The purpose of this study is to evaluate the effects and importance of different cross–section representations and energy group structures for steady state and transient analysis. More energy groups may be more accurate, but the calculation becomes much more expensive, hence a balance between accuracy and calculation effort must be find. This study is aimed at comparing a multi–group energy dynamics code, MGT (Multi–group TINTE) with TINTE (TIme Dependent Neutronics and TEmperatures). TINTE’s original version (version 204d) only distinguishes between two energy group structures, namely thermal and fast region with a polynomial reconstruction of cross–sections pre–calculated as a function of different conditions and temperatures. MGT is a TINTE derivative that has been developed, allowing a variable number of broad energy groups. The MGT code will be benchmarked against the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark: the PBMR–400 core design. This comparative study reveals the variations in the results when using two different methods for cross–section generation and multi–group energy structure. Inputs and results received from PBMR (Pty) Ltd. were used to do the comparison. A comparison was done between two–group TINTE and the equivalent two energy groups in MGT as well as between 4, 6 and 8 energy groups in MGT with the different cross–section generation methods, namely inline spectrum– and tabulated cross–section method. The characteristics that are compared are reactor power, moderation– and maximum fuel temperatures and k–effective (only steady state case). This study revealed that a balance between accuracy and calculation effort can be met by using a 4–group energy group structure. A larger part of the available increase in accuracy can be obtained with 4–groups, at the cost of only a small increase in CPU time. The changing of the group structures in the steady state case from 2 to 8 groups has a greater influence on the variation in the results than the cross–section generation method that was used to obtain the results. In the case of a transient calculation, the cross–section generation method has a greater influence on the variation in the results than on the steady state case and has a similar effect to the number of energy groups. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The purpose of this study is to evaluate the effects and importance of different cross–section representations and energy group structures for steady state and transient analysis. More energy groups may be more accurate, but the calculation becomes much more expensive, hence a balance between accuracy and calculation effort must be find. This study is aimed at comparing a multi–group energy dynamics code, MGT (Multi–group TINTE) with TINTE (TIme Dependent Neutronics and TEmperatures). TINTE’s original version (version 204d) only distinguishes between two energy group structures, namely thermal and fast region with a polynomial reconstruction of cross–sections pre–calculated as a function of different conditions and temperatures. MGT is a TINTE derivative that has been developed, allowing a variable number of broad energy groups. The MGT code will be benchmarked against the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark: the PBMR–400 core design. This comparative study reveals the variations in the results when using two different methods for cross–section generation and multi–group energy structure. Inputs and results received from PBMR (Pty) Ltd. were used to do the comparison. A comparison was done between two–group TINTE and the equivalent two energy groups in MGT as well as between 4, 6 and 8 energy groups in MGT with the different cross–section generation methods, namely inline spectrum– and tabulated cross–section method. The characteristics that are compared are reactor power, moderation– and maximum fuel temperatures and k–effective (only steady state case). This study revealed that a balance between accuracy and calculation effort can be met by using a 4–group energy group structure. A larger part of the available increase in accuracy can be obtained with 4–groups, at the cost of only a small increase in CPU time. The changing of the group structures in the steady state case from 2 to 8 groups has a greater influence on the variation in the results than the cross–section generation method that was used to obtain the results. In the case of a transient calculation, the cross–section generation method has a greater influence on the variation in the results than on the steady state case and has a similar effect to the number of energy groups. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The Generation IV International Forum (GIF) has initiated an international collaboration for the research and development of the Generation IV future nuclear energy systems. The Canadian PT-SCWR is Canada’s contribution to the GIF as a GEN-IV advanced energy system. The PT-SCWR is a pressure tube reactor type and considered as an evolution of the conventional CANDU reactor. The PT-SCWR is characterized by bi-directional coolant flow through the High Efficiency Re-entrant Channel (HERC). The Canadian SCWR is a unique design involving high pressure and temperature coolant, a light water moderator, and a thorium-plutonium fuel, and is unlike any operating or conceptual reactor at this time. The SCWR does share some features in common with the BWR configuration (direct cycle, control blades etc…), CANDU (separate low temperature moderator), and the HTGR/HTR (coolant with high propensity to up-scatter), and so it represents a hybrid of many concepts. Because of its hybrid nature there have been subtle feedback effects reported in the literature which have not been fully analyzed and are highly dependent on these unique characteristics in the core. Also given the significant isotopic changes in the fuel it is necessary to understand how the feedback mechanisms evolve with fuel depletion. Finally, given the spectral differences from both CANDU and HTR reactors further study on the few-energy group homogenization is needed. The three papers in this thesis address each one of these issues identified in literature. Models were created using the SCALE (Standardized Computer Analysis for Licensing Evaluation) code package. Through this work, it was found that the lattice is affected by more than one large individual phenomenon but that these phenomena cancel one another to have a small net final change. These phenomena are highly affected by the coolant properties which have major roles in neutron thermalization process since the PT-SCWR is characterized by a tight lattice pitch. It was observed that fresh and depleted fuel have almost similar behaviour with small differences due to the Pu depletion and the production of minor actinides, 233U and xenon. It was also found that a higher thermal energy barrier is recommended for the two-energy-group structure since the PT-SCWR is characterized by a large coolant temperature compared to the conventional water thermal reactors. Two, three and four optimum energy group structure homogenizations were determined based on the behaviour of the neutron multiplication factor and other reactivity feedback coefficients. Robust numerical computations and experience in the physics of the problem were used in the few-energy group optimization methodology. The results show that the accuracy of the expected solution becomes highly independent of the number of energy groups with more than four energy groups used. / Thesis / Doctor of Philosophy (PhD)