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1	Numerical modeling of friction stir welding : a comparison of Alegra and Forge3 / Oliphant, Alma H., January 2004 (has links) (PDF) Thesis (M.S.)--Brigham Young University. Dept. of Mechanical Engineering, 2004. / Includes bibliographical references (p. 83-85).
2	Detailed study of the transient rod pneumatic system on the annular core research reactor Fehr, Brandon M. 27 May 2016 (has links) Throughout the history of the Annular Core Research Reactor (ACRR), Transient Rod (TR) A has experienced an increased rate of failure versus the other two TRs (B and C). Either by pneumatic force or electric motor, the transient rods remove the poison rods from the ACRR core allowing for the irradiation of experiments. In order to develop causes for why TR A is failing (rod break) more often, a better understanding of the whole TR system and its components is needed. This study aims to provide a foundational understanding of how the TR pneumatic system affects the motion of the TRs and the resulting effects that the TR motion has on the neutronics of the ACRR. Transient rod motion profiles have been generated using both experimentally-obtained pressure data and by thermodynamic theory, and input into Razorback, a SNL-developed point kinetics and thermal hydraulics code, to determine the effects that TR timing and pneumatic pressure have on reactivity addition and reactivity feedback. From this study, accurate and precise TR motion profiles have been developed, along with an increased understanding of the pulse timing sequence. With this information, a safety limit within the ACRR was verified for different TR travel lengths and pneumatic system pressures. In addition, longer reactivity addition times have been correlated to cause larger amounts of reactivity feedback. The added clarity on TR motion and timing from this study will pave the way for further study to determine the cause for the increased failure rate of TR A. ACRR Sandia National Laboratories SNL Transient rod TRIGA
3	Seamless Level 2 / Level 3 Probabilistic Risk Assessment Using Dynamic Event Tree Analysis Osborn, Douglas M. 29 August 2013 (has links) No description available. Nuclear Engineering Probabilistic Risk Assessment Dynamic PRA MELCOR MACCS MACCS2 Dynamic Event Tree DET Uncertainty Nuclear SOARCA Severe Reactor Accident Station Blackout SBO Sandia National Laboratories Nuclear Regulatory Commission NRC SNL
4	Development and assessment of CFD models including a supplemental program code for analyzing buoyancy-driven flows through BWR fuel assemblies in SFP complete LOCA scenarios Artnak, Edward Joseph 31 January 2013 (has links) This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-of-coolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based control-volume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy. / text CFD CFD Model Buoyancy driven flow Buoyancy flow FLUENT FLUENT CFD SNL Sandia National Laboratories SFP LOCA Spent fuel pool loss of coolant accident SFP loss of coolant BWR fuel assembly CAD Model SolidWorks SFP Model SFP BWR fuel assembly NRC MATLAB MATLAB SFP Model 9x9 BWR fuel assembly MELCOR MELCOR SFP 250 million cells Hydraulic flow loss Loss coefficients Buoyancy flows MELCOR calibration Pressure loss BWR pressure loss SFP airflow SNL fuel assembly

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