The build–up of 6Li and 3He, that is, the strong thermal neutron absorbers or the so called "neutron poisons", in the beryllium reflector changes the physical characteristics of the reactor, such as reactivity, neutron spectra, neutron flux level, power distribution, etc.; furthermore,gaseous isotopes such as 3H and 4He induce swelling and embrittlement of the reflector. The SAFARI–1 research reactor, operated by Necsa at Pelindaba in South Africa, uses a beryllium reflector on three sides of the core, consisting of 19 beryllium reflector elements in total. This MTR went critical in 1965, and the original beryllium reflectors are still used. The individual neutron irradiation history of each beryllium reflector element, as well as the impact of beryllium poisoning on reactor parameters, were never well known nor investigated before. Furthermore, in the OSCAR{3 code system used in predictive neutronic calculations for SAFARI–1, beryllium reflector burn–up is not accounted for; OSCAR models the beryllium reflector as a non–burnable, 100% pure material. As a result, the poisoning phenomenon is not accounted for. Furthermore, the criteria and hence the optimum replacement time of the reflector has never been developed. This study presents detailed calculations, using MCNP, FISPACT and the OSCAR{3 code system, to quantify the influence of impurities that were originally present in the fresh beryllium reflector, the beryllium reflector poisoning phenomenon, and further goes on to propose the reflector's replacement criteria based on the calculated fluence and predicted swelling. Comparisons to experimental low power flux measurements and effects of safety parameters are also established. The study concludes that, to improve the accuracy and reliability of the predictive OSCAR code calculations, beryllium re flector burn–up should undoubtedly be incorporated in the next releases of OSCAR. Based on this study, the inclusion of the beryllium reflector burn–up chains is planned for implementation in the currently tested OSCAR–4 code system. In addition to beryllium reflector poisoning, the replacement criteria of the reflector is developed. It is however crucial that experimental measurements on the contents of 3H and 4He be conducted and thus swelling of the reflector be quantifed. In this way the calculated results could be verified and a sound replacement criteria be developed. In the absence of experimental measurements on the beryllium reflector, the analysis and quantifcation of the calculated results is reserved for future studies. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

Les travaux réalisés durant cette thèse rentrent dans cette problématique. Ils concernent d’une part le développement d’un schéma de calculs et d’évaluation des échauffements nucléaires générés dans le réacteur MARIA en utilisant les codes français de transport neutronique TRIPOLI-4 © et APOLLO-2. Les travaux dans ce volet ont concerné principalement les calculs des échauffements photoniques induits par les rayonnements gammas essentiellement. D’autre part des travaux expérimentaux ont été conduits durant cette thèse. Ils ont concerné la mesure des échauffements nucléaires dans des emplacements spécifiques du réacteur MARIA. Cela a permis une première validation des schémas de calcul adoptés. Des comparaisons C/E ont été effectuées. Elles sont présentées et discutées dans cette thèse. Cela a permis d’émettre des recommandations quant aux techniques de mesure des échauffements nucléaires dans le réacteur MARIA et les moyens de modélisation qui peuvent être associés. Les comparaisons calculs-expérience font l’objet du cinquième. Les écarts relevés entre les résultats des modélisations et les mesures des échauffements nucléaires pour différentes configurations de mesures (au moyen de GT et de calorimètre mono cellule KAROLINA) permettent de dégager grâce à ces premiers travaux de thèse des recommandations pertinentes pour les travaux futurs. / This thesis work presents a calculation scheme which enables evaluation of heat generation from nuclear reactions in the MARIA nuclear reactor by use French computational codes TRIPOLI-4 © (TRIPOLI-4 is a registered trademark of CEA) and Apollo-2. Particular attention was devoted to the heat induced by gamma radiation. The thesis also presents measurements of nuclear heating in selected locations inside MARIA MTR reactor. This allows reaching first steps of validation and qualification of computer calculations. Research and analysis presented in the thesis allow one to compare the results obtained by using proposed calculation scheme with the experimental measurement methods. Finally, further works and perspectives were proposed on the development of the calculations and experimental measurements of nuclear heating in nuclear reactors.Qualifying the calculations was possible by performing especially dedicated 7-day core measurement campaigns. Nuclear heating measurements were performed with gamma thermometers and specially designed KAROLINA calorimeter. All measurement devices used were mounted in a dedicated probe, designed and built for this purpose, which allowed for the adjustment of instruments position inside the MARIA core. The main scientific hypothesis of this work is that currently available Monte Carlo simulations of neutron and gamma transport can be used to correct and accurate calculations of prompt nuclear heating in nuclear reactor, whereas delayed component of nuclear heating can be determined experimentally. For this purpose new calculation scheme and improvements in nuclear heating measurements were implemented.

The aim of this master thesis is to show a connection among nuclear fuel safety, experiments led in research reactors and calculation codes. This thesis focuses on the calculation code Transuranus. There are represented four experiments, which were calculated in Transuranus. The fission gas release, elongation and growth of fuel were particularly monitored. Is is possible to set differences among versions v1m1j09 and v1m3j12 from achieved results, as well as the influence of selected Transuranus parameters on the results, so the thesis may bring new pieces of knowledge for improvement of safety analysis calculation by Transuranus.