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

Nuclear reactors with improved safety concepts are currently being studied within the nuclear engineering community, with a focus on passive safety features. One of these reactor concepts is the Very High Temperature gas-cooled Reactor (VHTR) of which the Reactor Cavity Cooling Systems (RCCS) is seen as an integral and crucial part of the passive safety concept. Considerable validation and development of the necessary software tools is required to perform analysis and designs of these future reactor concepts. The primary objective of this study is to establish a methodology for the creation of an integrated system level process model of a typical air-cooled RCCS in Flownex®, and to illustrate its applicability by simulating different scenarios that illustrate the operational characteristics of such a system. For this purpose, the existing RCCS conceptual design that is being studied by the KAERI was used as the case study. As a start, selected case studies were performed to verify that the Flownex® models were set up correctly to perform natural circulation flows, both in steady and transient conditions, and with radiation, convection and conduction taking part. These are the major typical physical phenomena in the RCCS. The models were compared with EES (Engineering Equation Solver) models of the same geometries and specifications. There was a good agreement between Flownex® and EES model results. After this verification, a simulation model of the integrated RCCS system was developed. The Flownex® models were applied to model selected possible operational scenarios. The major observations from the results are that: - The RCCS carries with it enough heat to the ambient such that the concrete wall temperature is maintained below the benchmark value of 65°C for the different boundary conditions imposed. - The RCCS maintains its functionality even with three quarters of the risers blocked or in the event that there is a break in one of the chimney pipes. / 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

Nuclear reactors with improved safety concepts are currently being studied within the nuclear engineering community, with a focus on passive safety features. One of these reactor concepts is the Very High Temperature gas-cooled Reactor (VHTR) of which the Reactor Cavity Cooling Systems (RCCS) is seen as an integral and crucial part of the passive safety concept. Considerable validation and development of the necessary software tools is required to perform analysis and designs of these future reactor concepts. The primary objective of this study is to establish a methodology for the creation of an integrated system level process model of a typical air-cooled RCCS in Flownex®, and to illustrate its applicability by simulating different scenarios that illustrate the operational characteristics of such a system. For this purpose, the existing RCCS conceptual design that is being studied by the KAERI was used as the case study. As a start, selected case studies were performed to verify that the Flownex® models were set up correctly to perform natural circulation flows, both in steady and transient conditions, and with radiation, convection and conduction taking part. These are the major typical physical phenomena in the RCCS. The models were compared with EES (Engineering Equation Solver) models of the same geometries and specifications. There was a good agreement between Flownex® and EES model results. After this verification, a simulation model of the integrated RCCS system was developed. The Flownex® models were applied to model selected possible operational scenarios. The major observations from the results are that: - The RCCS carries with it enough heat to the ambient such that the concrete wall temperature is maintained below the benchmark value of 65°C for the different boundary conditions imposed. - The RCCS maintains its functionality even with three quarters of the risers blocked or in the event that there is a break in one of the chimney pipes. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

