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1
Sommer, Christopher. "Fuel cycle design and analysis of SABR subrcritical advanced burner reactor /." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2008. http://hdl.handle.net/1853/24720.
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Bopp, Andrew T. "The calculation of fuel bowing reactivity coefficients in a subcritical advanced burner reactor." Thesis, Georgia Institute of Technology, 2013. http://hdl.handle.net/1853/50295.
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The United States' fleet of Light Water Reactors (LWRs) produces a large amount of spent fuel
each year; all of which is presently intended to be stored in a fuel repository for disposal. As these
LWRs continue to operate and more are built to match the increasing demand for electricity, the
required capacity for these repositories grows. Georgia Tech's Subcritical Advanced Burner Reactor
(SABR) has been designed to reduce the capacity requirements for these repositories and thereby
help close the back end of the nuclear fuel cycle by burning the long-lived transuranics in spent nuclear
fuel. SABR's design is based heavily off of the Integral Fast Reactor (IFR).
It is important to understand whether the SABR design retains the passive safety characteristics
of the IFR. A full safety analysis of SABR's transient response to various possible accident scenarios
needs to be performed to determine this. However, before this safety analysis can be performed, it is
imperative to model all components of the reactivity feedback mechanism in SABR. The purpose of
this work is to develop a calculational model for the fuel bowing reactivity coefficients that can be used
in SABR's future safety analysis. This thesis discusses background on fuel bowing and other reactivity
coefficients, the history of the IFR, the design of SABR, describes the method that was developed for
calculating fuel bowing reactivity coefficients and its validation, and presents an example of a fuel
bowing reactivity calculation for SABR.
3
Zakova, Jitka. "Advanced fuels for thermal spectrum reactors." Doctoral thesis, KTH, Reaktorfysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-103085.
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The advanced fuels investigated in this thesis comprise fuels non− conventional in their design/form (TRISO), their composition (high content of plutonium and minor actinides) or their use in a reactor type, in which they have not been used before (e.g. nitride fuel in BWR). These fuels come with a promise of improved characteristics such as safe, high temperature operation, spent fuel transmutation or fuel cycle extension, for which reasons their potentialis worth assessment and investigation. Their possible use also brings about various challenges, out of which some were addressed in this thesis. TRISO particle fuels with their superior retention abilities enable safe, high−temperature operation. Their combination with molten salt in the Advanced High Temperature Reactor (AHTR) concept moreover promises high operating temperature at low pressure, but it requires a careful selection of the cooling salt and the TRISO dimensions to achieve adequate safety characteristic, incl. a negative feedback to voiding. We show that an AHTR cooled with FLiBe may safely operate with both Pu oxide and enriched U oxide fuels. Pu and Minor Actinides (MA) bearing fuels may be used in BWR for transmutation through multirecycling; however, the allowable amounts of Pu and MA are limited due to the degraded feedback to voiding or low reactivity.We showed that the main positive contribution to the void effect in the fuelswith Pu and MA content of around 11 to 15% consist of the decreased thermalcapture probability in Pu-240, Pu-239 and Am-241 and increased fast and resonance fission probability of U-238, Pu239 and Pu-240. The total void worthmoreover increases during multirecycling, limiting the allowable amount ofMA to 2.45% in uranium−based fuels. An alternative, thorium−based fuel allows for 3.45% MA without entering the positive voiding regime at any point of the multirecycling. The increased alpha−heating associated with the use of transmutation fuels, is at level 24−31 W/kgFUEL in the uranium based fuels and 32−37 W/kgFUEL in the thorium−based configurations. The maximum value of the neutron emission, reached in the last cycle, is 1.7·106 n/s/g and 2·106 n/s/g for uranium and for thorium−based fuels, respectively. Replacing the standard UO2 fuel with higher−uranium density UN orUNZrO2 fuels in BWR shows potential for an increase of the in-core fuelresidence time by about 1.4 year. This implies 1.4% higher availability of the plant. With the nitride fuels, the total void worth increases and the efficiency of the control rods and burnable poison deteriorates, but no major neutronics issue has been identified. The use of nitride fuels in the BWR environment is conditioned by their stability in hot steam. Possible methods for stabilizing nitride fuels in water and steam at 300◦ C were suggested in a recent patentapplication.QC 20121004
4
Chand, Rashmi. "Advanced oxidative wastewater treatment using cavitational reactors." Thesis, Abertay University, 2008. https://rke.abertay.ac.uk/en/studentTheses/fdce9629-7b22-43c6-9162-d03848e5df3b.
