Browsing by Author "Farrell, MS"

Now showing 1 - 9 of 9
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    A conceptual design study of a low throughput reprocessing facility for nuclear fuel
    (Australian Atomic Energy Commission, 1967-11) Cairns, RC; May, JR; Baillie, MG; Farrell, MS
    The idea of reducing fuel reprocessing costs by changing reprocessing plant design philosophy is explained. It is shown how a significant reduction in unit reprocessing costs can lead to earlier recovery of nuclear material. A classification is given of some existing nuclear chemical reprocessing plants as a function of their maintenance philosophies. The feasibility of the rack concept is discussed for application to a conceptual low throughput reprocessing plant specifically designed for reprocessing fuel from the A.A.E.C.'s Dido-class reactor HIFAR. Laboratory and design development work is described. Preliminary cost estimates are given for a site at the Research Establishment, Lucas Heights, with maximum use of existing facilities, services, and plant. The study did not reveal any technical difficulties that would make the rack concept impractical. The concept of indirect maintenance for items of equipment which are likely to require frequent attention is technically feasible, and it appears possible to remove racks for repair of equipment by normal direct maintenance techniques. Additional development followed by plant construction and operation would be necessary to verify these conclusions and to establish any cost advantages. However, the cost estimates deduced at the start of the study did not change substantially during the course of the work.
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    Development of solvent extraction processes for the H.T.G.C.R. fuel cycle, Part 3 - chemical data for the extraction of actinides and fission products from aqueous beryllium sulphate solutions using amines.
    (Australian Atomic Energy Commission, 1968-01) Fardy, JJ; Farrell, MS
    Chemical data are presented for the actinides (uranium, plutonium and thorium), the fission products (cerium, zirconium, niobium and ruthenium) and beryllium in the extraction of these from beryllium sulphate solutions by amines, in particular Primene-JMT and Alamine-336. The data so obtained define the major chemical parameters requiring control in a solvent extraction process to decontaminate beryllium sulphate, and indicate that the use of the solvent Primene-JMT/Solvesso-100 will adequately remove the uranium, thorium, plutonium, cerium, zirconium and niobium without extracting significant quantities of beryllium. Ruthenium is not highly extractable and strontium and caesium are extractable only in trace quantity. Although this process will be adequate for primary decontamination of the beryllium sulphate, a further step to remove residual ruthenium, strontium and caesium may be required. The actinides with the extracted fission products are readily stripped from the amine phase with nitric acid, thus permitting recovery of the actinides by a conventional tributyl phosphate process. The amine phase is readily converted back to the free-base form using sodium carbonate. This is necessary to prevent recycling of the nitrate which would inhibit the extraction of thorium, plutonium and the fission products.
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    Diluents for the solvent extraction of thorium using tributyl phosphate: general consideration and the third phase problem.
    (Australian Atomic Energy Commission, 1958-09) Farrell, MS; Goldrick, JD
    An examination of three possible diluents available from Australian sources of supply, has shown that ‘odourless mineral spirits’ from Vacuum Oil Co. Pty. Ltd, is the purest in alkane content, and is the most suitable of the three for use with TBP for the chemical processing of irradiated fuel, for example by the Thorfex process. A small quantity of secoctyl alcohol will suppress third phase formation, but further work is required to ensure that its use is otherwise compatible with the process.
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    Dissolution of sintered thoria.
    (Australian Atomic Energy Commission, 1959-11) Farrell, MS
    Thoria, prepared by calcining thorium oxalate, and compacted and sintered, became increasingly more difficult to dissolve in HNO3 HF mixture as the sintering temperature was increased. The temperature of calcination of the oxalate is an important factor. Low calcination temperatures produced a more reactive thoria with a greater surface area. This thoria sintered more readily, producing denser compacts which had a smaller B.E.T. surface area and were more difficult to dissolve. Thoria produced by calcination at 600°C showed "mottling" when sintered at 170°C. These dark-coloured mottles dissolved much more slowly than the white matrix in which they were embedded. No material other than thoria was detected in an X-Ray analysis of the mottled material. No extra lines or deformations were visible, and although interstitial carbon was suspected, no evidence for this was obtained.
