ANSTO :: Browsing by Author "Reeve, KD"

Browsing by Author "Reeve, KD"

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An assessment of possible protective coatings for BeO and BeO based reactor fuel elements
(Australian Atomic Energy Commission, 1966-07) Reeve, KD
Beryllium oxide based reactor fuel elements may be required to operate in moist air at temperatures in the range 1000 - 1200oC. In this environment, unprotected fuel elements would lose excessive amounts of BeO as Be(OH)2. Methods of preventing this loss by means of non-reactive coatings are considered. By application of established coating principles and also criteria specific to this requirement, the most promising materials are found to be Al2O3 and MgAl2O4. MgAl2O4 would be difficult to apply to form a dense crack-free coating; Al2O3 could be applied by various methods, the best of which appear to be isostatic pressing, slip casting or spraying, and vapour deposition. It is recommended that a detailed thermal stress analysis be carried out on an Al2O3 coated fuelled BeO sphere, and that experimental work should be concentrated on isostatic coating.
A comparative study of two grades of BeO
(Australian Atomic Energy Commission, 1961-11) Reeve, KD; Ramm, EJ
Pechiney and Brush UOX BeO differ markedly in fabrication behaviour, only Brush UOX being readily sinterable. A comparative study of the two powders has shown few outstanding differences in powder properties. Both are of high purity but contain free and combined moisture to the extent of about 1.5 per cent. Pechiney BeO has a larger mean crystallite size (0.2 — 0.3μ) than Brush UOX (0.1 — 0.15μ) and a larger range of crystallite size, and both contain a small proportion of crystallites of size 1μ. and larger. The tap density of UOX is much lower than that of Pechiney, and its surface area is higher by a factor of two, UOX BeO can be hot—pressed or cold—pressed and sintered to high densities at temperatures of 1400º and 1500ºC respectively, which are approximately 300ºC lower than those required for Pechiney. Grain size of fabricated material increases with fabrication temperature in both grades, and at the same temperatures is larger for UOX than Pechiney. However, at comparable densities grain sizes are also comparable for the two materials, The bend strength at room temperature of hot—pressed Pechiney BeO reaches a maximum of approximately 30,000 p.s.i. when pressed at 1750ºC. Above this temperature the strength falls due to increasing crystallite size. Cold—pressed and sintered Brush UOX appears weaker, possibly due to weaker grain boundary cohesion. Cold—pressed and sintered BeO is considered to be worthy of most intensive study as an irradiation resistant material, and suggestions are given for improving its strength and homogeneity and decreasing its crystallite size.
Development and testing of corrosion-resistant alumina coatings for beryllia-based reactor fuel elements.
(Australian Atomic Energy Commission, 1971-05) Reeve, KD; Ramm, EJ; Webb, CE
The use of beryllia as the basis of an all-ceramic fuel element for high-temperature reactors cooled by ambient air depends on inhibition of corrosion of beryllium oxide by protecting the surface of the fuel element. Alumina is a promising coating material. The development of corrosion resistant alumina coatings for BeO spheres is described, and results of corrosion, accelerated corrosion and neutron irradiation tests are presented. The limitations of the coating are discussed in detail Porous coatings can fail by 'undermining', but dense coatings are probably satisfactory for at least two years' operation out-of-pile at 1200ºC and for much longer times at lower temperatures. Satisfactory operation at 1200ºC for one year has been proved in a long term test. The eventual failure mechanism will probably be associated with the growth of a two layer reaction zone. Neutron irradiation tests indicate that a dose of 1020 nvt and possibly 1.6—2 x 102º nvt should be acceptable for retention of coating-to-BeO bond during simultaneous exposure above 500-700ºC to fast neutrons and moist air.