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This thesis explores various novel ways of treatment of wastewater contaminated by toxic organic pollutants using single and combined advanced oxidative wastewater treatment technologies in conjunction with a variety of acoustic and hydrodynamic cavitational reactors. There have been many reports in the literature on the use of hydroxyl radicals as the core part of AOPs and hence, as the first objective, the amount of hydroxyl radical generation from different acoustic and hydrodynamic cavitational reactors was studied using the potassium iodide dosimeter. The results reveal that optimum concentrations of less toxic chloroalkanes (chloroform and dichloromethane) could be efficient alternatives to carbon tetrachloride for enhancement of hydroxyl radical generation in cavitational reactors. Increasing ultrasonic amplitudes and operating hydrodynamic cavitational pressures lead to higher rates of hydroxyl radical production. Having explored the efficiency of generation of hydroxyl radicals the capacity of the reactors to degrade the model pollutant phenol, via a modified classic Fenton reaction which uses zero valent iron catalysts (instead of iron salts) and hydrogen peroxide under acidic conditions was studied. This process, named the advanced Fenton process (AFP), is the main foundation of the phenolic wastewater treatment reported in this thesis. Phenol degradation was assessed using different frequencies of ultrasound where a comparison between 20, 300 and 520 kHz ultrasonic reactors showed that 300 kHz was by far the most efficient US reactor resulting in 100% phenol removal and 37% total organic carbon (TOC) mineralization in 25 min. The concept of Latent Remediation (LR) was discovered during investigations into innovative approaches towards development of cost/energy-effective methods to treat phenolic wastewater. LR consists of inputting only 15 min of either ultrasound or stirring to the reaction medium, which contains optimised amounts of hydrogen peroxide and iron catalyst, and then the silent-dark AFP phenol degradation was studied over 24 h. The excellent results revealed that >80% TOC mineralization was achieved after this time. It was also found that zero valent copper catalysts were effective for phenol degradation and offered an excellent alternative to iron in the AFP, however toxicity analysis on the 24, 48 and 72 h samples showed that zero valent iron exhibited decreased toxicity when compared to zero valent copper. Conventional granular/powdered activated carbons were replaced with activated carbon cloth and investigations on the potential use of this material for phenol removal/decomposition was studied in detail at different operating pHs (3, 5.5 and 9), temperatures (20, 40 and 80 °C), oxidants (H2O2/O3) in various reactors (pump, shaker and US bath). Another aspect of the AOP application, disinfection of natural waters, was studied employing hydrodynamic cavitation and ozonation in a novel Liquid Whistle Reactor system. Model markers of faecal coliforms, Escherichia coli, were chosen for the study and the combined technologies of hydrodynamic cavitation and stepwise ozonation proved be highly beneficial, resulting in ~ 6 log bacterial reduction revealing 99.9999% disinfection efficiency of the process.
5
Elshahat, Ayah Elsayed. "Enhancing nuclear energy sustainability using advanced nuclear reactors." Thesis, University of Manchester, 2015. https://www.research.manchester.ac.uk/portal/en/theses/enhancing-nuclear-energy-sustainability-using-advanced-nuclear-reactors(2c39b9ca-86a9-446f-8832-ae9469485a2d).html.
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The safety performance of nuclear power reactors is a very important factor in evaluating nuclear energy sustainability. Improving the safety performance of nuclear reactors can enhance nuclear energy sustainability as it will improve the environmental indicator used to evaluate the overall sustainability of nuclear energy. Great interest is given now to advanced nuclear reactors especially those using passive safety components. Investigation of the improvement in nuclear safety using advanced reactors was done by comparing the safety performance of a conventional reactor which uses active safety systems, such as Pressurized Water Reactor (PWR), with an advanced reactor which uses passive safety systems, such as AP1000, during a design basis accident, such as Loss of Coolant Accident (LOCA), using the PCTran as a simulation code. To assess the safety performance of PWR and AP1000, the “Global Safety Index” GSI model was developed by introducing three indicators: probability of accident occurrence, performance of safety system in case of an accident occurrence, and the consequences of the accident. Only the second indicator was considered in this work. A more detailed model for studying the performance of passive safety systems in AP1000 was developed. That was done using SCDAPSIM/RELAP5 code as it is capable of modelling design basis accidents (DBAs) in advanced nuclear reactors.
