Thesis (MScEng)--Stellenbosch University, 2012. / ENGLISH ABSTRACT: In light of the rapid depletion of the world’s oil reserves, concerns about energy security prompted the exploration of alternative sources of liquid fuels for transportation. One such alternative is the production of synthetic fuels with the indirect coal liquefaction process or Coal-To-Liquids (CTL) process. In this process, coal is burned in a gasifier in the presence of steam and oxygen to produce a synthesis gas or syngas, consisting mainly of hydrogen and carbon monoxide. The syngas is then converted to liquid fuels and a variety of useful chemicals in a Fischer Tropsch synthesis reactor. However, the traditional process for syngas production also produces substantial amounts of carbon dioxide. In fact, only about one third of the carbon in the coal feedstock ends up in the liquid fuel product using traditional CTL technology. If additional hydrogen was available, the carbon utilisation of the process could be improved significantly. The high temperature reactor (HTR) is a gas cooled Generation IV nuclear reactor ideally suited to provide electrical power and high temperature heat for the production of carbon neutral hydrogen via high temperature electrolysis. The integration of an HTR into a CTL process therefore provides an opportunity to improve the thermal and carbon efficiency of the CTL process significantly. This thesis presents a possible process flow scheme for a nuclear assisted CTL process. The system is evaluated in terms of its thermal or syngas production efficiency (defined as the ratio of the heating value of the produced syngas to the sum of the heating value of the coal plus the HTR heat input) as well as its carbon utilisation. If the hydrogen production plant is sized to produce only enough associated oxygen to supply in the needs of the gasification plant, syngas is produced at about 63% thermal efficiency, while 71.5% of the carbon is utilised in this process. It was found that the optimum HTR outlet temperature to produce hydrogen with a high temperature steam electrolysis process is 850°C. If enough process heat and electrical power are available and process equipment capacities are sufficient, the carbon utilisation of the process could be improved even further to values in excess of 90%. / AFRIKAANSE OPSOMMING: Die uitputting van die wêreld se olie-reserwes, asook kommer oor energiesekuriteit het daartoe gelei dat alternatiewe bronne van vloeibare brandstowwe vir vervoer ondersoek moes word. Een so 'n alternatief is die produksie van sintetiese brandstof d.m.v. die indirekte steenkool vervloeiing proses of sogenaamde Coal-To-Liquids (CTL) proses. In hierdie proses word steenkool in die teenwoordigheid van stoom en suurstof in 'n vergasser gebrand om 'n sintesegas of singas te produseer, wat hoofsaaklik uit waterstof en koolstofmonoksied bestaan. Die sintesegas word daarna omgeskakel na vloeibare brandstowwe en 'n verskeidenheid van nuttige chemikalieë in 'n Fischer-Tropsch-sintese reaktor. Ongelukkig produseer die tradisionele proses vir sintesegas produksie ook 'n beduidende hoeveelheid koolstofdioksied. Trouens, slegs sowat een derde van die koolstof in die steenkool roumateriaal eindig in die vloeibare brandstof produk indien van tradisionele CTL-tegnologie gebruik gemaak word. Indien addisionele waterstof beskikbaar was, kon die koolstofbenutting van die proses aansienlik verbeter word. Die hoë temperatuur reaktor (HTR) is 'n gas-verkoelde Generasie IV kernreaktor wat by uitstek geskik is om elektrisiteit en hoë temperatuur hitte te verskaf vir die produksie van koolstofneutrale waterstof d.m.v. hoë temperatuur elektrolise. Die integrasie van 'n HTR in 'n CTL-proses bied dus 'n geleentheid om die termiese- en koolstofdoeltreffendheid van die CTL-proses aansienlik te verbeter. In hierdie ondersoek word 'n moontlike proses vloeidiagram vir 'n kern-gesteunde CTL-proses voorgestel. Die stelsel is geëvalueer in terme van sy termiese- of sintesegas produksie doeltreffendheid (gedefinieer as die verhouding van die hittewaarde van die geproduseerde sintesegas gedeel deur die som van die hittewaarde van die steenkool en die HTR hitte-insette) sowel as sy koolstof-effektiwiteit. Indien die waterstof produksie-aanleg ontwerp word om net genoeg geassosieerde suurstof te voorsien om in die behoeftes van die vergassing-aanleg te voorsien, word sintesegas teen ongeveer 63% termiese doeltreffendheid vervaardig, terwyl 71.5% van die koolstof in hierdie proses benut word. Daar is bevind dat 850°C die optimum HTR uitlaat temperatuur is om waterstof d.m.v. hoë temperatuur stoom-elektrolise te vervaardig. Indien daar genoeg proses hitte en elektrisiteit beskikbaar is en die proses toerusting kapasiteite voldoende is, sou die koolstof-benutting van die proses tot meer as 90% verbeter kon word.

Basierend auf dem Reaktordynamikcode DYN3D für LWR, wurde die Codeversion DYN3D-HTR für das Blockkonzept eines graphit-moderierten, helium-gekühlten Hochtemperaturreaktors entwickelt. Diese Entwicklung umfasst die: • methodische Weiterentwicklung der 3D stationären Neutronenflussberechnung für hexagonale Geometrie (HTR-Brennelement-Blöcke), • Generierung von Wirkungsquerschnittsdaten unter Berücksichtigung der doppelten Heterogenität, • Modellierung der Wärmeleitung und des Wärmetransports in der Graphitmatrix. Die nodale SP3-Neutronentransport-Methode in DYN3D wurde auf hexagonale Brennelementgeometrie erweitert. Es wird eine Unterteilung der Hexagone in Dreiecke vorgenommen, so dass die Verfeinerung hexagonaler Strukturen untersucht werden kann. Die Verifikation erfolgte durch Vergleiche mit Monte-Carlo-Referenzlösungen. Für die Behandlung der doppelten Heterogenität der Brennelementstruktur bei Homogenisierung der Wirkungsquerschnitte wurden neue Methoden entwickelt. Zum einen wurde ein zweistufiges Homogenisierungsverfahren basierend auf der Methode der sog. Reactivity Equivalent Transformation (RPT) weiterentwickelt. Zum anderen ermöglichte die Verfügbarkeit des neuen Monte-Carlo-Codes SERPENT die Anwendung eines einstufigen Verfahrens, wobei die 3D heterogenen Strukturen in einem Rechenschritt konsistent erfasst werden können. Weiterhin wur-de in DYN3D ein 3D Wärmeleitungsmodell implementiert, das den radialen und axialen Wärmetransport in der Graphitmatrix beschreiben kann. DYN3D-HTR wurde schließlich anhand der Testfälle für Reaktivitätstransienten erprobt. Die Verifikation erfolgte durch Vergleich zwischen 3D und 1D Berechnung der Wärmeleitung. Schließlich wurde DYN3D mit dem CFD-Code ANSYS-CFX gekoppelt, um auch dreidimensionale Strömungen in Reaktorkernen berechnen zu können. Der Kern wird als poröser Körper modelliert. Die Kopplung wurde an anhand von 2 Testbeispielen, dem Auswurf eines Steuerstabes und einer lokalen Strömungsblockade in einem Brennelement, erprobt.

The closure of the nuclear fuel cycle is a topic of interest in the sustainability context of nuclear energy. The implication of such closure includes considerations of nuclear waste management. This originates from the fact that a closed fuel cycle requires recycling of useful materials from spent nuclear fuel and discarding of non-usable streams of the spent fuel, which are predominantly the fission products. The fission products represent the near-term concerns associated with final geological repositories for the waste stream. Long-lived fission products also contribute to the long-term concerns associated with such repository. In addition, an ultimately closed nuclear fuel cycle in which all actinides from spent nuclear fuels are incinerated will result in fission products being the only source of radiotoxicity. Hence, it is desired to develop a transmutation strategy that will achieve reduction in the inventory and radiological parameters of significant fission products within a reasonably short time. In this dissertation, a transmutation strategy involving the use of the VHTR is developed. A set of specialized metrics is developed and applied to evaluate performance characteristics. The transmutation strategy considers six major fission products: 90Sr, 93Zr, 99Tc, 129I, 135Cs and 137Cs. In this approach, the unique core features of VHTRs operating in equilibrium fuel cycle mode of 405 effective full power days are used for transmutation of the selected fission products. A 30 year irradiation period with 10 post-irradiation cooling is assumed. The strategy assumes no separation of each nuclide from its corresponding material stream in the VHTR fuel cycle. The optimum locations in the VHTR core cavity leading to maximized transmutation of each selected nuclides are determined. The fission product transmutation scenarios are simulated with MCNP and ORIGEN-S. The results indicate that the developed fission product transmutation strategy offers an excellent potential approach for the reduction of inventories and radiological parameters, particularly for long-lived fission products (93Zr, 99Tc, 129I and 135Cs). It has been determined that the in-core transmutation of relatively short-lived fission products (90Sr and 137Cs) has minimal advantage over a decay-only scenario for these nuclides. It is concluded that the developed strategy is a viable option for the reduction of radiotoxicity contributions of the selected fission products prior to their final disposal in a geological repository. Even in the cases where the transmutation advantage is minimal, it is deemed that the improvement gained, coupled with the virtual storage provided for the fission products during the irradiation period, makes the developed fission product transmutation strategy advantageous in the spent fuel management scenarios. Combined with the in-core incineration options for TRU, the developed transmutation strategy leads to potential achievability of engineering time scales in the comprehensive nuclear waste management.