The development and testing of Synroc C as a high level nuclear waste form
(Cambridge University Press, 2011-02-15) Reeve, KD; Levins, DM; Ramm, EJ; Woolfrey, JL; Buykx, WJ
The current status of SYNROC C research and development by the Australian Atomic Energy Commission is reviewed. A non-radioactive fabrication demonstration line designed to produce 10 cm o.d., 90 cm long, cylinders of SYNROC canned in stainless steel by the method of in-can hot pressing is being commissioned. Leach tests are proving the excellent leach resistance of SYNROC. Accelerated radiation damage testing using fast neutrons has simulated storage times of up to 6.7×105 years. Thermophysical properties of SYNROC have been measured over the temperature range 20–650°C. © Materials Research Society 1982
The development and testing of SYNROC for high level radioactive waste fixation
(Australian Atomic Energy Commission, 1981-02-23) Reeve, KD; Levins, DM; Ramm, RJ; Woolfrey, JL; Buykx, WJ; Ryan, RK; Champan, JF
Research and development on the SYNROC concept for high level radioactive waste fixation commenced at the Australian Atomic Energy Commission Research Establishment, Lucas Heights, in March 1979, in collaboration with a complementary program at The Australian National University (ANU). The present paper reports progress in the project's second year and reviews its current status. An inactive 30 kg-scale SYNROC fabrication line incorporating in-can hot pressing as the fabrication step has been built for operation in mid-1981. Atmospheric pressure and hydrothermal leach tests are demonstrating the excellent leach resistance of SYNROC. Accelerated radiation damage tests using fast neutrons are simulating damage in SYNROC for periods of close to 10/sup 6/ years. In supporting research, mineral phase development, impact friability and thermophysical properties of SYNROC are being studied.
Development of ceramic coatings for fission product retention in ceramic fuels.
(Australian Atomic Energy Commission, 1966-01) Smith, PD; Reeve, KD
A general discussion on glaze—type coatings for fission product retention within BeO spheres fuelled with PuO2 - ThO2 particles is presented. From these considerations, a glaze thickness of 0.004 inch was chosen as a basis for laboratory studies. Glaze development commenced with a conventional high temperature glaze; this was modified firstly by increasing the SiO2 content and then by progressively replacing some of the SiO2 by Ti02, Zr02. Ti02 - Zr02, BeO2 and BeO plus Ti02. Glaze structures varied from amorphous to predominantly crystalline. When applied to fuelled BeO cubes, some glazes failed to cover fuel particles and others tended to react with them, Two of the most promising glazes behaved poorly in fission product release experiments, probably because glaze—fuel interaction had allowed fuel migration to the glaze surface-It is concluded that glaze type coatings show little promise for the present applications.
Disposal options for high level nuclear waste
(Australian Atomic Energy Commission, 1982-01-18) Reeve, KD
The options for the management/disposal of high level nuclear waste are discussed; it is concluded that irretrievable disposal of solidified waste is a likely end step in all disposal schemes. Disposal will probably be either in mined repositories or in deep drill holes. Two solidified waste forms - borosilicate glass and SYNROC - are considered in some detail. It is concluded that SYNROC would provide a higher level of assurance of radionuclide retention in both disposal concepts. Deep drill-hole disposal using SYNROC is an attractive concept which should be given increased attention.
Fabrication of beryllia-coated, fuelled beryllia spheres for in-pile fission product release tests
(Australian Atomic Energy Commission, 1966-01) Reeve, KD; Clare, TE; Silver, JM; Bridgford, KC
Three sees of beryllia-coated, fuelled beryllia spheres were made for fission product release testing in a sweep capsule irradiation rig Results of various pre-irradiation tests are presented and discussed, and a summary of fission gas release results is included, Gas release rates were expected to vary inversely as the beryllia density However, the release rate was lowest for the loading with an intermediate density and was highest for that with the highest density One or more of several structural factors may have changed the behaviour from that expected.
Grinding studies on beryllium oxide powder.
(Australian Atomic Energy Commission, 1963-03) Reeve, KD; Ramm, EJ
Inhomogeneities in Brush UOX beryllium oxide observed in the powder and in cold pressed and sintered specimens have been removed by grinding the powder prior to fabrication, all grinding procedures reduced the densities obtained under standard sintering conditions, but some grain refinement was noted on sintering after short grinding periods. These effects are related to the introduction of alumina and silica impurities during ball milling. There is some indication that short grinding periods improve the strength of sintered specimens.