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Lange, Carsten. "Advanced nonlinear stability analysis of boiling water nuclear reactors." Doctoral thesis, Saechsische Landesbibliothek- Staats- und Universitaetsbibliothek Dresden, 2009. http://nbn-resolving.de/urn:nbn:de:bsz:14-qucosa-24954.
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This thesis is concerned with nonlinear analyses of BWR stability behaviour, contributing to a deeper understanding in this field. Despite negative feedback-coefficients of a BWR, there are operational points (OP) at which oscillatory instabilities occur. So far, a comprehensive and an in-depth understanding of the nonlinear BWR stability behaviour are missing, even though the impact of the significant physical parameters is well known. In particular, this concerns parameter regions in which linear stability indicators, like the asymptotic decay ratio, lose their meaning. Nonlinear stability analyses are usually carried out using integral (system) codes, describing the dynamical system by a system of nonlinear partial differential equations (PDE). One aspect of nonlinear BWR stability analyses is to get an overview about different types of nonlinear stability behaviour and to examine the conditions of their occurrence. For these studies the application of system codes alone is inappropriate. Hence, in the context of this thesis, a novel approach to nonlinear BWR stability analyses, called RAM-ROM method, is developed. In the framework of this approach, system codes and reduced order models (ROM) are used as complementary tools to examine the stability characteristics of fixed points and periodic solutions of the system of nonlinear differential equations, describing the stability behaviour of a BWR loop. The main advantage of a ROM, which is a system of ordinary differential equations (ODE), is the possible coupling with specific methods of the nonlinear dynamics. This method reveals nonlinear phenomena in certain regions of system parameters without the need for solving the system of ROM equations. The stability properties of limit cycles generated in Hopf bifurcation points and the conditions of their occurrence are of particular interest. Finally, the nonlinear phenomena predicted by the ROM will be analysed in more details by the system code. Hence, the thesis is not focused on rendering more precisely linear stability indicators like DR. The objective of the ROM development is to develop a model as simple as possible from the mathematical and numerical point of view, while preserving the physics of the BWR stability behaviour. The ODEs of the ROM are deduced from the PDEs describing the dynamics of a BWR. The system of ODEs includes all spatial effects in an approximated (spatial averaged) manner, e.g. the space-time dependent neutron flux is expanded in terms of a complete set of orthogonal spatial neutron flux modes. In order to simulate the stability characteristics of the in-phase and out-of-phase oscillation mode, it is only necessary to take into account the fundamental mode and the first azimuthal mode. The ROM, originally developed at PSI in collaboration with the University of Illinois (PSI-Illinois-ROM), was upgraded in significant points: • Development and implementation of a new calculation methodology for the mode feedback reactivity coefficients (void and fuel temperature reactivity) • Development and implementation of a recirculation loop model; analysis and discussion of its impact on the in-phase and out-of-phase oscillation mode • Development of a novel physically justified approach for the calculation of the ROM input data • Discussion of the necessity of consideration of the effect of subcooled boiling in an approximate manner With the upgraded ROM, nonlinear BWR stability analyses are performed for three OPs (one for NPP Leibstadt (cycle7), one for NPP Ringhals (cycle14) and one for NPP Brunsbüttel (cycle16) for which measuring data of stability tests are available. In this thesis, the novel approach to nonlinear BWR stability analyses is extensively presented for NPP Leibstadt. In particular, the nonlinear analysis is carried out for an operational point (OP), in which an out-of-phase power oscillation has been observed in the scope of a stability test at the beginning of cycle 7 (KKLc7_rec4). The ROM predicts a saddle-node bifurcation of cycles, occurring in the linear stable region, close to the KKLc7_rec4-OP. This result allows a new interpretation of the stability behaviour around the KKLc7_rec4-OP. The results of this thesis confirm that the RAM-ROM methodology is qualified for nonlinear BWR stability analysesDie vorliegende Dissertation leistet einen Beitrag zum tieferen Verständnis des nichtlinearen Stabilitätsverhaltens von Siedewasserreaktoren (SWR). Trotz der Tatsache, dass in diesem technischen System nur negative innere Rückkopplungskoeffizienten auftreten, können in bestimmten Arbeitspunkten oszillatorische Instabilitäten auftreten. Obwohl relativ gute Kenntnisse über die signifikanten physikalischen Einflussgrößen vorliegen, fehlt bisher ein umfassendes Verständnis des SWR-Stabilitätsverhaltens. Das betrifft insbesondere die Bereiche der Systemparameter, in denen lineare Stabilitätsindikatoren, wie zum Beispiel das asymptotische Decay Ratio (DR), ihren Sinn verlieren. Die nichtlineare Stabilitätsanalyse wird im Allgemeinen mit Systemcodes (nichtlineare partielle Differentialgleichungen, PDG) durchgeführt. Jedoch kann mit Systemcodes kein oder nur ein sehr lückenhafter Überblick über die Typen von nichtlinearen Phänomenen, die in bestimmten System-Parameterbereichen auftreten, erhalten werden. Deshalb wurde im Rahmen der vorliegenden Arbeit eine neuartige Methode (RAM-ROM Methode) zur nichtlinearen SWR-Stabilitätsanalyse erprobt, bei der integrale Systemcodes und sog. vereinfachte SWR-Modelle (ROM) als sich gegenseitig ergänzende Methoden eingesetzt werden, um die Stabilitätseigenschaften von Fixpunkten und periodischen Lösungen (Grenzzyklen) des nichtlinearen Differentialgleichungssystems, welches das Stabilitätsverhalten des SWR beschreibt, zu bestimmen. Das ROM, in denen das dynamische System durch gewöhnliche Differentialgleichungen (GDG) beschrieben wird, kann relativ einfach mit leistungsfähigen Methoden aus der nichtlinearen Dynamik, wie zum Beispiel die semianalytische Bifurkationsanalyse, gekoppelt werden. Mit solchen Verfahren kann, ohne das DG-System explizit lösen zu müssen, ein Überblick über mögliche Typen von stabilen und instabilen oszillatorischen Verhalten des SWR erhalten werden. Insbesondere sind die Stabilitätseigenschaften von Grenzzyklen, die in Hopf-Bifurkationspunkten entstehen, und die Bedingungen, unter denen sie auftreten, von Interesse. Mit dem Systemcode (RAMONA5) werden dann die mit dem ROM vorhergesagten Phänomene in den entsprechenden Parameterbereichen detaillierter untersucht (Validierung des ROM). Die Methodik dient daher nicht der Verfeinerung der Berechnung linearer Stabilitätsindikatoren (wie das DR). Das ROM-Gleichungssystem entsteht aus den PDGs des Systemcodes durch geeignete (nichttriviale) räumliche Mittelung der PDG. Es wird davon ausgegangen, dass die Reduzierung der räumlichen Komplexität die Stabilitätseigenschaften des SWR nicht signifikant verfälschen, da durch geeignete Mittlungsverfahren, räumliche Effekte näherungsweise in den GDGs berücksichtig werden. Beispielsweise wird die raum- und zeitabhängige Neutronenflussdichte nach räumlichen Moden entwickelt, wobei für eine Simulation der Stabilitätseigenschaften der In-phase- und Out-of-Phase-Leistungsoszillationen nur der Fundamentalmode und der erste azimuthale Mode berücksichtigt werden muss. Das ROM, welches ursprünglich am Paul Scherrer Institut (PSI, Schweiz) in Zusammenarbeit mit der Universität Illinois (USA) entwickelt wurde, ist in zwei wesentlichen Punkten erweitert und verbessert worden: • Entwicklung und Implementierung einer neuen Methode zur Berechnung der Rückkopplungsreaktivitäten • Entwicklung und Implementierung eines Modells zur Beschreibung der Rezirkulationsschleife (insbesondere wurde der Einfluss der Rezirkulationsschleife auf den In-Phase-Oszillationszustand und auf den Out-of-Phase-Oszillationszustand untersucht) • Entwicklung einer physikalisch begründeten Methode zur Berechnung der ROM-Inputdaten • Abschätzung des Einflusses des unterkühlten Siedens im Rahmen der ROM-Näherungen Mit dem erweiterten ROM wurden nichtlineare Stabilitätsanalysen für drei Arbeitspunkte (KKW Leibstadt (Zyklus 7) KKW Ringhals (Zyklus 14) und KKW Brunsbüttel (Zyklus 16)), für die Messdaten vorliegen, durchgeführt. In der Dissertationsschrift wird die RAM-ROM Methode ausführlich am Beispiel eines Arbeitspunktes (OP) des KKW Leibstadt (KKLc7_rec4-OP), in dem eine aufklingende regionale Leistungsoszillation bei einem Stabilitätstest gemessen worden ist, demonstriert. Das ROM sagt die Existenz eines Umkehrpunktes (saddle-node bifurcation of cycles, fold-bifurcation) voraus, der sich im linear stabilen Gebiet nahe der Stabilitätsgrenze befindet. Mit diesem ROM-Ergebnis ist eine neue Interpretation der Stabilitätseigenschaften des KKLc7_rec4-OP möglich. Die Resultate der in der Dissertation durchgeführten RAM-ROM Analyse bestätigen, dass das weiterentwickelte ROM für die Analyse des Stabilitätsverhaltens realer Leistungsreaktoren qualifiziert wurde
7
Can, Levent. "Analysis of coolant options for advanced metal cooled nuclear reactors." Thesis, Monterey, Calif. : Naval Postgraduate School, 2006. http://bosun.nps.edu/uhtbin/hyperion.exe/06Dec%5FCan%5FAP.pdf.