Im Rahmen dieser Dissertation werden eine Serie von Grundlagenexperimenten zur Ablagerung (Deposition) und Remobilisierung (Resuspension) von Aerosolpartikeln in turbulenten Strömungen beschrieben. Die Kernmotivation stellt die Quelltermanalyse von Druckentlastungsstörfällen von Hochtemperaturreaktoren (HTR) dar. Im Primärkreislauf früherer HTR-Forschungsanlagen wurden größere Mengen an radiologisch belastetem Graphitstaub gefunden. Dieser Staub scheint größtenteils durch Abrieb zwischen den graphitischen Kernstrukturen entstanden zu sein und verteilte sich während des fortlaufenden Reaktorbetriebs über sämtliche Oberflächen des Primärkreislaufs. Während eines Druckentlastungsstörfalls kann dieser Staub durch die Gasströmung remobilisiert und aus dem Primärkreislauf ausgetragen werden. Eine Quelltermanalyse solch eines Störfallszenarios erfordert die Kenntnis über die Menge und die räumliche Verteilung des Staubs, die radiologische Belastung sowie das Remobilisierungsverhalten in Bezug auf die zu erwartenden Strömungstransienten. Nach dem heutigen Stand von Wissenschaft und Technik kann die räumlich-zeitliche Verteilung des Staubs im Primärkreislauf für stationären Reaktorbetrieb unter Verwendung eindimensionaler Systemcodes abgeschätzt werden. Jedoch ist unbekannt, welcher Anteil des Staubinventars durch die Gasströmung remobilisiert und aus dem Primärkreislauf ausgetragen werden würde. Zur systematischen Untersuchung des Staubtransportverhaltens in turbulenten Strömungen wurden zwei kleinskalige Versuchsanlagen entwickelt und eine Serie von Depositions- und Resuspensionsexperimenten durchgeführt. Die partikelbeladene Strömung in der Heißgasumgebung des HTR-Primärkreislaufs wurde über die Verwendung von Ähnlichkeitskennzahlen auf eine Luftströmung bei Umgebungsbedingungen herunterskaliert. Die Strömung und die Partikel wurden mittels hochauflösender, bildgebender und nichtinvasiver Messverfahren räumlich und zeitlich vermessen, um eine umfangreiche Datenbasis für die Analyse der Partikeltransportprozesse zu erstellen. Inhaltlich lassen sich die durchgeführten Untersuchungen in drei Teile gliedern. Der erste Teil besteht aus zwei Studien über die Deposition und Resuspension monodisperser, sphärischer Einzelpartikel in einer ungestörten, horizontalen Kanalströmung. Die systematische Variation experimenteller Randbedingungen wie der Partikelgröße, der Oberflächenrauheit und der Strömungsgeschwindigkeit ermöglichte die Quantifizierung der einzelnen Einflussgrößen. Im zweiten und dritten Teil der Dissertation wurden die Deposition und Resuspension einer mehrschichtigen Ablagerung (Partikel-Multilayer) zwischen periodischen Stufen und in einer Kugelschüttung untersucht, um die komplexe Interaktion zwischen der turbulenten Strömung und der Multilayer-Ablagerung weiter zu erforschen. Die gewonnenen Erkenntnisse leisten einen Beitrag für die Quelltermanalyse des Staubtransports im HTR-Primärkreislauf und können für die Weiterentwicklung numerischer Strömungssimulationen des Partikeltransports in turbulenten Strömungen verwendet werden. / Aerosol particle deposition and resuspension experiments in turbulent flows were performed to investigate the complex particle transport phenomena and to provide a database for the development and validation of computational fluid dynamics (CFD) codes. The background motivation is related to the source term analysis of an accidental depressurization scenario