Thesis (M.Eng. (Mechanical )) -- University of Pretoria, 2007. / Includes bibliographical references (leaves 73-81)

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%. / Dissertation (MEng)--University of Pretoria, 2008. / Mechanical and Aeronautical Engineering / unrestricted

Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2010. / 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. / 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(Mtech (Industrial Engineering)--Cape Peninsula University of Technology, 2010 / 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.

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). / Dr. J. M. Meeuwis

Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2010. / 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. / 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.

Thesis (MScEng (Mechanical and Mechatronic Engineering))--University of Stellenbosch, 2011. / 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. / 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 (MBA)--Stellenbosch University, 2002. / 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. / 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.

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. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.

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. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.