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Thesis (M.S. in Applied Physics)--Naval Postgraduate School, December 2006.Thesis Advisor(s): Craig F. Smith "December 2006." Includes bibliographical references (p. 69-70). Also available in print.
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Allen, Kenneth S. "Advanced polymeric burnable poison rod assemblies for pressurized water reactors." [Gainesville, Fla.] : University of Florida, 2003. http://purl.fcla.edu/fcla/etd/UFE0000628.
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Pauli, Lisa M. "Containment building : architecture between the city and advanced nuclear reactors." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/62885.
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Thesis (M. Arch.)--Massachusetts Institute of Technology, Dept. of Architecture, 2011.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Page 127 blank Cataloged from student submitted PDF version of thesis.
Includes bibliographical references (p. 124-126).
Since the inception of nuclear energy research, the element thorium (Th) has been considered the superior fuel for nuclear reactions because of its potency, safety, abundance and reduced waste. Cold War agendas broke from the logic of efficient energy production to establish a nationwide network of reactors designed to enrich uranium fuel for a nuclear arsenal. Contemporary dilemmas of global warming, increasing fuel prices, carbon emissions, and anti-proliferation movements have brought the discussion of clean, safe nuclear power to the forefront of American energy policy; it is no longer tolerable or sustainable to rely on a uranium (U) nuclear network. The architectural typology of nuclear energy has not been addressed in America for 35 years and is one that belies the promise of clean energy's progress through technology and public intervention. Containment Building is an architectural response to nuclear technological advancement that challenges historical separation between nuclear power and the public. It is a self-sustained, thorium-powered nuclear plant sited in and powering New York City. It is a nuclear campus that programatically and urbanistically engages the public and contains radio isotope labs, a nuclear medicine and imaging facility, a food irradiation center, a wellness hotel and spa, an electric taxi charging station, and a plug-in park along the Hudson River waterfront.
by Lisa M. Pauli.
M.Arch.
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Chaleff, Ethan S. "The Radiative Heat Transfer Properties of Molten Salts and Their Relevance to the Design of Advanced Reactors." The Ohio State University, 2016. http://rave.ohiolink.edu/etdc/view?acc_num=osu1480539289737113.
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Sumner, Tyler Scott. "A safety and dynamics analysis of the subcritical advanced burner reactor: SABR." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2008. http://hdl.handle.net/1853/24636.
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Abejón, Orzáez Jorge. "Neutronics analysis of a modified Pebble Bed Advanced High Temperature Reactor." Columbus, Ohio : Ohio State University, 2009. http://rave.ohiolink.edu/etdc/view?acc%5Fnum=osu1238045558.
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Gonzalez, Vargas Jose Angel [Verfasser], and R. [Akademischer Betreuer] Stieglitz. "Advanced Reactor Physics Methods for Transient Analysis of Boiling Water Reactors / Jose Angel Gonzalez Vargas ; Betreuer: R. Stieglitz." Karlsruhe : KIT-Bibliothek, 2017. http://d-nb.info/1148551336/34.
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Demirdöven, Nurettin 1974. "Hydrogen : what fuel cell vehicles and advanced nuclear reactors have in common." Thesis, Massachusetts Institute of Technology, 2005. http://hdl.handle.net/1721.1/32281.
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Thesis (S.M.)--Massachusetts Institute of Technology, Engineering Systems Division, Technology and Policy Program, 2005.Vita.
Includes bibliographical references.
