High-temperature gas-cooled reactors (HTGR) are passively safe, efficient, and economical solutions to the world’s energy crisis. HTGRs are capable of generating high temperatures during normal operation, introducing design challenges related to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential to ensure the adequacy of safety features, such as the reactor cavity cooling system (RCCS). Modeling abilities of system analysis codes, used to develop an understanding of light water reactor phenomenology, need to be proven for HTGRs. RELAP5-3D v2.3.6 is used to generate two reactor plant models for a code-to-code and a code-to-experiment benchmark problem. The code-to-code benchmark problem models the Russian VGM reactor for pressurized and depressurized pressure vessel conditions. Temperature profiles corresponding to each condition are assigned to the pressure vessel heat structure. Experiment objectives are to calculate total thermal energy transferred to the RCCS for both cases. Qualitatively, RELAP5-3D’s predictions agree closely with those of other system codes such as MORECA and Thermix. RELAP5-3D predicts that 80% of thermal energy transferred to the RCCS is radiant. Quantitatively, RELAP5-3D computes slightly higher radiant and convective heat transfer rates than other system analysis codes. Differences in convective heat transfer rate arise from the type and usage of convection models. Differences in radiant heat transfer stem from the calculation of radiation shape factors, also known as view or configuration factors. A MATLAB script employs a set of radiation shape factor correlations and applies them to the RELAP5-3D model. This same script is used to generate radiation shape factors for the code-toexperiment benchmark problem, which uses the Japanese HTTR reactor to determine temperature along the outside of the pressure vessel. Despite lacking information on material properties, emissivities, and initial conditions, RELAP5-3D temperature trend predictions closely match those of other system codes. Compared to experimental measurements, however, RELAP5-3D cannot capture fluid behavior above the pressure vessel. While qualitatively agreeing over the pressure vessel body, RELAP5-3D predictions diverge from experimental measurements elsewhere. This difference reflects the limitations of using a system analysis code where computational fluid dynamics codes are better suited.

The failure of graphite moderator bricks in an Advanced Gas cooled Reactor (AGR) is potentially a serious problem. This thesis describes the generation of self stress in the moderator brick during irradiation and the derivation of a simple analytical model to predict the magnitude of this stress. The magnitude of the self stress in the brick is affected by the variations in the material properties of the graphite used for the brick and this is also examined, developing a statistical approach to the analysis. Property variations between bricks are considered but no allowance has been made for material property variations within a brick. Finally, the thesis compares the self stress in one of the critical peak rated moderator bricks to the strength of the irradiated oxidised material on a statistical basis and predicts the failure probability of a brick due to self stress to be extremely low at 25.5 full power years (FPY). However, the failure probability rises steeply and for the peak rated bricks at 29 FPY it approaches 100%.

The High Temperature Integrated Multi-Modular Thermal Reactor is a small modular reactor that uses an enhanced conductivity BeO-UO2 fuel with supercritical CO2 coolant to drive turbo-machinery in a direct Brayton cycle. The core consists of several self-contained pressurized modules, each containing fuel elements in pressurized channels surrounded by a graphite moderator, and Brayton cycle turbo-machinery. Each module is subcritical by itself, and when several modules are brought into proximity of one another, a single critical core is formed. The multi-modular approach and use of BeO-UO2 fuel with graphite moderator and supercritical CO2 coolant leads to an inherently safe system capable of high efficiency operation. The pressure channel design and multi-modular approach eliminates engineering challenges associated with large pressure vessels. The subcriticality of the modules ensures inherent safety during construction, transportation, and after decommissioning. Serpent, a continuous-energy Monte-Carlo reactor physics burnup calculation code, was used to develop a critical configuration of the subcritical modules using UO2 fuel enriched with 5 wt% 235U with a 5 wt% BeO additive. The core lifetime was found to be 14.6 years operating at 10 MWth, though the U enrichment and power can be altered to achieve desired core lifetimes. Negative fuel and moderator temperature coefficients of reactivity were found that could maintain safety during operation. The multi-modular design was found to be beneficial compared to a core with all fuel elements in one module. Batch battery type refueling was found to be beneficial and the feasibility of controlling the reactor was demonstrated through the use of control shells that surround each module. The HT-IMMTR design is an inherently safe, highly efficient, economically competitive, and most important, feasible reactor design that takes advantage of proven technologies to facilitate the demonstration of a successful commercial deployment.

