomic Energy Commission, 1972-11) Alfredson, PGMost types of nuclear power reactors use fuel in the form of high density-uranium dioxide pellets clad in Zircaloy. Sinterable uranium dioxide powder is usually produced via the ammonium diuranate (ADU) route. This involves dissolution of uranium ore concentrates (yellow cake) in nitric acid, purification by solvent extraction using tributyl phosphate in kerosene, precipitation of ADU, filtration, drying, calcination and reduction with hydrogen to give uranium dioxide powder. The AAEC has carried out pilot plant development of these processes to demonstrate the production of nuclear grade uranium dioxide from Australian yellow cake and to improve the processes and technology wherever possible.
- ItemPreliminary design and cost considerations for a plant to produce nuclear purity uranium dioxide from Australian ore concentrates(Australian Atomic Energy Commission, 1971-03) Charlton, BG; Alfredson, PGDesign considerations are outlined for plants for the production of nuclear purity uranium dioxide with capacities of 100, 200 and 500 tonnes U/year. The cost of the process equipment is not greatly affected by various process alternatives; equipment performance, which affects product quality, consistency of power properties and plant reliability, is important in determining the recommended process. This involves the following steps: batch dissolution, continuous solvent extraction in mixer-settlers, single-stage precipitation of ADU, thickening and spray drying of ADU, and calcination-reduction in continuous pulsed fluidised bed reactors. Estimates of the cost of recovery of free acid and combined nitrate from the raffinate and filtrate waste streams indicated that the value of the recovered acid would be greater that the processing cost only in the case of free acid recovery from solvent extraction raffinate in the 500 tonne/year plant. However if acid recovery is necessary for plant effluent control, the processing cost can be largely offset by the value of the recovered acid.
- ItemPreliminary design and cost considerations for a plant to produce nuclear purity uranium dioxide from Australian ore concentrates(Austtalian Atomic Energy Commission, 1971-03-01) Charlton, BG; Alfredson, PGDesign considerations are outlined for plants for the production of nuclear purity uranium dioxide with capacities of 100, 200 and 500 tonnes U/year. The cost of the process equipment is not greatly affected by various process alternatives; equipment performance, which affects product quality, consistency of power properties and plant reliability, is important in determining the recommended process. This involves the following steps: batch dissolution, continuous solvent extraction in mixer-settlers, single-stage precipitation of ADU, thickening and spray drying of ADU, and calcination-reduction in continuous pulsed fluidised bed reactors. Estimates of the cost of recovery of free acid and combined nitrate from the raffinate and filtrate waste streams indicated that the value of the recovered acid would be greater that the processing cost only in the case of free acid recovery from solvent extraction raffinate in the 500 tonne/year plant. However if acid recovery is necessary for plant effluent control, the processing cost can be largely offset by the value of the recovered acid.
- ItemThe production of sinterable uranium dioxide from ammonium diuranate in a pulsed fluidised bed reactor - interim report(Australian Atomic Energy Commission, 1970-12) Fane, AG; Alfredson, PGResults and operational experience are reported for the batchwise production of uranium dioxide for ammonium diuranate in a pulsed fluidised bed reactor. Alternative proposals for batch/continuous operation are assessed and compared with continuous operation. The future development programme is outlined.
- ItemThe production of sinterable uranium dioxide from ammonium diuranate in a pulsed fluidised bed reactor - interim report(Australian Atomic Energy Commission, 1970-12-01) Fane, AG; Alfredson, PGResults and operational experience are reported for the batchwise production of uranium dioxide from amanium diuranate in a pulsed fluidised bed reactor. Alternative proposals for the batch/continous operation are assessed and compared with continuous operation. The future development programme is outlined.
