The double ended guillotine break leading to density-driven air ingress has been identified as a low probability yet high consequence event for Very High Temperature Reactor (VHTR). The lower plenum of the VHTR contains the core support structure and is composed of graphite. During an air ingress event, oxidation of the graphite structure under high temperature conditions in an oxygen containing environment could degrade the integrity of the core support structure. Following this large break, air from the reactor containment will begin to enter the lower plenum via two mechanisms: diffusion or density driven stratified flow. The large difference in time scales between the mechanisms leads to the need to perform high fidelity experimental studies to investigate the dominant the air ingress mechanism. A scaled test facility has been designed and built that allows the acquisition of velocity measurements during stratification after a pipe break. A non-intrusive optical measurement technique provides full-field velocity measurements profiles of the two species Particle Image Velocimetry (PIV). The data allow a more developed understanding of the fundamental flow features, the development of improved models, and possible mitigation strategies in such a scenario.Two brine-water experiments were conducted with different break locations. Flow fronts were analyzed and findings concluded that the flow has a constant speed through the pipe after the initial lock exchange. The time in which the flow enters the lower plenum is an important factor because it provides the window of opportunity for mitigation strategies in an actual reactor scenario. For both cases the flow of the heavier density liquid (simulating air ingress from the reactor containment) from the pipe enters the reactor vessel in under 6 seconds. The diffusion velocity and heavy flow front of the stratified flow layer were compared for the SF6/He gas case. It is seen that the diffusion plays less of a role as the transport mechanism in comparison to the density-driven stratified flow since the velocity of the diffusion is two orders of magnitude smaller than the velocity of the stratified flow mechanism. This is the reason for the need for density-driven stratified flow investigations following a LOCA. These investigations provided high-quality data for CFD validation in order for these models to depict the basic phenomena occurring in an air ingress scenario.

Two topics from nuclear engineering field are included in this dissertation. One study is the air-ingress phenomenon during a loss of coolant accident (LOCA) scenario, and the other is a 5-by-5 bundle assembly problem under a design of PWRs. The objectives are to investigate the Kelvin-Helmholtz instability of the gravity-driven stratified flows inside a coaxial pipe and the effects caused by two types of spacers at the downstream of the rod bundle problem. Richardson extrapolation is used for the grid independent study. Simulation results give good agreements with the experiments. Wavelet analysis and Proper Orthogonal Decomposition (POD) are used to study the flow behaviors and flow patterns. For the air-ingress phenomenon, Brunt-Vaisala frequency, or buoyancy frequency, predicts a frequency of 2.34 Hz, which is confirmed by the dominant frequency of 2.4 Hz obtained from the wavelet analysis between times 1.2 s and 1.85 s. For the rod bundle study, the dominant frequency at the center of the subchannel is given as 2.4 Hz with a secondary dominant frequency of 4 Hz and a much minor frequency of 6 Hz. Generally, wavelet analysis has much better performance than POD in the air-ingress phenomenon that is a strongly transient scenario; they both appropriate for the rod bundle study. Based on this study, when the fluid pair in a real condition is used, the time which air intrudes into the reactor is predictable.

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This study investigates the development of a strategy to simulate nuclear graphite oxidation with Computational Fluid Dynamics (CFD) to determine an estimate of graphite lost. The task was achieved by comparing the results of the CFD approach with a number of different experiments. For molecular diffusion, simulated results were compared to analytical solutions. Mass flow rates under conditions of natural convection were sourced from the 2002 NACOK experiment. Experimental data from the KAIST facility were sourced for the basic oxidation of graphite in a controlled environment. Tests included the reactions of carbon with oxygen and with carbon dioxide. Finally, the tests at NACOK from 2004 and 2005 were chosen for comparison for the simulation of oxidation. The 2005 test considered two reacting pebble bed regions at different temperatures. The 2004 test included multiple detailed structural graphite. Comparison of results indicated that the phenomenon of diffusion can be correctly simulated. The general trends of the mass flow rates under conditions of natural convection were obtained. Surface reaction rates were defined with user functions in Fluent. Good comparisons of the simulated and the KAIST experimental results were obtained. For the 2005 NACOK comparison, the pebble bed regions were simulated with a porous medium approach. Results showed that correct trends and areas of oxidation were estimated. The 2004 tests were with a combination of a porous medium and surface reaction approaches. More detailed oxidation experimental data would possibly improve the accuracy of the results. This research has shown that the CFD approach developed in the present study can identify areas of maximum oxidation although the accuracy needs to be improved. Both the porous and detailed surface reaction approaches produced consistent results. The limitations of the approach were discussed. These included transient phenomena which were estimated with steady state simulations, and the effects of change in geometry were not considered. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2010.

This study investigates the development of a strategy to simulate nuclear graphite oxidation with Computational Fluid Dynamics (CFD) to determine an estimate of graphite lost. The task was achieved by comparing the results of the CFD approach with a number of different experiments. For molecular diffusion, simulated results were compared to analytical solutions. Mass flow rates under conditions of natural convection were sourced from the 2002 NACOK experiment. Experimental data from the KAIST facility were sourced for the basic oxidation of graphite in a controlled environment. Tests included the reactions of carbon with oxygen and with carbon dioxide. Finally, the tests at NACOK from 2004 and 2005 were chosen for comparison for the simulation of oxidation. The 2005 test considered two reacting pebble bed regions at different temperatures. The 2004 test included multiple detailed structural graphite. Comparison of results indicated that the phenomenon of diffusion can be correctly simulated. The general trends of the mass flow rates under conditions of natural convection were obtained. Surface reaction rates were defined with user functions in Fluent. Good comparisons of the simulated and the KAIST experimental results were obtained. For the 2005 NACOK comparison, the pebble bed regions were simulated with a porous medium approach. Results showed that correct trends and areas of oxidation were estimated. The 2004 tests were with a combination of a porous medium and surface reaction approaches. More detailed oxidation experimental data would possibly improve the accuracy of the results. This research has shown that the CFD approach developed in the present study can identify areas of maximum oxidation although the accuracy needs to be improved. Both the porous and detailed surface reaction approaches produced consistent results. The limitations of the approach were discussed. These included transient phenomena which were estimated with steady state simulations, and the effects of change in geometry were not considered. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2010.

