Safety is a paramount concern in the operation of training and test reactors. A major component of a reactor is the maintenance of safe thermal hydraulic operating conditions. If the temperature of the water coolant exceeds the boiling point, the heat transfer out of the fuel rods into the coolant will greatly decrease and will need to rely upon other safety feedbacks and systems to avoid an accident condition. TRIGA thermal hydraulic systems are currently modeled using a finite differencing code, TRACE/SNAP, developed by the Nuclear Regulatory Commission. While the code is currently certified, it has shortcomings that this work improves upon, notably the simplification of the more complex flow geometries by using circular pipes and a heat transfer correlation that is valid across all flow regimes observed during operation of the TRIGA. A computational fluid dynamics code, FLUENT, along with real-time thermocouple probe measurements of the channel were used to solve both of these major issues. A high resolution model of four adjacent flow channels was created to provide a numerical experimental data set for enhancing the correlations used in the TRACE model. The hot flow channel is connected to three surrounding channels where crossflow occurs causing a more complex flow pattern than the isolated single channel system used in TRACE/SNAP. To calibrate the FLUENT model, a thermocouple probe was designed and placed in the TRIGA core in the center of the flow channel. The reactor was operated over the full range of licensed power levels to obtain a fully encompassing data set of coolant temperatures. The FLUENT model was then adjusted so that the temperatures at the location of the probe in the model matched those from the experimental measurements. Based on the results from the FLUENT testing, data was extracted to develop a new heat transfer coefficient correlation and loss factor coefficient correlation due to the non-circular geometry and fuel rod end fittings for use in the TRACE/SNAP code. These adjustments were then implemented into TRACE/SNAP to improve the code for future users performing safety analysis on TRIGA reactors. / text

RELAP5-3D has been coupled with both DAKOTA and STAR-CCM+ in order to expand the capability of the thermal-hydraulic code and facilitate complex studies of desired systems. In the first study, RELAP5-3D was coupled with DAKOTA to perform a sensitivity study of the South Texas Project (STP) power plant during steady-state and transient scenarios. The coupled software was validated by analyzing the simulation results with respect of the physical expectations and behavior of the power plant, and thermal-hydraulic parameters which caused greatest sensitivity where identified: inlet core temperature and reactor thermal power. These variables, along with break size and discharge coefficients, were used for further investigation of the sensitivity of the RELAP5-3D LOCA transient simulation under three difference cases: two inch break, six inch break, and guillotine break. Reactor thermal power, core inlet temperature, and break size were identified as producing the greatest sensitivity; therefore, future research would include uncertainty quantification for these parameters. In the second study, a small scale experimental facility, designed to study the thermal hydraulic phenomena of the Reactor Cavity Cooling System (RCCS) for a Very High Temperature Reactor (VHTR), was used as a model to test the capabilities of coupling Star-CCM+ and RELAP5-3D. This chapter discusses the capabilities and limitations of the STAR-CCM+/RELAP5-3D coupling, and a simulation, on the RCCS facility, was performed using STAR-CCM+ to study the flow patterns where expected complex flow phenomena occur and RELAP5-3D for the complete system. The code showed inability to perform flow coupling simulations and it is unable, at this time, to handle closed loop systems. The thermal coupling simulation was successful and showed congruent qualitative results to physical expectations. The locations of large fluid vortices were located specifically in the pipes closest to the inlet of the bottom manifold. In conclusion, simulations using coupled codes were presented which greatly improved the capabilities of RELAP5-3D stand-alone and computational time required to perform complex thermal-hydraulic studies. These improvements show greatly benefit for industrial applications in order to perform large scale thermal-hydraulic systems studies with greater accuracy while minimizing simulation time.

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.

A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT) - type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs. / UOIT

A new correlation for the prediction of the Countercurrent Flow Limitation (CCFL) in a large diameter tube with a falling water lm is proposed. Dierent from previous correlations, it predicts the onset of ooding by considering the relative velocities of the working uids and the lm thickness of the liquid layer. This provides a more complete accounting of the physical forces contributing to CCFL. This work has been undertaken in order to provide a better estimate of CCFL for reactor safety codes such as MELCOR, MAAP, and SCDAP/RELAP. Experiments were conducted to determine the CCFL for a 3-inch inner diameter smooth tube with an annular liquid lm and air injection from the bottom. The size of the test section and the range of working uid ow rates were based on a scaling analysis of the surge line of a PressurizedWater Reactor pressurizer. An experimental facility was designed and constructed based on this analysis in order to collect data on the CCFL phenomenon. In order to capture some of the physical phenomena at the onset of ooding visual pictures were taken at high speed. These pictures provided a new understanding of the process of transition to ooding. The facility also produced a new set of ooding data. This can also lead to a more comprehensive mechanistic model.