No description available.

Graphite is used as a moderating material and as a structural component in a number of current generation nuclear reactors. During reactor operation stresses develop in the graphite components, causing them to deform. If significant numbers of graphite components were to fail in this manner, the material’s effectiveness as a neutron moderator will be reduced, and the reactor’s safe operation may be compromised. It is therefore important to understand how the microstructure of graphite affects the material’s response to these stresses. Despite much research into the effects of stress on nuclear grade graphite, there remain gaps in our understanding of this process, and there are a number of frequently observed limitations in the current research. Many existing studies either focus on the bulk material, ignoring the important changes at the microlevel; or focus on residual stresses due to the lack of available in-situ data. An experimental programme was designed to study stress-induced changes to the microstructures of Gilsocarbon and Pile Grade A graphite used in UK nuclear reactors. Particular focus was paid to the deformation of the pore structure, since graphite is highly porous and the porosity has a significant effect on the strength and structural integrity of the graphite components. A compression rig was used to simulate the build-up of operational stresses, during which confocal laser microscopy and X-ray tomography were performed to quantify changes to the pore structure at the microlevel; while X-ray diffraction was performed to study deformation of the crystal lattice and quantify the build-up of lattice strains. Pore properties of interest included pore area, surface area, volume, eccentricity, orientation, angularity and separation. Crystal lattice properties of interest included layer spacing, unit cell and crystallite size parameters, lattice strains and Bacon Anisotropy Factor. The experimental and analytical techniques were designed to significantly enhance our current understanding of how graphite responds to stress, with each observation made using a novel technique or improving the effectiveness of existing techniques. These studies have enabled significant novel observations and discussions of the stress-induced deformation behaviour of nuclear grade graphite to be made.

Social and environmental justice for a growing and developing global population requires significant increases in energy use. A possible means of contributing to this energy increase is to incinerate plutonium from spent fuel of pressurised light water reactors (Pu(PWR)) in high-temperature reactors such as the Pebble Bed Modular Reactor Demonstration Power Plant 400 MWth (PBMR-DPP-400). Previous studies showed that at low temperatures a 3 g Pu(PWR) loading per fuel sphere or less had a positive uniform temperature reactivity coefficient (UTC) in a PBMR DPP-400. The licensing of this fuel design is consequently unlikely. In the present study it was shown by diffusion simulations of the neutronics, using VSOP-99/05, that there is a fuel design containing thorium and plutonium that achieves a negative maximum UTC. Further, a fuel design containing 12 g Pu(PWR) loading per fuel sphere achieved a negative maximum UTC as well as the other PBMR (Ltd.) safety limits of maximum power per fuel sphere, fast fluence and maximum temperatures. It is proposed that the low average thermal neutron flux, caused by reduced moderation and increased absorption of thermal neutrons due to the higher plutonium loading, is responsible for these effects. However, to fully understand the mechanisms involved a detailed quantitative analysis of the roll of each factor is required. A 12 g Pu(PWR) loading per fuel sphere analysis shows a burn-up of 180.7 GWd/tHM which is approximately double the proposed PBMR (Ltd.) low enriched uranium fuel burn-up. The spent fuel has only a decrease of 24.5 % in the Pu content which is sub-optimal with respect to proliferation and waste disposal objectives. Incinerating Pu(PWR) in the PBMR-DPP 400 MWth is potentially licensable and economically feasible and should be considered for application by industry. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2012