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Hitesh Bindra / Helium-cooled, graphite moderated reactors have been considered for a future fleet of high temperature and high efficiency nuclear power plants. Nuclear-grade graphite is used in these reactors for structural strength, neutron moderation, heat transfer and, within a helium environment, has demonstrated stability at temperatures well above HTGR operating conditions. However, in the case of an air ingress accident, the oxygen introduced into the core can affect the integrity of the fuel graphite matrix. In this work a combination of computational models and mixed effects experiments were used to better understand the air ingress process and its potential effects on the heat removal capabilities of an HTGR design following an air-ingress accident. Contributions were made in the understanding of the air-ingress phenomenon, its potential effects on graphite, and in experimental and computational techniques. The first section of this thesis focuses on experimental and computational studies that were undertaken to further the understanding of the Onset of Natural Convection (ONC) phenomenon expected to occur inside of an HTGR following an air ingress accident. The effects of two newly identified factors on ONC – i.e., the existence of the large volume of stagnate helium in a reactor's upper plenum, and the possibility of an upper head leak – were investigated. Mixed-effects experimental studies were performed to determine the changes induced in nuclear grade graphite exposed to high-temperature, oxidizing flow of varying flow rates. Under all scenarios, the thermal diffusivity of the graphite test samples was shown to increase. Thermal conductivity changes due to oxidation were found to be minor in the tested graphite samples – especially compared to the large drop in thermal conductivity the graphite is expected to experience due to irradiation. Oxidation was also found to increase the graphite's surface roughness and create a thin outer layer of decreased density. The effects of thermal contacts on the passive cooling ability of an HTGR were experimentally investigated. Conduction cool down experiments were performed on assemblies consisting of a number of rods packed into a cylindrical tube. Experimental conditions were then modeled using several different methodologies, including a novel graph laplacian approach, and their results compared to the experimentally obtained temperature data. Although the graph laplacian technique shows great promise, the 2–D Finite Element Model (FEM) provided the best results. Finally, a case study was constructed in which a section of a pebble bed reactor consisting of a number of randomly packed, spherical fuel particles was modeled using the validated FEM technique. Using a discrete elements model, a stable, randomly packed geometry was created to represent the pebble bed. A conduction cool down scenario was modeled and the results from the FEM model were compared to best possible results obtainable from a more traditional, homogeneous 1–D approximation. When the graphite in the bed was modeled as both oxided and irradiated, the homogeneous method mispredicted the maximum temperature given by the 3–D, FEM model by more than 100°C.

Der VHTR als Weiterentwicklung des HTR gilt als eines von sechs aussichtsreichen Reaktorkonzepten für Kernkraftwerkte der Generation IV. Im Rahmen dieser Arbeit wird ein CFD-Modell des HTR-Moduls entwickelt und durch die Simulation eines postulierten Lufteinbruchszenarios die Anwendbarkeit unter Beweis gestellt. Zunächst wird eine Bestandsaufnahme bestehender HTR-Rechenprogramme vorgestellt und die Methodik CFD in ihren Grundzügen erläutert. Anhand der Grundgleichungen werden die zur Berechnung des Störfalls zu modellierenden, HTR-spezifischen Parameter diskutiert, die verwendeten empirischen Korrelationen vorgestellt und die umfangreiche Validierung des entwickelten Modellansatzes zusammengefasst. Anschließend wird die Anwendbarkeit des HTR-Modells auf ein konkretes Lufteinbruchszenario eines HTR-Moduls gezeigt. Dabei werden die einzelnen Phasen des Szenarios anhand der Simulationsergebnisse intensiv diskutiert. Abschließend erfolgt eine Diskussion der Modellunsicherheiten und der numerischen Fehler.

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Der VHTR als Weiterentwicklung des HTR gilt als eines von sechs aussichtsreichen Reaktorkonzepten für Kernkraftwerkte der Generation IV. Im Rahmen dieser Arbeit wird ein CFD-Modell des HTR-Moduls entwickelt und durch die Simulation eines postulierten Lufteinbruchszenarios die Anwendbarkeit unter Beweis gestellt. Zunächst wird eine Bestandsaufnahme bestehender HTR-Rechenprogramme vorgestellt und die Methodik CFD in ihren Grundzügen erläutert. Anhand der Grundgleichungen werden die zur Berechnung des Störfalls zu modellierenden, HTR-spezifischen Parameter diskutiert, die verwendeten empirischen Korrelationen vorgestellt und die umfangreiche Validierung des entwickelten Modellansatzes zusammengefasst. Anschließend wird die Anwendbarkeit des HTR-Modells auf ein konkretes Lufteinbruchszenario eines HTR-Moduls gezeigt. Dabei werden die einzelnen Phasen des Szenarios anhand der Simulationsergebnisse intensiv diskutiert. Abschließend erfolgt eine Diskussion der Modellunsicherheiten und der numerischen Fehler.