The sudden loss of the plasma magnetic confinement, known as disruption, is one of the major issue in a nuclear fusion machine as JET (Joint European Torus), Disruptions pose very serious problems to the safety of the machine. The energy stored in the plasma is released to the machine structure in few milliseconds resulting in forces that at JET reach several Mega Newtons. The problem is even more severe in the nuclear fusion power station where the forces are in the order of one hundred Mega Newtons. The events that occur during a disruption are still not well understood even if some mechanisms that can lead to a disruption have been identified and can be used to predict them. Unfortunately it is always a combination of these events that generates a disruption and therefore it is not possible to use simple algorithms to predict it. This thesis analyses the possibility of using neural network algorithms to predict plasma disruptions in real time. This involves the determination of plasma parameters every few milliseconds. A plasma boundary reconstruction algorithm, XLOC, has been developed in collaboration with Dr. D. Ollrien and Dr. J. Ellis capable of determining the plasma wall/distance every 2 milliseconds. The XLOC output has been used to develop a multilayer perceptron network to determine plasma parameters as ?i and q? with which a machine operational space has been experimentally defined. If the limits of this operational space are breached the disruption probability increases considerably. Another approach for prediction disruptions is to use neural network classification methods to define the JET operational space. Two methods have been studied. The first method uses a multilayer perceptron network with softmax activation function for the output layer. This method can be used for classifying the input patterns in various classes. In this case the plasma input patterns have been divided between disrupting and safe patterns, giving the possibility of assigning a disruption probability to every plasma input pattern. The second method determines the novelty of an input pattern by calculating the probability density distribution of successful plasma patterns that have been run at JET. The density distribution is represented as a mixture distribution, and its parameters arc determined using the Expectation-Maximisation method. If the dataset, used to determine the distribution parameters, covers sufficiently well the machine operational space. Then, the patterns flagged as novel can be regarded as patterns belonging to a disrupting plasma. Together with these methods, a network has been designed to predict the vertical forces, that a disruption can cause, in order to avoid that too dangerous plasma configurations are run. This network can be run before the pulse using the pre-programmed plasma configuration or on line becoming a tool that allows to stop dangerous plasma configuration. All these methods have been implemented in real time on a dual Pentium Pro based machine. The Disruption Prediction and Prevention System has shown that internal plasma parameters can be determined on-line with a good accuracy. Also the disruption detection algorithms showed promising results considering the fact that JET is an experimental machine where always new plasma configurations are tested trying to improve its performances.

M.Phil. (Energy Studies) / SAFARI-1 is a 20 MW research reactor, which is over 45 years old, and is expected to reach the end of its operating life between 2020 and 2030. The aim of this work is to investigate various alternative conceptual core layouts of the SAFARI-1 reactor in order to facilitate more e ective utilization of the reactor, while potentially expanding its operating lifetime. The spatial and energy neutron distribution is one of the most signi cant parameters in the characterization of such an alternative core layout. This neutron distribution is a result of basic physics processes such as particle matter interactions, nuclear reactions, material properties, e ects of temperature and the time evolution of the system. This study focuses on the steady-state neutron distribution within the highly heterogeneous and complex geometry of the reactor core for the various alternative core layouts. This work has searched for and found a di erent inhomogeneous neutron distribution within the core, arising from a di erent core layout, which can nonetheless still achieve e ciency in the operation for various design purposes, but with a lower power output. Via numerical analysis with the OSCAR-4 code system, the safety and utilization requirements for the SAFARI-1 reactor are evaluated and quantied in terms of its steady-state neutron ux distribution. A SAFARI-1 reference core, obtained via an equilibrium cycle calculation, was used to generate a set of safety and utilization targets against which alternative designs may be measured. Alternative core layouts were developed by using a parametric study to scope the size and power level of potential candidate conceptual cores with the aim of minimizing the power level while adhering to the safety requirements. Utilization parameters of interest include isotope production capability, thermal ux levels in beam tubes and production levels in the silicon doping facility...