High burnup irradiation testing of spherical beryllium oxide based fuel elements for a conceptual high temperature air-cooled reactor
(Australian Atomic Energy Commission, 1973-05) Hanna, GL; Reeve, KD
Fuelled beryllium oxide spheres, with a thin layer of porous BeO separating the fuelled core from the unfuelled shell, were irradiated at 500ºC, 750ºC and 1000ºC to burnups of 13.4 to 15.6 per cent (U + Th) and fast neutron doses of 1.4 x 10 20 to 1.9 x 1020 nvt. The four spheres irradiated at 1000ºC were apparently undamaged but the shell had fractured in all those irradiated at 500ºC and 750ºC. Failure is believed to have been caused by enhanced expansion of the inner regions of the unfuelled shell arising from exposure to both fast neutrons and energetic β-particles. Further work necessary to prove a fuel element design for the proposed application is briefly outlined.
Impurities and their effects in nuclear-grade beryllia powders.
(Australian Atomic Energy Commission, 1966-01) Reeve, KD
The criteria for sinterability in high purity beryllia powders are discussed. Of common impurities in less pure powders, silicon, aluminium, calcium, magnesium andiron are considered to have the major effects on sinterability. Of these, Si and Al have detrimental effects, and Ca, Mg and Fe beneficial effects. It is suggested that for an impure powder to be sinterable, the combined (Si + Al) content should be no higher than the combined (Ca + Mg + Fe) content, both expressed in p.p.m.
The preparation of spheroidal UO2 - ThO2 particles
(Australian Atomic Energy Commission, 1963-03) Reeve, KD; Jones, KA
A self—abradory process is described for the small-scale preparation of 150 - 200 micron spheroidal particles of various UO2- ThO2 compositions. The particles can be sintered to high densities before or after dispersion in beryllium oxide. Because of the high compaction pressure used in making particles, they are strong enough to resist abrasion and crushing during mixing with beryllium oxide powder, after sintering, the particles consist of a (U, Th)02 solid solution with a small range of composition, but the overall composition does not vary from one particle to another. The types of porosity observed after sintering are consistent with the occurrence of two competitive mechanisms during spheroidisation, namely particle abrasion and particle build-up.
Sintering studies on ceramic fuel materials
(Australian Atomic Energy Commission, 1964-01) Reeve, KD; Jones, KA
Satisfactory dense crack—free dispersions of spheroidal U02 — Th02 particles in BeO can be produced by "co-sintering", a process in which unsintered particles are hydrostatically pressed in BeO and the dispersion is sintered in one step, the large voids and cracks in dispersions of fully sintered particles in BeO following sintering are related to mismatched sintering shrinkage between the two phases; they are probably indicative of a high interfacial energy between the two phases. Co—sintering has been used for individual fabrication of specimens for irradiation testing, but is not immediately applicable to large scale fabrication of fuel element shapes, Possible modifications of the method are discussed.
Some oxide ceramics as reactor materials
(Melbourne University Press on behalf of The Australian Atomic Energy Commission, 1958-06-02) Reeve, KD
Although the conventional advantages of ceramics lie in their favourable high temperature properties reactor technology has not yet made use of these properties, mainly because of their uncertain behaviour under irradiation. The fabrication and properties of beryllia, uranium dioxide and thoria, and of mixed oxide systems, are discussed.
The stability of fissile-fertile oxide solid solutions in air
(Australian Atomic Energy Commission, 1966-10) Reeve, KD
The stability of (UTh)02, (PuTh)02, and (PuUTh)02 solid solutions when heated to high temperatures in air is reviewed and discussed. (PuTh)02 is chemically stable under these conditions, but compositions containing uranium oxidise to give either a non-stoichiometric fluorite phase containing excess oxygen, or to this phase plus orthorhombic U3O8. U3O8 formation is to be avoided if maximum dimensional stability is required. (UTh)02 compositions containing 50 m/o or more of Th02 do not form an orthorhombic phase under any conditions of oxidation. Information on the extent of non-stoichiometry and the effects of excess oxygen on unit cell volume, bulk volume, and the rate of uranium loss at high temperatures is also reviewed.

Browsing by Author "Reeve, KD"

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