Thesis (MSc)--Stellenbosch University, 2011. / ENGLISH ABSTRACT: In the current design of the high temperature gas cooled reactor, a small fraction of coated fuel particles will be defective. Hence, 110Ag may be released from the fuel spheres into the coolant gas (helium) and plate out on the cooler surfaces of the main power system. This poses a radiation risk to operating personnel as well as general public. The objectives of this thesis were to design and construct an apparatus in which silver-109 atoms may be produced and deflected in an inhomogeneous and homogeneous magnetic field, compare experimental and theoretical results, and make a recommendation based on the findings of this thesis to the idea of removing silver-110 atoms from the helium fluid by deflecting them with an inhomogeneous magnetic field onto target plates situated on the inner perimeter of a helium pipe. The experimental results for the deflection of the collimated Ag- atoms with the round-hole collimators showed a deflection of 1.77° and 2.05° of the Ag- atoms due to an inhomogeneous magnetic field when the target plate was positioned 13 and 30 mm away from the magnet, respectively. These values were considerably greater than 0.01° and 0.02° that were calculated for the average velocity of atoms, v = 500 m/s. The case where Ag- atoms were collimated with a pair of slits and the target plate positioned 13mm away from the magnet showed the following: An inhomogeneous magnetic field changes the rectangular shape of the beam to a roughly elliptical shape. The beam of Ag- atoms was not split into two separate beams. This was caused by the beam of Ag- atoms consisting of atoms travelling at different speeds. The maximum deflection of Ag- atoms was 1.16° in the z direction and 1.12° in the x direction. These values were also significantly greater than 0.01 mm calculated at v = 500 m/s. This huge difference between the theoretical and experimental results raised a conclusion that the size of each Ag deposit depended mostly on the exposure time that was given to it. It was noticed that the beam of Ag- atoms was not split into two separate beams, in both cases. The conclusion was that the technique of removing Ag- atoms from the helium stream by means of an inhomogeneous magnetic field may not be effective. This is due to the inability of the inhomogeneous magnetic field to split the beam of Ag- atoms into two separate beams in a vacuum of ~10-5 mbar. It would be even more difficult for an inhomogeneous magnetic field to split the beam of Ag- atoms in helium, due to the Ag- atoms having a shorter mean free path in helium compared to a vacuum. / AFRIKAANSE OPSOMMING: In die huidige ontwerp van die hoë temperatuur gas afgekoelde reaktor, is 'n klein fraksie van omhulde brandstof deeltjies foutief. 110Ag kan dus vrygestel word vanaf die brandstof sfere in die verkoelingsgas (helium) wat dan op die koeler oppervlaktes van die hoofkragstelsel presipiteer. Hierdie 110Ag deeltjies hou 'n bestraling risiko vir die bedryfpersoneel sowel as vir die algemene publiek in. Die doelwitte van hierdie verhandeling is eerstens om 'n apparaat te ontwerp en konstrueer wat silwer-109 atome produseer en nie-homogene en homogene magnetiese velde deflekteer,. Tweedens om die eksperimentele en teoretiese resultate met mekaar te vergelyk. Derdens om 'n aanbeveling te maak gebasseer op die bevindinge van hierdie verhandeling rakende die verwydering van silwer-110 atome uit die helium vloeistof deur hulle met 'n nie-homogene magneetveld te deflekteer op die teikenplate binne-in 'n helium pyp. Die eksperimentele resultate vir die defleksie van die gekollimeerde Ag-atome met die ronde gat kollimators toon ‘n defleksie van 1.77° en 2.05° van die Ag-atome as gevolg van ‘n nie-homogene magneetveld wanneer die teikenplaat 13mm en 30mm, onderskeidelik, vanaf die magneet geposisioneer