Browsing by Author "Hilditch, RJ"

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    Effects of irradiation on beryllia-based fuels.
    (Australian Atomic Energy Commission, 1967-09) Hickman, BS; Rotsey, WB; Hilditch, RJ; Veevers, K
    Dispersions of (UTh)O2 in beryliia, containing 1.7 per cent to 25 per cent (UTh)C>2 in three fuel particle sizes, coarse (150 —200μ)> medium (33 — 35μ and fine (<10 and <5º) were irradiated to burnups of 3—10 per cent of heavy metal atoms in the range 300-900ºC, in both fast and thermal fluxes. Changes in volume, lattice parameter, line breadth, and modulus of rupture were measured. Volume changes in the fine dispersions were ascribed wholly to fission fragment damage and were about 50 per cent greater than those caused by fast neutrons alone; they increased with increasing fission fragment flux, and decreased as irradiation temperature increased. Volume changes in medium and coarse dispersions were about 25 per cent greater than those caused by fast neutrons alone; the enhancement of the damage is attributed to the additional β flux. As fuel particle size increased, deterioration in strength under irradiation was more marked. This was attributed to more intense fission fragment damage in the recoil zone around larger particles causing volume increases which exceeded those of the remainder of the matrix. For maximum initial strength and retention of strength under irradiation the fuel particle size should not exceed 5μ, and the inter-particle spacing should not exceed 30.
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    The irradiation behaviour of beryllium based dispersion fuels - a preliminary irradiation experiment.
    (Australian Atomic Energy Commission, 1962-09) Hanna, GL; Hickman, BS; Hilditch, RJ
    The effects of fission fragment damage on vacuum hot pressed fuel specimens of (U Th) Be13 dispersed in a beryllium matrix were examined by irradiation in a predominantly thermal neutron flux. Damage equivalent to that caused by 4 x 10 19 to 11 x 10 19 fissions per cm3 (depending on specimen composition) was achieved at temperatures between 435º and 530ºC. All specimens increased in volume on irradiation. The increases ranged from 0.1 per cent, to 5 per cent., depending on the volume fraction of fuel phase and the number of fissions per cm3. Some of the volume change — possibly up to 0.7 per cent. — was due to thermal effects alone. Release of fission gases was as high as 2 per cent, in some cases and was generally higher than would be expected from recoil in specimens having no open porosity. The fractional release was greater in specimens which experienced a high volume increase. Microstructures showed no significant change on irradiation. All specimens were slightly porous before irradiation and it is considered that the swelling of specimens was due to the growth of existing pores and that the release of fission gases was facilitated by an increase in open porosity.
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    The irradiation behaviour of beryllium oxide dispersion fuels.
    (Australian Atomic Energy Commission, 1963-03) Hanna, GL; Hickman, BS; Hilditch, RJ
    Specimens of beryllium oxide based dispersion fuels containing between three and twenty-six volume per cent, of U02— Th02 solid solution were irradiated to fission densities of 2 to 14 x 1019 fissions/cm3 of total specimen (equivalent burn—ups of 80 to 230 per cent.) at temperatures of 600 - 850ºC. The experiment was primarily designed to investigate fission product damage although some fast neutron damage did occur in the matrix, the specimens showed excellent resistance to fission product damage; dimensional changes were small, fission product escape was generally only that expected by recoil and there was no sign of cracking due to thermal stresses although these reached estimated values of about 30,000 p.s.i. in some specimens, Metallographic examination showed that some weakening of the matrix grain boundaries had occurred and some preliminary x-ray results suggested that the matrix was in a state of strain. It is suggested that these effects could be due either to fast neutron damage in the matrix or swelling of the fuel particles. The experiment did not provide any conclusive evidence for the superiority of coarse fuel particles (100 - 180μ.) over fine fuel particles (< 10μ) although the dimensional changes and the degree of matrix strain were higher in the latter specimens.
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    The irradiation behaviour of cold-pressed and sintered beryllium oxide dispersion fuels.
