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.

High Temperature Gas-cooled Reactors (HTGRs) can provide clean electricity,as well as process heat that can be used to produce hydrogen for transportation and other sectors. A prototypic HTGR, the Next Generation Nuclear Plant (NGNP),will be built at Idaho National Laboratory.The need for HTGR analysis tools and methods has led to the addition of gas-cooled reactor (GCR) capabilities to the light water reactor code MELCOR. MELCOR will be used by the Nuclear Regulatory Commission licensing of the NGNP and other HTGRs. In the present study, new input techniques have been developed for MELCOR HTGR analysis. These new techniques include methods for modeling radiation heat transfer between solid surfaces in an HTGR, calculating fuel and cladding geometric parameters for pebble bed and prismatic block-type HTGRs, and selecting appropriate input parameters for the reflector component in MELCOR. The above methods have been applied to input decks for a water-cooled reactor cavity cooling system (RCCS); the 400 MW Pebble Bed Modular Reactor (PBMR), the input for which is based on a code-to-code benchmark activity; and the High Temperature Test Facility (HTTF), which is currently in the design phase at Oregon State University. RCCS results show that MELCOR accurately predicts radiation heat transfer rates from the vessel but may overpredict convective heat transfer rates and RCCS coolant flow rates. PBMR results show that thermal striping from hot jets in the lower plenum during steady-state operations, and in the upper plenum during a pressurized loss of forced cooling accident, may be a major design concern. Hot jets could potentially melt control rod drive mechanisms or cause thermal stresses in plenum structures. For the HTTF, results will provide data to validate MELCOR for HTGR analyses. Validation will be accomplished by comparing results from the MELCOR representation of the HTTF to experimental results from the facility. The validation process can be automated using a modular code written in Python, which is described here.