of a High Temperature Reactor (HTR). During the operation of former HTR pilot plants, larger amounts of radio-contaminated graphite dust were found in the primary circuit. This dust most likely arose due to abrasion between the graphitic core components and was deposited on the inner wall surfaces of the primary circuit. In case of an accident scenario, such as a depressurization of the primary circuit, the dust may be remobilized and may escape the system boundaries. The estimation of the source term being discharged during such a scenario requires fundamental knowledge of the particle deposition, the amount of contaminants per unit mass as well as the resuspension phenomena. Nowadays, the graphite dust distribution in the primary circuit of an HTR can be calculated for stationary conditions using one-dimensional reactor system codes. However, it is rather unknown which fraction of the graphite dust inventory may be remobilized during a depressurization of the HTR primary circuit. Two small-scale experimental facilities were designed and a set of experiments was performed to investigate particle transport, deposition and resuspension in turbulent flows. The facility design concept is based on the fluid dynamic downscaling of the helium pressure boundary in the HTR primary circuit to an airflow at ambient conditions in the laboratory. The turbulent flow and the particles were recorded by high-resolution, non-invasive imaging techniques to provide a spatio-temporal insight into the particle transport processes. The different investigations of this thesis can be grouped into three categories. Firstly, the deposition and resuspension of monodisperse single particles in a horizontal turbulent channel flow was studied. The systematic variation of the experimental boundary conditions allows for the quantification of the influences of particle size, surface roughness, and fluid velocity. In the second and third part of this thesis, the deposition and resuspension of a particle multilayer between periodic steps and in a pebble bed was studied to explore the complex interaction between the turbulent flow and the particles, respectively. The findings of this thesis are a contribution to the source term analysis of HTR related accidental depressurizations. Furthermore, the database can be applied to CFD code developments for the numerical simulation of particle transport processes in turbulent flows.

Der VHTR als Weiterentwicklung des HTR gilt als eines von sechs aussichtsreichen Reaktorkonzepten für Kernkraftwerkte der Generation IV. Im Rahmen dieser Arbeit wird ein CFD-Modell des HTR-Moduls entwickelt und durch die Simulation eines postulierten Lufteinbruchszenarios die Anwendbarkeit unter Beweis gestellt. Zunächst wird eine Bestandsaufnahme bestehender HTR-Rechenprogramme vorgestellt und die Methodik CFD in ihren Grundzügen erläutert. Anhand der Grundgleichungen werden die zur Berechnung des Störfalls zu modellierenden, HTR-spezifischen Parameter diskutiert, die verwendeten empirischen Korrelationen vorgestellt und die umfangreiche Validierung des entwickelten Modellansatzes zusammengefasst. Anschließend wird die Anwendbarkeit des HTR-Modells auf ein konkretes Lufteinbruchszenario eines HTR-Moduls gezeigt. Dabei werden die einzelnen Phasen des Szenarios anhand der Simulationsergebnisse intensiv diskutiert. Abschließend erfolgt eine Diskussion der Modellunsicherheiten und der numerischen Fehler.

