Very High Temperature Reactors are complex reactors and various system codes have been developed to design different aspects such as neutronics, thermal hydraulics etc. Flownex is one of the system codes and it has been used to model the flow and heat transfer for a pebble fuel element and pebble-bed reactor. Although Flownex has been used to model the High Temperature Test Reactor, the prismatic block was modelled in a simplified manner. The aim of this study was to develop a more integrated model for a single block. A one sixth block was modelled in Flownex and the results were validated by comparing the results with results obtained using the Computational Fluid Dynamics (CFD) code STAR-CCM+. The conduction heat transfer through the prismatic blocks containing the fuel elements in a Very High Temperature Reactor is of crucial importance for the proper operation of the reactor under normal operating conditions and upset conditions. In this study, a model developed in a system code, Flownex is discussed. The model comprised of a collection of 1-D solid conduction heat transfer, convection heat transfer and pipe elements that were arranged in such a manner to represent the heat transfer and fluid flow in the prismatic block using a network approach. The validity of the model was investigated by comparing the heat transfer and temperature distribution in the block for various scenarios with the corresponding values obtained using a detailed CFD model of one twelfth of a prismatic block. Cubical and triangular block verification cases were conducted in Flownex and the results were validated by STAR-CCM+. The results were very comparable; however one issue has to be addressed. The one sixth integrated prismatic block was then modelled for a steady state and the results were also comparable. The outlet helium temperatures predicted by the STAR-CCM+ model was 542.94 C, at the same time the Flownex model predicted 542.98 C. Although the Flownex model did not provide the same detail as the STAR-CCM+ model the agreement between the results obtained with the two codes was satisfactory. Based on these findings it was concluded that Flownex could be used to build a representative integrated network model for a prismatic block reactor. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

Very High Temperature Reactors are complex reactors and various system codes have been developed to design different aspects such as neutronics, thermal hydraulics etc. Flownex is one of the system codes and it has been used to model the flow and heat transfer for a pebble fuel element and pebble-bed reactor. Although Flownex has been used to model the High Temperature Test Reactor, the prismatic block was modelled in a simplified manner. The aim of this study was to develop a more integrated model for a single block. A one sixth block was modelled in Flownex and the results were validated by comparing the results with results obtained using the Computational Fluid Dynamics (CFD) code STAR-CCM+. The conduction heat transfer through the prismatic blocks containing the fuel elements in a Very High Temperature Reactor is of crucial importance for the proper operation of the reactor under normal operating conditions and upset conditions. In this study, a model developed in a system code, Flownex is discussed. The model comprised of a collection of 1-D solid conduction heat transfer, convection heat transfer and pipe elements that were arranged in such a manner to represent the heat transfer and fluid flow in the prismatic block using a network approach. The validity of the model was investigated by comparing the heat transfer and temperature distribution in the block for various scenarios with the corresponding values obtained using a detailed CFD model of one twelfth of a prismatic block. Cubical and triangular block verification cases were conducted in Flownex and the results were validated by STAR-CCM+. The results were very comparable; however one issue has to be addressed. The one sixth integrated prismatic block was then modelled for a steady state and the results were also comparable. The outlet helium temperatures predicted by the STAR-CCM+ model was 542.94 C, at the same time the Flownex model predicted 542.98 C. Although the Flownex model did not provide the same detail as the STAR-CCM+ model the agreement between the results obtained with the two codes was satisfactory. Based on these findings it was concluded that Flownex could be used to build a representative integrated network model for a prismatic block reactor. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

A new whole-core transport method is described for 3-D hexagonal geometry. This is an extension of a stochastic-deterministic hybrid method which has previously been shown highly accurate and efficient for eigenvalue problems. Via Monte Carlo, it determines the solution to the transport equation in sub-regions of reactor cores, such as individual fuel elements or sections thereof, and uses those solutions to compose a library of response expansion coefficients. The information acquired allows the deterministic solution procedure to arrive at the whole core solution for the eigenvalue and the explicit fuel pin fission density distribution more quickly than other transport methods. Because it solves the transport equation stochastically, complicated geometry may be modeled exactly and therefore heterogeneity even at the most detailed level does not challenge the method. In this dissertation, the method is evaluated using comparisons with full core Monte Carlo reference solutions of benchmark problems based on gas-cooled, graphite-moderated reactor core designs. Solutions are given for core eigenvalue problems, the calculation of fuel pin fission densities throughout the core, and the determination of incremental control rod worth. Using a single processor, results are found in minutes for small cores, and in no more than a few hours for a realistically large core. Typical eigenvalues calculated by the method differ from reference solutions by less than 0.1%, and pin fission density calculations have average accuracy of well within 1%, even for unrealistically challenging core configuration problems. This new method enables the accurate determination of core eigenvalues and flux shapes in hexagonal cores with efficiency far exceeding that of other transport methods.