The current waste management strategy for spent nuclear fuel (SNF) mandated by the U.S. Congress is the disposal of high-level waste (HLW) in a geological repository at Yucca Mountain. Ongoing efforts on closed-fuel cycle options and difficulties in opening and safeguarding such a repository have led to investigations of alternative waste management strategies. One potential strategy would make use of fuels containing transuranic (TRU) nuclides in nuclear reactors. This would prolong reactor operation on a single fuel loading and by doing so, would reduce current HLW stockpiles. The analysis has already shown that high-temperature gas-cooled reactors (HTGRs) and their Generation IV extensions, very-high-temperature reactors (VHTRs), have encouraging performance characteristics that will allow for prolonged operation with no intermediate refueling, as well as for transmutation of TRUs. The objective of this research was to show that TRU-fueled VHTRs have the possibility of prolonged operation on a single fuel loading while retaining their Generation IV safety features. In addition, this research evaluated performance characteristics, and identified operational domains of these systems, as well as the possibility of HLW reduction. A whole-core, 3-D model of a power size prismatic VHTR with a detailed temperature distribution was developed for calculations with the SCALE 5.1 code package. Results of extensive criticality and depletion calculations with multiple fuel loadings showed that VHTRs are capable and suitable for autonomous operation when loaded with TRU fuel.

At the brink of nuclear waste repository crises, viable alternatives for the long term radiotoxic wastes are seriously being considered worldwide. Minor actinides serve as one of these targeted wastes. Partitioning and transmutation in fission reactors is one possible incineration option and could potentially serve as a source of nuclear fuel required for sustainability of energy resources. The objective of this research was to evaluate the neutronic performance of the pebble-bed Very High Temperature Reactor (VHTR) configurations with various fuel loadings. The configuration adjustments and design sensitivity studies specifically targeted the achievability of spectral variations. The development of several realistic full-core 3D models and validation of all modeling techniques used was a major part of this research effort. In addition, investigating design sensitivities helped identify the parameters of primary interest. The full-core 3D models representing the prototype and large scale cores were created for use with SCALE 5.0 and SCALE 5.1 code systems. Initially the models required the external calculation of a Dancoff correction factor; however, the recent release of SCALE 5.1 encompassed inherent double heterogeneity modeling capabilities. The full core 3D models with multi-heterogeneity treatments are in agreement with available pebble-bed High Temperature Test Reactor data and were validated through benchmark studies. Analyses of configurations with various fuel loadings have indicated promising performance and safety characteristics. It was found that through small configuration adjustments, the pebble-bed design can be tweaked to produce desirable spectral shifts. The future operation of Generation IV nuclear energy systems would be greatly facilitated by the utilization of minor actinides as a fuel component. This would offer development of new fuel cycles, and support sustainability of a fuel source.

The Very High Temperature Reactor (VHTR) system behavior should be predicted during normal operating conditions and during transient conditions. To predict the VHTR system behavior there is an urgent need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. This analysis was performed by coupling a RELAP5-3D© VHTR model to a Fluent outlet plenum model. The RELAP5 VHTR model outlet conditions provide the inlet boundary conditions to the Fluent outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled without having to model the entire reactor with a computationally expensive code such as Fluent. The two codes were successfully coupled. The values of pressure, mass flow rate and temperature across the coupled boundary showed only slight differences. The coupling tool used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.

A new methodology for the accurate and efficient determination of steady state thermal hydraulic parameters for prismatic high temperature gas reactors is developed. Two conceptual reactor designs under investigation by the nuclear industry include the General Atomics GT-MHR and the Department of Energy MHTGR-350. Both reactors use the same hexagonal prismatic block, TRISO fuel compact, and circular coolant channel array design. Steady state temperature, pressure, and mass flow distributions are determined for the base reference designs and also for a range of values of the important parameters. Core temperature distributions are obtained with reduced computational cost over more highly detailed computational fluid dynamics codes by using efficient, correlations and first-principles-based approaches for the relevant thermal fluid and thermal transport phenomena. Full core 3-D heat conduction calculations are performed at the individual fuel pin and lattice assembly block levels. The fuel compact is treated as a homogeneous medium with heat generation. A simplified 1-D fluid model is developed to predict convective heat removal rates from solid core nodes. Downstream fluid properties are determined by performing a channel energy balance down the axial node length. Channel exit pressures are then compared and inlet mass flows are adjusted until a uniform outlet pressure is reached. Bypass gaps between assembly blocks as well as coolant channels are modeled. Finite volume discretization of energy, and momentum conservation equations are then formed and explicitly integrated in time. Iterations are performed until all local core temperatures stabilize and global convective heat removal matches heat generation. Several important observations were made based on the steady state analyses for the MHTGR and GT-MHR. Slight temperature variation in the radial direction was observed for uniform radial powers. Bottom-peaked axial power distributions had slightly higher peak temperatures but lower core average temperatures compared to top and center-peaked power distributions. The same trend appeared for large bypass gap sizes cases compared to smaller gap widths. For all cases, peak temperatures were below expected normal operational limits for TRISO fuels. Bypass gap flow for a 3 mm gap width was predicted to be between 10 and 11% for both reactor designs. Single assembly hydrodynamic and temperature results compared favorably with those available in the literature for similar prismatic HTGR thermal hydraulic, computational fluid dynamics analyses. The method developed here enables detailed local and core wide thermal analysis with minimal computational effort, enabling advanced coupled analyses of high temperature reactors with thermal feedback. The steady state numerical scheme also offers a potential for select transient scenario modeling and a wide variety of design optimization studies.

