Dissertations / Theses on the topic 'Pebble bed reactors – Temperature'

Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2010.<br>ENGLISH ABSTRACT: Regulators and designers of nuclear reactors regard knowledge of the pebble fuel temperature as important, due to the role that it plays in maintaining structural integrity and the production of neutrons. By using special fuel assemblies fitted with measuring equipment it is possible to measure the fuel temperature in stationary fuel reactors. This, however, is not possible in the pebble bed modular reactor due to its dynamic core. Designers of the pebble bed modular reactor have reserved special inspection channel borings inside the center reflector for fuel temperature measurement. By means of optical fibers and interferometry, the temperature can be measured inside such a channel. Currently the only way to control the fuel surface and core temperature is by measuring the gas inlet and outlet temperatures. This thesis attempts to determine the pebble temperature by measuring the temperature in a reflector channel. This is done by constructing an electrically heated pebble bed experimental setup simulating a cutout section of a pebble bed modular reactor core. An additional computational fluid dynamics simulation of the experimental setup was also performed. This thesis also attempts to determine if there is a measureable temperature peak that can indicate where a pebble was in contact with the reflector surface. This could then be used in future studies to determine the pebble fuel velocity as it moves down the reactor core. The computational fluid dynamics results were validated by experimental measurements. In the computational fluid dynamics model and experimental setup, it was found that there was indeed a measureable temperature difference on the temperature gradient along the reflector wall. The heat being conducted away from the pebble through the contact area can explain this. These differences were only observed when the channel was moved closer to the pebbles and it is therefore advised that some redesigning of the channel should be done if the in-core temperature is to be accurately interpreted by the designers at PBMR (Pty) Ltd.<br>AFRIKAANSE OPSOMMING: Reguleerders en ontwerpers van kern reaktore beskou die kennis van die korrel brandstof temperatuur as belangrik. Dit is weens die rol wat die brandstof temperatuur speel met die behoud van strukturele integriteit en die produksie van neutrone binne-in die reaktor. Met behulp van spesiale brandstof montasies toegerus met die meetings instrumentasie, is dit moontlik om die brandstof temperatuur in stilstaande brandstof reaktore te meet. Dit is egter nie moontlik in die korrel bed modulêre reaktor nie, as gevolg van sy dinamiese kern. Ontwerpers van die korrel bed modulêre reaktor het spesiale kanale in die binnekant van die middel reflektor vir brandstof temperatuur meeting gereseveer. Deur middel van optiese vesel en interferometrie, kan die temperatuur binne so 'n kanaal gemeet word. Tans is die enigste manier om die brandstof-oppervlak temperatuur te berekern, net moontlik deur gebruik te maak van die gemete gas inlaat-en uitlaat temperature van die reaktor. Hierdie tesis poog om vas te stel of die korrel brandstof temperatuur deur die meet van die oppervlak temperatuur in 'n reflektor-kanaal bepaal kan word. Dit word gedoen deur 'n elektriese verhitte korrel bed eksperimentele opstelling te bou wat 'n gedeelte van 'n korrel bed modulêre reaktor simuleer. 'n Bykomende numeriese simulasie van die eksperimentele opstelling was ook uitgevoer. Hierdie werk het ook probeer om vas te stel of daar 'n meetbare temperatuur piek op die temperatuur profiel aandui kan word waar 'n korrel in kontak is met die reflektor se oppervlak. Dit kan dan in toekomstige studies gebruik word om te bepaal wat die korrel brandstof spoed was soos dit in die reaktor beweeg. Die numerise simulasie uitslae was deur eksperimentele metings bevestig. In die numerise simulasie model en die eksperimentele opstelling, is daar gevind dat daar inderdaad 'n meetbare temperatuur verskil op die temperatuurgradiënt teen die reflektor oppervlak is. Dit kan verduidelik word as gevolg van die hitte wat weg van die korrel gelei word deur middel van die kontak area. Hierdie verskille was slegs waargeneem wanneer die kanaal nader aan die korrels geskuif is en dit word as n aanbeveling aan PBMR (Pty) Ltd gemaak om sommige herontwerpe aan die kanaal te doen indien die in-kerntemperatuur gemeet wil word en akkuraat geinterpreteer wil word.

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1999.<br>Includes bibliographical references (leaf 46).<br>The Modular Pebble Bed High Temperature Gas Reactor (MPBR) being designed by the Massachusetts Institute of Technology and the Idaho National Engineering and Environmental Laboratory operates with an online refueling system. This leads to an increased risk of proliferation because the fuel pebbles can be diverted while the plant is operating. In order to show that the MPBR dose not pose a proliferation risk the fuel content was determined for different burnups up to 94 MWD/kg. This data shows that the fuel is very poor nuclear weapon material. Safeguard systems were also designed in agreement with the International Atomic Energy Agency's standards to prevent diversion of significant quantities of fissile material.<br>by Jennifer Marie Anderson.<br>S.B.

The increasing demand for energy and the effect on climate change are some of the big drivers in support of the nuclear renaissance. A great amount of energy is spent on studies to determine the contribution of nuclear power to the future energy supply. Many countries are investing in generation III and IV reactors such as the Westinghouse AP1000 because of its passive cooling system, which makes it attractive for its safety. The pebble bed high temperature gas cooled reactors are designed to be intrinsically safe, which is one of the main drivers for developing these reactors. A pebble bed reactor is a high temperature reactor which is helium–cooled and graphitemoderated using spherical fuel elements that contain triple–coated isotropic fuel particles (TRISO). The success of its intrinsic safety lies in the design of the fuel elements that remain intact at very high temperatures. When temperatures significantly higher than 1600 °C are reached during accidents, the fuel elements with their inherent safety features may be challenged. A pebble bed reactor has an online fuelling concept, where fuel is circulated through the core. The fuel is loaded at the top of the core and through gravity, moves down to the bottom where it is unloaded to either be discarded or to be re–circulated. This is determined by the burnup measuring system. By circulating the fuel spheres more than once through the reactor a flattened axial power profile with lower power peaking and therefore lower maximum fuel temperatures can be achieved. This is an attractive approach to increase the core performance by lowering the important fuel operating parameters. However, the circulation has an economic impact, as it increases the design requirements on the burnup measuring system (faster measuring times and increased circulation). By adopting a multi–pass recycling scheme of the pebble fuel elements it is shown that the axial power peaking can be reduced The primary objective for this study is the investigation of the influences on the core design with regards to the number of fuel passes. The general behaviour of the two concepts, multi–pass refuelling and a once–through circulation, are to be evaluated with regards to flux and power and the maximum fuel temperature profiles. The relative effects of the HTR–Modul with its cylindrical core design and the PBMR 400 MW with its annular core design are also compared to verify the differences and trends as well as the influences of the control rods on core behaviour. This is important as it has a direct impact on the safety of the plant (that the fuel temperatures need to remain under 1600 °C in normal and accident conditions). The work is required at an early stage of reactor design since it influences design decisions needed on the fuel handling system design and defuel chute decay time, and has a direct impact on the fuel burnup–level qualification. The analysis showed that in most cases the increase in number of fuel passes not only flattens the power profile, but improves the overall results. The improvement in results decreases exponentially and from ten passes the advantage of having more passes becomes insignificant. The effect of the flattened power profile is more visible on the PBMR 400 MW than on the HTR–Modul. The 15–pass HTR–Modul design is at its limit with regards to the measuring time of a single burnup measuring system. However, by having less passes through the core, e.g. tenpasses, more time will be available for burnup measurement. The PBMR 400 MW has three defuel chutes allowing longer decay time which improves measurement accuracy, and, as a result could benefit from more than six passes without increasing the fuel handling system costs. The secondary objective of performing a sensitivity analysis on the control rod insertion positions and the effect of higher fuel enrichment has also been achieved. Control rod efficiency is improved when increasing the excess reactivity by means of control rod insertion. However, this is done at lower discharge burnup and shut down margins. Higher enrichment causes an increase in power peaking and more fuel–passes will be required to maintain the peaking and temperature margins than before.<br>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.<br>Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

Thesis (MScEng (Electrical and Electronic Engineering))--University of Stellenbosch, 2009.<br>The PBMR has called for research into the possibility of distributed in-core temperature measurement. In this thesis, several methods for distributed temperature measurement in high-pressure, -radiation and -temperature environments have been investigated by means of a literature study. The literature study has revealed FBG temperature sensors as the most feasible solution to the temperature measurement challenge. Various parameters affecting the propagation of light in optical fibres and consequently the FBG reflection profile was researched. The differential equations describing FBG structures were solved and implemented in Matlab in order to simulate WDM of a distributed FBG sensing system. Distributed sensing with apodized FBGs written in sapphire optical fibre show the most promise of becoming a solution to the measurement challenge. However, practical testing of sapphire FBGs exposed to the environment in the PBMR core is required. With this long-term goal in mind, a general test platform for FBG temperature sensors was assembled. A heater controller was built for a specialized fibre heating element capable of controlling the temperature of a single FBG up to 1600 C. Temperature measurement using wavelength division multiplexing of apodized FBGs written in silica optical fibre were demonstrated in the test platform with great success. The measured results corresponded very well with the theory. Finally, the implementation of FBGs in the PBMR is discussed and recommendations are made for future work

