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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

One of the future reactor technologies defined by the Generation-IV International Forum (GIF) is the Lead-Cooled Fast Reactor (LFR). An advantage with this reactor technology is that steam production is accomplished by means of heat exchangers located within the primary reactor vessel, which decreases costs and increases operational safety. However, a crack in a heat exchanger tube may create steam (void) into the coolant and this process has the potential to introduce reactivity changes, which may cause criticality issues. This fact motivates the development of a methodology to detect such voids. This thesis comprises theoretical investigations on a possible route to detect voids by studying changes of the neutron spectrum in a small LFR as a function of various types of in-core voids .The methodology includes a combination of fission chambers loaded with U-235 and Pu-242 operating in various positions. It is shown that such a combination results in information that can be made independent on reactor power, a feasible property in order to detect the relatively small spectral changes due to void. A sensitivity analysis of various combinations of detectors, fuel burnup and void has also been included in the investigation. The results show that the proposed methodology yields a reasonably large sensitivity to voids down to (1-2) % of the coolant volume. The results obtained so far point in the direction that the proposed methodology is an interesting subject for further studies.

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