is. Hierdie waardes is aansienlik groter as die teoretiese defleksies van 0.01° en 0.02o wat bereken is vir ‘n gemiddelde snelheid van 500 m/s vir die atome. Die geval waar Ag-atome met 'n paar splete gekollimeer is en die teikenplaat 13 mm weg van magneet geposisioneer is, is die volgende resultate verkry: 'n nie-homogene magneetveld verander die reghoekige vorm van die bondel na 'n rowwe elliptiese vorm. Die bondel Ag-atome is nie volkome twee afsonderlike bundels verdeel nie. Dit is omdat die bondel van Ag-atome bestaan uit atome wat teen verskillende snelhede beweeg. Die maksimum defleksie van Ag-atome is 1.16° in die z-rigting en 1.12° in die x-rigting. Hierdie waardes is ook aansienlik groter as 0.01° bereken teen 500 m/s. Hierdie groot verskil tussen die teoretiese en eksperimentele resultate dui daarop dat die grootte van elke Ag neerslag grootliks afhanklik is van die blootstellingstyd wat daaraan gegee is. Daar is vasgestel dat die straal van Ag-atome in beide gevalle nie in twee afsonderlike bondels verdeel nie. Die gevolgtrekking is dat die tegniek van die verwydering van Ag-atome uit die helium stroom deur middel van 'n nie-homogene magneetveld nie effektief is nie. Dit is te wyte aan die onvermoë van die nie-homogene magneetveld om die bondel Ag-atome te verdeel in twee afsonderlike bondels in 'n vakuum van ~ 10-5 mbar. Dit sou selfs nog moeiliker vir 'n nie-homogene magnetiese veld wees om die bundel Ag-atome in helium te verdeel, weens die korter gemiddelde beskikbare pad van Ag-atome in helium wanneer dit met 'n vakuum vergelyk word.

The premise of the work presented here is to use a common analytical tool, Computational Fluid dynamics (CFD), along with a difference turbulence models. Eddy viscosity models as well as state-of-the-art Large Eddy Simulation (LES) were used to study the flow past bluff bodies. A suitable CFD code (CFX5.6b) was selected and implemented. Simulation of turbulent transport for the gas through the gaps of the randomly distributed spherical fuel elements (pebbles) was performed. Although there are a number of numerical studies () on flows around spherical bodies, none of them use the necessary turbulence models that are required to simulate flow where strong separation exists. With the development of high performance computers built for applications that require high CPU time and memory; numerical simulation becomes one of the more effective approaches for such investigations and LES type of turbulence models can be used more effectively. Since there are objects that are touching each other in the present study, a special approach was applied at the stage of building computational domain. This is supposed to be a considerable improvement for CFD applications. Zero thickness was achieved between the pebbles in which fission reaction takes place. Since there is a strong pressure gradient as a result of high Reynolds Number on the computational domain, which strongly affects the boundary layer behavior, heat transfer in both laminar and turbulent flows varies noticeably. Therefore, noncircular curved flows as in the pebble-bed situatio n, in detailed local sense, is interesting to be investigated. Since a compromise is needed between accuracy of results and time/cost of effort in acquiring the results numerically, selection of turbulence model should be done carefully. Resolving all the scales of a turbulent flow is too costly, while employing highly empirical turbulence models to complex problems could give inaccurate simulation results. The Large Eddy Simulation (LES) method would achieve the requirements to obtain a reasonable result. In LES, the large scales in the flow are solved and the small scales are modeled. Eddy viscosity and Reynolds stress models were also be used to investigate the applicability of these models for this kind of flow past bluff bodies at high Re numbers.

