Made available in DSpace on 2014-10-09T12:31:34Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:00:27Z (GMT). No. of bitstreams: 1 02235.pdf: 3818796 bytes, checksum: 65e547d59f55d89221cf5fb934675699 (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:31:34Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:00:27Z (GMT). No. of bitstreams: 1 02235.pdf: 3818796 bytes, checksum: 65e547d59f55d89221cf5fb934675699 (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Existing containment models for a loss of coolant accident at many nuclear power plants were created in the 1970s using older computer technology and thermal hydraulic models which were available at that time. While conservative, these models may not present the detail necessary to identify conditions which may be used to produce additional design margin for the plant. After exploring containment and critical flow modeling, the basis for the use of GOTHIC in this analysis was established. A GOTHIC model was then created to simulate the loss of coolant accident results shown in an Updated Final Safety Analysis Report analysis for the North Anna Power Station. This model was used to examine the effects of increased nodalization in a subcompartment on the existing containment model. It is shown that adding multidimensional sub-nodes to areas of interest can provide valuable detail which was absent in the UFSAR model. Simulations are able to show the localized pressure spike around a LOCA pipe break that quickly dissipates, leaving significantly lower pressures in what was once an averaged, single, lumped-parameter node. This suggests that additional design margin may exist depending on where the pipe break is assumed to occur. / Master of Science

Under simulated accident conditions, tees in the primary coolant loop of a Pressurized Water Reactor (PWR) can deviate from their original design purpose and become separators that effectively remove core heat sink capacity. This method of primary coolant removal is a phenomelogical subset of phase separation known as liquid entrainment, whereby liquid is forced from its original path by the inertia of the gas. A comprehensive literature review revealed common deficiencies in previous studies. The Westinghouse AP600 advanced reactor design was chosen to assess the validity of entrainment models. Following a systematic scaling analysis of the prototypic design a model separate effects test was proposed and constructed at Oregon State University. Just under 100 tests were run to fill the deficiencies found in the literature review. New data from the Air-water Test Loop for Advanced Thermal-hydraulic Studies (ATLATS) could not be predicted by published correlations. A new theoretical model for predicting liquid entrainment onset and steady state entrainment was developed. Comparison with all available data shows a marked improvement for predicting the mass flow rate out the vertical branch. / Graduation date: 2003

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Made available in DSpace on 2014-10-09T12:36:18Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:59:29Z (GMT). No. of bitstreams: 1 03967.pdf: 2109353 bytes, checksum: 57437fbcbe44e88577f3719ed9ed26be (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:31:20Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:00:17Z (GMT). No. of bitstreams: 1 01383.pdf: 5253778 bytes, checksum: a56e7795fb89767fd64b2c7c22a1e16a (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:31:36Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:00:33Z (GMT). No. of bitstreams: 1 02281.pdf: 4701050 bytes, checksum: 60f7f41ef9b4e9378ba1df67374b6843 (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP