Global ETD Search

1	Multi application small light water reactor containment analysis and design Haugh, Brandon Patrick 28 May 2002 (has links) This thesis presents the assessment of the Multi Application Small Light Water Reactor (MASLWR) containment design during steady-state and transient conditions. The MASLWR project is a joint effort between Idaho National Environmental and Engineering Laboratory (INEEL), NEXANT Bechtel, and Oregon State University. The project is funded under a Nuclear Energy Research Initiative (NERI) grant from the Department of Energy (DOE). The GOTHIC code was used to simulate the full scale prototype and the Oregon State University MASLWR test facility. Detailed models of the full scale prototype and OSU test facility were generated in GOTHIC. GOTHIC condensation heat transfer models produced heat transfer coefficients that vary by an order of magnitude. This had a significant impact on the pressurization rate and peak pressure achieved within containment. A comparison of the GOTHIC calculation results for the full scale prototype and the test facility model shows reasonable agreement with respect to containment pressure trends and safety system mass flow rates. / Graduation date: 2003 Light water reactors Nuclear reactors -- Containment
2	Modeling of once-through steam generator thermal-hydraulics during a loss of coolant accident Kamboj, Brij Kumar 05 1900 (has links) No description available. Two-phase flow Nuclear reactors Cooling Nuclear reactors Containment
3	RELAP5-3D modeling of ADS blowdown of MASLWR facility Bowser, Christopher Jordan 13 June 2012 (has links) Oregon State University has hosted an International Atomic Energy Agency (IAEA) International Collaborative Standard Problem (ICSP) through testing conducted on the Multi-Application Small Light Water (MASLWR) facility. The MASLWR facility features a full-time natural circulation loop in the primary vessel and a unique pressure suppression device for accident scenarios. Automatic depressurization system (ADS) lines connect the primary vessel to a high pressure containment (HPC) which dissipates steam heat through a heat transfer plate thermally connected to another vessel with a large cool water inventory. This feature drew the interest of the IAEA and an ICSP was developed where a loss of feedwater to the steam generators prompted a depressurization of the primary vessel via a blowdown through the ADS lines. The purpose of the ICSP is to evaluate the applicability of thermal-hydraulic computer codes to unique experiments usually outside of the validation matrix of the code itself. RELAP5-3D 2:4:2 was chosen to model the ICSP. RELAP5-3D is a best-estimate code designed to simulate transient fluid and thermal behavior in light water reactors. Modeling was conducted in RELAP5-3D to identify the strengths and weaknesses of the code in predicting the experimental trends of the IAEA ICSP. This extended to nodalization sensitivity studies, an investigation of built-in models and heat transfer boundary conditions. Besides a qualitative analysis, a quantitative analysis method was also performed. / Graduation date: 2013 Nuclear Thermal Light water reactors -- Safety measures Pressurized water reactors -- Blowdown Nuclear reactors -- Containment
4	Particle orbits and diffusion in torsatrons Potok, Robert Edward January 1980 (has links) Thesis (Sc.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1980. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by Robert Edward Potok. / Sc.D. Nuclear Engineering. Fusion reactors Nuclear reactors Containment Plasma confinement Particles (Nuclear physics)
5	High pressure condensation heat transfer in the evacuated containment of a small modular reactor Casey, Jason R. 19 December 2012 (has links) At Oregon State University the Multi-Application Small Light Water Reactor (MASLWR) integral effects testing facility is being prepared for safety analysis matrix testing in support of the NuScale Power Inc. (NSP) design certification progress. The facility will be used to simulate design basis accident performance of the reactor's safety systems. The design includes an initially evacuated, high pressure capable containment system simulated by a 5 meter tall pressure vessel. The convection-condensation process that occurs during use of the Emergency Core Cooling System has been characterized during two experimental continuous blowdown events. Experimental data has been used to calculate an average heat transfer coefficient for the containment system. The capability of the containment system has been analytically proven to be a conservative estimate of the full scale reactor system. / Graduation date: 2013 Condensation Nuclear Small Modular Reactor Containment Light water reactors -- Safety measures Nuclear reactors -- Containment Condensation Heat -- Transmission
6	Application of the Stimulus-Driven Theory of Probabilistic Dynamics to the hydrogen issue in level-2 PSA / Application de la Stimulus Driven Theory of Probabilistic Dynamics (SDTPD) au risque hydrogène dans les EPS de niveau 2. Peeters, Agnes 05 October 2007 (has links) Les Etudes Probabilistes de Sûreté (EPS) de niveau 2 en centrale nucléaire visent à identifier les séquences d’événements pouvant correspondre à la propagation d’un accident d’un endommagement du cœur jusqu’à une perte potentielle de l’intégrité de l’enceinte, et à estimer la fréquence d’apparition des différents scénarios possibles.<p>Ces accidents sévères dépendent non seulement de défaillances matérielles ou d’erreurs humaines, mais également de l’occurrence de phénomènes physiques, tels que des explosions vapeur ou hydrogène. La prise en compte de tels phénomènes dans le cadre booléen des arbres d’événements s’avère difficile, et les méthodologies dynamiques de réalisation des EPS sont censées fournir une manière plus cohérente d’intégrer l’évolution du processus physique dans les changements de configuration discrète de la centrale au long d’un transitoire accidentel.<p>Cette thèse décrit l’application d’une des plus récentes approches dynamiques des EPS – la Théorie de la Dynamique Probabiliste basée sur les Stimuli (SDTPD) – à différents modèles de déflagration d'hydrogène ainsi que les développements qui ont permis cette applications et les diverses améliorations et techniques qui ont été mises en oeuvre.<p><p>Level-2 Probabilistic Safety Analyses (PSA) of nuclear power plants aims to identify the possible sequences of events corresponding to an accident propagation from a core damage to a potential loss of integrity of the containment, and to assess the frequency of occurrence of the different scenarios.<p>These so-called severe accidents depend not only on hardware failures and human errors, but also on the occurrence of physical phenomena such as e.g. steam or hydrogen explosions. Handling these phenomena in the classical Boolean framework of event trees is not convenient, and dynamic methodologies to perform PSA studies are expected to provide a more consistent way of integrating the physical process evolution with the discrete changes of plant configuration along an accidental transient.<p>This PhD Thesis presents the application of one of the most recently proposed dynamic PSA methodologies, i.e. the Stimulus-Driven Theory of Probabilistic Dynamics (SDTPD), to several models of hydrogen explosion in the containment of a plant, as well as the developed methods and improvements.<p> / Doctorat en Sciences de l'ingénieur / info:eu-repo/semantics/nonPublished Sciences de l'ingénieur Physique Nuclear energy Nuclear power plants -- Accidents Nuclear power plants -- Safety measures Nuclear reactors -- Safety measures Nuclear reactors -- Containment Energie nucléaire Centrales nucléaires -- Accidents Réacteurs nucléaires -- Confinement PSA STIMOCCO reliability fiabilité stimulus dynamic benchmark SARNET sûreté étude safety analysis probabiliste explosion deflagration niveau 2 severe accident accident grave cell-to-cell mapping combustion EPS cellules Petri dynamique probabilistic level-2 nuclear energy hydrogen énergie nucléaire centrale nucléaire nuclear power plant hydrogène délai delay

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