Irradiation experiments for the pebble bed modular reactor PBMR fuel (coated fuel particles and pebble fuel) are planned at the South African First Atomic Reactor Installation (SAFARI-1). The experiments are conducted to investigate the behavior of the fuel under normal operating and accelerated/accident simulating conditions because the safe operation of the reactor relies on the integrity of the fuel for retention of radioactivity. For fuel irradiation experiments, the accurate knowledge and analysis of the neutron spectrum of the irradiation facility is required. In addition to knowledge of the neutron spectrum in the irradiation facility, power distributions and knowledge of nuclear heating values has to be acquired. The SAFARI-1 reactor boosts operating fluid temperatures of about 300 K. On the contrary, the PBMR can reach temperatures in up to about 1370 K under normal operating conditions. This calls for design of high temperature irradiation rigs for irradiation of the PBMR fuel in the SAFARI-1 reactor. The design of this instrument (rig) should be such that to create an isolated high temperature environment in the SAFARI-1 reactor, to achieve the requirements of the PBMR fuel irradiation program. The design of the irradiation rig is planned such that the rig should fit in the existing irradiation channels of the SAFARI-1 reactor, a time and cost saving from the licensing perspective. This study aims to establish the know-how of coated particle and pebble modeling in using the Monte Carlo N-Particle code (MCNP5). The study also aims to establish the know-how of rig design. In this study, the Necsa in-house code Overall System for the Calculation of Reactors (OSCAR-3), a software known as OScar 3-Mcnp INTerface (OSMINT) linking OSCAR-3 and MCNP5, also developed at Necsa, as well as MCNP5 code developed and maintained by the Los Alamos team, are used to calculate neutronic and power distribution parameters that are important for fuel irradiations and for rig design. This study presents results and data that can be used to make improvements in the design of the rig or to confirm if the required operational conditions can be met with the current preliminary rig design. Result of the neutronic analysis are presented for the SAFARI-1 core, core irradiation channel B6 (where the PBMR fuel irradiation rig is loaded for the purpose of this study), the rig structure and the pebble fuel are presented. Furthermore results of the power distribution and nuclear heating values in the reactor core, the irradiation channel B6, the rig structures and the pebble fuel is also presented. The loading of the PBMR fuel irradiation rig in core position B6 reduces the core reactivity due to the fact that the loading of the rig displaces the water moderator in channel B6 introducing vast amounts of helium. This impacts on the keff value because there will be less neutron thermalization and reproduction due to the decreased population of thermal neutrons. The rig is found to introduce a negative reactivity insertion of 46 pcm. The loading of this rig in the core leads to no significant perturbations on the core power distribution. The core hottest channel is still localized in core channel C6 both with RIG IN and RIG OUT cases. A power tilt is observed, with the south side of the core experiencing reduced assembly averaged fission power, with correspondingly small compensations from the assemblies on the north side of the core. The perturbations on the core assembly averaged fluxes are more pronounced in the eight assemblies surrounding B6. Core position B6 suffers an 18% neutron flux depression with the loading of the rig. The fluxes in core positions A5, A6, A7, B5, B7 and C7 are increased when the rig is loading. The largest increases are noted as 12% in A7, 9% in A6 and 6% in A5 and B7. All the eight core positions surrounding B6 experience reduced photon fluxes with the loading of the rig. Core position B6 shows a flux depression of up to 20%, with 10% reduction in core position A6. The remainder seven positions surrounding B6 shows flux depressions of no more than 5%. Further on, due to decreased moderation effects, the axial neutron flux in core position B6 is reduced by 20% when the rig is loaded. The energy dependent neutron flux in B6 decreases by 50% in the thermal energy range with corresponding increases of up to 50% in the resonance and fast energy regions. The axial and the energy dependent photon flux in core position B6 decreases by up to 20% when the rig is loaded. The magnitude of the neutron and photon fluxes is found to have a direct proportion on the neutron and photon heating values. While the amount of neutron heating in core position B6 increases by one order of magnitude, when the rig is loaded, the photon heating values increases by up to 60% in the region spanning ±10cm about the core centerline. The amount of photon heating in the rig structural materials dominates neutron heating, except in the helium regions of the rig, where neutron heating dominates photon heating. In the fuel region of the pebble, fission heating (3803W) largely dominates photon heating (119W). / Thesis (M.Sc. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2009