The current project was conducted in order to select an optimized open Thorium/Uranium fuel cycle for the Pebble Bed Modular Reactor (PBMR) concept in motivation for the 100MWth PBMR Power Plant. A sensitivity study on the heavy metal loading of the fuel sphere was performed to accomplish this task. The effect on various parameters was evaluated to determine the influence of varying the Heavy Metal (HM) from 6 gHM/sphere to 20 gHM/sphere and at different feed fuel enrichments from 6 to 18 wt%. The findings also gave insight into the fuel cycle performance and the Uranium ore savings that is envisaged. The 20 gHM/sphere HM loading performed the best amongst the other HM fuel loadings. It also showed the least Uranium ore requirement namely; 15 % less than that required by the Uranium cycle. Recommendations for further research based on the results obtained in the study, are made in the closing chapters.<br>Thesis (M.Sc. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The objective of this study was the development of a systems-CFD model of the PBMR reactor that used the minimum number of grid points to achieve grid independence. The number of grid points is reduced by increasing the accuracy of the discretisation scheme in the reactor model. Any reduction in the number of grid points leads to an increase in calculation speed, which is critical for systems simulation codes that are used for optimisation or transient simulations. While some previous reactor models had been developed for systems simulation codes, their discretisation schemes have not been optimised to use the minimum number of grid points and some heat transfer phenomena were neglected without knowing the effect. Therefore, there was a need to optimise discretisation schemes as well as investigate the effect of including certain heat transfer mechanisms. Modelling methods for several phenomena were developed and implemented in a reactor model in the Flownex systems simulation code, which is used to simulate the PBMR. Subjects of investigation included pebble bed convection discretisation, fuel sphere discretisation, the effect of the radiation heat transfer modelling approach as well as conjugate conduction and radiation across the helium riser channels in the reactor side reflector. After testing the phenomenological models in isolation, the comprehensive reactor model was tested by simulating the SANA experiment and HTR-10 reactor experiments published by the IAEA. Several sensitivity studies were performed to assess the effect of physical as well as numerical parameters. Two reactor discretisation schemes were also evaluated, namely the control-volume based scheme and the element-based scheme. The control-volume based scheme was found to provide a simpler and more intuitive framework for implementing mathematical models, but not to increase accuracy directly. The most significant finding was that the newly developed second-order accurate convection heat transfer scheme gives the greatest improvement in calculation speed by requiring the least number of pebble bed increments. The other important finding was that the methods currently used in many reactor simulation codes for fuel sphere discretisation and radiation heat transfer approximation are appropriate and give adequate accuracy.<br>Thesis (Ph.D. (Mechanical Engineering))--North-West University, Potchefstroom Campus, 2007

Thesis (MScEng)--Stellenbosch University, 2013.<br>ENGLISH ABSTRACT: Production of graphite dust inside the Pebble Bed Modular Reactor (PBMR) influences the reactor operation negatively. Graphite is used as a moderator in the reactor core and the formation and transportation of graphite dust away from the reactor core decreases the amount of moderator which in turn has a negative impact on the reactor operation. High levels of radioactive dust may also contaminate reactor components which may pose a health risk to maintenance personnel. In this study a pressure vessel was designed and used to measure the wear of a graphite pebble in helium at elevated temperatures. By means of a multi-linear regression analysis a proper mathematical function was established in order to relate graphite wear to certain tribological parameters. These parameters were identified through a literature study. Discrete Element Modelling (DEM) was used to simulate the gravitational flow of graphite pebbles through the reactor core. The experimentally determined mathematical function was incorporated into the DEM simulation to estimate the annual mass of graphite dust to be produced by the PBMR pebble bed as a result of pebble-pebble interaction and pebble-wall interaction during refuelling.<br>AFRIKAANSE OPSOMMING: Die vorming van grafiet stof binne die korrelbed-modulêre reaktor (PBMR) beïnvloed die werking daarvan negatief. Grafiet word gebruik as 'n moderator in die reaktor kern en die vorming en vervoer van grafietstof weg van die reaktor kern lei tot 'n afname in die hoeveelheid moderator en dit het 'n negatiewe impak op die werking van die reaktor. Hoë vlakke van radioaktiewe grafietstof kontamineer ook reaktorkomponente wat 'n gesondheidsrisiko vir onderhoudspersoneel inhou. In hierdie studie was 'n drukvat ontwerp en gebruik om die slytasie van 'n grafietkorrel in helium by verhoogde temperature te meet. 'n Multi-lineêre regressie analise is dan gebruik om 'n wiskundige funksie daar te stel wat die verband tussen grafietslytasie en die eksperimentele parameters vas stel. Hierdie parameters was met behulp van 'n literatuurstudie geïdentifiseer. Diskrete Element Modellering (DEM) was gebruik om die gravitasionele vloei van grafietkorrels in die reaktor te modelleer. Die eksperimenteel bepaalde wiskundige funksie word in die DEM simulasie ge-inkorporeer om 'n skatting te maak van die jaarlikse massa grafietstof wat gevorm sal word in die PBMR korrelbed as 'n gevolg van korrel-korrel interaksie en korrel-wand interaksie gedurende hersirkulasie.

Thesis (MScEng)--Stellenbosch University, 2012.<br>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.<br>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.

Thesis (Nucl.E. and S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, February 2002.<br>Includes bibliographical references (v. 1, leaves 152-171).<br>The applicability of the Monte Carlo code MCNP4B to the neutronic modeling of pebblebed reactors was investigated. A modeling methodology was developed based on an analysis of critical experiments carried out at the HTR-PROTIEUS and ASTRA facilities, and the critical loading of the HTR-10 reactor. A body-centred cubic lattice of spheres with a specified packing fraction approximates the pebble bed, and exclusion zones offset the contribution of partial spheres generated by the geometry routines in MCNP4B at the core boundaries. The coated fuel particles are modeled in detail and are distributed over the fuelled region of the fuel sphere using a simple cubic lattice. This method predicted the critical core loading accurately in all cases. The calculation of control-rod worths in the more decoupled tall annular ASTRA core gave results within 10% compared to the reported experiments. An approximate method was also developed for the MCNP4B modeling of pebble-bed reactors with burnup. The nuclide densities of homogenized layers in the VSOP94 reactor model are transferred to the corresponding MCNP4B model with the lattice of spheres represented explicitly. The method was demonstrated on the PBMR equilibrium core, and used for a parallel study of burnup k- and isotopics on a single pebble. Finally, a study was carried out of the proliferation potential of a modular pebble-bed reactor for both normal and off-normal operation. VSOP94 analysis showed that spent fuel from pebble-bed reactors is proliferation resistant at high discharge burnup, because of its unfavourable plutonium isotopic composition and the need to divert -157,000 pebbles to accumulate sufficient 239Pu for a nuclear weapon. The isotopics of first-pass fuel pebbles are more favourable, but even more pebbles (-258,000) would be needed. However, a supercell MOCUP model was used to demonstrate that -20,000 pebbles would be needed if loaded with depleted uranium. But the associated reactivity loss would necessitate a compensatory increase in core height of approximately 50 cm. Such a change in core loading, as well as the properties of the special pebbles, would be noticed in a safeguarded facility.<br>by Julian Robert Lebenhaft.<br>Nucl.E.and S.M.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.<br>Cataloged from PDF version of thesis.<br>Includes bibliographical references (p. 73-74).<br>The primary objective of this thesis was to validate the sustained counter air diffusion (SCAD) method at preventing natural circulation onset in diffusion dominated air ingress accidents. The analysis presented in this thesis starts with a vertically oriented rupture of a coaxial pipe. Air enters into the reactor cavity at a rate dictated by diffusion, until the buoyancy force is strong enough to initiate natural convective flow through the reactor. The SCAD method, developed by Yan et al. reduces the buoyancy force in a high temperature gas reactor (HTGR), during the lengthy diffusion phase, by injecting minute amounts of helium into the top of the reactor to set up a counter helium-air diffusion circuit. By delaying the onset of natural circulation, air enters the reactor only at diffusion transport rates, instead of much higher natural convection transport rates. Thus, the air ingress rate is reduced by several orders of magnitude. Without the continuous convective driven supply of "fresh" air the threat of oxidizing graphite components is significantly reduced. To validate SCAD a small scale simulated Pebble Bed Reactor (PBR) was constructed and a series of air ingress experiments with and without helium injection were conducted. In addition, Computational Fluid Dynamic (CFD) simulations were performed using FLUENT @ to model the experiment and gain further insight into the behavior of the flow field leading up to the onset of natural circulation. In order to have the CFD predicted natural circulation onset time better match the experimentally determined onset time, the initial helium fraction in the numerical model had to be reduced by 15%. This reduction is within the uncertainty of the experimental set-up. This change helped display an important feature of the behavior of air ingress accidents. With the initial helium fraction in the simulated reactor at 100% the first half of the transient is a very slow completely diffusion dominated transport phase. The second half of the transient had an air transport rate that had an increasing natural convective transport contribution leading up to the onset of natural circulation and complete natural convective transport. Reducing the initial helium fraction by only 15% caused that initial very slow, pure diffusion transport phase to be bypassed and the natural circulation onset time was dictated by the combined effects of free convection and diffusion transport, not simply diffusion. A full scale PBR experiencing a similar accident will have the core entirely filled with helium. Thus, for a vertically oriented double ended guillotine (DEG) large-break loss of coolant accident (LB-LOCA) the subsequent air ingress rate will be dictated by the slow diffusion of air into the reactor cavity, for most of the transient. For the helium injection tests, even at the at the lowest tested injection rate, both the experiment and the CFD simulation showed that natural circulation was prevented over a time period twice as long as the time to onset. The tests showed that without helium injection, natural circulation started after about 117 minutes on average. With helium injection, natural circulation did not start after 240 minutes when the experiment was terminated. Additional injection tests were run where after 240 minutes the helium injection was terminated, but data continued to be taken. In these tests natural circulation was initiated in approximately 120 minutes after termination of helium injection confirming the helium injection flow was preventing natural circulation from starting. The lowest tested helium injection rate corresponded to 0.01% of the test assembly's total volume per minute, demonstrating how small of a flow rate is needed for the SCAD method to work. Minimal helium injection is not intended to be an emergency core cooling system but rather a system to prevent or delay natural circulation which would result in a large amount of air ingress. The system response was formulated non-dimensionally to quantify the impact SCAD has on the driving parameters that impact the onset of natural circulation, namely the buoyancy force, mass flow rate, and density ratio between the hot and cold leg. The results showed that SCAD suppresses the buoyancy force and forces a mass flow (transport) rate that causes any changes in the hot leg density to be counter-acted by density changes in the cold leg. The transport rate that is established is orders of magnitude less than the natural circulation transport rate. Using the driving nondimensional parameters, a methodology was also developed in order to formulate a correlation to estimate the minimum injection rate (MIR) of helium to prevent the onset of natural circulation. In order to properly derive a correlation for the MIR, further experiments and/or simulations are required over different geometrical configurations. The non-dimensional analysis showed that Yan's MIR estimate was conservative for the experimental configuration, and would be conservative for a full scale PBR. Therefore, Yan's MIR calculation was used to provide an order of magnitude estimate for the helium injection rate in a full scale PBR. The resulting MIR of helium for a full scale PBR was 5.36 g/hr, which corresponds to storing only 11.6 kg of helium on-site to prevent the onset of natural circulation for three full months. The experiment and CFD simulations were performed using an inverted U-tube which simulates a vertically oriented pipe configuration. If the pipe break occurs in a horizontal configuration, the air ingress phenomena could be substantially different depending on the break size and orientation. Thus, this thesis concludes that the method is capable of preventing natural circulation onset as long as air ingress occurs at transport rates comparable to diffusion after the break occurs.<br>by Joseph Paul Yurko.<br>S.M.

