Journal of Nuclear Energy. Part B. Reactor Technology最新文献

Specimens of three heat-treatable steels with the following nominal compositions: (1) 17% Cr, 2% Ni, 0·1% C; (2) 13% Cr, 00·2 % C; and (3) 1% Cr, 00·2% Mo, 00·3% C; and (4) a non heat-treatable steel 17% Cr, 00·1% C, stabilized with niobium, were irradiated with fast neutrons at temperatures of 700° and 500-6000°C.

As a result of irradiation at 5000°-6000°C the mechanical properties of all the heat-treatable steels were changed. No such changes were observed in the non heat-treatable steel. A metallographic investigation indicated that the increase in tensile strength of the steels during irradiation at 500°-600°C was due to changes in structure. The authors suggest that the observed micro-structural changes were caused by the occurrence of “displacement spikes” in the irradiated material.

The paper gives an account of some general conclusions reached from an examination of all available data on the solubility of metals in liquid metals. Solubility values have been determined from published binary-phase diagrams. Solubilities show periodic variation with increase in atomic number of the solute, and this periodicity is broadly independent of the nature of the liquid metal.

A correlation of solubility (x) is found with the solute lattice energy which, in turn, is proportional to the latent heat of fusion (L_f). Using as solutes any two transition elements which are horizontally adjacent in the Periodic Table, the value of (log_ex₂ - log_ex_i)/(L_f1 - L_f2) is nearly proportional to the absolute temperature for nine of the lower melting liquid metals. This fact may be used to estimate solubility values for which no measurements exist. A number of such estimates are given.

As part of a study of liquid metal slurries as blanket materials for liquid metal fuelled reactors, thoria prepared by calcination of the oxalate has been mixed vigorously with inhibited lead, lead-bismuth eutectic and bismuth at temperatures in the range 500-600°C to form slurries containing up to 18 per cent (by volume) solids. Solidified samples have been examined for reactions between the corrosion inhibitors (zirconium and magnesium) and thoria, immediately after preparation and after circulation in thermal convection loops and in a pumped loop.

Magnesium at eutectic concentration in lead and bismuth reacted with thoria to produce intermetallic compounds which plugged thermal convection loops. Magnesium at normal inhibitor levels did not reduce thoria to give a thorium concentration in excess of the solubility limit in bismuth at 400°C.

Thoria concentrations of up to 7 per cent in bismuth were not found to effect the rheological properties of the liquids in these experiments, nor to alter corrosion inhibition effects by the ZrN film technique. The rheology of lead slurries was influenced principally by thoria settling upwards. Apart from settling effects the slurries were physically stable.

This paper describes the design, development and results of performance of mechanical and electromagnetic pumps for circulating sodium and sodium-potassium alloys. These pumps have been used during the past two years in both experimental and pilot plants.

The detailed design of a nuclear reactor, especially if it is of a new type, must be preceded by experimental work to refine the calculated parameters. A new method is proposed for determining the critical mass and the neutron flux distribution. A model of the reactor is used that is free from fissile material. The fuel channels are filled with a neutron absorber, of cross-section equivalent to that of the intended fuel, and the creation of fast fission neutrons is simulated by a neutron source which is moved along the various channels in turn. Resulting thermal neutron flux distributions are measured with thermal neutron detectors. The required critical mass can be deduced if the absolute source strength and neutron flux are known. The method has been checked on a reactor that operates with uranium hexafluoride. A critical mass was predicted that was in good agreement with the value found on starting up the reactor. The method may be of use in preliminary studies of new reactors in assisting in the choice of lattice dimensions and other parameters. It is simple, and involves neither fissile materials nor high neutron fluxes.