Social and environmental justice for a growing and developing global population requires significant increases in energy use. A possible means of contributing to this energy increase is to incinerate plutonium from spent fuel of pressurised light water reactors (Pu(PWR)) in high-temperature reactors such as the Pebble Bed Modular Reactor Demonstration Power Plant 400 MWth (PBMR-DPP-400). Previous studies showed that at low temperatures a 3 g Pu(PWR) loading per fuel sphere or less had a positive uniform temperature reactivity coefficient (UTC) in a PBMR DPP-400. The licensing of this fuel design is consequently unlikely. In the present study it was shown by diffusion simulations of the neutronics, using VSOP-99/05, that there is a fuel design containing thorium and plutonium that achieves a negative maximum UTC. Further, a fuel design containing 12 g Pu(PWR) loading per fuel sphere achieved a negative maximum UTC as well as the other PBMR (Ltd.) safety limits of maximum power per fuel sphere, fast fluence and maximum temperatures. It is proposed that the low average thermal neutron flux, caused by reduced moderation and increased absorption of thermal neutrons due to the higher plutonium loading, is responsible for these effects. However, to fully understand the mechanisms involved a detailed quantitative analysis of the roll of each factor is required. A 12 g Pu(PWR) loading per fuel sphere analysis shows a burn-up of 180.7 GWd/tHM which is approximately double the proposed PBMR (Ltd.) low enriched uranium fuel burn-up. The spent fuel has only a decrease of 24.5 % in the Pu content which is sub-optimal with respect to proliferation and waste disposal objectives. Incinerating Pu(PWR) in the PBMR-DPP 400 MWth is potentially licensable and economically feasible and should be considered for application by industry. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2012

23000 tonnes de déchets graphites seront générés lors du démantèlement de la première filière de réacteurs en France (9 réacteurs Uranium Naturel Graphite Gas, UNGG). Ces déchets radioactifs sont classés dans la catégorie Faible Activité Vie Longue (FAVL). Dans le cadre de la loi, l’agence nationale pour la gestion des déchets radioactifs (Andra) étudie un concept de stockage à faible profondeur. Cette étude s’intéresse plus particulièrement au carbone 14, qui est un des principaux radionucléides à vie longue (5730 ans) dans les déchets graphite, mais aussi au tritium qui est l’un des principaux contributeurs de la radioactivité à court terme. Ces deux radionucléides ont la particularité d’exister sous différentes formes, aussi bien en phase gaz (14CO2, HT,…) qu’en phase liquide (14CO32-, HTO,…). Leur spéciation va influencer leur migration du stockage vers l’environnement. Des expériences de lixiviation en milieu alcalin (NaOH 0,1mol.L-1, simulant les conditions de stockage), ont été réalisées sur des échantillons de graphites irradiés provenant de deux réacteurs : SLA2 et G2, afin de quantifier leur relâchement et de définir leur spéciation. Les études montrent que le carbone se trouve aussi bien en phase gaz qu’en phase liquide. Dans la phase gaz, le relâchement est faible (< 0,1%), et correspond à des formes oxydables. Le carbone 14 est relâché majoritairement en phase liquide : 65% de la fraction d’inventaire relâchée est sous forme de carbone 14 inorganique, et 35% de carbone 14 organique. Deux formes de tritium ont été identifiées dans la phase gaz : HTO et HT/Tritium Organiquement Lié. Plus de 90% du tritium en phase gaz se trouve sous forme HT/TOL, mais ce relâchement est faible (<0,1%). Majoritairement le tritium est en phase liquide sous forme HTO. / 23000 tons of graphite wastes will be generated during dismantling of the first generation of French reactors (9 gas cooled reactors). These wastes are classified as Long Lived Low Level wastes (LLW-LL). As requested by the law, the French National Radioactive Waste Management Agency (Andra) is studying concepts of low-depth disposals.In this work we focus on carbon 14, the main long-lived radionuclide in graphite waste (5730y), but also on tritium, which is the main contributor to the radioactivity in the short term. Carbon 14 and tritium may be released from graphite waste in many forms in gaseous phase (14CO2, HT…) or in solution (14CO32-, HTO…). Their speciation will strongly affect their migration from the disposal site to the environment. Leaching experiments, in alkaline solution (0.1 M NaOH simulating repository conditions) have been performed on irradiated graphite, from Saint-Laurent A2 and G2 reactors, in order to quantify their release and characterize their speciation. The studies show that carbon 14 exists in both gaseous and aqueous phases. In the gaseous phase, release is weak (<0.1%) and corresponds to oxidizable species. Carbon 14 is mainly released into liquid phase, as both inorganic and organic species. 65% of released fraction is inorganic and 35% organic carbon. Two tritiated species have been identified in gaseous phase: HTO and HT/Organically Bond Tritium. More than 90% of tritium in that phase corresponds to HT/OBT. But release is weak (<0.1%). HTO is mainly in the liquid phase.