The steam generator plays a primary role in the safety and performance of a pressurized water reactor nuclear power plant. The cost to utilities is in the order of millions of Rands a year as a direct result of damage to steam generators. The damage results in lower efficiency or even plant shutdown. It is necessary for the utility and for academia to have models of nuclear components by which research and analysis may be performed. It must be possible to analyse steam generator performance for both day-to-day operational analysis as well as in the case of extreme accident scenarios. The homogeneous model for two-phase flow is simpler in its implementation than the two-fluid model, and therefore suffers in accuracy. Its advantage lies in its quick turnover time for development of models and subsequent analysis. It is often beneficial for a modeller to be able to quickly set up and analyse a model of a system, and a trade-off between accuracy and time-management is thus required. Searches through available literature failed to provide answers to how the homogeneous model compares with the two-fluid model for operational and safety analysis. It is expected to see variations between the models, from the analysis of the mathematics, but it remains to be shown what these differences are. The purpose of this study was to determine how the homogeneous model for two-phase flow compares with the two-fluid model when applied to a u-tube steam generator of a typical pressurized water reactor. The steam generator was modelled in both RELAP5 and in Flownex. A custom script was written for Flownex in order to implement the Chen correlation for boiling heat transfer. This was significantly less detailed than RELAP5’s solution of a matrix of flow regimes and heat transfer correlations. The geometry of the models were based on technical drawings from Koeberg Nuclear Power Plant, and were simplified to a one-dimensional model. Plant data obtained from Koeberg was used to validate the models at 100%, 80% and 60% power output. It was found that the overall heat transfer rate predicted with the RELAP5 two-fluid model was within 1.5% of the measured data from the Koeberg plant. The results generated by the homogeneous model for the overall heat transfer were within 4.5% of the measured values. However, the differences in the detailed temperature distributions and heat transfer coefficient values were quite significant at the inlet and outlet ends of the tube bundle, at the bottom tube sheet of the steam generator. In this area the water-level was not accurately modelled by the homogeneous model, and therefore there was an under-prediction in heat transfer in that region. Large differences arose between the Flownex and RELAP5 solutions due to difference in the heat transfer correlations used. The Flownex model exclusively implemented the Chen correlation, while RELAP5 implements a flow regime map correlated to a table of heat transfer correlations. It was concluded that the results from the homogeneous model for two-phase flow do not differ significantly when compared with the two-fluid model when applied to the u-tube steam generator at the normal operating conditions. Significant differences do, however, occur in lower regions of the boiler where the quality is lower. We conclude that the homogeneous model offers significant advantage in simplicity over the two-fluid model for normal operational analysis. This may not be the case for detailed accident analysis, which was beyond the scope of this study. / Thesis (MIng (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2013.

The steam generator plays a primary role in the safety and performance of a pressurized water reactor nuclear power plant. The cost to utilities is in the order of millions of Rands a year as a direct result of damage to steam generators. The damage results in lower efficiency or even plant shutdown. It is necessary for the utility and for academia to have models of nuclear components by which research and analysis may be performed. It must be possible to analyse steam generator performance for both day-to-day operational analysis as well as in the case of extreme accident scenarios. The homogeneous model for two-phase flow is simpler in its implementation than the two-fluid model, and therefore suffers in accuracy. Its advantage lies in its quick turnover time for development of models and subsequent analysis. It is often beneficial for a modeller to be able to quickly set up and analyse a model of a system, and a trade-off between accuracy and time-management is thus required. Searches through available literature failed to provide answers to how the homogeneous model compares with the two-fluid model for operational and safety analysis. It is expected to see variations between the models, from the analysis of the mathematics, but it remains to be shown what these differences are. The purpose of this study was to determine how the homogeneous model for two-phase flow compares with the two-fluid model when applied to a u-tube steam generator of a typical pressurized water reactor. The steam generator was modelled in both RELAP5 and in Flownex. A custom script was written for Flownex in order to implement the Chen correlation for boiling heat transfer. This was significantly less detailed than RELAP5’s solution of a matrix of flow regimes and heat transfer correlations. The geometry of the models were based on technical drawings from Koeberg Nuclear Power Plant, and were simplified to a one-dimensional model. Plant data obtained from Koeberg was used to validate the models at 100%, 80% and 60% power output. It was found that the overall heat transfer rate predicted with the RELAP5 two-fluid model was within 1.5% of the measured data from the Koeberg plant. The results generated by the homogeneous model for the overall heat transfer were within 4.5% of the measured values. However, the differences in the detailed temperature distributions and heat transfer coefficient values were quite significant at the inlet and outlet ends of the tube bundle, at the bottom tube sheet of the steam generator. In this area the water-level was not accurately modelled by the homogeneous model, and therefore there was an under-prediction in heat transfer in that region. Large differences arose between the Flownex and RELAP5 solutions due to difference in the heat transfer correlations used. The Flownex model exclusively implemented the Chen correlation, while RELAP5 implements a flow regime map correlated to a table of heat transfer correlations. It was concluded that the results from the homogeneous model for two-phase flow do not differ significantly when compared with the two-fluid model when applied to the u-tube steam generator at the normal operating conditions. Significant differences do, however, occur in lower regions of the boiler where the quality is lower. We conclude that the homogeneous model offers significant advantage in simplicity over the two-fluid model for normal operational analysis. This may not be the case for detailed accident analysis, which was beyond the scope of this study. / Thesis (MIng (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2013.