This thesis reports on two technology and policy issues directly related to hydrogen economy. The first issue concentrates on the end-use application of hydrogen as a transportation fuel, and deals with the following question: what is the place of hydrogen fuel cell vehicles among the new, more-efficient advanced vehicle technologies. Our analysis indicates that fuel cell vehicles using hydrogen from fossils fuels offer no significant energy efficiency advantage over hybrid vehicles in urban driving cycle. Therefore, there is a strong justification for federal support for hybrid vehicles that will achieve similar results, quicker. The second issue focuses on another important technology and policy question related to large scale hydrogen production: are there any comparative efficiency, cost and/or political advantages of using an advanced nuclear reactor coupled to a thermochemical conversion plant to produce hydrogen with respect using a conventional nuclear reactor coupled to an electrolysis plant? The results suggest that given the existing technical and cost uncertainties, developing an advanced nuclear reactor technology solely for the use of thermochemical hydrogen production is not good energy (R&D) policy. Electrolysis is a more promising alternative provided a more efficient electrolysis technology can be coupled to an advanced nuclear energy (i.e. electricity) source at a reasonable cost. Therefore, large R&D investment in thermochemical hydrogen production should be balanced with a similar R&D in large scale electrolysis technologies that are relatively easier to deploy and have lower engineering risks.
by Nurettin Demirdöven.
S.M.
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Andrews, Nathan Christopher. "Application of advanced fuel concepts for use in innovative pressurized water reactors." Thesis, Massachusetts Institute of Technology, 2015. http://hdl.handle.net/1721.1/103733.
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Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.Cataloged from PDF version of thesis.
Includes bibliographical references (pages 198-202).
This work addresses several specific knowledge gaps that exist in the use of alternative fuel and cladding combinations in a pressurized water reactor (PWR) environment. In the switch from a UO2 with zirconium-based cladding to any other combination, there is a multitude of questions that need to be answered. This work examines three of these knowledge gaps: (1) the disposition of weapons-grade plutonium in thorium and silicon carbide cladding, (2) economics of accident tolerant fuel (ATF) claddings and (3) breeding of plutonium in uranium nitride fuel. Burning weapons-grade plutonium in a standard pressurized water reactor (PWR) using thoria as a fuel matrix has been compared to using urania. Two cladding options were considered: a 0.76 mm thick silicon carbide ceramic matrix composite (SiC CMC) and 0.57 mm thick standard Zircaloy cladding. A large benefit was found in using thoria compared to urania in terms of plutonium percentage and mass burned. A slightly smaller mass of plutonium is required in a core with SiC CMC cladding, due to its lower neutron absorption compared to Zircaloy. The thorium system was also better from a non-proliferation viewpoint, resulting in less fissile mass at discharge and more fissile mass burned over an assembly's lifetime. A limited safety comparison was made for two reactivity insertion accidents: (1) highest worth rod ejection accident (REA) and (2) main steam line break (MSLB). The MSLB accident demonstrated a safe value for the minimum departure from nucleate boiling ratio. The maximum enthalpy added to the fuel during the REA was also below current regulatory limits for PWRs. This indicates that the more negative moderator temperature coefficients of thoria-plutonia and urania-plutonia fuel, compared to a typical PWR design, were not limiting. For an ATF cladding to replace zirconium alloys, it must be economically viable by having similar fuel cycle costs to today's materials. Four proposed materials are examined: stainless steel (SS), FeCrAl alloy, molybdenum (Mo) and SiC CMC, each having its own development time and costs. The chosen cladding thicknesses were dependent on strength and manufacturing constraints. It was found that all options may end up requiring higher enrichment than zirconium-based claddings for the same fuel cycle length. If the present value of avoiding a reactor accident with a large radioactivity release is estimated using past experience for LWR large accidents and if it is assumed that ATF cladding is able to prevent such release, there is a definite net economic benefit relative to typical Zircaloy cladding only in using SiC, since it only results in a small fuel cycle cost increase. There is only a marginal benefit in using SiC to prevent a core-only loss without radioactivity release (TMI-type) accident and a large loss using metallic ATF concepts. The thermal hydraulic and neutronic feasibility of a nitride fueled pressurized water reactor (PWR) breeder design were examined. Because of its higher fuel density, nitride fuel would be preferable to traditional oxide fuel in attempting to achieve breeding in a PWR. The design chosen uses large hexagonal assemblies with 14 inner seed pin rows and 4 outer blanket pin rows. In this design, reactor grade plutonium of 12.75 wtHM was used as fuel. Nitride was also simulated as being 100% N-15, to limit neutronic penalties and C-14 production. The as specified assembly model only achieved a fissile inventory ratio (FIR) value above 1.0 when the thimble regions were assumed to be voided, which lowers the H/HM ratio in the assembly. This led to FIR values above 1.0 for the oxide, 85% theoretical density nitride (N85) and 95% theoretical density nitride (N95). All were at an FIR of 1.03 at 35 MWd/kgHM. However, the single batch discharge burnup of the voided assembly in MWd/kgHM was 32.2 for N95, 24.5 for N85, while only 15.6 for the oxide.
by Nathan Christopher Andrews.