Irradiation experiments for the pebble bed modular reactor PBMR fuel (coated fuel particles and pebble fuel) are planned at the South African First Atomic Reactor Installation (SAFARI-1). The experiments are conducted to investigate the behavior of the fuel under normal operating and accelerated/accident simulating conditions because the safe operation of the reactor relies on the integrity of the fuel for retention of radioactivity. For fuel irradiation experiments, the accurate knowledge and analysis of the neutron spectrum of the irradiation facility is required. In addition to knowledge of the neutron spectrum in the irradiation facility, power distributions and knowledge of nuclear heating values has to be acquired. The SAFARI-1 reactor boosts operating fluid temperatures of about 300 K. On the contrary, the PBMR can reach temperatures in up to about 1370 K under normal operating conditions. This calls for design of high temperature irradiation rigs for irradiation of the PBMR fuel in the SAFARI-1 reactor. The design of this instrument (rig) should be such that to create an isolated high temperature environment in the SAFARI-1 reactor, to achieve the requirements of the PBMR fuel irradiation program. The design of the irradiation rig is planned such that the rig should fit in the existing irradiation channels of the SAFARI-1 reactor, a time and cost saving from the licensing perspective. This study aims to establish the know-how of coated particle and pebble modeling in using the Monte Carlo N-Particle code (MCNP5). The study also aims to establish the know-how of rig design. In this study, the Necsa in-house code Overall System for the Calculation of Reactors (OSCAR-3), a software known as OScar 3-Mcnp INTerface (OSMINT) linking OSCAR-3 and MCNP5, also developed at Necsa, as well as MCNP5 code developed and maintained by the Los Alamos team, are used to calculate neutronic and power distribution parameters that are important for fuel irradiations and for rig design. This study presents results and data that can be used to make improvements in the design of the rig or to confirm if the required operational conditions can be met with the current preliminary rig design. Result of the neutronic analysis are presented for the SAFARI-1 core, core irradiation channel B6 (where the PBMR fuel irradiation rig is loaded for the purpose of this study), the rig structure and the pebble fuel are presented. Furthermore results of the power distribution and nuclear heating values in the reactor core, the irradiation channel B6, the rig structures and the pebble fuel is also presented. The loading of the PBMR fuel irradiation rig in core position B6 reduces the core reactivity due to the fact that the loading of the rig displaces the water moderator in channel B6 introducing vast amounts of helium. This impacts on the keff value because there will be less neutron thermalization and reproduction due to the decreased population of thermal neutrons. The rig is found to introduce a negative reactivity insertion of 46 pcm. The loading of this rig in the core leads to no significant perturbations on the core power distribution. The core hottest channel is still localized in core channel C6 both with RIG IN and RIG OUT cases. A power tilt is observed, with the south side of the core experiencing reduced assembly averaged fission power, with correspondingly small compensations from the assemblies on the north side of the core. The perturbations on the core assembly averaged fluxes are more pronounced in the eight assemblies surrounding B6. Core position B6 suffers an 18% neutron flux depression with the loading of the rig. The fluxes in core positions A5, A6, A7, B5, B7 and C7 are increased when the rig is loading. The largest increases are noted as 12% in A7, 9% in A6 and 6% in A5 and B7. All the eight core positions surrounding B6 experience reduced photon fluxes with the loading of the rig. Core position B6 shows a flux depression of up to 20%, with 10% reduction in core position A6. The remainder seven positions surrounding B6 shows flux depressions of no more than 5%. Further on, due to decreased moderation effects, the axial neutron flux in core position B6 is reduced by 20% when the rig is loaded. The energy dependent neutron flux in B6 decreases by 50% in the thermal energy range with corresponding increases of up to 50% in the resonance and fast energy regions. The axial and the energy dependent photon flux in core position B6 decreases by up to 20% when the rig is loaded. The magnitude of the neutron and photon fluxes is found to have a direct proportion on the neutron and photon heating values. While the amount of neutron heating in core position B6 increases by one order of magnitude, when the rig is loaded, the photon heating values increases by up to 60% in the region spanning ±10cm about the core centerline. The amount of photon heating in the rig structural materials dominates neutron heating, except in the helium regions of the rig, where neutron heating dominates photon heating. In the fuel region of the pebble, fission heating (3803W) largely dominates photon heating (119W). / Thesis (M.Sc. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2009