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    Laboratory development of the grind-leach process for the H.T.G.C.R. fuel cycle, Part 1 - dissolution of urania-thoria fuel particles in nitric acid solutions.
    (Australian Atomic Energy Commission, 1965-09) Farrell, MS; Isaacs, SR
    The dissolution of 5 w/o UO2 in ThO2 fuel particles in nitric acid — fluoride solutions is a function of the Th02 dissolution rate. The Th02 dissolution rate has an apparent first order dependency on nitric acid concentration and a fractional order dependency on fluoride ion concentration. The apparent activation energy of the dissolution of the Th02 in 13M nitric acid — 0-05M fluoride is 19 kcal/mole. The presence of metal ions such as zirconium, aluminium, and beryllium inhibits the dissolution of the thoria. The order of inhibiting power is ZrlV > AlIII > BeII and this order is the same as the stability of their simplest fluoro — complexes, The addition of small quantities of sulphate also seriously inhibits the dissolution, precipitating thorium as Th(S04)2 , 4H20 and quantitatively removing the fluoride.
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    Laboratory development of the grind-leach process for the H.T.G.C.R. fuel cycle, Part 4 - leaching and dissolution of beryllia-based fuels
    (Australian Atomic Energy Commission, 1967-07) Shying, ME; Lee, EJ; Farrell, MS
    This report covers the chemical development of an acid-leach head-end process for the separation of (U,Th)02 from a beryllia matrix. The work is related to a feasibility study of the H.G.T.C.R. fuel cycle. Although more than 99 per cent of the actinides can be recovered in a single batch nitric acid leach, more than 30 per cent of the beryllia is dissolved. When it was found that sulphuric acid selectively dissolved the beryllia, a two-step process using nitric and sulphuric acids in sequence was developed, with recycling of the actinide-rich heel to the next batch. Preliminary results from multi-batch experiments indicate that it might be possible to decrease the beryllia loss to the nitrate stream to 10 per cent and still recover 99 per cent of the actinides. Further development of the process would involve the use of proven reactor material irradiated to high burn-up. Suggestions for further research are given.
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    Laboratory development of the grind-leach process for the H.T.G.C.R. fuel cycle, Part II - dissolution of beryllia in nitric acid solutions
    (Australian Atomic Energy Commission, 1966-06) Farrell, MS; Isaacs, SR; Shying, ME
    the development of the grind—leach process for the processing of beryllia—based fuels requires a knowledge of the dissolution of beryllia in nitric acid. A kinetic study using powdered specimens has proved suitable for the investigation of this system. The parameters studied include particle size, agitation, temperature, acid concentration, the effect of the addition of fluoride and aluminium, and the effect of neutron irradiation of the beryllia. The dissolution of beryllia in nitric acid is controlled by a chemical reaction at the surface of the solid and has an apparent activation energy of 18 kcal/mole.
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    The recovery of beryllium from beryllium oxide matrix fuels through its oxyacetate
    (Australian Atomic Energy Commission, 1965-07) Farrell, MS; Szego, LE; Yates, PB
    The laboratory development of a process for the recovery and decontamination of beryllium from fuel processing solutions is described. Azeotropic distillation is used in converting aqueous beryllium solutions to beryllium oxyacetate. The beryllium oxyacetate is then purified by recrystallization. Using a feed solution containing 3 β curies of fission products per mole of beryllium, a decontamination factor of 3 to 4 x 103 can be obtained with a 90 per cent, recovery of beryllium. Some solubility and conductance data are given for beryllium oxyacetate and its hydrolysis products. The hydrolysis reaction is briefly discussed.
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    Reprocessing of homogeneous beryllium-based reactor fuel - a suggested scheme for the selective aqueous dissolution of the matrix
    (Australian Atomic Energy Commission, 1962-08) Farrell, MS; Temple, RB
    The matrix of a dilute homogeneous H.T.G.C, reactor fuel employing metallic beryllium as a moderator can be selectively dissolved by a caustic soda solution containing salicylate ion. At least 99 per cent, of the uranium and thorium can be recovered as insoluble solids, but in the case of irradiated material the uranium loss might be higher. Some decontamination of the resulting beryllium solution from fission products and Pa233 can also be obtained. A tentative chemical flowsheet is proposed on the basis of the results obtained.