Ph. D.
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Silva, Rodney Aparecido Busquim e. "Implications of advanced computational methods for reactivity initiated accidents in nuclear reactors." Universidade de São Paulo, 2015. http://www.teses.usp.br/teses/disponiveis/3/3139/tde-20072016-142605/.
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Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). The three cores have the same general design and control rod assembly (CRA) positions, and the ejected CRA has similar worth and at the same rod ejection pace. The CRE is assessed under both hot zero power (HZP) and hot full power (HFP) conditions. The analyses indicate that RIA modeling and simulation should be carried out through a systematic coding and configuration approaches, otherwise the results will not capture the true transient behavior of the core under analysis. The simulation of one code depends on the appropriate configuration of parameters generated by the other code and on the correct determination of the TH/NK mapping weight factors for the various mesh regions in each of the models. From the design point of view, the standalone codes predict milder magnitude of power and reactivity increase compared to the coupled P3D/R5 simulation. The magnitudes of reduced peak power and reactivity become larger as the core size shrinks. The HFP simulation shows that the three NPPs have the same transient peak value, but the post-transient steady power is lower for a smaller core. The HZP analysis indicates that the transient peak is lower for the smaller core, but the post-transient power occurs at the same level. The three-dimensional (3D) power distributions are different among the HFP and HZP cases, but do not depend on the size of the core. The results indicate: i) HFP: core power increases in the area surrounding the ejected rod/bank assembly, and this increase becomes lower as the NPPs shrinks however, the power is well-distributed after the transient; and ii) HZP: the area surrounding the CRA stays hotter, but the 3D peak assembly factor becomes lower, during and after the transients, as the NPPs shrinks. These features confirm that the smaller cores yield a safer response to a given inserted reactivity compared to larger cores. A stochastic extended Kalman filter (EKF) algorithm is implemented to estimate the reactivity based on the reactor power profile, after the addition of random noise. The inverse point kinetics (IPK) deterministic method is also implemented and the results of the application of EKF and IPK are compared to the P3D/R5 simulation. The following sophisticated strategies made the EKF algorithm robust and accurate: the system is modeled by a set of continuous time nonlinear stochastic differential equations; the code uses a time step directly based on the power measured and applies that to the model for online discretization and linearization; filter tuning goes automatically up from the first time step; and the state noise covariance matrix is updated online at each time step. It was found that the IPK reactivity has higher noise content compared to the EKF reactivity for all cases. Thus, the EKF presents superior and more accurate results. Furthermore, under a small reactivity insertion, the IPK reactivity varies widely from positive to negative values: this variation is not observed within the EKF. A sensitivity analysis for three distinct standard deviation (SD) noise measurements suggests that EKF is superior to IPK method, independent of the noise load magnitude. As the noise content increases, the error between the IPK and P3D/R5 reactivity also increases. A sensitivity analysis for five distinct carry-over effects of different random noise loads indicates that the random addition of different noise loads to the reactor power does not change the overall performance of both algorithms.Este trabalho aplica métodos computacionais avançados para simular a ejeção de barras de controle (CRE) em uma planta térmica nuclear (NPP). São avaliados o impacto da ocorrência de acidentes iniciados por reatividade (RIAs) na reatividade total, na distribuição da potência em três dimensões (3D) e na determinação da reatividade. As ferramentas utilizadas são: o código termo-hidráulico (TH) RELAP5 (R5), o código neutrônico (NK) PARCS (P3D), a versão acoplada P3D/R5, e o ambiente computacional MATLAB. Este estudo considera três reatores nucleares de diferentes tamanhos: NPP1 (2772 MWT); NPP2 (530 MWt); e NPP3 (1061 MWt). Os três núcleos possuem projeto similar e idêntica posição dos grupos das barras de controle (CRA), além do mesmo valor de reatividade diferencial das CRA ejetadas e idêntica velocidade de ejeção. A ocorrência da CRE é avaliada sob condições de hot zero power (HZP) e de hot full power (HFP). As análises indicam que a modelagem e a simulação de RIAs devem ser realizadas sistematicamente, caso contrário os resultados não irão refletir o comportamento em