The moisture separator forms part of the moisture separator reheater (MSR) component used in a steam cycle in a nuclear power plant, to reduce the risk of erosion of the low-pressure (LP) turbine and to improve cycle efficiency. The performance and optimisation of moisture separators is well studied in literature; however, there have been few investigations on the impact of moisture separator degradation on MSR performance. To investigate this impact a mathematical model, representing the steam flow through the MSR, is developed and used to simulate and analyse the impact of degradation conditions. The mathematical model was developed for design conditions, calibrated and validated against manufacturer specifications. The model was then augmented to include two moisture separator degradation conditions. The first degradation condition is the partial blockage of separator vane channels due to fouling, and the second is separator material deterioration resulting in steam bypass of the moisture separator. The model uses known properties of the MSR inlet steam and predicts the properties of steam exiting the MSR, given the simulated degradation of the moisture separator. The outcomes of the model simulations demonstrated that partial blockage of moisture separator vane channels increases steam velocity though the separator and consequently improves MSR performance, but with a noted pressure drop. The velocity increased until a theoretical upper limit, above which re-entrainment of droplets back into the steam flow reduces MSR performance. It was concluded that there is margin in the separator surface area design, where a minimal reduction in separator surface area (represented in the model as blockage of the vane channels) would improve the performance of the MSR, while still allowing for a buffer against the re-entrainment velocity upper limit. Equally, an unexpected improvement in MSR performance may be an indication of blockage of separator vane channels that, if not monitored and managed, could surpass the critical velocity limit where re-entrainment adversely affects the MSR performance. The simulation results demonstrated that steam bypass of the moisture separator is a credible degradation condition which affects MSR performance. It was found that steam bypass of the moisture separator leads to a decline in the quality of steam exiting the separator and a decline in MSR performance. The simulation of a fully bypassed moisture separator showed that the MSR performance declines by more than three times the design value when compared to the scenario where there is no bypass of the moisture separator.

South Africa is poised to expand significantly its nuclear power generation industry. Considering that the current South African nuclear safety regulatory approach is applied to regulate the operation and maintenance of one mature nuclear power plant, it is expected that significant adaptation of this approach will occur for the regulatory system to accommodate the planned industry expansion. This dissertation tests the hypothesis that the optimal nuclear safety regulatory approach for South Africa’s planned nuclear industry can already be determined by systematically comparing the suitability of various alternatives in use in the international nuclear industry. Investigating the validity of this hypothesis improves the understanding of the possibilities available for future nuclear safety regulation in South Africa and aids preparations and decision-making in this regard. Research was conducted on the various nuclear safety regulatory approaches applied internationally and on what determines the suitability of each approach in different circumstances. The characteristics of South Africa’s current and planned nuclear power generation industry were investigated. Applying multi-criteria decision making analysis methodology, a test was developed and used to systematically assess the relative suitability of the various regulatory approaches to the South African context. The three primary approaches to nuclear safety regulation considered were the prescriptive approach, the performance based approach and the goal-setting approach. Based on currently available information, the test results show that the goal-setting regulatory approach is the optimal approach for South Africa’s planned nuclear power industry. However research findings also show that the state level bilateral cooperation the South African government would pursue to develop South Africa’s fleet approach to the 9,6 gigawatt nuclear new build programme may have sufficient influence on South Africa’s nuclear industry to change South Africa’s optimal nuclear safety regulatory approach or make this plant specific. The benefits of aligning South Africa’s nuclear safety regulatory approach with the approach applied in the fleet vendor company’s country of origin may outweigh other considerations. The vendor company for South Africa’s nuclear new build programme is not yet known. Even though systematic comparison of the suitability of various regulatory approaches shows that the goal-setting nuclear safety regulatory approach is the optimal approach for South Africa, the hypothesis is shown to be false. The optimal nuclear safety regulatory approach for South Africa’s planned nuclear industry cannot already be determined, since bilateral cooperation with the nuclear new build fleet vendor company’s country of origin may be the dominant factor in shaping South Africa’s nuclear safety regulatory approach. In the interim and in the event that strategic regulatory alignment for the new build fleet is not embarked upon, the research findings and test results have an important implication: Applying the goal-setting approach as the dominant nuclear safety regulatory approach can optimize nuclear safety regulation of South Africa’s nuclear industry.

The Fukushima Daiichi Nuclear Power Plant accident created renewed international interest in the behaviour of spent fuel subsequent to a complete loss of water inventory in a spent fuel pool (SFP). The study conducted in this dissertation serves as a starting point in gaining an understanding of the thermal hydraulics and associated heat transfer processes involved when spent fuel assemblies (SFAs) become uncovered in air. The complete loss of cooling in a SFP is a complex 3-D problem, hence several simplifications were necessary in this research to narrow the scope. Further, due to the complexity of this topic, the results obtained serves purely as a first order approximation. Accordingly, the Flownex systems CFD code (version 8.6.1.2630) was used to simulate the thermal response of the uncovered SFAs in the SFP of a typical Pressurised Water Reactor (PWR) during a severe accident scenario. Two network models were developed. The first was to identify the dominant heat transfer mechanisms with-in the spent fuel pool and it therefore accounted for a range of physics. This included convective heat transfer through the composite SFA channel walls, conduction along the vertical axial direction of the SFAs and through the inner and outer rack wall as well as through the fuel building (FB) roof and side walls. The model also took into account the radiative heat transfer from the cladding surface of the composite SFAs to the FB roof and side walls. This network model informed that the heat transfer with-in the SFA during the considered extreme accident scenario is dominated by radiative heat transfer. This informed the development of an improved 2-D network model using conduction elements which were specially calibrated in this work to account for radiative heat loss. An effective conduction for the fuel volume which is dependent on temperature was determined and was used to assess the severe accident. Transient results showed that the spent fuel may reach cladding oxidation temperature within circa 10.5 hrs after a complete loss of inventory.