It is a great challenge in the design of the PBMR to accurately predict gas flow and heat transfer in the reactor. Understanding the heat transfer at the core-reflector interface in particular is a very important aspect as the reactivity of the control rods housed in the reflectors is highly temperature dependent. It is also very important because the core-reflector interface is on the critical path for heat removal during accident conditions. PBMR has developed an OECD/NEA coupled neutronic/thermal-hydraulic benchmark to aid in the understanding of the different modelling approaches currently employed at PBMR. A comparison of THERMIX-KONVEK and DIREKT results showed large temperature differences at the core-reflector interfaces. Further investigation showed that these differences are as a result of the numerical methods used i.e. Cell-Centred (CC) vs. Vertex-Centered (VC). The present study extended this comparison to Star-CD (CC) and Flownex (VC) which are also used to simulate the reactor at PBMR. An ID MATLAB program that mimics the CC and VC numerical methods was verified against Star-CD and Flownex. This program was then used to model an ID version of the OECD/NEA benchmark. Results were compared with DIREKT and THERMIX-KONVEK. Although the results compared well, there were significant errors at the core-reflector interfaces. The findings of this study were that different numerical methods will predict different temperatures, heat fluxes and (temperature-dependent) sink terms. It was also shown that in addition to the differences resulting from numerical methods, differences were seen between Star-CD and DIREKT and Flownex and THERMIX-KONVEK in the region of the core-reflector boundary. In general, for complicated simulations like that of the pebble bed, the numerical basis of software used to simulate the problem needs to be understood for the problem to be correctly modelled. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2010.

It is a great challenge in the design of the PBMR to accurately predict gas flow and heat transfer in the reactor. Understanding the heat transfer at the core-reflector interface in particular is a very important aspect as the reactivity of the control rods housed in the reflectors is highly temperature dependent. It is also very important because the core-reflector interface is on the critical path for heat removal during accident conditions. PBMR has developed an OECD/NEA coupled neutronic/thermal-hydraulic benchmark to aid in the understanding of the different modelling approaches currently employed at PBMR. A comparison of THERMIX-KONVEK and DIREKT results showed large temperature differences at the core-reflector interfaces. Further investigation showed that these differences are as a result of the numerical methods used i.e. Cell-Centred (CC) vs. Vertex-Centered (VC). The present study extended this comparison to Star-CD (CC) and Flownex (VC) which are also used to simulate the reactor at PBMR. An ID MATLAB program that mimics the CC and VC numerical methods was verified against Star-CD and Flownex. This program was then used to model an ID version of the OECD/NEA benchmark. Results were compared with DIREKT and THERMIX-KONVEK. Although the results compared well, there were significant errors at the core-reflector interfaces. The findings of this study were that different numerical methods will predict different temperatures, heat fluxes and (temperature-dependent) sink terms. It was also shown that in addition to the differences resulting from numerical methods, differences were seen between Star-CD and DIREKT and Flownex and THERMIX-KONVEK in the region of the core-reflector boundary. In general, for complicated simulations like that of the pebble bed, the numerical basis of software used to simulate the problem needs to be understood for the problem to be correctly modelled. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2010.