Very high temperature reactor (VHTR) is one of the candidates for Generation IV reactor. It can be continuously operated with average core outlet temperature between 900°C and 950°C, so the core temperature is one of the key features in the design of VHTR. Bypass flow in the prismatic core of VHTR is not a designed feature but it is inevitable due to the combination of several causes and considerably affects the core temperature. Although bypass flow has been studied extensively, the current status of research on thermal/hydraulic core flow of VHTR is far from completion. Present study is the starting of bypass flow characteristic investigation using small-scale model that will fulfill understandings of bypass flow in the prismatic core of VHTR. Bypass flow experiments are conducted by using three small-scale models of prismatic blocks. They are stacked in a test section to form bypass gaps of single-layer blocks as exist in prismatic core of VHTR. Three bypass gap widths set in air and water flow experiments are 6.1, 4.4 and 2.7 mm. Experimental data shows that bypass flow fraction depends on bypass gap width and downstream condition of prismatic blocks, while pressure drop of flow through bypass gaps depends on bypass gap width only. Bypass flow simulations are performed by using STAR-CCM+ software after meshing parameters were determined from simulation exercises and grid independent study. Three turbulence models are employed in all bypass flow simulations which are stopped at physical time of 100 seconds marching by implicit unsteady scheme. Bypass flow fraction, coolant channel Reynolds number and bypass gap Reynolds number from air flow and water flow simulations with 6.1-mm bypass gap width are very close to experimental data. This is because bypass flow fractions from experiments at this bypass gap width are matched in construction of the simulation models. Discrepancies between results from simulations and experiments for remaining gaps increase when bypass gap width becomes smaller. Finally, guidelines for bypass flow experiments and simulations are drawn from the data in present study to improve bypass flow study in the future.

Very High Temperature Rector (VHTR) had been designated as one of those promising reactors for the Next Generation (IV) Nuclear Plant (NGNP). For a prismatic core VHTR, one of the most crucial design considerations is the bypass flow and crossflow effect. The bypass flow occurs when the coolant flow into gaps between fuel blocks. These gaps are formed as a result of carbon expansion and shrinkage induced by radiations and manufacturing and installation errors. Hot spots may appear in the core if the large portion of the coolant flows into bypass gaps instead of coolant channels in which the cooling efficiency is much higher. A preliminary three dimensional steady-state CFD analysis was performed with commercial code STARCCM+ 6.04 to investigate the bypass flow and crossflow phenomenon in the prismatic VHTR core. The k-ε turbulence model was selected because of its robustness and low computational cost with respect to a decent accuracy for varied flow patterns. The wall treatment used in the present work is two-layer all y+ wall treatment to blend the wall laws to estimate the shear stress. Uniform mass flow rate was chose as the inlet condition and the outlet condition was zero gauge pressure outlet. Grid independence study was performed and the results indicated that the discrepancy of the solution due to the mesh density was within 2% of the bypass flow fraction. The computational results showed that the bypass flow fraction was around 12%. Furthermore, the presence of the crossflow gap resulted in a up to 28% reduction of the coolant in the bypass flow gap while mass flow rate of coolant in coolant channels increased by around 5%. The pressure drop at the inlet due to the sudden contraction in area could be around 1kpa while the value was about 180 Pa around the crossflow gap region. The error analysis was also performed to evaluate the accumulated errors from the process of discretization and iteration. It was found that the total error was around 4% and the variation for the bypass flow fraction was within 1%.

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