Current regulations specify the minimum number of operators required per nuclear power plant. However, these requirements are based on the operation of large nuclear power plants, which are not inherent safe and can result in a meltdown. For newly developed small nuclear reactors, the current number of operators seems to be excessive causing the technology to be less competitive. Before the number of required operators can be optimised, it should be demonstrated that human errors will not endanger or cause risk to the plant or public. For this study, a small pebble bed High Temperature Reactor (HTR) Nuclear Power Plant (NPP), the Th-100, was evaluated. The inherent safety features of this type of nuclear reactor include independent barriers for fission product capture and passive heat dissipation during a loss of coolant. The control and instrumentation architecture include two independent protection systems. The Control and Limitation System is the first protection system to react if the reactor parameters exceed those of the normal operational safe zone. If the Control and Limitation System fail to maintain the reactor within the safe zone, the Reactor Protection System would at that time operate and force the reactor to a safe state. Both these automated protection systems are installed in a control room local to the reactor building, protected from adverse conditions. In addition, it is connected to a semi-remote control room, anticipated as a multi-unit control room to include the monitoring and control of the auxiliary systems. Probable case studies of human error associated with multi-unit control rooms were evaluated against the logic of the Control and Limitation System. Fault Tree Analysis was used to investigate all possible failures. The evaluation determined the reliability of the Control and Limitation System and highlighted areas which design engineers should take into account if a higher reliability is required. The scenario was expanded, applying the same methods, to include the large release of fission products in order to verify the reliability calculations. The probability of a large release of fission products compared with studies done on other nuclear installations revealed to be much less for the evaluated HTR as was expected. As the study has proved that human error cannot have a negative influence on the safety of the reactor, it can be concluded that the first step has been met which is required, when applying for a waiver to utilise a multi-unit control room for the small pebble bed HTR NPP. Also, from the study, it is recommended that a practical approach be applied for the evaluation of operator duties on a live plant, to optimise the number of operators required. This in turn will position the inherent safe HTR competitively over other power stations. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

Current regulations specify the minimum number of operators required per nuclear power plant. However, these requirements are based on the operation of large nuclear power plants, which are not inherent safe and can result in a meltdown. For newly developed small nuclear reactors, the current number of operators seems to be excessive causing the technology to be less competitive. Before the number of required operators can be optimised, it should be demonstrated that human errors will not endanger or cause risk to the plant or public. For this study, a small pebble bed High Temperature Reactor (HTR) Nuclear Power Plant (NPP), the Th-100, was evaluated. The inherent safety features of this type of nuclear reactor include independent barriers for fission product capture and passive heat dissipation during a loss of coolant. The control and instrumentation architecture include two independent protection systems. The Control and Limitation System is the first protection system to react if the reactor parameters exceed those of the normal operational safe zone. If the Control and Limitation System fail to maintain the reactor within the safe zone, the Reactor Protection System would at that time operate and force the reactor to a safe state. Both these automated protection systems are installed in a control room local to the reactor building, protected from adverse conditions. In addition, it is connected to a semi-remote control room, anticipated as a multi-unit control room to include the monitoring and control of the auxiliary systems. Probable case studies of human error associated with multi-unit control rooms were evaluated against the logic of the Control and Limitation System. Fault Tree Analysis was used to investigate all possible failures. The evaluation determined the reliability of the Control and Limitation System and highlighted areas which design engineers should take into account if a higher reliability is required. The scenario was expanded, applying the same methods, to include the large release of fission products in order to verify the reliability calculations. The probability of a large release of fission products compared with studies done on other nuclear installations revealed to be much less for the evaluated HTR as was expected. As the study has proved that human error cannot have a negative influence on the safety of the reactor, it can be concluded that the first step has been met which is required, when applying for a waiver to utilise a multi-unit control room for the small pebble bed HTR NPP. Also, from the study, it is recommended that a practical approach be applied for the evaluation of operator duties on a live plant, to optimise the number of operators required. This in turn will position the inherent safe HTR competitively over other power stations. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