In this thesis advanced fuel cycles for the incineration, i.e. deep–burn, of weapons–grade plutonium, reactor–grade plutonium from pressurised light water reactors and reactor–grade plutonium + the associated Minor Actinides in the 400 MWth Pebble Bed Modular Reactor Demonstration Power Plant was simulated with the VSOP 99/05 diffusion code. These results were also compared to the standard 9 g/fuel sphere U/Pu 9.6% enriched uranium fuel cycle. The addition of the Minor Actinides to the reactor–grade plutonium caused an unacceptable decrease in the burn–up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which is intended for direct disposal in a deep geological repository, without chemical reprocessing. All the Pu fuel cycles failed the adopted safety limits in that either the maximum fuel temperature of 1130°C, during normal operation, or the maximum power of 4.5 kW/sphere was exceeded. All the Pu cycles also produced positive Uniform Temperature Reactivity Coefficients, i.e. the coefficient where the temperature of the fuel and the graphite moderator in the fuel spheres are varied together. these positive temperature coefficients were experienced at low temperatures, typically below 700°C. This was due to the influence of the thermal fission resonance of 241Pu. The safety performance of the weapons–grade plutonium was the worst. The safety performance of the reactor–grade plutonium also deteriorated when the heavy metal loading was reduced from 3 g/sphere to 2 g or 1 g. In view of these safety problems, these Pu fuel cycles were judged to be not licensable in the PBMR DPP–400 reactor. Therefore a redesign of the fuel cycle for reactor–grade plutonium, the power conversion system and the reactor geometry was proposed in order to solve these problems. The main elements of these proposals are: v 1. The use of 3 g reactor–grade plutonium fuel spheres should be the point of departure. 232Th will then be added in order to restore negative Uniform Temperature Reactivity Coefficients. 2. The introduction of neutron poisons into the reflectors, in order to suppress the power density peaks and thus the temperature peaks. 3. In order to counter the reduction in burn–up by this introduction of neutron poisons, a thinning of the central reflector was proposed. / Thesis (PhD (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.

In this thesis advanced fuel cycles for the incineration, i.e. deep–burn, of weapons–grade plutonium, reactor–grade plutonium from pressurised light water reactors and reactor–grade plutonium + the associated Minor Actinides in the 400 MWth Pebble Bed Modular Reactor Demonstration Power Plant was simulated with the VSOP 99/05 diffusion code. These results were also compared to the standard 9 g/fuel sphere U/Pu 9.6% enriched uranium fuel cycle. The addition of the Minor Actinides to the reactor–grade plutonium caused an unacceptable decrease in the burn–up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which is intended for direct disposal in a deep geological repository, without chemical reprocessing. All the Pu fuel cycles failed the adopted safety limits in that either the maximum fuel temperature of 1130°C, during normal operation, or the maximum power of 4.5 kW/sphere was exceeded. All the Pu cycles also produced positive Uniform Temperature Reactivity Coefficients, i.e. the coefficient where the temperature of the fuel and the graphite moderator in the fuel spheres are varied together. these positive temperature coefficients were experienced at low temperatures, typically below 700°C. This was due to the influence of the thermal fission resonance of 241Pu. The safety performance of the weapons–grade plutonium was the worst. The safety performance of the reactor–grade plutonium also deteriorated when the heavy metal loading was reduced from 3 g/sphere to 2 g or 1 g. In view of these safety problems, these Pu fuel cycles were judged to be not licensable in the PBMR DPP–400 reactor. Therefore a redesign of the fuel cycle for reactor–grade plutonium, the power conversion system and the reactor geometry was proposed in order to solve these problems. The main elements of these proposals are: v 1. The use of 3 g reactor–grade plutonium fuel spheres should be the point of departure. 232Th will then be added in order to restore negative Uniform Temperature Reactivity Coefficients. 2. The introduction of neutron poisons into the reflectors, in order to suppress the power density peaks and thus the temperature peaks. 3. In order to counter the reduction in burn–up by this introduction of neutron poisons, a thinning of the central reflector was proposed. / Thesis (PhD (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.