<p>The use of hydrogen is predicted to increase substantially in the future, both as chemical feedstock and also as energy carrier for transportation. The annual world production of hydrogen amounts to some 50 million tonnes and the majority is produced using fossil fuels like natural gas, coal and naphtha. High temperature nuclear reactors (HTRs) represent a novel way to produce hydrogen at large scale with high efficiency and less carbon footprint. The aim of this master thesis has been to evaluate the feasibility of HTRs for hydrogen production by analyzing both the reactor concept and its potential to be used in certain hydrogen niche markets. The work covers the production, storage, distribution and use of hydrogen as a fuel for vehicles and aviation and as chemical feedstock for the oil refining and ammonia production industry.</p><p>The study indicates that HTRs may be suitable for hydrogen production under certain conditions. However, the use of hydrogen as an energy carrier necessitates a widespread hydrogen infrastructure (e.g. pipe-lines, refuelling stations and large scale storage), which is associated with major energy losses. Both mentioned industries could benefit from nuclear-based hydrogen with less infrastructural changes, but the potential market is by far smaller than if hydrogen is used as an energy carrier. A maximum of about 60 HTRs of 600 MWth worldwide has been estimated for the ammonia production industry. The Swedish refineries are likely too small to utilize the HTR but in the larger refineries HTR might be applicable.</p>

The use of hydrogen is predicted to increase substantially in the future, both as chemical feedstock and also as energy carrier for transportation. The annual world production of hydrogen amounts to some 50 million tonnes and the majority is produced using fossil fuels like natural gas, coal and naphtha. High temperature nuclear reactors (HTRs) represent a novel way to produce hydrogen at large scale with high efficiency and less carbon footprint. The aim of this master thesis has been to evaluate the feasibility of HTRs for hydrogen production by analyzing both the reactor concept and its potential to be used in certain hydrogen niche markets. The work covers the production, storage, distribution and use of hydrogen as a fuel for vehicles and aviation and as chemical feedstock for the oil refining and ammonia production industry. The study indicates that HTRs may be suitable for hydrogen production under certain conditions. However, the use of hydrogen as an energy carrier necessitates a widespread hydrogen infrastructure (e.g. pipe-lines, refuelling stations and large scale storage), which is associated with major energy losses. Both mentioned industries could benefit from nuclear-based hydrogen with less infrastructural changes, but the potential market is by far smaller than if hydrogen is used as an energy carrier. A maximum of about 60 HTRs of 600 MWth worldwide has been estimated for the ammonia production industry. The Swedish refineries are likely too small to utilize the HTR but in the larger refineries HTR might be applicable.

The goal of this work is to investigate the capabilities of two different commercial codes, OpenFOAM and ANSYS CFX, to predict forced convection heat transfer in low Prandtl number fluids and investigate the sensitivity of these predictions to the type of code and to several input parameters.The goal of the work is accomplished by predicting forced convection heat transfer in two different experimental setups with the codes OpenFOAM and ANSYS CFX using three different turbulence models and varying the input parameters in an extensive sensitivity analysis. The computational results are compared two the experimental data and analyzed for qualitative and quantitative parameters, such as shape of velocity and temperature profiles, thickness of the boundary layers and wall temperatures.The results show that predictions of the temperature and velocity field are generally sufficient to good, however, the sensitivity especially to the turbulent Prandtl number has to be taken into account when computing forced convection heat transfer in low Prandtl number fluids. The results also show that methods applied to OpenFOAM cannot directly be applied to ANSYS CFX. / <p>QC 20130531</p> / GENIUS

No description available.