At Koeberg Nuclear Power Station, the reactor thermal power limit is one of the most important quantities specified in the operating licence, which is issued to Eskom by the National Nuclear Regulator (NNR). The reactor thermal power is measured using different methodologies, with the most important being the Secondary Heat Balance (SHB) test which has been programmed within the central Koeberg computer and data processing system (KIT). Improved accuracy in the SHB will result in a more accurate representation of the thermal power generated in the core. The input variables have a significant role to play in determining the accuracy of the measured power. The main aim of this thesis is to evaluate the sensitivity of the SHB to the changes in all input variables that are important in the determination of the reactor power. The guidance provided by the Electric Power Research institute (EPRI) is used to determine the sensitivity. To aid with the analysis, the SHB test was duplicated using alternate software. Microsoft Excel VBA and Python were used. This allowed the inputs to be altered so that the sensitivity can be determined. The new inputs included the uncertainties and errors of the instrumentation and measurement systems. The results of these alternate programmes were compared with the official SHB programme. At any power station, thermal efficiency is essential to ensure that the power station can deliver the maximum output power while operating as efficiently as possible. Electricity utilities assign performance criteria to all their stations. At Koeberg, the thermal performance programme is developed to optimize the plant steam cycle performance and focusses on the turbine system. This thesis evaluates the thermal performance programme and turbine performance. The Primary Heat Balance (PHB) test also measures reactor power but uses instrumentation within the reactor core. Due to its location inside the reactor coolant system, the instrumentation used to calculate the PHB is subject to large temperature fluctuations and therefore has an impact on its reliability. To quantify the effects of these fluctuations, the sensitivity of the PHB was determined. The same principle, which was used for the SHB sensitivity analysis, was applied to the PHB. The impact of each instrument on the PHB test result was analysed using MS Excel. The use of the software could be useful in troubleshooting defects in the instrumentation. A sample of previously authorised tests and associated data were used in this thesis. The data for these tests are available from the Koeberg central computer and data processing system.

The Transition Radiation Detector is a sub-detector of the ALICE experiment that is used primarily as an electron detector and a trigger mechanism. The TRD currently has 521 individual chambers distributed over 18 super modules. Each chamber houses a radiator, a drift region and a multi-wire proportional chamber with the readout electronics. When charge is absorbed in the anode wires of the multi-wire proportional chamber, it creates a common-mode effect, this common-mode effect manifests itself as a drop in the signal produced by the surrounding readout electronics where no particle has traversed, called the baseline. Capacitors have been installed in a layout to produce a low-pass filter (RC circuit) to decrease the impact of the common-mode effect. These capacitors were installed for a pad row pair, creating capacitor coupling for a high voltage supply segment. However, these capacitors were prone to failure, causing dead chambers that could not be used to acquire data for the remainder of the run, so it was decided to remove them. With their removal, the extent of the common-mode effect on the baseline had to be understood and corrected for in order to better calibrate the detector system. The University of Cape Town has one chamber, an L3C0 chamber. This chamber was used to collect two datasets with the same parameters, one with the 2.2 nF capacitors installed and the other without the 2.2 nF capacitors, to study the effect. It is found that the drop in baseline is only experienced by anode wires with the same capacitor coupling. Furthermore, it is observed that there is a linear relationship between the charge absorbed by the anode wires and the drop in baseline, thus charge absorbed by the anode wires can be summed should they have the same capacitor coupling. It is also found that the drop in baseline is 2.5 times larger in the dataset without capacitors. The final part of the thesis corrects for this common-mode effect, using a correction factor determined from the dataset

Thesis (Ph. D.)--Ohio State University, 2006. / Title from first page of PDF file. Includes bibliographical references (p. 205-215).

Thesis (M.S.)--University of Michigan, 1982.

Thesis, PlanB (M.S.)--University of Wisconsin--Stout, 2005. / Includes bibliographical references.

Thesis (M. S.)--Nuclear Engineering, Georgia Institute of Technology, 2010. / Committee Chair: Abdel-Khalik, Said I.; Committee Member: Ghiaasiaan, S. Mostafa; Committee Member: Hertel, Nolan E. Part of the SMARTech Electronic Thesis and Dissertation Collection.

Thesis (doctoral)--University of Helsinki, 2007. / Includes bibliographical references. Also available on the World Wide Web.