Thesis (MScEng)--Stellenbosch University, 2012. / ENGLISH ABSTRACT: The focus of this project was on the potential application of waste heat recovery and utilisation (WHR&U) systems in pebble bed modular reactor (PBMR) technology. The background theory provided in the literature survey showed that WHR&U systems have attracted the attention of many researchers over the past two decades, as using waste heat improves the system overall efficiency, notwithstanding the cost of extra plant. PBMR waste heat streams were identified and investigated based on the amount of heat rejected to the environment. WHR&U systems require specially designed heat recovery equipment, and as such the used and/or spent PBMR fuel tanks were considered by the way of example. An appropriately scaled system was designed, built and tested, to demonstrate the functioning of such a cooling system. Two separate and independent cooling lines, using natural circulation flow in a particular form of heat pipes called thermosyphon loops were used to ensure that the fuel tank is cooled when the power conversion unit has to be switched off for maintenance, or if it fails. A theoretical model that simulates the heat transfer process in the as-designed WHR&U system was developed. It is a one-dimensional flow model assuming quasi-static and incompressible liquid and vapour flow. An experimental investigation of the WHR&U system was performed in order to validate the theoretical model results. The experimental results were then used to modify the theoretical heat transfer coefficients so that they simulate the experiments more accurately. Three energy conversion devices, the dual-function absorption cycle (DFAC), the organic Rankine cycle (ORC) and the Stirling engine (SE), were identified as suitable for transforming the recovered heat into a useful form, depending on the source temperatures from 60 ºC to 800 ºC. This project focuses on a free-piston SE with emphasis on the thermo-dynamic performance of a SE heat exchanger. It was found that a heat exchanger with a copper woven wire mesh configuration has a relatively large gas-to-metal and metal-to-liquid heat transfer area. Tube-in-shell heat exchanger configurations were tested, with the working fluid flowing in ten copper inner pipes, while a coolant flows through the shell tube. A lumped parameter model was used to describe the thermo-fluid dynamic behaviour of the SE heat exchanger. In order to validate the theoretical results, a uni-directional flow experimental investigation was performed. The theoretical model was adjusted so that it simulated the SE heat exchanger. It was found that after this correction the theoretical model accurately predicts the experiment. Finally, a dynamic analysis of the SE heat exchanger experimental set-up was undertaken to show that, although vibrating, the heat exchanger setup assembly was indeed acceptable from a vibrational and fatigue point of view. / AFRIKAANSE OPSOMMING: Die hoofoogmerk met hierdie projek was die moontlike aanwending van afvalhitteherwinningen- benutting-(WHR&U-) stelsels in modulêre-gruisbedreaktor-(PBMR-) tegnologie. Agtergrondteorie in die literatuurondersoek toon dat WHR&U-stelsels al menige navorser se belangstelling geprikkel het, hetsy vanweë die moontlike ekonomiese voordele wat dit inhou óf vir besoedelingsvoorkoming, bo-en-behalwe die koste van bykomende toerusting. Die PBMRafvalhittestrome is ondersoek en bepaal op grond van die hoeveelheid hitte wat dit na die omgewing vrystel. Om in die prosesbehoeftes van WHR&U-stelsels te voorsien, moet goed ontwerpte, doelgemaakte hitteherwinningstoerusting in ʼn verkoelings- en/of verhittingsproses gebruik word, dus is die PBMR as voorbeeld gebruik vir die konsep. ʼn Toepaslik geskaleerde WHR&U-stelsel is dus ontwerp, gebou en getoets om die geldigheid van die stelselontwerp te toon. Twee onafhanklike verkoelingslyne, wat van natuurlike konveksie gebruik maak, in die vorm van hitte-pype of termoheuwel lusse, was gebruik om te verseker dat verkoeling verskaf word wanneer die hoof lus breek of instandhouding nodig hê. ʼn Teoretiese model is ontwikkel wat die hitteoordragproses in die ontwerpte WHR&U-stelsel simuleer. Dié model was ʼn eendimensionele vloeimodel wat kwasistatiese en onsamedrukbare vloeistof- en dampvloei in die WHR&U-stelsel-lusse veronderstel. ʼn Eksperimentele ondersoek is op die WHR&U-stelsel uitgevoer ten einde die teoretiese model se resultate te bevestig. Die eksperimentele resultate was dus geneem om die teoretiese hitteoordragkoëffisiënte aan te pas sodat dit die eksperimente kon simuleer. Drie