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Irradiation Damage Aspects of Dispersion Fuel Elements for the H.T.G.C. Reactor

Description: The concept of a dispersion fuel element is discussed with particular reference to irradiation damage. The application of this concept to the A.A.E.C. H.T.G.C. reactor system is outlined and the limitations imposed by irradiation damage considerations are discussed. The maximum desirable heavy metal - beryllium ratio (i.e. U+Th:Be) for the various systems under consideration should be about 1:55 for the system (U,Th)Be13 in Be, 1:13 for the system (U,Th)O2 in Be, and 1:8 for the system (U,Th)O2… more
Date: June 1962
Creator: Hickman, B. S. (Brian Stuart)
Partner: UNT Libraries Government Documents Department
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A Preliminary Review of the Design and Feasibility of Prestressed Concrete Pressure Vessels for Nuclear Reactors

Description: The design of prestressed concrete pressure vessels is discussed and some approximate design formulae are developed. The design and performance of vessels reported in the literature are reviewed and an approximate comparison is made of steel and concrete pressure vessels for a particular case. Concrete vessels are attractive for moderate temperatures and pressures because of the large size of vessel which can be built and the non-explosive mode of failure. However it is unlikely that large cost… more
Date: June 1963
Creator: Holt, N. D.
Partner: UNT Libraries Government Documents Department
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An Experimental Determination of the Diffusion Length of Thermal Neutrons in Beryllium Oxide

Description: The diffusion length of thermal neutrons in beryllium oxide of effective density 2.86 g cm-3 has been measured as 29.9 +- 0.8 cm. Using published experimental values for the diffusion constant of beryllium oxide, a value of [sigma] a = 9.0 +- 0.5 mb is deduced for the effective 2200m/s microscopic absorption cross-section.
Date: June 1963
Creator: Brittliff, E.; Duerden, P. & McCulloch, D. B.
Partner: UNT Libraries Government Documents Department
open access

The Microbiology of Heavy Water in the HIFAR Reactor

Description: The high flux research reactor HIFAR contains ten tons of heavy water which acts as moderator and primary coolant. Over an eighteen months period regular microbiological examinations have been carried out on samples of heavy water taken from various parts of the circuit. The heavy water circuit provides an interesting opportunity for the study of microorganisms because of the high isotopic purity (greater than 99.6 per cent.), and the high chemical purity of the heavy water in the reactor. Furt… more
Date: June 1962
Creator: Davis, P. S. & McPherson, G. G.
Partner: UNT Libraries Government Documents Department
open access

Reactions of Preoxidized Beryllium Powder in Moist Carbon Dioxide

Description: Breakaway corrosion of Be in moist CO2 can be avoided if the Be is fabricated using preoxidized powder. The powder is preoxidized by heating in dry O/sub 2/. Preoxidation of Be powder was measured as a function of temperature and time of heating in O/sub 2/. The subsequent reactions of the preoxidized powder in moist CO/sub 2/ at 700 deg C were studied and the effect of increasing amounts of added oxide was measured. A model is proposed to explain the inhibition of corrosion by added oxide. (au… more
Date: June 1962
Creator: Adams, R. B.; Price, G. H. & Stuart, W. I.
Partner: UNT Libraries Government Documents Department
open access

Design of Concentric Tubular Reactor Fuel Elements for Uniform Coolant Conditions

Description: Concentric tubular reactor fuel element geometries to give equal coolant outlet temperatures are presented. Oscillations from tube to tube in thickness and temperatures generally occur but it is possible to eliminate them by choice of the centre element. This may be a fuel rod or a non-heat—producing rod with or without a surrounding annulus of fuel. The geometries and temperatures are dependent on the voidage and on a non-dimensional parameter equivalent to a Biot number based on the channel e… more
Date: June 1962
Creator: Binns, Ian M.
Partner: UNT Libraries Government Documents Department
open access

Effect of Neutron Irradiation on Beryllium Metal

Description: This report summarises all the results obtained to date from a programme on the effects of neutron irradiation on the properties of beryllium metal. Results are presented on changes in density and mechanical properties in material fabricated by various routes and irradiated to fast neutron doses from 1019 nvt to 6 x 1023 nvt and at temperatures in the range 75ºC — 700ºC, Summaries of electron microscopy observations and electrical resistivity measurements, which are reported in more detail else… more
Date: June 1963
Creator: Hickman, B. S. (Brian Stuart) & Stevens, G. T.
Partner: UNT Libraries Government Documents Department
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