CIVINS<br>The Modular Pebble Bed Reactor is a small high temperature helium cooled reactor that is being considered for both electric power and hydrogen production. Pebble bed reactors are being developed in South Africa, China and the US. To make smaller 120 Mwe reactors economically competitive with larger 1500 Mwe traditional light water reactors changes in the way these plants are built are needed. Economies of production need to be sufficiently large to compete with economies of scale. MIT (Berte) has been working on developing a modular design and construction strategy for several years. This thesis builds on that work by demonstrating the technical feasibility of implementing the modularity approaches previously developed.

Thesis(Mtech (Industrial Engineering)--Cape Peninsula University of Technology, 2010<br>In the nuclear industry, Quality Management Systems are extremely important, especially if one wishes to improve public acceptance of radioactive solutions. There is normally minimum communication between the public and scientists, especially in nuclear science. People are not comfortable with nuclear technology, based on the past history of the Chernobyl catastrophe. Consequently, it is difficult to discuss important and sensitive issues like disposing of nuclear waste. Quality Management Systems can improve public confidence and communication. Integrated Management Systems in the project planning stage of the project can be a proactive step towards preventing unnecessary delays and costs. There is a perception that quality is implemented or executed at the implementation stage of the Project Life cycle. Most people believe that a Quality Management System is quality control only and forget the aspect of Quality assurance. The project managers are more concerned with finishing the project and saving costs. Quality holds together the three pillars of project management, which are schedule, costs and scope. There are a plethora of things that can go wrong if the Quality Management System is not implemented on time, like scope changes that are not captured, monitored and controlled. This can lead to scope creep, unnecessary costs and schedule overruns. If there is no cost control, the project can also overrun its budget and consequently be stopped. PBMR is the only company that is active in new nuclear projects in South Africa, except Koeberg, which was commissioned about thirty years ago.

Through the years different models have been proposed to calculate the total effective thermal conductivity in packed beds. The purpose amongst others of these models is to calculate the temperature distribution and heat flux in high temperature pebble bed reactors. Recently a new model has been developed at the North–West University in South Africa and is called the Multi–Sphere Unit Cell (MSUC) model. The unique contribution of this model is that it manages to also predict the effective thermal conductivity in the near wall region by taking into account the local variation in the porosity. Within the MSUC model the thermal radiation has been separated into two components. The first component is the thermal radiation exchange between spheres in contact with one another, which for the purpose of this study is called the short range radiation. The second, which is defined as the longrange radiation, is the thermal radiation between spheres further than one sphere diameter apart and therefore not in contact with each other. Currently a few shortcomings exist in the modelling of the long–range radiation heat transfer in the MSUC model. It was the purpose of this study to address these shortcomings. Recently, work has been done by Pitso (2011) where Computational Fluid Dynamics (CFD) was used to characterise the long–range radiation in a packed bed. From this work the Spherical Unit Nodalisation (SUN) model has been developed. This study introduces a method where the SUN model has been modified in order to model the long–range radiation heat transfer in an annular reactor packed with uniform spheres. The proposed solution has been named the Cylindrical Spherical Unit Nodalisation (CSUN, pronounced see–sun) model. For validation of the CSUN model, a computer program was written to simulate the bulk region of the High Temperature Test Unit (HTTU). The simulated results were compared with the measured temperatures and the associated heat flux of the HTTU experiments. The simulated results from the CSUN model correlated well with these experimental values. Other thermal radiation models were also used for comparison. When compared with the other radiation models, the CSUN model was shown to predict results with comparable accuracy. Further research is however required by comparing the new model to experimental values at high temperatures. Once the model has been validated at high temperatures, it can be expanded to near wall regions where the packing is different from that in the bulk region.<br>Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.

Thesis (MBA)--Stellenbosch University, 2002.<br>ENGLISH ABSTRACT: The Pebble Bed Modular Reactor is a new and inherently safe concept for a nuclear power generation plant. In order to obtain the necessary licenses to build and operate this reactor, numerous design and safety analyses need to be performed. The results of these analyses must be supported with substantial proof to provide the nuclear authorities with a sufficient level of confidence in these results to be able to supply the required licences. Beside the obvious need for a sufficient level of confidence in the safety analyses, the analyses concerned with investment protection also need to be reliable from the investors’ point of view. The process to be followed to provide confidence in these analyses is the verification and validation process. It is aimed at presenting reliable material against which to compare the results from the simulations. This material for comparison will consist of a combination of results from experimental data, extracts from actual plant data, analytical solutions and independently developed solutions for the simulation of the event to be analysed. Besides comparison with these alternative sources of information, confidence in the results will also be built by providing validated statements on the accuracy of the results and the boundary conditions with which the simulations need to comply. Numerous standards exist that address the verification and validation of computer software, for instance by organisations such as the American Society of Mechanical Engineers (ASME) and the Institute of Electrical and Electronics Engineers (IEEE). The focal points of the verification and validation of the design and safety analyses performed on typical PBMR modes and states, and the requirements imposed by both the local and overseas nuclear regulators, are not entirely enveloped by these standards. For this reason, PBMR developed a systematic and disciplined approach for the preparation of the Verification and Validation Plan, aimed at capturing the essence of the analyses. This approach aims to make a definite division between software development and the development of technical analyses, while still using similar processes for the verification and validation. The reasoning behind this is that technical analyses are performed by engineers and scientists who should only be responsible for the verification and validation of the models and data they use, but not for the software they are dependent on. Software engineers should be concerned with the delivery of qualified software to be used in the technical analyses. The PBMR verification and validation process is applicable to both hand calculations and computer-aided analyses, addressing specific requirements in clearly defined stages of the software and Technical Analysis life cycle. The verification and validation effort of the Technical Analysis activity is divided into the verification and validation of models and data, the review of calculational tasks, and the verification and validation of software, with the applicable information to be validated, captured in registers or databases. The resulting processes are as simple as possible, concise and practical. Effective use of resources is ensured and internationally accepted standards have been incorporated, aiding in faith in the process by all stakeholders, including investors, nuclear regulators and the public.<br>AFRIKAASE OPSOMMING: Die Modulêre Korrelbedreaktor is ’n nuwe konsep vir ’n kernkragsentrale wat inherent veilig is. Dit word deur PBMR (Edms.) Bpk. ontwikkel. Om die nodige vergunnings om so ’n reaktor te kan bou en bedryf, te bekom, moet ’n aansienlike hoeveelheid ontwerp- en veiligheidsondersoeke gedoen word. Die resultate wat hierdie ondersoeke oplewer, moet deur onweerlegbare bewyse ondersteun word om vir die owerhede ’n voldoende vlak van vertroue in die resultate te gee, sodat hulle die nodigde vergunnings kan maak. Benewens die ooglopende noodsaak om ’n voldoende vlak van vertroue in die resultate van die veiligheidsondersoeke te hê, moet die ondersoeke wat met die beskerming van die beleggers se beleggings gepaard gaan, net so betroubaar wees. Die proses wat gevolg word om vertroue in die resultate van die ondersoeke op te bou, is die proses van verifikasie en validasie. Dié proses is daarop gerig om betroubare vergelykingsmateriaal vir simulasies voor te lê. Hierdie vergelykingsmateriaal vir die gebeurtenis wat ondersoek word, sal bestaan uit enige kombinasie van inligting wat in toetsopstellings bekom is, wat in bestaande installasies gemeet is, wat analities bereken is; asook dit wat deur ’n derde party onafhanklik van die oorspronklike ontwikkelaars bekom is. Vertroue in die resultate van die ondersoeke sal, behalwe deur vergelyking met hierdie alternatiewe bronne van inligting, ook opgebou word deur die resultate te voorsien van ’n gevalideerde verklaring wat die akkuraatheid van die resultate aantoon en wat die grensvoorwaardes waaraan die simulasies ook moet voldoen, opsom. Daar bestaan ’n aansienlike hoeveelheid internasionaal aanvaarde standaarde wat die verifikasie en validasie van rekenaarsagteware aanspreek. Die standaarde kom van instansies soos die Amerikaanse Vereniging vir Meganiese Ingenieurs (ASME) en die Instituut vir Elektriese en Elektroniese Ingenieurs (IEEE) – ook van Amerika. Die aandag wat deur die Suid-Afrikaanse en oorsese kernkragreguleerders vereis word vir die toestande wat spesifiek geld vir korrelbedreaktors, word egter nie geheel en al deur daardie standaarde aangespreek nie. Daarom het die PBMR maatskappy ’n stelselmatige benadering ontwikkel om verifikasie- en validasieplanne voor te berei wat die essensie van die ondersoeke kan ondervang. Hierdie benadering is daarop gemik om ’n duidelike onderskeid te maak tussen die ontwikkeling van sagteware en die ontwikkeling van tegniese ondersoeke, terwyl steeds gelyksoortige prosesse in die verifikasie en validasie gebruik sal word. Die rede hiervoor is dat tegniese ondersoeke uitgevoer word deur ingenieurs en wetenskaplikes wat net vir verifikasie en validasie van hulle eie modelle en die gegewens verantwoordelik gehou kan word, maar nie vir die verifikasie en validasie van die sagteware wat hulle gebruik nie. Ingenieurs wat spesialiseer in sagteware-ontwikkeling behoort verantwoordelik te wees vir die daarstelling van sagteware wat deur die reguleerders gekwalifiseer kan word, sodat dit in tegniese ondersoeke op veiligheidsgebied gebruik kan word. Die verifikasie- en validasieproses van die PBMR is sowel vir handberekeninge as vir rekenaarondersteunde-ondersoek geskik. Hierdie proses spreek spesifieke vereistes in onderskeie stadiums gedurende die lewenssiklusse van die ontwikkeling van sagteware en van tegniese ondersoeke aan. Die verifikasie- en validasiewerk vir tegniese ondersoeksaktiwiteite is verdeel in die verifikasie en validasie van modelle en gegewens, die nasien van berekeninge en die verifikasie en validasie van sagteware, waarby die betrokke inligting wat gevalideer moet word, versamel word in registers of databasisse. Die prosesse wat hieruit voortgevloei het, is so eenvoudig as moontlik, beknop en prakties gehou. Hierdeur is ’n effektiewe benutting van bronne verseker. Internasionaal aanvaarde standaarde is gebruik wat die vertroue in die proses deur alle betrokkenes, insluitende beleggers, die owerhede en die publiek, sal bevorder.