energieomsettingstoestelle, naamlik die dubbel funksie absorpsie siklus (DFAC), die organiese Rankine siklus (ORC) en die Stirling enjin (SE), is as geskikte toestelle uitgewys om die herwonne hitte op grond van brontemperature tussen 60 ºC en 800 ºC in ʼn bruikbare vorm om te sit. Hierdie tesis het op vryesuier-SE’s gekonsentreer, met klem op die hitteruiler. Meer bepaald is die termodinamiese werkverrigting van ʼn SE-hitteruiler ondersoek. Daar is bevind dat ʼn hitteruiler met ʼn geweefde koperdraadmaas-samestelling oor ʼn betreklik groot gas-totmetaal- en metaal-tot-vloeistof-oordragoppervlakte beskik. Die verhitter en verkoeler is in ʼn buis-in-mantel-vorm ontwerp, met die werksvloeistof wat deur tien koperbinnepype vloei en ʼn koelmiddel deur die mantelbuis. ʼn Saamgevoegde-parameter-model is gebruik om die termodinamiese gedrag van die SEhitteruiler te beskryf. Ten einde die teoretiese resultate te bevestig, is ʼn eenrigtingvloeiproefondersoek uitgevoer. Die teoretiese model is aangepas sodat dit die SE-hitteruiler kon simuleer. Ná die nodige verstellings is daar bevind dat die teoretiese model die proefneming akkuraat voorspel. Laastens was ʼn dinamiese ontleding van die SE-hitteruiler ook onderneem om te toon dat, hoewel dit vibreer, die hitteruiler proef samestel inderdaad veilig is.

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.

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

The Pebble Bed Modular Reactor (PBMR) is one of the most technologically advanced developments in South Africa. In order to build a commercially viable demonstration power plant, all the specifically and uniquely designed equipment must first be qualified. All the prototype equipment is tested at the Helium Test Facility (HTF) at Pelindaba. One of the largest components that are tested is the Core Unloading Device (CUD). The main function of the CUD is to unload fuel from the bottom of the reactor core to enable circulation of the fuel core. The CUD housing vessel forms part of the reactor pressure boundary. Pebble-directing valves and other moving machinery are installed inside its machined inner surface. It is essential that the interior surfaces of the CUD are case hardened to provide a corrosion- and wear-resistant layer. Cold welding between the moving metal parts and the machined surface must also be prevented. Nitriding is a case hardening process that adds a hardened wear- and corrosion-resistant layer that will also prevent cold welding of the moving parts in the helium atmosphere. Only a few nitriding furnaces exist that can house a forging as large as the CUD of the PBMR. Commercial nitriding furnaces in South Africa are all too small and have limited flexibility in terms of the nitriding process. The nitriding of a vessel as large as the CUD has not yet been carried out commercially. The aim of this work was to design and develop a custom-made nitriding plant to perform the nitriding of the first PBMR/HTF CUD. Proper process control is essential to ensure that the required nitrided case has been obtained. A new concept for a gas nitriding plant was developed using the nitrided vessel interior as the nitriding process chamber. Before the commencement of detail design, a laboratory test was performed on a scale model vessel to confirm concept feasibility. The design of the plant included the mechanical design of various components essential to the nitriding process. A special stirring fan with an extended length shaft was designed, taking whirling speed into account. Considerable research was performed on the high temperature use of the various components to ensure the safe operation of the plant at temperatures of up to 600°C. Nitriding requires the use of hazardous gases such as ammonia, oxygen and nitrogen. Hydrogen is produced as a by-product and therefore safety was the most important design parameter. Thermohydraulic analyses, i.e. heat transfer and pressure drop calculations in pipes, were also performed to ensure the successful process design of the nitriding plant. The nitriding plant was subsequently constructed and operated to verify the correct design. A large amount of experimental and operating data was captured during the actual operation of the plant. This data was analysed and the thermohydraulic analyses were verified. Nitrided specimens were subjected to hardness and layer thickness tests. The measured temperature of the protruding fan shaft was within the limits predicted by Finite Element Analysis (FEA) models. Graphs of gas flow rates and other operation data confirmed the inverse proportionality between ammonia supply flow rate and measured dissociation rate. The design and operation of the nitriding plant were successful as a nitride layer thickness of 400 μm and hardness of 1 200 Vickers hardness (VHN) was achieved. This research proves that a large pressure vessel can successfully be nitrided using the vessel interior as a process chamber. / Thesis (M.Ing. (Mechanical Engineering))--North-West University, Potchefstroom Campus, 2010.

