Irradiation Damage Aspects of Dispersion Fuel Elements for the H.T.G.C. Reactor (open access)

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

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 in BeO. The disadvantages of keeping uranium and thorium in separate particles are discussed and it is suggested that to minimize irradiation damage effects, the fuel particles should consist of solid solutions of the uranium and thorium compounds.
Date: June 1962
Creator: Hickman, B. S. (Brian Stuart)
System: The UNT Digital Library
A Preliminary Review of the Design and Feasibility of Prestressed Concrete Pressure Vessels for Nuclear Reactors (open access)

A Preliminary Review of the Design and Feasibility of Prestressed Concrete Pressure Vessels for Nuclear Reactors

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 savings will be made by using prestressed concrete instead of steel for the pressure vessels.
Date: June 1963
Creator: Holt, N. D.
System: The UNT Digital Library
An Experimental Determination of the Diffusion Length of Thermal Neutrons in Beryllium Oxide (open access)

An Experimental Determination of the Diffusion Length of Thermal Neutrons in Beryllium Oxide

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.
System: The UNT Digital Library
Binary and Ternary Systems Involving Beryllium Oxide - a Literature Survey (open access)

Binary and Ternary Systems Involving Beryllium Oxide - a Literature Survey

Data relating to the binary and ternary systems involving beryllium oxide are presented . The survey deals with all available literature up to December 1962.
Date: June 1963
Creator: Kairaitis, D.
System: The UNT Digital Library
The Microbiology of Heavy Water in the HIFAR Reactor (open access)

The Microbiology of Heavy Water in the HIFAR Reactor

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. Furthermore, during its passage through the reactor core the water and suspended bacteria are subjected to intense irradiation, the neutron flux being approximately 10 14 neutrons cm-2 sec-1. The presence of bacteria in the heavy water circuit has been demonstrated and experimental results and methods used are discussed. Some evidence is presented to show that the ion—exchange resin bed contributes nutrients to support bacterial growth.
Date: June 1962
Creator: Davis, P. S. & McPherson, G. G.
System: The UNT Digital Library
Reactions of Preoxidized Beryllium Powder in Moist Carbon Dioxide (open access)

Reactions of Preoxidized Beryllium Powder in Moist Carbon Dioxide

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. (auth)
Date: June 1962
Creator: Adams, R. B.; Price, G. H. & Stuart, W. I.
System: The UNT Digital Library
Design of Concentric Tubular Reactor Fuel Elements for Uniform Coolant Conditions (open access)

Design of Concentric Tubular Reactor Fuel Elements for Uniform Coolant Conditions

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 equivalent diameter.
Date: June 1962
Creator: Binns, Ian M.
System: The UNT Digital Library
Effect of Neutron Irradiation on Beryllium Metal (open access)

Effect of Neutron Irradiation on Beryllium Metal

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 elsewhere., are also given, It is concluded that all the observed property changes can be interpreted in terms of the distribution of helium which is produced by fast neutron transmutation reactions in beryllium and that damage due to defect production is negligible for irradiation temperatures of 75ºC and above. Density changes duetoheiium bubble formation are shown to be very small but serious deterioration of mechanical properties can occur. The mechanical property changes and the distribution of helium are shown to be very dependent on material history and on the irradiation temperature. The standard Lucas Heights hot extruded material is shown to retain good mechanical properties for irradiation temperatures above 550ºC but serious loss of low temperature ductility is found to occur for …
Date: June 1963
Creator: Hickman, B. S. (Brian Stuart) & Stevens, G. T.
System: The UNT Digital Library