Thesis (MScIng)--University of Stellenbosch, 2008.<br>ENGLISH ABSTRACT: The focus of this project was the application of a passive device in the form of a loop thermosyphon as a reactor cavity cooling system (RCCS) for a Pebble Bed Modular Reactor. An extensive literature review showed that loop thermosyphons have been widely researched, both theoretically and experimentally. In the review attention has specifically been given to matters such as safety, instability, control and mathematical modelling. One of the objectives of the project was to build one of the axially symmetric sections of Dobson’s (2006) proposed full scale RCCS using a scaled down version consisting of a single loop heated by a section of the reactor pressure vessel and cooled by a tank of water. The second objective was to derive a theoretical model that could be used in a computer code to simulate the experiment. The theory and experiment would then be compared in order to verify the code. The mathematical model created used the following three major assumptions: quasistatic flow, incompressible liquid and vapour and one dimensionality. The conservation equations in the form of a set of difference equations with the appropriate closure equations were then solved explicitly. It was found that the theoretical results were heavily influenced by the surface optical properties as well as the heat transfer coefficients. The emissivity influenced the transition point from single to two-phase flow as well as the condenser outlet temperature. The single phase heat transfer coefficients influenced the condenser outlet temperature significantly while it was found that for two phase flow the combination of the available boiling and condensation heat transfer coefficients had only minor effects on the end results. A stainless steel and aluminium thermosyphon loop was built using water as the working fluid. A stainless steel heater plate provided the heat input while a 200 L water tank was the heat sink. Temperature and flow rate measurements were recorded as a function of time with various heating/cooling transients from start-up to steady state for three operating modes. The three operating modes were single phase, two-phase and heat pipe mode. It was found that the theoretical temperatures correspond reasonably well with the experimental temperatures. The time predicted by the theoretical model to reach the operating temperature was however somewhat longer than for the experimental. This is to be expected when considering that there is some uncertainty pertaining to the heat transfer coefficients as well as surface emissive properties. The correspondence of the theoretical and experimental fin temperatures was poor due to significant thermal stratification of the air separating the heater plate and fins. Several shortcomings in the theoretical model as well as the experimental setup were identified and discussed. The conclusion was reached that this exploratory study showed that the loop thermosyphon is a viable option for the RCCS and that the mathematical model is a viable theoretical simulation tool. Several recommendations were made for further study to address and overcome the shortcomings identified in the theoretical and experimental models in order to prove this conclusion. Amongst these is the determination of better material surface properties and heat transfer coefficients and improved mass flow rate measurement. Investigating scaling issues, natural convection outside the loop and updating of the computer program is also recommended.<br>AFRIKAANSE OPSOMMING: Die fokus van hierdie projek was die toepassing van passiewe apparatuur, in die vorm van ‘n geslote lus termoheuwel, as ‘n reaktor kamer verkoellings stelsel vir die korrel bed modulêre reaktor. Die literatuur studie wys dat hierdie tegnologie reeds breedvoerig ondersoek is teoreties sowel as eksperimenteel. In die literatuur oorsig word aandag spesifiek gegee aan veiligheid, onstabiliteit, beheer en modelleering. Een van die doelwitte van die projek was om ‘n klein skaalse model te bou van een van die aksiaal simmetriese seksies van Dobson (2006) se voorgestelde volskaalse reaktor kamer verkoellings stelsel. Die model bestaan uit n enkele lus verhit deur ‘n seksie van die reaktor drukvat en verkoel deur ‘n tenk vol water. Die tweede doelwit was die afleiding van ‘n teoretiese model wat in ‘n rekenaar program gebruik kan word om die eksperiment te simuleer. Die teoretiese en eksperimentele data kan dan vergelyk word om die geldigheid van die program te toets. Die volgende aanames is gemaak tydens die afleiding van die wiskundige model: kwasi-statiese vloei, onsamedrukbare vloeistof en gas en een dimensionalitiet. Die behouds wette is in die vorm van ‘n stel differensie vergelykings met die toepasbare sluitings vergelykings eksplisiet opgelos. Dit is bevind dat die teoretiese resultate swaar beinvloed is deur die materiaal oppervlak eienskappe sowel as die warmteoordrag koëffisiënte. Die emisiviteit beinvloed die oorgangs punt van enkel na twee fase vloei sowel as die kondenser uitlaat temperatuur. Die enkel fase warmteoordrag koëffisiënt het n beduidende invloed op die kondenser uitlaat temperatuur terwyl dit voorkom asof die spesifieke kombinasie van die koking en kondensasie warmteoordrag koëffisiënte minimale invloed op die resultate het in die twee fase gebied. Vlekvrye staal en aluminium is gebruik om die lus te bou met water as die verkoelings middel. Warmte is toegevoeg tot die stelsel deur ’n vlekvrye staal verhittings plaat terwyl ‘n 200 L water tenk die warmte onttrek het. Temperatuur en massa vloei tempo is aangeteken as ‘n funksie van tyd vir verskeie verhitting/verkoellings oorgangs gedragte vanaf begin tot bestendige toestand vir drie bedryfs modusse. Die drie bedryfs modusse was enkel fase, twee fase en hitte pyp modus. Dit is bevind dat die teoretiese temperature redelik goed ooreengekom het met die eksperimentele waardes. Die tyd wat dit neem om by die bedryfs temperatuur te kom soos voorspel deur die teorie is egter langer as wat in die eksperiment gevind is. Dit is te verstane wanneer die onsekerheid in die warmteoordrag koëffisiënte en materiaal oppervlak eienskappe in ag geneem word. Die fin temperature het ‘n swakker ooreenkoms getoon as gevolg van beduidende termiese stratifikasie van die lug tussen die fin en verhittings plaat. Verskeie tekortkominge in die teoretiese model en eksperimentele opstelling is geïdentifiseer en bespreek. Die gevolgtrekking is gemaak dat die ondersoek bewys dat geslote lus termoheuwels ‘n lewensvatbare opsie is vir ‘n reaktor kamer verkoellings stelsel en dat die wiskundige model lewensvatbaar is vir teoretiese simulasie. Verskeie aanbevelings word egter gemaak om die tekortkominge in die teoretiese en eksperimentele modelle aan te spreek om so doende die gevolgtrekking te staaf. Dit word aanbeveel dat beter waardes vir die materiaal oppervlak eienskappe en warmteoordrag koëffisiënte gevind word en verbeterde massa vloei meetings gedoen word. Dit word verder aanbeveel om skaleering asook natuurlike konveksie buite die lus te ondersoek en om die rekenaar program by te werk.