The Pebble Bed Modular Reactor (PBMR) is one of the most technologically advanced developments in South Africa. In order to build a commercially viable demonstration power plant, all the specifically and uniquely designed equipment must first be qualified. All the prototype equipment is tested at the Helium Test Facility (HTF) at Pelindaba. One of the largest components that are tested is the Core Unloading Device (CUD). The main function of the CUD is to unload fuel from the bottom of the reactor core to enable circulation of the fuel core. The CUD housing vessel forms part of the reactor pressure boundary. Pebble-directing valves and other moving machinery are installed inside its machined inner surface. It is essential that the interior surfaces of the CUD are case hardened to provide a corrosion- and wear-resistant layer. Cold welding between the moving metal parts and the machined surface must also be prevented. Nitriding is a case hardening process that adds a hardened wear- and corrosion-resistant layer that will also prevent cold welding of the moving parts in the helium atmosphere. Only a few nitriding furnaces exist that can house a forging as large as the CUD of the PBMR. Commercial nitriding furnaces in South Africa are all too small and have limited flexibility in terms of the nitriding process. The nitriding of a vessel as large as the CUD has not yet been carried out commercially. The aim of this work was to design and develop a custom-made nitriding plant to perform the nitriding of the first PBMR/HTF CUD. Proper process control is essential to ensure that the required nitrided case has been obtained. A new concept for a gas nitriding plant was developed using the nitrided vessel interior as the nitriding process chamber. Before the commencement of detail design, a laboratory test was performed on a scale model vessel to confirm concept feasibility. The design of the plant included the mechanical design of various components essential to the nitriding process. A special stirring fan with an extended length shaft was designed, taking whirling speed into account. Considerable research was performed on the high temperature use of the various components to ensure the safe operation of the plant at temperatures of up to 600°C. Nitriding requires the use of hazardous gases such as ammonia, oxygen and nitrogen. Hydrogen is produced as a by-product and therefore safety was the most important design parameter. Thermohydraulic analyses, i.e. heat transfer and pressure drop calculations in pipes, were also performed to ensure the successful process design of the nitriding plant. The nitriding plant was subsequently constructed and operated to verify the correct design. A large amount of experimental and operating data was captured during the actual operation of the plant. This data was analysed and the thermohydraulic analyses were verified. Nitrided specimens were subjected to hardness and layer thickness tests. The measured temperature of the protruding fan shaft was within the limits predicted by Finite Element Analysis (FEA) models. Graphs of gas flow rates and other operation data confirmed the inverse proportionality between ammonia supply flow rate and measured dissociation rate. The design and operation of the nitriding plant were successful as a nitride layer thickness of 400 μm and hardness of 1 200 Vickers hardness (VHN) was achieved. This research proves that a large pressure vessel can successfully be nitrided using the vessel interior as a process chamber. / Thesis (M.Ing. (Mechanical Engineering))--North-West University, Potchefstroom Campus, 2010.