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Title: The effect of radiation creep on the stresses in the moderator graphite of a nuclear reactor
Author: Head, John Lawrence
Awarding Body: University of London
Current Institution: Imperial College London
Date of Award: 1970
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The effects of neutron irradiation of graphite are briefly described and an explanation is given of the generation of stresses in the graphite components of a nuclear reactor core. A creep law is proposed for graphite under multi-axial stress and the equations governing the stresses, strains and deformations of graphite components are presented. Simplified equations are obtained, appropriate to the moderator blocks of the Magnox reactor and the Advanced Gas-cooled Reactor. Computer programs are described which solve the simplified equations, stepwise in time, using the proposed creep law and obtaining the creep strains by iteration. Stress and strain histories are given for moderator blocks irradiated under environmental conditions typical of both types of reactor. The modification of the program for the analysis of a fuel tube of a High Temperature Reactor is described and results are given of analyses of fuel tubes of the type used in the experimental reactor, Dragon. A comparison is made between the predicted residual stresses in a particular fuel tube irradiated in Dragon and the measured residual stresses in the same fuel tube. Analyses of hypothetical fuel tubes made from different types of graphite enable conclusions to be drawn regarding the relative suitability of these graphites for fuel tubes of this type. A description is given of a program developed for the analysis of the "interacting tubular" fuel pin, one of several types of fuel pin currently being considered for use in future commercial High Temperature Reactors. This program incorporates a subroutine to recalculate the axial distributions of coolant and fuel pin surface temperature. This was found to be essential, as deformations of this type of fuel pin influence the heat transfer. Results are given from an analysis of a fuel pin of this type. Finally, a brief description is given of a finite element program, suitable for the analysis of the alternative "teledial" fuel pin and multi-channel moderator blocks of the High Temperature Reactor.
Supervisor: Not available Sponsor: Not available
Qualification Name: Thesis (Ph.D.) Qualification Level: Doctoral
EThOS ID:  DOI: Not available