La dénitrification autotrophe à l’aide de composés réduit de soufre est une approche intéressante pour le traitement biologique des contaminations azotées et des effluents pauvre en matière organique. La dénitrification autotrophe utilise des composés inorganiques comme sources d'énergie et de carbone. L'absence de matière organique élimine le besoin de post-traitements pour éliminer l'excès de carbone organique et limite la formation sous-produits d’oxydation dans le cadre de la production d’eau potable. Les eaux usées provenant des industries métallurgiques et minières ont généralement un faible pH, des températures basses et des concentrations élevées en métaux lourds. L'élimination biologique de l'azote est un défi parce que les bactéries dénitrifiantes prospèrent habituellement à pH neutre et à températures ambiantes (20-30 °C).Le but de cette thèse était de développer un procédé robuste de dénitrification à base de soufre dans des bioréacteurs à pH acide, températures psychrophiles (&lt; 20 °C) et concentrations élevées en nickel. Le procédé a été optimisé au préalable avec des essais biologiques étudiant l'influence de la source de soufre (S2O32-, S0 biogénique et le synthétisé chimiquement), de la taille des particules de S0 (poudre et lentilles), de la culture dénitrifiante (cultures pures et mixtes de Thiobacillus) et de la température (6-30 °C) sur la cinétique de la dénitrification. L'utilisation de S2O32- et d’une culture pure de T. denitrificans ont permis d’atteindre des rendements de dénitrification les plus élevés. Le soufre élémentaire biogénique a été testé pour la première fois comme donneur d'électrons pour la dénitrification, montrant des taux de dénitrification 1.7 fois plus élevés que ceux obtenue avec de la poudre de S0 synthétisé chimiquement. Les taux de la dénitrification avec le S2O32- augmentent exponentiellement avec la température et les calculs avec l'équation d'Arrhenius donnent une énergie d'activation apparente Ea de 76.6 kJ/mol.Deux réacteurs à lit fluidisé (FBR) ont été utilisés pour étudier la dénitrification avec S2O32- à différents pH (5.25-7.00) et températures décroissantes (3-20 °C). Des rendements de dénitrification &gt; 99% ont été observés pour eaux usés présentant des pH compris entre 5.75 et 5.30. L'addition d'une unité de carbonatation fournissant au biofilm du CO2 comme source de carbone supplémentaire, permettant une dénitrification complète à un pH de 4.75. Dans le même FBR, des taux de charge d'azote élevés (jusqu'à 3,3 kg N-NO3-/m3 d) avec le thiosulfate ont été maintenu à des températures aussi basses que 3 °C. L'impact de deux composés du Nickel (NiEDTA2- et NiCl2) sur la dénitrification à base de soufre a été étudiée dans deux FBR en parallèle à 20 (± 2) °C et des concentrations de nickel variant dans la gamme de 5-200 mg Ni/L. Dans des bioessais discontinues, 25-100 mg Ni/L de NiCl2 ont inhibée l'élimination de NO3- de 7-16%, alors qu'aucune inhibition n'a été observée avec NiEDTA2-. L'EDTA non complexée a inhibée la dénitrification à des concentrations supérieures à 100 mg/L. Les deux composés de Ni ont montré aucun effet négatif sur la dénitrification en FBR aux concentrations testées. Le bilan massique du nickel, la caractérisation de la phase solide et la modélisation thermodynamique ont révélé que des précipités de nickel ont été principalement éliminés avec l'effluent. Les phosphates, sulfures et oxydes de nickel ont été déterminés comme les principaux précipités de nickel et étaient principalement amorphe.Les FBRs se sont révélés être bioprocédés robustes pour l'élimination de l'azote à pH acide, pour des températures psychrophiles et des concentrations élevées de nickel. Les résultats de cette étude sont d'un grand intérêt pour le traitement des eaux souterraines et minières contaminés par les nitrates dans les régions froides du monde et également pour les eaux usées industrielles acides et chargées en métaux lourds<br>Autotrophic denitrification driven by reduced sulfur compounds is a promising and cost-effective biological nitrogen removal process, recommended for the treatment of organic-deficient waters, e.g. groundwater and several industrial wastewaters. Autotrophic denitrifiers utilize inorganic compounds as sources of energy and carbon. The lack of organics eliminates the need of post-treatments to remove excess organic carbon and limits the formation of harmful organic byproducts (e.g. trihalomethanes, THM), resulting in a clean and safe treatment also for drinking water. Wastewaters from mining and metal-finishing industry commonly feature low pH and temperatures as well as high heavy metal concentrations. Nitrogen removal from these waters is a technical challenge, since denitrifying bacteria usually thrives at circumneutral pH and ambient temperatures (20-30°C).The aim of this study was to develop a robust and efficient sulfur-based denitrification bioreactor process able to tolerate acidic pH, psychrophilic temperatures (&lt; 20°C) and high nickel concentrations. The process was preliminary optimized in batch bioassays investigating the influence of sulfur source, i.e. thiosulfate (S2O32-) and biogenic and chemically synthesized elemental sulfur (S0), S0 particle size (powder and lentils), denitrifying culture (pure and mixed cultures of Thiobacillus) and temperature (6-30°C) on denitrification kinetics. The use of S2O32- and a pure culture of Thiobacillus denitrificans resulted in the highest denitrification rates. Biogenic S0 was tested for the first time as electron donor for autotrophic denitrification, showing 1.7-fold faster NO3- removal than that achieved with chemically synthesized S0 powder. The rates of thiosulfate-driven denitrification exponentially increased with temperature, being modeled according to the Arrhenius equation with an apparent activation energy Ea of 76.6 kJ/mol and a temperature coefficient Q10 of 3.0.Fluidized-bed reactors (FBRs) were used to investigate continuous thiosulfate-driven denitrification under decreasing feed pH (5.25-7.00) and temperatures (3-20°C). Denitrification efficiencies &gt; 99% were observed at feed and effluent pH as low as 5.75 and 5.30, respectively. At lower feed pH values, the denitrification activity rapidly decreased due to an inorganic carbon deficiency. The addition of a carbonation unit providing CO2 as supplemental carbon source to the FBR biofilm allowed complete denitrification even at a pH of 4.75. In the same FBR, high-rate (up to 3.3 kg N-NO3-/m3 d) thiosulfate-driven denitrification was maintained at temperatures as low as 3°C. The impact of two Ni compounds, i.e. NiEDTA2- and NiCl2, on sulfur-based denitrification was investigated in a parallel FBR at 20 (±2)°C and feed Ni concentrations in the range of 5-200 mg Ni/L. Preliminary batch bioassays were carried out to assess Ni and free EDTA toxicity on sulfur-based denitrification. In batch bioassays, 25-100 mg Ni/L of NiCl2 inhibited NO3- removal by 7-16%, whereas no inhibition was observed with NiEDTA2-. Free EDTA inhibited sulfur-based denitrification at concentrations exceeding 100 mg/L. Both Ni compounds showed no detrimental effects on sulfur-based denitrification in FBR at the tested concentrations. Nickel mass balance, solid-phase characterization and thermodynamic modeling revealed that nickel precipitates were mostly washed out with the effluent, due to the slow Ni precipitation kinetics and high upflow velocities in the FBR. Nickel phosphate, sulfide and oxide were indicated as the main nickel precipitates and were mostly amorphous.FBRs were shown to be powerful and robust biofilm systems for nitrogen removal under acidic pH, psychrophilic temperatures and high nickel concentrations. The results of this study are of great interest for the treatment of NO3- contaminated ground and mining waters in cold regions (e.g. Canadian and Scandinavian regions) as well as acidic and heavy-metal-laden wastewaters

Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2011.<br>ENGLISH ABSTRACT: The feasibility of a closed loop thermosyphon for the Reactor Cavity Cooling System of the Pebble Bed Modular Reactor has been the subject of many research projects. Difficulties identified by previous studies include the hypothetical inaccuracies of heat transfer coefficient correlations available in literature. The aim of the research presented here is to develop inside-pipe heat transfer correlations that are specific to the current design of the RCCS. In order to achieve this, a literature review is performed which identifies reactors which employ closed loop thermosyphons and natural circulation. The literature review also explains the general one-dimensional two-fluid conservation equations that form the basis for numerical modelling of natural circulation loops. The literature review lastly discusses available heat transfer coefficient correlations with the aim of identifying over which ranges and under which circumstances these correlations are considered accurate. The review includes correlations commonly used in natural circulation modelling in the nuclear industry in aims of identifying correlations applicable to the modelling of the proposed RCCS. One of the objectives of this project is to design and build a one-third-height-scale model of the RCCS. Shortcomings of previous experimental models were assessed and, as far as possible, compensated for in the design of the model. Copper piping is used, eliminating material and surface property uncertainties. Several sight glasses are incorporated in the model, allowing for the visual identification of two-phase flow regimes. An orifice plate is used allowing for bidirectional flow measurement. The orifice plate, thermocouples and pipe-in-pipe heat exchangers are calibrated in-situ to minimize experimental error and aid repeatability. Twelve experiments are performed with data logging occurring every ten seconds. The results presented here are limited to selected single and two-phase flow operating mode results. Error analyses and repeatability of experimental measurements for single and two-phase operating modes as well as cooling water mass flow rates are performed, to show repeatability of experimental results. These results are used to mathematically determine the experimental inside-pipe heat transfer coefficients for both the evaporator and condenser sections. Trends in the heat transfer coefficient profiles are identified and the general behaviour of the profiles is thoroughly explained. The RCCS is modelled as a one-dimensional system. Correlations for the friction factor, heat transfer coefficient, void fraction and two-phase frictional multiplier are identified. The theoretical heat transfer coefficients are calculated using the mathematical model and correlations identified in the literature review. Fluid parameters are evaluated using experimentally determined temperatures and mass flow rates. The resulting heat transfer coefficient profiles are compared to experimentally determined profiles, to confirm the hypothesis that existing correlations do not accurately predict the inside-pipe heat transfer coefficients. The experimentally determined coefficients are correlated to 99% confidence intervals. These generated correlations, along with identified and established twophase heat transfer coefficient correlations, are used in a mathematical model to generate theoretical coefficient profiles. These are compared to the experimentally determined coefficients to show prediction accuracy.<br>AFRIKAANSE OPSOMMING: Die haalbaarheid van ‘n natuurlike sirkulasie geslote lus vir die Reaktor Holte Verkoeling Stelsel (RHVS) van die Korrelbed Modulêre Kern-Reaktor (KMKR) is die onderwerp van talle navorsings projekte. Probleme geïdentifiseer in vorige studies sluit in die hipotetiese onakkuraatheid van hitte-oordrag koëffisiënt korrelasies beskikbaar in literatuur. Die doel van die navorsing aangebied is om binne-pyp hitte-oordrag koëffisiënt korrelasies te ontwikkel spesifiek vir die huidige ontwerp van die RHVS. Ten einde dit te bereik, word ‘n literatuurstudie uitgevoer wat kern-reaktors identifiseer wat gebruik maak van natuurlike sirkulasie lusse. Die literatuurstudie verduidelik ook die algemene een-dimensionele twee-vloeistof behoud vergelykings wat die basis vorm vir numeriese modellering van natuurlike sirkulasie lusse. Die literatuurstudie bespreek laastens beskikbare hitte-oordrag koëffisiënt korrelasies met die doel om te identifiseer vir welke massavloei tempo waardes en onder watter omstandighede hierdie korrelasies as korrek beskou is. Die ontleding sluit korrelasies in wat algemeen gebruik word in die modellering van natuurlike sirkulasie in die kern industrie met die hoop om korrelasies vir gebruik in die modellering van die voorgestelde RHVS te identifiseer. Een van die doelwitte van die projek is om ‘n een-derde-hoogte-skaal model van die RHVS te ontwerp en te bou. Tekortkominge van vorige eksperimentele modelle is geidentifiseer en, so ver as moonlik, voor vergoed in die ontwerp van die model. Koper pype word gebruik wat die onsekerhede van materiaal en opperkvlak eindomme voorkom. Verkseie deursigtige polikarbonaat segmente is ingesluit wat visuele identifikasie van twee-fase vloei regimes toelaat. ‘n Opening plaat word gebruik om voorwaartse en terugwaartse vloeimeting toe te laat. Die opening plaat, termokoppels en hitte uitruilers is gekalibreer in plek om eksperimentele foute te verminder en om herhaalbaarheid te verseker. Twaalf eksperimente word uitgevoer en data word elke tien sekondes aangeteken. Die resultate wat hier aangebied word, is beperk tot geselekteerde enkel- en tweefase vloei meganismes van werking. Fout ontleding en herhaalbaarheid van eksperimentele metings, om die herhaalbaarheid van eksperimentele resultate te toon. Hierdie is gebruik om wiskundig te bepaal wat die eksperimentele binne-pyp hitte-oordrag koëffisiënte is vir beide die verdamper en kondenseerder afdelings. Tendense in die hitte-oordrag koëffisiënt profiele word geïdentifiseer en die algemene gedrag van die profiles is deeglik verduidelik. Die RHVS is gemodelleer as 'n een-dimensionele stelsel. Korrelasies vir die wrywing faktor, hitte-oordrag koëffisiënte, leegte-breuk en twee-fase wrywings vermenigvuldiger word geïdentifiseer. Die teoretiese hitte-oordrag koëffisiënte word bereken deur middle van die wiskundige model en korrelasies wat in literatuur geidentifiseer is. Vloeistof parameters is geëvalueer met eksperimenteel bepaalde temperature en massa-vloei tempos. Die gevolglike hitte-oordrag koëffisiënt profiles is vergelyk met eksperimentele profiele om die hipotese dat die bestaande korrelasies nie die binne-pyp hitte-oordrag koëffisiënte akkuraat voorspel nie, te bevestig. Die eksperimenteel bepaalde koëffisiënte is gekorreleer en die gegenereerde korrelasies, saam met geïdentifiseerde twee-fase hitte-oordrag koëffisiënt korrelasies, word gebruik in 'n wiskundige model om teoretiese koëffisiënt profiele te genereer. Dit word dan vergelyk met die eksperimenteel bepaalde hitteoordrag koëffisiënte om die akkuraatheid van voorspelling te toon. Tekortkominge in die teoretiese en eksperimentele model word geïdentifiseer en aanbevelings gemaak om hulle aan te spreek in die toekoms.

Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2010.<br>ENGLISH ABSTRACT: A reactor cavity cooling system (RCCS) is used in the PBMR to protect the concrete citadel surrounding the reactor from direct nuclear radiation impingement and heat. The speci ed maximum operating temperature of the concrete structure is 65 ±C for normal operating conditions and 125 ±C for emergency shut-down conditions. A conceptual design of an entirely passive RCCS suitable for the PBMR was done by using closed loop thermosyphon heat pipes (CLTHPs) to remove heat from a radiation heat shield over a horizontal distance to an annular cooling dam placed around the PBMR. The radiation shield is placed in the air space between the Reactor Pressure Vessel (RPV) and the concrete citadel, 180 mm from the concrete citadel. A theoretical heat transfer model of the RCCS was created. The theoretical model was used to develop a computer program to simulate the transient RCCS response during normal reactor operation, when the RCCS must remove the excess generated heat from the reactor cavity and during emergency shut-down conditions, when the RCCS must remove the decay heat from the reactor cavity. The main purpose of the theoretical model is to predict the surface temperature of the concrete citadel for di erent heat generation modes in the reactor core and ambient conditions. The theoretical model assumes a 1D geometry of the RCCS. Heat transfer by both radiation and convection from the RPV to the radiation heat shield (HS) is calculated. The heat shield is modelled as a n. The n e ciency was determined with the experimental work. Conduction through the n is considered in the horizontal direction only. The concrete structure surface is heated by radiation from the outer surface of the heat shield as well as by convection heat transfer from the air between the heat shield and the concrete structure surface. The modelling of the natural convection closed loop thermosyphon heat pipes in the RCCS is done by using the Boussinesq approximation and the homogeneous ow model. An experiment was built to verify the theoretical model. The experiment is a full scale model of the PBMR in the horizontal, or main heat transfer, direction, but is only a 2 m high section. The experiments showed that the convection heat transfer between the RPV and the HS cannot be modelled with simple natural convection theory. A Nusselt number correlation developed especially for natural convection in enclosed rectangles found in literature was used to model the convection heat transfer. The Nusselt number was approximately 3 times higher than that which classic convection theory suggested. An optimisation procedure was developed where 121 di erent combinations of n sizes and heat pipe sizes could be used to construct a RCCS once a cooling dam size was chosen. The purpose of the optimisation was to nd the RCCS with the lowest total mass. A cooling dam with a diameter of 50 m was chosen. The optimal RCCS radiation heat shield that operates with the working uid only in single phase has 243 closed loop thermosyphon heat pipes constructed from 62.72 mm ID pipes and 25 mm wide atbar ns. The total mass of the single phase RCCS is 225 tons. The maximum concrete structure temperature is 62.5 ±C under normal operating conditions, 65.8 ±C during a PLOFC emergency shut-down condition and 80.9 ±C during a DLOFC emergency shut-down condition. In the case where one CLTHP fails and the adjacent two must compensate for the loss of cooling capacity, the maximum concrete structure temperature for a DLOFC emergency shut-down will be 87.4 ±C. This is 37.6 ±C below the speci ed maximum temperature of 125 ±C. The RCCS design is further improved when boiling of the working uid is induced in the CLTHP. The optimal RCCS radiation heat shield that operates with the working uid in a liquid-vapour mixture, or two phase ow, has 338 closed loop thermosyphon heat pipes constructed from 38.1 mm ID pipes and 20 mm wide atbar ns. The total mass of the two phase RCCS is 198 tons, 27 tons less than the single phase RCCS. The maximum concrete structure temperature is 60 ±C under normal operating conditions, 2.5 ±C below that of the single phase RCCS. During a PLOFC emergency shut-down condition, the maximum concrete structure temperature is 62.3 ±C, 3.5 ±C below that of the single phase RCCS and still below the normal operating temperature of the single phase RCCS. By inducing two phase ow in the CLTHP, the maximum temperature of the working uid is xed equal to the saturation temperature of the working uid at the vacuum pressure. This property of water is used to limit the concrete structure temperature. This e ect is seen in the transient response of the RCCS where the concrete structure temperature increases until boiling of the working uid starts and then the concrete structure temperature becomes constant irrespective of the heat load on the RCCS. An increased heat load increases the quality of the working uid liquid-vapour mixture. Working uid qualities approaching unity causes numerical instabilities in the theoretical model. The theoretical model cannot capture the heat transfer to a control volume with a density lower than approximately 20 kg/m3. This limits the extent to which the two phase RCCS can be optimised. Recommendations are made relating to future work on how to improve the theoretical model in particular the convection modelling in the reactor cavities as well as the two phase ow of the working uid. Further recommendations are made on how to improve the basic design of the heat shield as well as the cooling section of the CLTHPs.<br>AFRIKAANSE OPSOMMING: 'n Reaktor lug spasie verkoelingstelsel (RLSVS) word in die PBMR gebruik om die beton wat die reaktor omring te beskerm teen direkte stralingskade en hitte. Die gespesi seerde maksimum temperatuur van die beton is 65 ±C onder normale bedryfstoestande en 125 ±C gedurende die noodtoestand afskakeling van die reaktor. 'n Konseptuele ontwerp van 'n geheel en al passiewe RLSVS geskik vir die PBMR is gedoen deur gebruik te maak van geslote lus termo-sifon (GLTSe) om hitte van die stralingskerm te verwyder oor a horisontale afstand na 'n ringvormige verkoelingsdam wat rondom die reaktor geposisioneer is. Die stralingskerm word in die lug spasie tussen die reaktor drukvat (RDV) en die beton geplaas, 180 mm vanaf die beton. 'n Teoretiese hitteoordrag model van die RLSVS was geskep. Die teoretiese model was gebruik vir die ontwikkeling van 'n rekenaar program wat die transiënte gedrag van die RLSVS sal simuleer gedurende normale bedryfstoestande, waar die oorskot gegenereerde hitte verwyder moet word vanuit die reaktor lug spasie, asook gedurende noodtoestand afskakeling van die reaktor, waar die afnemingshitte verwyder moet word. Die primêre doel van die teoretiese model is om the oppervlak temperatuur van die beton te voorspel onder verskillende bedryfstoestande asook verskillende omgewingstoestande. Die teoretiese model aanvaar 'n 1D geometrie van die RLSVS. Hitte oordrag d.m.v. straling asook konveksie vanaf die RDV na die stralingskerm word bereken. The stralingskerm word gemodelleer as 'n vin. Die vin doeltre endheid was bepaal met die eksperimente wat gedoen was. Hitte geleiding in die vin was slegs bereken in die horisontale rigting. Die beton word verhit deur straling vanaf die agterkant van die stralingskerm asook deur konveksie vanaf die lug tussen die stralingskerm en die beton. The modellering van die natuurlike konveksie GLTS hitte pype word gedoen deur om gebruik te maak van die Boussinesq benadering en die homogene vloei model. 'n Eksperiment was vervaardig om the teoretiese model te veri eer. Die eksperiment is 'n volskaal model van die PBMR in die horisontale, of hoof hitteoordrag, rigting, maar is net 'n 2 m hoë snit. Die eksperimente het gewys dat die konveksie hitte oordrag tussen die RDV en die stralingskerm nie met gewone konveksie teorie gemodelleer kan word nie. 'n Nusselt getal uitdrukking wat spesi ek ontwikkel is vir natuurlike konveksie in geslote, reghoekige luggapings wat in die literatuur gevind was, was gebruik om die konveksie hitteoordrag te modelleer. Die Nusselt getal was ongeveer 3 maal groter as wat klassieke konveksie teorie voorspel het. 'n Optimeringsprosedure was ontwikkel waar 121 verskillende kombinasies van vin breedtes en pyp groottes wat gebruik kan word om 'n RLSVS te vervaardig nadat 'n toepaslike verkoelingsdam diameter gekies is. Die doel van die optimering was om die RLSVS te ontwerp wat die laagste totale massa het. 'n Verkoelingsdam diameter van 50 m was gekies. Die optimale RLSVS stralingskerm, waarvan die vloeier slegs in die vloeistof fase bly, bestaan uit 243 GLTSe wat van 62.72 mm binne diameter pype vervaardig is met 25 mm breë vinne. The totale massa van die enkel fase RLSVS is 225 ton. Die maksimum beton temperatuur is 62.5 ±C vir normale bedryfstoestande, 65.8 ±C vir 'n PLOFC noodtoestand afskakeling en is 80.9 ±C vir 'n DLOFC noodtoestand afskakeling. In die geval waar een GLTS faal gedurende 'n DLOFC noodtoestand afskakeling en die twee naasgeleë GLTSe moet kompenseer vir die vermindering in verkoelings kapasiteit, is die maksimum beton temperatuur 87.4 ±C. Dit is 37.6 ±C laer as die gespesi seerde maksimum temperatuur van 125 ±C. Die RLSVS ontwerp kan verder verbeter word wanneer die vloeier in die GLTSe kook. Die optimale RLSVS stralingskerm met die vloeier wat kook, of in twee fase vloei is, bestaan uit 338 GLTSe wat van 38.1 mm binne diameter pype vervaardig is met 20 mm breë vinne. The totale massa van die twee fase vloei RLSVS is 198 ton, 27 ton ligter as die enkel fase RLSVS. Die maksimum beton temperatuur is 60 ±C vir normale bedryfstoestande, 2.5 ±C laer as die enkel fase RLSVS. Gedurende 'n PLOFC noodtoestand afskakeling is die maksimum beton temperatuur 62.3 ±C, 3.5 ±C laer as die enkel fase RLSVS en nogtans onder die maksimum beton temperatuur van die enkel fase RLSVS vir normale bedryfstoestande. Deur om koking te veroorsaak in die GLTS word die maksimum temperatuur van die vloeier vasgepen gelyk aan die versadigings temperatuur van die vloeier by die vakuüm druk. Hierdie einskap van water word gebruik om 'n limiet te sit op die maksimum temperatuur van die beton. Hierdie e ek kan gesien word in die transiënte gedrag van die RLSVS waar die beton temperatuur styg tot en met koking plaasvind en dan konstant raak ongeag van die hitte belasting op die RLSVS. 'n Toename in die hitte belasting veroorsaak net 'n toename in die kwaliteit van die vloeistof-gas mengsel. Mengsel kwaliteite van 1 nader veroorsaak numeriese onstabiliteite in die teoretiese model. The teoretiese model kan nie die hitteoordrag beskryf na 'n kontrole volume wat 'n digtheid het laer as ongeveer 20 kg/m3. Hierdie plaas 'n limiet op die optimering van die twee fase RLSVS. Aanbevelings was gemaak met betrekking tot toekomstige werk aangaande die verbetering van die teoretiese model met spesi eke klem op die modellering van konveksie in die reaktor asook die modellering van twee fase vloei. Verdere aanbevelings was gemaak aangaande die verbetering van die stralingskerm ontwerp asook die ontwerp van die verkoeling van die GLTSe.