Next generation nuclear power plants (GEN-IV) will be capable of not only producing energy in a reliable, safe and sustainable way, but they will also be capable of reducing the amount of nuclear waste, which has been accumulated over the lifetime of current-generation nuclear power plants, through transmutation. Due to the use of new and different coolants, existing computational tools need to be tested, further developed and improved in order to thermal-hydraulically design these power plants.This work covers two different non-unity Prandtl number fluids which are considered as coolants in GEN-IV reactors, liquid lead/lead-bismuth-eutectic and supercritical water. The study investigates different turbulence modeling strategies, such as Large Eddy Simulation (LES) and Reynolds-Averaged Navier-Stokes (RANS) modeling, and their applicability to these proposed coolants. It is shown that RANS turbulence models are partly capable of predicting wall heat transfer in annular flow configurations. However, improvements in these prediction should be possible through the use of advanced turbulence modeling strategies, such as the use of separate thermal turbulence models. A large blind benchmark study of heat transfer in supercritical water showed that the available turbulence modeling strategies are not capable of predicting deteriorated heat transfer in a 7-rod bundle at supercritical pressures. New models which take into account the strong buoyancy forces and the rapid change of the molecular Prandtl number near the wall occurring during the transition of the fluid through the pseudocritical point need to be developed. One of these strategies to take into account near-wall buoyancy forces is the use of advanced wall functions, which cannot only help in modeling these kind of flows, but also decrease computational time by 1 to 2 orders of magnitude. Different advanced wall function models were implemented in the open-source CFD toolbox OpenFOAM and their performance for different flows in sub- and supercritical conditions were evaluated. Based on those results, the wall function model UMIST-A by Gerasimov is recommended for further investigation and specific modeling tactics are proposed.Near-wall temperature and velocity behavior is important to and influenced by the wall itself. The thermal inertia of the wall influences the temperature in the fluid. However, a more important issue is how temperature fluctuations at the wall can induce thermal fatigue. With the help of LES thermal mixing in a simplified model of a control rod guide tube was investigated, including the temperature field inside the control rod and guide tube walls. The WALE sub-grid turbulence model made it possible to perform LES computations in this complex geometry, because it automatically adapts to near-wall behavior close to the wall, without the use of ad-hoc functions. The results for critical values, such as the amplitude and frequency of the temperature fluctuations at the wall, obtained from the LES computations are in good agreement with experimental results.The knowledge gained from the aforementioned investigations is used to optimize the flow path in a small, passively liquid-metal-cooled pool-type GEN IV reactor, which was designed for training and education purposes, with the help of 3D CFD. The computations were carried out on 1/4 of the full geometry, where the small-detail regions of the heat exchangers and the core were modeled using a porous media approach. It was shown that in order to achieve optimal cooling of the core without changing the global geometry a ratio of close to unity of the pressure drop over the core and the heat exchanger needs to be achieved. This is done by designing a bottom plate which channels enough flow through the core without choking the flow in the core. Improved cooling is also achieved by reducing heat losses from the hot leg through the flow shroud to the cold leg by applying thermal barrier coating similar to methods used in gas turbine design. / Nästa generations kärnkraftverk (GEN-IV) kan inte bara producera el på ett pålitligt, säkert och hållbart sätt, utan det kan också reducera mängden kärnavfall, som har producerats under tiden som man använt nuvarande generationen kärnkraftverk, genom att transmutera avfallen. Framtidens kärnkraftverk använder andra kylmedel än nuvarande kraftverk som t.ex. flytande bly, gas eller superkritiskt vatten. Det betyder att många beräkningsverktyg måste testas, utvecklas och förbättras så att man kan genomföra termohydrauliska designberäkningar. Den här avhandlingen omfattar två olika kylmedel, flytande bly och superkritiskt vatten, som har ett Prandtl-tal som skiljer sig från 1 och kommer att användas i GEN-IV reaktorer. Studien undersöker olika strategier för att modellera turbulens som Large Eddy Simulation (LES) och Reynolds-Averaged Navier-Stokes (RANS) och hur man kan använda dessa strategierna i beräkningar av strömning och värmetransfer i den nya kylvätskan. Undersökningen visar att RANS turbulensmodeller delvis kan förutsäga värmeöverföringen vid en vägg i en ringformad strömningsgeometri. Förbättringar av förutsägelsen ska vara möjlig genom användning av avancerade strategier för turbulensmodellering, t.ex. termiska turbulensmodeller. En stor prestandajämförelse för värmeöverföring i superkritiskt vatten visade att ingen av nuvarande strategier för turbulensmodellering kan förutsäga försämrad värmeöverföring i en 7-stavknippet under superkritiskt tryck. Nya modeller, som omfattar de starka flytkrafterna och den snabba förändringen av den molekulära Prandtl-tal vid väggen som uppstår när vätskan går genom pseudokritiska punkten, måste utvecklas. Avancerade väggfunktioner är en av strategierna som kan ta hänsyn till dessa fenomen. Väggfunktioner kan inte bara hjälpa till att modellera de typer av flöden som behövs utan kan också hjälpa till att sänka beräkningstiden med en eller två tiopotenser. Olika avancerade väggfunktioner i open-source beräkningsverktyget OpenFOAM implementerades och deras prestation i sub- och superkritiska vattenflödar värderades. Baserat på detta rekommenderas Gerasimovs modell för ytterligare utredning. Dessutom läggs olika strategier fram för att utöka modellens validitet till flöde med superkritiskt vatten i sammanband med försämrad och förbättrad värmeöverföring. Kunskap om beteendet av temperatur och hastighet i väggens närhet är viktigt för väggens integritet, detta då väggen även påverkar beteendet. Väggens termiska tröghet påverkar flödets temperatur och hastighet. Dock är ett ännu viktigare problem, som kan uppträda, är att temperaturfluktuationer kan framkalla termisk utmattning i en vägg. Med användning av LES utreds termisk blandning av varmt och kallt vatten i en simplifierad modell av ett styrstavsledrör, inklusive temperaturfältet i styrstaven och ledrörsväggen. Användningen av WALE LES-turbulensmodellen gör det möjligt att utföra beräkningar i den komplexa geometrin, detta eftersom modellen anpassar sig automatiskt till fenomenen nära väggen utan användning av ad-hoc funktioner. LES resultaten för alla värden som är viktiga för att bestämma utmattningsbeteende, som amplitud och frekvens av temperaturfluktuationer i väggens närhet och i väggen själv, är i god överensstämmelse med resultaten från experiment från KTH i samma geometri.Kunskapen som vunnits genom ovannämnda utredningar användes för att optimera den termohydrauliska designen av en liten, pool-typ GEN-IV reaktor som är passivt kyld med flytande bly. Reaktorn är designad som en utbildnings- och träningsreaktor och optimeringen genomfördes med hjälp av 3D CFD. Beräkningarna genomfördes på en fjärdedel av reaktorns hela geometrin. Regioner med små detaljer, som de åtta värmeväxlarna och reaktorns kärna, modellerades genom porösa material. Det visar sig att för att ha en optimal kylning av kärnan, utan att förändra reaktorns globala geometri, måste förhållandet mellan tryckförlust i reaktorkärnan och värmeväxlarna vara nära 1. Detta uppnås genom att designa plattan vid ingången till kärnan så att tillräckligt med bly flödar genom kärnan utan att kväva flödet i denna. Ytterligare en förbättring i reaktorkylningen uppnås genom att reducera värmeförlusten genom väggen som skiljer varm och kall vätska. Detta görs med en strategi som förekommer i gasturbinteknologin, genom att man lägger till ett tunt skikt av termiskt isolerande material på väggen, som reducerar värmeöverföring med ungefär 50%. / <p>QC 20151123</p> / THEMFA / GENIUS / THINS

