By Anthony M. Judd
This ebook is a useful source for either graduate-level engineering scholars and training nuclear engineers who are looking to extend their wisdom of quickly nuclear reactors, the reactors of the long run! The booklet is a concise but finished creation to all features of quickly reactor engineering. It covers themes together with neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the functionality of ceramic and steel fuels below irradiation, structural adjustments, and fission-product migration; the consequences of irradiation and corrosion on structural fabrics, irradiation swelling; warmth move within the reactor center and its influence on center layout; coolants together with sodium and lead-bismuth alloy; coolant circuits; pumps; warmth exchangers and steam turbines; and plant keep watch over. The ebook contains new discussions on lead-alloy and gasoline coolants, steel gasoline, using reactors to eat radioactive waste, and accelerator-driven subcritical structures.
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Extra info for An Introduction to the Engineering of Fast Nuclear Reactors
This is equivalent to finding the reactivity of a reactor that may not be exactly critical. Alternatively the composition (for example the concentration of plutonium in the core) or the dimensions (say the radius of the core) may be altered to make k = 1. e. with the transverse planes covered by a hexagonal or triangular mesh). Again, in the initial stages of design, diffusion theory is adequate but when it comes to the details transport theory is necessary. A typical transport theory code would use a nodal formulation of the transport equation with hexagonal nodes, each node corresponding to an individual core position occupied by a fuel subassembly, a control rod or an incineration target, etc.
9 to find the eigenvalue k of a system it is very useful to have a means of estimating the effect of small changes. This is particularly so in the case of temperature coefficients. For example, as explained later, a change in temperature makes small alterations to certain group cross-sections by means of the Doppler effect. A method of estimating the resultant change in k is needed, and this is provided by perturbation theory. Suppose the perturbation in which we are interested results in an increase δ cg in the capture cross-section for group g in a small region dV at a point r in the reactor.
The outcome of each event in the history of each neutron is chosen at random in accordance with the known probabilities. This is done for large numbers of different neutron histories, so large a number that when they are all put together they make up an estimate of the actual neutron distribution. The method depends on a random number generator that produces a sequence of numbers Rn distributed uniformly at random in the range (0,1). For a “source-type” calculation (for example a subcritical reactor driven by a neutron source) the procedure is straightforward.
An Introduction to the Engineering of Fast Nuclear Reactors by Anthony M. Judd