Le contrôle thermique des réactions dans les réacteurs à lits fixes nécessite la maîtrise du transfert thermique en proche paroi. Afin d’optimiser leur conception et en particulier de maintenir un chemin réactionnel le plus proche possible de l’isothermicité, un modèle de transfert thermique pertinent dans un tel milieu est indispensable. Deux premières thèses au Laboratoire d'Energétique et de Mécanique Théorique et Appliquée ont permis de mettre au point un modèle dispersif utilisable à cœur de réacteur. La présente thèse met en évidence un effet de paroi non négligeable causé par la variation de porosité du lit à l’approche de la paroi. Ce travail propose un raccordement du cœur, où le modèle thermique précédent est légitime, à la paroi, par une couche homogénéisée. Cette approche est validée par une caractérisation expérimentale des paramètres du modèle sur un dispositif de laboratoire en utilisant une méthode d’inversion originale reposant sur une approche Bayesienne. Une validation de cette caractérisation dans une géométrie plus proche d’un réacteur industriel est également effectuée<br>Thermal control for chemical reactions conducted in fixed-bed reactors requires a fine knowledge about heat transfer in the near-wall region of the bed. In order to optimize the process design, to minimize the mass of catalyst that is needed, one must attempt to maintain the operational path of the reactor the closest of the isotherm functioning. A pertinent and reliable heat transfer model is then required. Two thesis were realized in the Laboratoire d'Energétique et de Mécanique Théorique et Appliquée in order to construct a corresponding dispersive model. This model can be used at the core of the reactor, where the porous medium can be homogenized. This thesis shows a wall effect that cannot be neglected. This wall effect is directly caused by the important variation of the porosity distribution near the wall of the bed. This work proposes a junction between a core layer, where the previous model is still valid, and the wall through an homogenized near-wall layer. This approach is validated by the mean of an experimental characterization of the model parameters, using an original inversion technique based on a Bayesian approach. A validation of this characterization in another geometry is also done. This geometry is closer to an industrial reactor

Pebble-fueled high temperature gas-cooled reactor (HTGR) technology was first developed by the Federal Republic of Germany in the 1950s. More recently, the design has been embraced by the People's Republic of China and the Republic of South Africa. Unlike light water reactors that generate heat from fuel assemblies comprised of fuel rods, pebble-fueled HTGRs utilize thousands of 60-mm diameter fuel spheres (pebbles) comprised of thousands of TRISO particles. As this reactor type is deployed across the world, adequate methods for safeguarding the reactor must be developed. Current safeguards methods for the pebble-fueled HTGR focus on extensive, redundant containment and surveillance (C/S) measures or a combination of item-type and bulk-type material safeguards measures to deter and detect the diversion of fuel pebbles. The disadvantages to these approaches are the loss of continuity of knowledge (CoK) when C/S systems fail, or are compromised, and the introduction of material unaccounted for (MUF). Either vulnerability can be exploited by an adversary to divert fuel pebbles from the reactor system. It was determined that a solution to maintaining CoK is to develop a system to identify each fuel pebble that is inserted and removed from the reactor. Work was performed to develop and evaluate the use of inert microspheres placed in each fuel pebble, whose random placement could be used as a fingerprint to identify the fuel pebble. Ultrasound imaging of 1 mm zirconium oxide microspheres was identified as a possible imaging system and microsphere material for the new safeguards system concept. The system concept was evaluated, and it was found that a minimum of three microspheres are necessary to create enough random fingerprints for 10,000,000 pebbles. It was also found that, over the lifetime of the reactor, less than 0.01% of fuel pebbles can be expected to have randomly the same microsphere fingerprint. From an MCNP 5.1 model, it was determined that less than fifty microspheres in each pebble will have no impact on the reactivity or temperature coefficient of reactivity of the reactor system. Finally, using an ultrasound system it was found that ultrasound waves can penetrate thin layers of graphite to image the microsphere fingerprint.

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.<br>MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2012

A high temperature molten salt test loop that utilizes FLiNaK (LiF-NaF-KF) at 700ºC has been proposed by Oak Ridge National Laboratory (ORNL) to study molten salt flow characteristics through a pebble bed for applications in high temperature thermal systems, in particular the Pebble Bed – Advanced High Temperature Reactor (PB-AHTR). The University of Tennessee Nuclear Engineering Department has been tasked with developing and testing pressure instrumentation for direct measurements inside the high temperature environment. A nickel diaphragm based direct contact pressure sensor is developed for use in the salt. Capacitive and interferometric methods are used to infer the displacement of the diaphragm. Two sets of performance data were collected at high temperatures. The fiber optic, Fabry-Perot interferometric sensor was tested in a molten salt bath. The capacitive pressure sensor was tested at high temperatures in a furnace under argon cover gas.

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.<br>MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2015

SpinChem AB's rotating bed reactor (RBR) is a small cylindrical centrifuge that allowsfor eective convective reactions between chemical substances. The aim of my master'sdissertation was to develop a measurement system that could estimate the uid velocity insidethe RBR. This report covers a quick pre-study of dierent uid measurement methods,details how we designed and constructed the system using integrated circuits and other electroniccomponents, and illustrates calibration results and nal velocity measurements. Thesystem design was based on the architecture of hot-wire constant temperature anemometry(CTA), a technique that measures the uid velocity over a heated metal wire. The resultsof this project is the measurement system itself, and the calibration and velocity measurementsgathered from experimentation. Two issues we encountered was that the calibrationprocess resulted in an overestimation of the velocity in the order of 4000 cm/s, and thatthe response to changes in velocity rates was quite slow. However, a slight tuning of thetwo calibration constants A0 &amp; B0 smoothed the velocity estimation out to 40 cm/s, whichis more what we would expect. This illustrated how sensitive the results were to errors inthe calibration. Further investigation of the slow response, and a more stringently executedcalibration setup, could make the CTA a potential candidate for estimating the uid velocityinside the RBR.

The fission product releases from spherical fuel elements used in modern high temperature gas cooled reactors are one of the first source terms used in describing the safety of planned nuclear plants during normal and accident conditions. The verification and validation of the model and code used to predict the gaseous fission product behaviour and release from spherical fuel elements for the PBMR were documented in this dissertation. The PBMR is the latest design in high temperature gas cooled reactor technology utilizing spherical fuel elements based on the LEU TRISO-coated particle design. Fission products can be divided into relatively short-lived noble gas and halogens, and relatively long-lived metallic fission and activation products. Each group is described by its own models and sets of transport parameters. The noble gases and halogen fission product releases from the fuel elements are direct indications of fuel performance and are modelled by the Booth equation. The fission product release legacy code NOBLEG for noble gases and halogens was developed previously to calculate this diffusion model for high temperature reactors. The model and code are verified and validated for use in PBMR design and analyses under normal operating conditions. The history of irradiation experiments conducted on coated fuel particles and spherical fuel elements was investigated, and the most suitable irradiation tests with their post irradiation investigations were identified for the purpose of validation of the model and code. The model used to determine gaseous fission product behaviour and release from spherical fuel elements is described in detail. The application of this model in the code is verified mathematically with the Booth model, and by inspection of the source code. The thermohydraulic model used by NOBLEG to calculate fuel temperatures is verified with code to code comparisons with the core neutronics code VSOP. The irradiation tests HFR-K5 and -K6 were selected to validate the gaseous fission product code NOBLEG. An investigation was done into the development of NOBLEG to calculate gaseous fission product release under oxidizing conditions caused by water ingress events. New relationships were derived from water vapour injection tests done during the irradiation experiment HFR-K6, that allows NOBLEG to estimate the increase in gaseous fission product release under oxidizing conditions. A new model was proposed to explain peculiarities observed during the water injection tests.<br>Prof. P.P. Coetzee

