<p>The ability to accurately predict the void fraction in a nuclear reactor plays a vital role in the field of nuclear safety. Specifically, high volumetric fractions of vapour in a reactor core can lead to a severe degradation in the ability of the coolant to remove heat from the fuel. Unfortunately, the behaviour of two-phase flow is complex and not particularly well understood, making it infeasible to describe the fluid behaviour in the core analytically. Instead, computer codes which rely heavily on temporal and spatial averaging, as well as empirical correlations, are used to provide a 'best-estimate' of the core thermalhydraulic behaviour. <br /><br />In this study, both the system code RELAP5-3D and the sub channel code ASSERT are used to simulate sub channel void fraction measurements conducted on full scale electrically heated BWR type bundles. RELAP was demonstrated to be capable of predicting the void fraction at the sub channel level to within ±O.l 0 of the reported value 82.6% ofthe time, while ASSERT was successful 95.6% of the time. When the accuracy levels are broken down by sub channel geometry, the comer and side sub channels tended to be the most inaccurately predicted, and this was determined to have been caused by the strong influence of sub channel nnxmg. <br /><br />Uncertainty bands around the simulated points were predicted using the CIAU method developed at the University of Pisa. In the cases simulated with RELAP, the simulated uncertainty bounds enveloped a portion of the experimental uncertainty bounds 95.4% of the time while the ASSERT assessment saw coverage 85.4% of the time.</p> / Master of Applied Science (MASc)
| Identifer | oai:union.ndltd.org:mcmaster.ca/oai:macsphere.mcmaster.ca:11375/9122 |
| Date | 06 1900 |
| Creators | Leung, Kenneth |
| Contributors | Novog, D. R., Engineering Physics |
| Source Sets | McMaster University |
| Detected Language | English |
| Type | thesis |
Page generated in 0.002 seconds