<p>Among the six GEN-IV reactor concepts recommended by the Gen-IV International Forum, supercritical water-cooled reactors (SCWR) have gained significant interests due to its economic advantage, technology and experience continuity. In the last few years, extensive R&D activities have been launched covering the various aspects of SCWR development, especially in thermal-hydraulic analysis. In Canada, most R&D projects are led by AECL or NRCan.</p> <p>SCWR design and development require the modification of simulation codes used for design and safety demonstration of subcritical water-cooled reactors. This study modifies the subchannel code COBRA-TF, applicable to only subcritical water-cooled reactors, to a new version COBRA-TF-SC, applicable to both supercritical and subcritical water-cooled reactors. Supercritical water property data tables and supercritical water property formulations are implemented. Supercritical water heat transfer and pressure drop correlations are also added. The saturation curve in the subcritical model is extended by introducing a pseudo two-phase region at supercritical pressures to avoid any numerical instabilities consistent with other studies.</p> <p>Some simple fuel bundle experimental data on the flow and temperature distribution are used to evaluate the code. The fuel bundle experiment is simulated with both COBRA-TF-SC and AECL's ASSERT-PV-SC. The COBRA-TF-SC predicted results show good agreement with the experimental data and results obtained from ASSERT-PV-SC, demonstrating good feasibility and accuracy of this code. COBRA-TF-SC is then used to predict the detailed thermalhydraulics behaviour of the 62-element Canadian SCWR fuel bundle design. The advantage of COBRA-TF-SC is that it can accommodate transcritical flow conditions whereas the existing subchannel codes for SCWRs cannot.</p> / Master of Applied Science (MASc)