This work details modelling buoyancy-driven viscous flow and heat transfer through heterogeneous saturated packed pebble beds via a set of volume-averaged conservation equations in which local thermal disequilibrium is accounted for. The latter refers to the two phases considered viz. solid and fluid, differing in temperature. This is effected by describing each phase with its own governing equation. Further to the aforementioned, the governing equation set is written in terms of intrinsic volume-averaged material properties that are fully variant with respect to temperature. The heterogeneous solid phase is described with a porosity field varying from 0.39 to 0.99. The intent of the stated upper bound is to explicitly model typical packed bed near-wall phenomena such as wall-channelling and pebble-wall heat transfer as true to reality as possible, while maintaining scientific rigour. The set of coupled non-linear partial differential equations is solved via a locally preconditioned artificial compressibility method, where spatial discretisation is effected with a compact finite volume edge-based discretisation method. The latter is done in the interest of accuracy. Stabilisation is effected via JST scalar-valued artificial dissipation. This is the first instance in which an artificial compressibility algorithm is applied to modelling heat and fluid flow through heterogeneous porous materials. As a result of the aforementioned, calculation of the acoustic velocities, stabilisation scaling factors and allowable time-step sizes were revised. The developed technology is demonstrated by application to the modelling of SANA test cases, i.e. natural convective flow inside a heated porous axisymmetric cavity. Predicted results are shown to be within 12% of experimental measurements in all cases, while having an average deviation of only 3%.<br>Dissertation (MEng)--University of Pretoria, 2008.<br>Mechanical and Aeronautical Engineering<br>unrestricted

Nuclear safety is the main concern for the licensing of nuclear power plants, not only in the Republic of South Africa but also worldwide. The design of the nuclear power plant plays an important role in the licensing process, which includes probabilistic and deterministic analysis of a set of design or Licensing basis events. This study was about the deterministic analysis for the sensitivity of licensing basis events radiological consequences to different radiological pathways. The study was done for the Pebble Bed Modular Reactor (PBMR), which is a nuclear power plant, still in its early phase of design approaching its detailed design phase. An abnormal event or an accident could lead to a release of radioactive particles and gases from a Pebble Bed Modular Reactor and could give rise to radiation exposure to workers and the surrounding population. Therefore nuclear events due to PBMR, which are Licensing Basis Events or Design Basis Accidents, must be analysed in order to demonstrate that accidental and routine releases of radioactivity are kept As Low As Reasonably Achievable (ALARA) and that the design basis meets offsite dose requirements with adequate safety margins. In this work, it is also shown that collectively the risk criteria are satisfied in the fundamental safety requirements of National Nuclear Regulator (NNR) of the Republic of South Africa (RSA) and similar risk criteria of the other countries in which it has to be employed. Furthermore the various pathways through which radioactivity can reach the public are analysed. The focus of the study was to determine which pathways deliver the greatest radiation exposure if there is an accident due to an event happening in PBMR and also to provide a LBE analysis process as a step in confirming that the design meets the licensing requirements.<br>Thesis (M.Sc. ARST) North-West University, 2004

碩士<br>國立清華大學<br>動力機械工程學系<br>104<br>The aim of this study is to design a uniform temperature distribution in scale-up fixed bed reactors with inserting the internal heat conducting part and different heating methods by numerical simulation. The performance of sorbents, the relevant capture efficiency and the number of running cycles can be improved due to the absence of hot spots. The simulation results show that the performance of desorption is only 58 % at 850 ℃ for the original design as the diameter of the reaction tube is increased from 1 inch to 4 inches. The reactor can be effectively improved by adding axial fins in hollow reactors because the temperature distribution becomes more uniform. The results also show that the best hollow ring radius will vary with the material, the number of fins and heating method. The method of internal heating is better than that of external heating for fixed material and the number of fins.

Public participation can be defined as ...”a process leading to a joint effort by stakeholders, technical specialists, the authorities and the proponent who work together to produce better decisions than if they had acted independently" (Greyling, 1999, p. 20). In South Africa, public participation processes are legally driven and form part a statutory part of environmental impact assessments. Many environmental impact assessments have been undertaken in South Africa, but the environmental impact assessment undertaken for the proposed construction of a demonstration module pebble bed modular reactor was perhaps one of the biggest studies undertaken to date from a public participation process point of view (Smit, 2003). The main aim of this mini-dissertation was to evaluate the public participation process followed for the environmental impact assessment of the demonstration module pebble bed modular reactor at Koeberg in the Western Cape Province against the criteria for fairness and competence as set out by Webler (In: Renn et al., 1995). Despite the fact that this work is eleven years old, it is still regarded as a benchmark for the evaluation of public participation processes in environmental decision making (Abelson et al., 2003). Webler (In: Renn et al., 1995) developed a normative theory for fairness and competence in public participation based on the theory of ideal speech of German sociologist Jürgen Habermas. Habermas’ main contribution to science was his theory of universal pragmatics (Author unknown, 2005). Universal pragmatics is a theory aimed at explaining how language is used to ensure mutual understanding and agreement. Webler (In: Renn et al., 1995) argues that the conditions of universal pragmatics, if applied to public participation, points towards the concepts of fairness (providing everyone with the opportunity to participate) and competence [providing participants (called interested and affected parties (I&APs) with the opportunity to make, question and validate speech acts]. Habermas advocates that each statement (or speech act) makes at least one validity claim and that there is a presupposition that the speaker can validate each claim to the satisfaction of all communication partners, should this be necessary (Perold, 2006). Furthermore, Habermas identifies four different types of validity claims, each having to do with a specific type of statement. In his theory, communicative speech acts have to do with comprehensibility; constantive speech acts with truth/correctness; regulative speech acts with normative rightness and representative speech acts with sincerity. Webler (In: Renn et al., 1995) developed a set of criteria to evaluate the fairness and competence in public participation. This set of criteria was applied to the public participation process of the case study. The study found that the process followed in the case study did not fare particularly well in either fairness or competence, but that fairness was slightly better than competence. The most alarming finding was that little attempt was made to ensure that validity claims – especially constantive (truth and factual information) – were validated or redeemed as this left the door open for misinterpretation, politics and incorrectness. It was also found that I&APs were, for the most, prevented from participating in the decision-making process. This finding may or may not be interpreted as negative as the public participation consultant never made a claim towards power sharing as well as the fact that there are widely differing opinions regarding the level to which public participation should take place. It was suggested that at least some elements of power sharing be incorporated into future processes, that validity claims – especially constantive (theoretical/factual) and therapeutic (regarding feelings and emotions) – must be able to stand up to scrutiny and should be validated. Finally, it was suggested that more attention be given to representative speech acts (statements regarding emotions, perceptions and feelings).<br>Dr. J. M. Meeuwis

Distributed power generation through renewable sources of energy has the potential of meeting the challenge of providing electricity access to the off-grid population, estimated to be around 1.2 billion residing across the globe with 300 million in India, in a sustainable way. Technological solutions developed around these energy challenges often involve thermal systems that convert heat available from sources like solar, biomass, geothermal or unused industrial processes into electricity. Conventional steam based thermodynamic cycle at distributed scale (< 1 MWe) suffers from low efficiency driving scientific research to develop new, scalable, efficient and economically viable power cycles. This PhD work conducts one such study which provides a database of thermal power blocks optimized for the lowest initial investment cost to developers of distributed power plants. The work is divided in two steps; a) feasibility study of various thermodynamic cycles for distributed power generation covering different operating temperature regimes and b) perform their detailed thermo-economic modelling for the heat sources mentioned above. Thermodynamic cycles are classified into three temperature domains namely, low (< 450 K), medium (< 600 K) and high (< 1000 K) T cycles. Any fluid whose triple point temperature is below the typical ambient temperatures is a potential working fluid in the power cycle. Most of the organic and the inorganic fluids satisfy this criterion and can be perceived as potential power cycle fluids. The general notion is that organic fluids are more suited for low or medium temperature cycles whereas inorganic fluids for high temperature ones. Organic fluids can further be classified into hydrofluorocarbon and hydrocarbon. While the former has high global warming potential (GWP), the latter is flammable in nature. Their mixture in certain compositions is found to obviate both the demerits and perform equally well on thermodynamic scales for low T cycles. On the similar lines, mixture of HCs and inorganic fluids, such as propane+CO2 and isopentane+CO2 are found to be more appropriate for medium T applications if the issues like pinch temperature in the regenerator arising due to temperature glide are taken care of. In the high temperature domain, high efficiency Brayton cycle (supercritical CO2) and transcritical condensing cycles are studied with the latter being 2 % more efficient than the former. However, application of the condensing cycle is limited to low temperature ambient locations owing to low critical temperature of CO2 (304 K). In the same cycle configuration, mixture of CO2 and propane (52 and 48%) with a critical temperature of ~ 320 K is observed to retain the thermodynamic performance with the increased heat rejection temperature matched to the tropical ambient conditions. However, these cycles are plagued by the high operating pressures (~300 bar) calling for high temperature steel making the power block uneconomical. In this regard, the advanced CO2 cycles are developed wherein the optimum operating pressures are limited to 150 bar with an increased cycle efficiency of 6 % over the S-CO2 cycle. Feasibility study carried out on these cycles in the Indian context indicates the low and medium T cycles to be better suited for distributed power generation over the high T cycles. In the second part of work, a comprehensive study is performed to optimize the low and the medium T cycles on a thermo-economic basis for the minimum specific investment cost ($/We). Such a study involves development of component level models which are then integrated to form the system of interest, thus, following a bottom-up approach. A major emphasis is given on the development of scroll expander and low cost pebble bed thermal energy storage system that are the reported in the literature as the areas with high uncertainties while connecting them to the system. Subsequently, the key design parameters influencing the specific cost of power from an air-cooled ORC are identified and used to formulate a 7-dimensional space to search for the minimum costs for applications with a) geothermal/waste or biogas heat sources and b) solar ORCs. Corresponding maps of operating parameters are generated to facilitate distributed power engineers in the design of economic systems within constraints such as available heat source temperatures, maximum expander inlet pressures imposed, etc. Further, the effect of power scaling on these specific costs is evaluated for ORC capacities between 5 and 500 kWe.