A thesis submitted to the Faculty of Science, University of the Witwatersrand, Johannesburg in fulfilment of the requirements for the degree of Doctor of Philosophy. Johannesburg, 2015 / Spent nuclear fuel storage is gradually becoming a nightmare for nuclear reactors which were commissioned in the 1980s. This leaves the nuclear facility management with the dilemma of having to choose between pursuing the cask storage option to relieve the demand pressure on the spent fuel pool, or to opt for the more radical but unpopular option of shutting down the reactor compromising the energy supply, and South Africa is no exception. In a bid to minimise the risk of reactor shut down, the Nuclear Analysis Section (NAS) of Eskom launched the present study of investigating the design requirements of spent fuel casks suitable for the storage and transportation of spent fuel assemblies that have an initial enrichment of up to 5 wt% and much higher burnup of between 50 and 60 GWD/MTU. The aim of the present study is to investigate the suitability of the existing casks for use in 5 wt% enriched fuel, given that they are licensed for a maximum enrichment of 3.5 wt%. As a result of the huge number of casks required, there is potentially a risk of shortage of cask storage space and, therefore, it was prudent that the study also investigates the most optimum storage array that will maximise the storage space, while keeping the effective neutron multiplication factor (keff) below the internationally recommended value of 0.95 [IAEA, 2014]. As such, it is also necessary to identify parameters which have the greatest effect on the neutron multiplication factor. These include determining the effect of changes in moderator and fuel temperature on the neutron multiplication factor and also what the effect of an increase in the concentration in 10B of the boral plate will have on the neutron multiplication factor. / M T 2016

Typescript (photocopy). / Digitized by Kansas Correctional Industries

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

Lattice bum-up calculations in thermal reactors are complicated by the necessity for use of transport theory to represent fuel rods, control rods, and burnable absorbers, by many time-dependent variables which must be considered in the analysis, and by geometric complexity which introduces time-dependent, spatial-spectral variations. Representation of lattice structure in a core is further complicated by fuel materials and loading patterns which can be non-symmetric, and by the type of material used as the moderator. The incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI) is an example of such a reactor. In this design, the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional bum-up and fuel depletion analysis codes have been found to be inadequate for these calculations, largely for the reasons mentioned above and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. A more sophisticated codes such as the transport lattice type code WIMS is suitable for the terrestrial commercial reactors. However it lacks some materials, such as ZrH, needed in special applications and it is not capable of performing calculations with highly enriched fuel. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the Thermionic Fuel Elements (TFE) is needed to be developed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations within the TFEs. This required the modification of the MCNP itself to perform the additional task of accomplishing burn-up calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up. The of burnable absorber Gadolinium which adds complications both in the physical model and the numerical analysis requires frequent spatial and spectral reevaluations as a function of burn-up. This difficulty is overcome by the application of MCNPBURN by assuming that the burnable poison is uniformly mixed in the fuel. MCNPBURN was benchmarked using a standard Pressurized Water Reactor fuel element against the LEOPARD and WIMS computer codes. The results from MCNPBURN show good agreement with LEOPARD and WIMS. The maximum difference between MCNPBURN and either of the two codes was approximately 9%. The differences can be accounted for by the appropriate treatment of the accumulated fission product. Application of the MCNPBURN for the ATI reactor core, which consists of 165 TFEs and operates at 375 kW of thermal power, showed a system lifetime greater than the projected lifetime of 7 years at full power. The average efficiency is about 5.86% and the change in the overall efficiency over the life time is 0.2%. The percentage of fuel mass burned is estimated to be about 3.6% of the initial mass. Another calculation includes the influence of burnable poisons mixed in the peak pins to flatten the overall core radial power distribution was performed. The efficacy of this change is quite apparent in reducing the power effectively in the peak pins though it may give rise in power elsewhere in the core. / Graduation date: 1994

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