Pursuant to the energy policy act of 2005, the High Temperature Gas-Cooled Reactor (HTGR) has been selected as the Very High Temperature Reactor (VHTR) that will become the Next Generation Nuclear Plant (NGNP). Although plans to build a demonstration plant at Idaho National Laboratories (INL) are currently on hold, a cooperative agreement on HTGR research between the U.S. Nuclear Regulatory Commission (NRC) and several academic investigators remains in place. One component of this agreement relates to validation of systems-level computer code modeling capabilities in anticipation of the eventual need to perform HTGR licensing analyses. Because the NRC has used MELCOR for LWR licensing in the past and because MELCOR was recently updated to include gas-cooled reactor physics models, MELCOR is among the system codes of interest in the cooperative agreement. The impetus for this thesis was a code-to-experiment validation study wherein MELCOR computer code predictions were to be benchmarked against experimental data from a reduced-scale HTGR testing apparatus called the High Temperature Test Facility (HTTF). For various reasons, HTTF data is not yet available from facility designers at Oregon State University, and hence the scope of this thesis was narrowed to include only computational studies of the HTTF and its prototype, General Atomics’ Modular High Temperature Gas-Cooled Reactor (MHTGR). Using the most complete literature references available for MHTGR design and using preliminary design information on the HTTF, MELCOR input decks for both systems were developed. Normal and off-normal system operating conditions were modeled via implementation of appropriate boundary and inititial conditions. MELCOR Predictions of system response for steady-state, pressurized conduction cool-down (PCC), and depressurized conduction cool-down (DCC) conditions were checked against nominal design parameters, physical intuition, and some computational results available from previous RELAP5-3D analyses at INL. All MELCOR input decks were successfully built and all scenarios were successfully modeled under certain assumptions. Given that the HTTF input deck is preliminary and was based on dated references, the results were altogether imperfect but encouraging since no indications of as yet unknown deficiencies in MELCOR modeling capability were observed. Researchers at TAMU are in a good position to revise the MELCOR models upon receipt of new information and to move forward with MELCOR-to-HTTF benchmarking when and if test data becomes available.

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.