Browsing by Subject "Prismatic"
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Item Physics-Based 3D Multi-Directional Reloading Algorithm for Deep Burn HTR Prismatic Block Systems(2011-10-21) Lewis, Tom Goslee, IIITo assure nuclear power sustainability, ongoing efforts on advanced closed-fuel cycle options and adapted open cycles have led to investigations of various strategies involving utilization of Transuranic (TRU) nuclides in nuclear reactors. Due to favorable performance characteristics, multiple studies are focused on transmutation options using High Temperature Gas-cooled Reactors (HTGRs). Prismatic HTGRs allow for 3-Dimensional (3D) fuel shuffling and prior shuffling algorithms were based on experimental block movement and/or manual block shuffle patterns. In this dissertation, a physics based 3D multi-directional reloading algorithm for prismatic deep burn very high temperature reactors (DB-VHTRs) was developed and tested to meet DB-VHTR operation constraints utilizing a high fidelity neutronics model developed for this dissertation. The high fidelity automated neutronics model allows design flexibility and metric tracking in spatial and temporal dimensions. Reduction of TRUs in DB-VHTRs utilizing full vectors of TRUs from light water reactor spent nuclear fuel has been demonstrated for both a single and two-fuel composition cores. Performance of the beginning-of-life and end-of-life (EOL) domains for multi-dimensional permutations were evaluated. Utilizing a two-fuel assembly permutation within the two-fuel system domain for a Single-Fuel vector, the developed shuffling algorithm for this dissertation has successfully been tested to meet performance objectives and operation constraints.Item Prismatic modular reactor analysis with melcor(2009-05-15) Zhen, NiHydrogen, a more sustainable source of energy, is a potential substitute for hydrocarbon fuel for power generation. The Very High Temperature gas-cooled Reactor (VHTR) concept can produce hydrogen with high efficiency and in large quantities. The US Department of Energy plans to build a VHTR as a next-generation hydrogen/electricity production plant. This reactor concept is very different from that of commercial reactors in the US. In order to acquire licensing eligibility for VHTRs, analysis tools need to be validated and applied to design and evaluate VHTRs under operation conditions and accident scenarios. In this thesis, MELCOR, a severe accident code, was used to analyze one of the VHTR designs ? a prismatic core Next Generation Nuclear Plant (NGNP). The NGNP is based on General Atomics? (GA) Gas Turbine ? Modular Helium Reactor (GT-MHR) 600 MW design. According to the current literature survey, more data is available for the GT-MHR than for the NGNP. Therefore, for the purposes of extending MELCOR capabilities and code validation, a model of the GT-MHR reactor pressure vessel (RPV) was developed. Based on the currently available data, a model of the NGNP RPV was then developed through modifying the GT-MHR RPV model. For both RPV models, coolant outlet temperature under normal operating conditions corresponds well to the data from literature. The reactor cavity cooling systems (RCCS), which passively removes heat from the RPV wall to the outside atmosphere, was then added to this GT-MHR RPV model. With this model addition, the heat removal rate of the RCCS under normal operating conditions was calculated to correspond well to the data from references. Pressurized conduction cooldown (PCC), one of the important postulated accident scenarios for a prismatic core reactor, was simulated with the complete model. MELCOR has been demonstrated to have the ability of modeling a prismatic core VHTR. The calculated outlet temperature and mass flow rate under normal operation correspond well to references. However, the calculation for the heat distribution in the graphite and fuel is unsatisfactory which requires MELCOR modification for the PCC simulation. For future work, a complete model of the NGNP under normal operation conditions will be developed when additional data becomes available.