Thesis (MScEng)--Stellenbosch University, 2012. / ENGLISH ABSTRACT: A high temperature reactor (HTR) generates heat inside of the reactor core through nuclear fission, from where the heat is transferred through the core and heats up the reactor pressure vessel (RPV). The heat from the RPV is transported passively through the reactor cavity, where it is cooled by the reactor cavity cooling system (RCCS), through the concrete confinement structure and ultimately into the environment. The concrete confinement structure can withstand temperatures of up to 65°C for normal operating conditions and temperatures of up to 125°C during an emergency. This project endeavours to research the heat transfer between an HTR’s RPV and the outside of the concrete confinement structure by utilising three investigative approaches: experimental, computational fluid dynamics (CFD) and analytical. The first approach, an experimental analysis, required the development of an experi- mental model. The model was used to perform experiments and gather temperature data that could be used to verify the accuracy of the CFD simulations. The second approach was a CFD analysis of the experimental model, and the external concrete temperatures from the simulation were compared with the temperatures measured with the experimen- tal model. Finally, an analytical analysis was performed in order to better understand CFD and how CFD solves natural convection-type problems. The experiments were performed successfully and the measurements taken were com- pared with the CFD results. The CFD results are in good agreement with the Dry experiments, but not with the Charged experiments. It was identified that the inaccurate results for the CFD simulations of the Charged experiments arose due to convective heat leakage through gaps in the heat shield and between the heat shield and the sides of the experimental model. A computer program was developed for the analytical analysis and it was established that the program could successfully solve the natural convection in a square cavity - as required. / AFRIKAANSE OPSOMMING: ’n Hoë temperatuur reaktor (HTR) genereer hitte binne die reaktor kern deur kernsplyting en die hitte word dan deur die kern versprei en verhit die reaktor se drukvat. Die hitte van die reaktor drukvat word dan passief deur die reaktorholte versprei, waar dit deur die reaktorholte se verkoelingstelsel afgekoel word, en deur die beton beskermingstruktuur gelei word en uiteindelik die omgewing bereik. Die beton beskermingstruktuur kan temperature van tot 65°C onder normale operasietoestande van die reaktor weerstaan, en temperature van tot 125°C tydens ’n noodgeval. Hierdie projek poog om die hitte-oordrag tussen ’n HTR-reaktor drukvat en die buitekant van die beton beskermingstruktuur te on- dersoek deur gebruik te maak van drie ondersoekbenaderings: eksperimenteel, numeriese vloei dinamika (NVD) en analities. Die eerste benadering, ’n eksperimentele analise, het die ontwikkeling van ’n eksper- imentele model vereis. Die model is gebruik om eksperimente uit te voer en temperatu- urmetings te neem wat gebruik kon word om die akkuraatheid van die NVD simulasies te bevestig. Die tweede benadering was ’n NVD-analise van die eksperimentele model, en die eksterne betontemperature verkry van die simulasies is vergelyk met die gemete temperature van die eksperimente. Uiteindelik is ’n analitiese analise uitgevoer ten einde NVD beter te verstaan en hoe NVD natuurlike konveksie-tipe probleme sal oplos. Die eksperimente is suksesvol uitgevoer en die metings is gebruik om die NVD resultate mee te vergelyk. Die NVD resultate van die Droë eksperimente het goeie akkuraatheid getoon. Dit was nie die geval vir die Gelaaide eksperimente nie. Daar is geïdentifiseer dat die verskille in resultate tussen die NVD en die eksperimente aan natuurlike konveksie hitte verliese deur gapings in die hitteskuld en tussen die hitteskuld en die kante van die eksperimentele model toegeskryf kan word. ’n Rekenaarprogram is geskryf vir die analitiese ontleding en die program kon suksesvol die natuurlike konveksie in ’n vierkantige ruimte oplos.

The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source that can be utilized in almost all existing uranium–based reactors and can significantly help in conserving limited uranium reserves. Furthermore, the elimination of proliferation risks associated with thorium–based fuel cycles is a key reason for re–evaluating the possible utilization of thorium in high temperature reactors. In addition to the many advantages that thorium–based fuel has over uranium–based fuel, there are vast thorium resources in the earth's crust that up until the present have not been exploited optimally. This study focuses on determining the amount of uranium ore that can be saved using thorium as a nuclear fuel in HTR's. Four identical 200 MWth high temperature reactors are considered which make use of different fuel cycles. These fuel cycles range from the conventional uranium fuel cycle to a thorium–based fuel cycle in which no U–238 is present and have been simulated using the VSOP–A system of computer codes. This study also considers the effect that protactinium, an isotope that occurs in thorium–based fuel cycles, will have on the decay heat production in the case of a depressurized loss of coolant (DLOFC) accident. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source that can be utilized in almost all existing uranium–based reactors and can significantly help in conserving limited uranium reserves. Furthermore, the elimination of proliferation risks associated with thorium–based fuel cycles is a key reason for re–evaluating the possible utilization of thorium in high temperature reactors. In addition to the many advantages that thorium–based fuel has over uranium–based fuel, there are vast thorium resources in the earth's crust that up until the present have not been exploited optimally. This study focuses on determining the amount of uranium ore that can be saved using thorium as a nuclear fuel in HTR's. Four identical 200 MWth high temperature reactors are considered which make use of different fuel cycles. These fuel cycles range from the conventional uranium fuel cycle to a thorium–based fuel cycle in which no U–238 is present and have been simulated using the VSOP–A system of computer codes. This study also considers the effect that protactinium, an isotope that occurs in thorium–based fuel cycles, will have on the decay heat production in the case of a depressurized loss of coolant (DLOFC) accident. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.