- ItemRadioactive waste management(Australian Atomic Energy Commission, 1975-08) Alfredson, PG; Levins, DMPresent and future methods of managing radioactive wastes in the nuclear industry are reviewed. In the stages from uranium mining to fuel fabrication, the main purpose of waste management is to limit and control dispersal into the environment of uranium and its decay products, particularly radium and radon. Nuclear reactors produce large amounts of radioactivity but release rates from commercial power reactors have been low and well within legal limits. The principal waste from reprocessing is a high activity liquid containing essentially all the fission products along with the transuranium elements. Most high activity wastes are currently stored as liquids in tanks but there is agreement that future wastes must be converted into solids. Processes to solidify wastes have been demonstrated in pilot plant facilities in the United States and Europe. After solidification, wastes may be stored for some time in man-made structures at or near the Earth's surface. The best method for ultimate disposal appears to be placing solid wastes in a suitable geological formation on land.
- ItemReview of processes for the production of hafnium-free zirconium.(Australian Atomic Energy Commission, 1970-10) Royston, D; Alfredson, PGThe three main industrial processes for the production of hafnium-free zirconium are described in terms of their head-end, zirconium-hafnium separation and zirconium metal forming steps. Possible improvements and alternative processes are outlined. Zirconium-hafnium separation schemes based on selective reduction of the chlorides or distillation and sublimation techniques show the most promise for future development in competition with the established hexone-thiocyanate and TBP-nitric acid solvent extraction schemes. Head-end steps involving direct chlorination of zircon in fluidised beds or caustic fusion and metal production via electrowinning warrant further development.
- ItemReview of recent developments in uranium extraction technology(Australian Atomic Energy Commission, 1978-12) Alfredson, PG; Crawford, RE; Ring, RJDevelopments in uranium ore processing technology since the AAEC Symposium on Uranium Processing in July 1972 are reviewed. The main developments include the use of: - autogenous or semi-autogenous grinding, - benefication techniques such as radiometric sorting, flotation, magnetic and gravity separation, - strong acid and ferric bacterial leaching processes, - solution mining and heap leaching operations - horizontal belt filters for solid-liquid separation, - continuous ion exchange processes for use with solutions containing up to 8wt % solids, - hydrogen peroxide and ammonia for the precipitation of uranium to improve product yield and purity, and - the recovery of by-product uranium from the manufacture of phosphoric acid and copper processing operations.
- ItemSpray drying of ammonium diuranate slurries(Australian Atomic Energy Commission, 1972-10) Levins, DM; Alfredson, PG; Hirst, RC; MacBride, PRSlow and fast settling ammonium diuranate (ADU) slurries were dried in a one metre diameter laboratory spray drier using either pneumatic or centrifugal atomisation. Uniform, fine, dry powders were obtained with both atomisation techniques but pneumatic atomisation was preferred. Spray drying did not adversely affect the properties of the subsequent UO2 powder and pellets. It was not necessary to treat the ADU feed slurry to remove ammonium nitrate in solution. A theoretical model of droplet evaporation during spray drying is developed, taking into account droplet size distribution, gas-spray hydrodynamics and heat transfer rates. This model satisfactorily accounts for the observed operational characteristics of the drier.
- ItemThermal denitration of uranyl nitrate in a fluidised bed reactor(Australian Atomic Energy Commission, 1974-07) Fane, AG; Charlton, BG; Alfredson, PGCommissioning and operating experience are described for the thermal denitration of uranyl nitrate in a 0.1 m diameter fluidised bed reactor. The effects of operating temperature, uranyl nitrate concentration and feed rate, nozzle air to liquid flow ratio, and the addition of sulphate to the feed, on the characteristics of the product and equipment performance were examined. Particle growth was a predominant feature which was strongly influenced by operating temperature. Changes in the main process variables exerted a minor influence on other properties of the product. The addition of sulphate to the uranyl nitrate feed solution produced an increase in surface area, and a decrease in pour and tap density. The wall-to-bed heat transfer coefficient was in the range 190 to 265W nf 2K-1, and shown to be an inverse function of the average particle size.