    (Australian Atomic Energy Commission, 1964-06) Hanna, GL; Hilditch, RJ
    Coarse and fine dispersions of (U.Th)O2 in BeO, prepared by cold-pressing and sintering, were exposed to fission fragment damage by irradiation in a thermalised neutron flux. Irradiation temperatures were between 610ºC and 720ºC and fission densities of 3 x 1019 to 2 x 1020 fissions per cm3 of compact were achieved. The dimensional stability of all specimens was very good and the greatest volume expansion was one per cent. Estimated thermal stresses ranged up to 32,000 p.s.i. but no thermal stress failure occurred. The fine and the most dilute coarse dispersions released only 0.1 per cent, of the fission gases but the remaining coarse dispersions released from 0.6 to 4.9 per cent. Coarse fuel particles were generally cracked after irradiation and by microscopy, fission fragment damage to the fuel — BeO interfaces was observed in the high burn—up specimens. Fission gas bubbles developed in coarse fuel particles on post—irradiation annealing at 1250ºC and 1500ºC.
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    The irradiation behaviour of hot-pressed dispersions of (U,Th)Be13 in beryllium - second irradiation experiment.
    (Australian Atomic Energy Commission, 1963-11) Hanna, GL; Hickman, BS; Hilditch, RJ
    Fuel specimens having uranium—thorium beryHides as the fissile bearing phase were irradiated at 400 to 700 ºC to burn—ups between 7 and lla/0 uranium in a predominantly thermal flux, two specimens of massive (U.;Th)Be 13 exhibited good dimensional stability and low fission gas release. Dispersion specimens containing 20, 35, and 50v/o (U,Th)Be13 swelled by 3 to 18 per cent0 and released up to 33 per cent, of the gaseous fission products. The results indicate that the mixed beryllide is an inherently good fuel material and lead to the conclusion that the poor irradiation stability of hot—pressed specimens may not be typical of (U,Th) Be 13— Be dispersions prepared by other techniques.
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    Irradiation of uranium metal tubes and rods in a 4V hole in HIFAR
    (Australian Atomic Energy Commission, 1961-10) Hickman, BS; Smith, R; Hilditch, RJ; Mercer, WL
    This report covers all aspects of the irradiation and post-irradiation examination of the specimens in the first test of metallic G.E.G.B. uranium in the HIFAR materials testing reactor. A preliminary report (NPCC/FEWP/P.667) issued in February, 1961, gave the results of the initial measurements and visual inspection of the specimens. The earlier report divided the specimens into three main groups into which they appeared to fall conveniently at the time. Further detailed examination, including a large amount of metallographic study, has suggested that it is better to divide the specimens into four groups with groups 1 and 2 including all results of direct application to power reactor fuel element operation. Certain sections of the present report repeat some of the information already given in the preliminary report. This Material has been unclassified
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    The strength of irradiated beryllium oxide fuelled with UO2-ThO2
    (Australian Atomic Energy Commission, 1965-07) Hanna, GL; Hilditch, RJ
    The mechanical strength of hot-pressed and cold-pressed and sintered dispersions of (U,Th)O2 in BeO was measured before and after irradiation using the diametral compression test. The addition of more than 5 volume per cent. (U,Th)02 lowered the unirradiated strength to about 60 per cent, of the strength of BeO. Irradiation resulted in a marked drop in strength of all coarse dispersions. Hot-pressed fine dispersions behaved similarly to coarse dispersions of the same composition but the cold-pressed fine dispersions exhibited an apparent increase of strength on irradiation. High temperatures favoured the retention of strength during irradiation,
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    Studies of uranium-sodium suspensions part 1 construction and operation of experimental loop
    (Australian Atomic Energy Commission, 1961-08) Bett, FL; Hilditch, RJ; Mepham, RG; Kluss, TN
    An experimental uranium-sodium suspension loop has been operated for 4,320 hours at the A.A.E.C. Research Establishment. This report describes the design, construction, commissioning, and operation of the loop to the point where a complete stable suspension was obtained.
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    Studies of uranium-sodium suspensions part 2 hydrodynamic, metallurgical and mechanical effects: general conclusions
    (Australian Atomic Energy Commission, 1961-08) Bett, FL; Hilditch, RJ; Mepham, RG; Gillespie, PA
    Hydrodynamic behaviour of a 0.76 atomic per cent, uranium in sodium suspension has been explored for velocities up to 8.3 ft/sec, and temperatures up to 625°C. The influence of velocity and temperature on suspension behaviour is described. Mechanical and metallurgical effects associated with circulation of the suspension in an austenitic stainless steel loop in contact with beryllium are listed. The behaviour of the suspension is explained in terms of hindered settling theory. A reactor loop metal illustration possible use of the uranium-sodium suspension as a reactor fuel/coolant is presented and recommendations for further work are given.