Experimental thermal-hydraulic study of a rectangular supercritical CO2 natural-circulation loop with a horizontal heated channel was conducted at different steady-state conditions. These included different system pressures and three different inlet temperatures, with different inlet and outlet valve openings. Approximately, 450 experimental steady-state data-points were collected. The data include measurements of pressure-drop along the heated channel, pressure-drop across inlet and outlet valves, applied heat on the heated channel, pressure, temperature and flow-rate. Steady-state curves of mass flow-rate versus power, outlet temperature versus power, and detailed information of frictional pressure drop and local head loss coefficients were produced. Comparison showed that for the available experimental set-up, computed frictional pressure-drops fell within 1-1.20 of the Blasius formula prediction. Moreover, flow oscillations were observed in several cases when outlet temperature of CO2 was higher than the pseudo-critical temperature on the negative slope part of the mass flow-rate versus power curve.

Made available in DSpace on 2014-10-09T12:54:44Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:59Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:54:44Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:59Z (GMT). No. of bitstreams: 0 / Este trabalho descreve o estudo e testes de um sistema capaz de identificar e classificar os transientes, ou estados transitórios, de sistemas termo-hidráulicos, utilizando a técnica de redes neurais artificiais do tipo mapas de características auto-organizáveis, com o objetivo de sua implantação nas novas gerações de reatores nucleares. A técnica desenvolvida neste trabalho consiste no uso de múltiplas redes para fazer a classificação e identificação dos estados transitórios, sendo cada uma especialista em um respectivo transitório do sistema, que competem entre si por meio do erro de quantização, que é uma medida gerada por estas redes neurais. Esta técnica se mostrou eficiente, apresentando características muito promissoras no que diz respeito ao desenvolvimento de novas funcionalidades em futuros projetos. Uma dessas características consiste no potencial de que a rede, além de responder qual estado transitório está em curso, também pode oferecer informações adicionais sobre esse transitório. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP

Density wave oscillations (DWO) are a class of two-phase flow instabilities which can pose significant safety concerns to boiling water reactors (BWR). During an anticipated transient without scram (ATWS) while operating in the proposed extended operating domain MELLLA+, natural circulation conditions can potentially lead to DWO-type instabilities which have the capability to develop into cycles of fuel surface dryout and rewet, damaging core integrity. In order to provide data on these phenomena, a series of tests were performed at the KATHY facility during which DWO was developed with and without simulated neutronic feedback. In this dissertation, the data provided by these tests is analyzed to determine the onset conditions for DWO. Following this, several models are assessed for their capability in predicting this stability boundary compared to the experimental results. The models were chosen in order to provide a suitably large range of prediction methodologies. Two analytical drift-flux models developed with and without thermal equilibrium are shown, with respective differences compared. A computational model of the full KATHY natural circulation loop is built using the 1D thermal-hydraulics code TRACE. This is adapted with a point-kinetics model for neutronic feedback for experimental comparison. With both the analytical models and the TRACE model, a series of parametric studies are performed showing the effects of inlet/outlet flow restrictions, pressure, channel geometry, and axial power profile on the stability boundary. Finally, two machine learning neural network-based models are developed and trained on various subsets of the experimental data. The results from each model showed certain benefits and drawbacks based on model complexity and physicality. / Doctor of Philosophy / Certain conditions in the core of a boiling water reactor (BWR) can lead to unstable flows due to the high ratio between the power and the coolant flow rate. These instabilities, called density wave oscillations (DWO), have been shown to occur during a specific accident scenario known as an anticipated transient without scram (ATWS) when the reactor is operating in a lower flow domain called MELLLA+. In this accident, pump flow through the core is halted, but the reactor is not shut down. This can lead to serious safety concerns if left unaddressed. To analyze these instabilities, the KATHY facility performed a series of tests with and without power feedback from simulated neutron response. In this dissertation, the onset conditions from these tests are given and compared to several models for predicting the stability boundary. Two analytical models proposed by Ishii and Saha are compared and the effect of certain parameters on the stability is assessed. Next, a model of the KATHY loop is built using the thermal-hydraulics code TRACE both with and without simulated power feedback. Finally, two types of machine learning models are developed to determine their accuracy in predicting the instability conditions. The overall performance of each is compared and their benefits and drawbacks are highlighted.

Under normal operating conditions a CANDU reactor pressure tube (PT) is insulated from its outer calandria tube (CT) by a CO2 gas annulus. If the primary loop coolant flow is compromised the PT can overheat and, if still pressurized, balloon into contact with the CT. At this point the moderator acts as an emergency heat sink. If the heat transferred from the CT to the moderator exceeds the critical heat flux (CHF) the CT can overheat, begin to strain due to the contact pressure, and eventually fail. A mechanistic model is presented that describes ballooning contact of the PT and CT, the resulting thermal contact conductance, heat flux to the moderator, and, if CHF is exceeded, the development of film boiling and potential CT strain. The goal is to create a software package that predicts fuel channel failure during a pressure tube overheat event. / Thesis / Master of Applied Science (MASc) / Computer software was developed to predict CANDU fuel channel failure in the event of a total station blackout. The model created successfully predicted the available experimental data.