Fundamental understanding of the behaviour of nuclear fuel has been of great importance. Enhancing this knowledge not only by means of experimental observations, but also via multi-scale modelling is of current interest. The overall goal of this thesis is to understand the impact of atomic interactions on the nuclear fuel material properties. Two major topics are tackled in this thesis. The first topic deals with non-stoichiometry in uranium dioxide (UO2) to be addressed by empirical potential (EP) studies. The second fundamental question to be answered is the effect of the atomic fraction of americium (Am), neptunium (Np) containing uranium (U) and plutonium (Pu) mixed oxide (MOX) on the material properties.UO2 has been the reference fuel for the current fleet of nuclear reactors (Gen-II and Gen-III); it is also considered today by the Gen-IV International Forum for the first cores of the future generation of nuclear reactors on the roadmap towards minor actinide (MA) based fuel technology. The physical properties of UO2 highly depend on material stoichiometry. In particular, oxidation towards hyper stoichiometric UO2 – UO2+x – might be encountered at various stages of the nuclear fuel cycle if oxidative conditions are met; the impact of physical property changes upon stoichiometry should therefore be properly assessed to ensure safe and reliable operations. These physical properties are intimately linked to the arrangement of atomic defects in the crystalline structure. The first paper evaluates the evolution of defect concentration with environment parameters – oxygen partial pressure and temperature by means of a point defect model, with reaction energies being derived from EP based atomic scale simulations. Ultimately, results from the point defect model are discussed, and compared to experimental measurements of stoichiometry dependence on oxygen partial pressure and temperature. Such investigations will allow for future discussions about the solubility of different fission products and dopants in the UO2 matrix at EP level.While the first paper answers the central question regarding the dominating defects in non-stoichiometry in UO2, the focus of the second paper was on the EP prediction of the material properties, notably the lattice parameter of Am, Np containing U and Pu MOX as a function of atomic fractions.The configurational space of a complex U1-y-y’-y’’PuyAmy’Npy’’O2 system, was assessed via Metropolis-Monte Carlo techniques. From the predicted configuration, the relaxed lattice parameter of Am, Np bearing MOX fuel was investigated and compared with available literature data. As a result, a linear behaviour of the lattice parameter as a function of Am, Np content was observed, as expected for an ideal solid solution. These results will allow to support and increase current knowledge on Gen-IV fuel properties, such as melting temperature, for which preliminary results are presented in this thesis, and possibly thermal conductivity in the future. / Doctorat en Sciences de l'ingénieur et technologie / info:eu-repo/semantics/nonPublished

Cílem práce bylo vytvořit základní koncept integrovaného výměníku tepla pro solí chlazený reaktor vyvíjený společností Terrestrial Energy s využitím programu Promex. První kapitola se zabývá historií a současnou situací v oblasti výzkumu malých modulárních reaktorů chlazených fluoridovými solemi. Ve druhé kapitole jsou popsány vlastnosti fluoridových solí a konstrukčních materiálů. Poslední kapitola se zabývá simulací tepelného výměníku pomocí programu Promex, validací modelu, transformací protiproudého výměníku na výměník s U trubkami a vizualizací výměníku s použitím CAD Invetoru.