Large coal-fired power stations are designed to be run predominantly at full load and optimum conditions. The behaviour of plants, operating at low load and varying conditions, is getting more and more attention due to the introduction of variable renewable generation on the grid. Consequently, the need for a fully transient high-fidelity system based model has grown, as this will enable one to study the behaviour of plants under such non-ideal conditions. This report details the development of a feedwater heater, deaerator and turbine component for such a high-fidelity transient system model using the Flownex Simulation Environment, a onedimensional thermohydraulic network solver. The components have been modelled all with the aim of using minimal design input data. The feedwater heater component model includes transient effects and thermodynamic relations to represent aspects such as heater performance, level control and transient inertia. In determining the heat transfer characteristics, the model makes use of plant-performance data and correlates the amount of heat transfer by using the feedwater mass flow as the load indicating parameter. This approach eliminates the need for specific geometrical details to calculate the effective heat transfer area. The level control is modelled by using a level representation built from using heat exchanger design methods. The turbine component is modelled by using Fuls’ Semi-Ellipse law or the pressure drop modelling and Ray’s semi-empirical method for the efficiency modelling. The model also contains transient effects, which include thermal inertia due to the shaft and casing, and rotational inertia due to the shaft. The deaerator component is modelled by adapting the model presented by Banda, and modifying the model to work under various conditions. This involved using curve fit methods in Flownex to use input data to model the pressure drop over the main condensate valve. Each of the mentioned components was validated and verified with plant data and finally packaged into a compound component which is a component consisting of a subnetwork in Flownex. These compound components further contain design inputs which are easily accessible by the user. The component models were integrated into larger networks in which various scenarios can be run. A short transient scenario was run on the low-pressure feedwater train of a specific power station. The scenario involved a turbine trip where the bled steam valves for the heaters were closed suddenly. The speed of the valves closing was however unknown and after closing the valves in approximately 10 seconds, results agreed relatively well with plant data. This illustrated the short transient capabilities of the feedwater heater component model. The three component models (feedwater heater, turbine and deaerator) were finally integrated into a regenerative Rankine cycle and was set up using minimal design data. The boiler, condenser and condensate pump were set as boundary conditions in the network but all extraction points for the network were connected. Steady-state results were obtained for various load cases and the main temperature, flow and pressure results were compared. Results agree well with plant data, even at low load conditions

Boiler feed pump turbines (BFPTs) are in use at a number of Eskom power stations. They utilise bled steam extracted from the main turbine in order to drive multistage centrifugal pumps which supply the boilers with feedwater. With an increase of renewables in the energy mix, the need for Eskom’s coal-fired power stations to run for extended periods at very low loads has arguably never been this great. Various systems affect the ability of these generation units to run economically at low loads. One such system is the boiler feed pump turbine and its associated pumps. A station was selected from Eskom’s fleet based on access to information and the station being a relatively typical plant. The Unit (a boiler and turbogenerator set) selected for study was one with the most thorough instrumentation available for remote monitoring. The BFPT system of this Unit was modelled in Flownex, a one-dimensional thermofluid process modelling package. The model included individual pump stages, steam admission valves and a stage-by-stage turbine model utilising custom stage components. These turbine stage components represent each stage with nozzles and other standard Flownex components. The boundary conditions of the system were set as functions of generator load in order to represent typical values for use in case studies. The relationships between load and boundary conditions were based on large samples of data from the station’s data capture system (DCS). A corresponding standby electric feed pump system was also modelled in Flownex for a comparative case study. After model validation, a number of case studies were performed, demonstrating the functionality of the model and also providing specific results of value to the station in question. These results include the minimum generator load possible with different steam supplies; maximum condenser back pressure before plant availability is affected; the viability of changing the pump leak-off philosophy; and the effect of electric feed pump use on power consumption. The main recommendations from the case studies were as follows: i. to stroke the steam admission valves as per the design charts, ii. to test the operation of the BFPT down to 40 % generator load, iii. to keep the pump leak-off philosophy unchanged, iv. to maintain the cooling water system and condensers sufficiently to avoid poor condenser vacuum, v. to reconsider the decommissioning of the “cold reheat” steam supply, vi. and, to favour use of the BFPT over the electric feed pumps at all generator loads.