regime transitório do núcleo. A simulação de um modelo em um código depende da apropriada configuração de parâmetros gerados pelo outro código e da determinação adequada do mapeamento TH/NK para as várias malhas dos modelos. Do ponto de vista de projeto, a utilização de códigos independentes resulta em cálculos de potência e reatividade conservadores em comparação com os resultados utilizando-se P3D/R5. Os picos de potência e de reatividade são menores à medida que o núcleo encolhe. A simulação em condições de HFP resulta em valores de pico de potência similares durante transitório para as três NPPs, mas a potência de pós-transitórios é menor para o menor núcleo. A análise em condições de HZP também indica que o valor máximo durante o transitório é menor para o menor núcleo, mas o pós-transitórios ocorre aos mesmos níveis de potência das demais NPPS. A distribuição de potência em 3D também apresenta resultados distintos para condições de HFP e HZP, mas tais resultados são independentes do tamanho do núcleo: i) HFP: há um aumento da potência do núcleo em torno da CRE, mas tal comportamento diminui para núcleos menores - no entanto, a potência é bem distribuída após o transitório; e ii) HZP: há aumento de potência na área do CRE, mas o pico de potência em 3D é menor durante e depois dos transitórios para núcleos menores. Tais características indicam que os núcleos menores respondem de forma mais segura quando da inserção de reatividade em comparação a reatores de maiores dimensões. O método estocástico de filtragem de Kalman estendido (EKF) foi codificado para estimar a reatividade com base no perfil de potência da NPP, após a adição de ruído aleatório. O método determinístico da cinética pontual inversa (IPK) também foi implementado e os resultados da aplicação dos algoritmos do EKF e IPK foram comparados com os resultados da simulação do P3D/R5. As seguintes estratégias, implementadas neste trabalho, possibilitaram a aplicação robusta e precisa do EKF: o sistema foi modelado por um conjunto de equações diferenciais não-lineares estocásticas de tempo contínuo; o algoritmo obtém o passo de tempo diretamente da potência medida e aplica-o ao modelo para a discretização e linearização online; o ajuste do filtro ocorre automaticamente a partir do primeiro passo de tempo; e a matriz de covariância do ruído no estado é atualizada online. Verificou-se que a reatividade calculada pelo método IPK possui maior nível de ruído quando comparada ao EKF para todos os casos estudados. Portanto, o EKF apresenta resultados superiores e mais precisos. Além disso, sob uma pequena inserção de reatividade, a reatividade calculada pelo método IPK varia consideravelmente de valores positivos para negativos: esta variação não é observada com o EKF. Uma análise de sensibilidade para três desvios padrão (SD) sugere que o algoritmo EKF é superior ao método IPK, independente da magnitude do ruído. Com o aumento da magnitude do ruído, o erro entre as reatividades calculadas pelo IPK e pelo P3D/R5 aumenta. A análise de sensibilidade para cinco ruídos aleatórios indica que a adição de ruído na potência do reator não altera o desempenho global de ambos os algoritmos.
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Andrews, Michael Robert. "The interaction of deposition promoters with AGR fuel cladding surfaces." Thesis, University of Newcastle Upon Tyne, 1998. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.244466.
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Hanson, Gregory Richard. "Microwave reflectometry on the advanced toroidal facility to measure density fluctuations and their radial correlation lengths." Diss., Georgia Institute of Technology, 1991. http://hdl.handle.net/1853/17280.
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MACEDO, LUIZ A. "Atuação de um sistema passivo de remoção de calor de emergência de reatores avançados em escoamento bifásico e com alta concentração de não-condensáveis." reponame:Repositório Institucional do IPEN, 2008. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11622.
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Made available in DSpace on 2014-10-09T12:53:58Z (GMT). No. of bitstreams: 0Made available in DSpace on 2014-10-09T14:09:27Z (GMT). No. of bitstreams: 0
Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
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Eul, Ryan C. "The impact of passive safety systems on desirability of advanced light water reactors." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41267.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering; and, (S.M.)--Massachusetts Institute of Technology, Engineering Systems Division, Technology and Policy Program, 2006.Includes bibliographical references (leaves 121-123).
This work investigates whether the advanced light water reactor designs with passive safety systems are more desirable than advanced reactor designs with active safety systems from the point of view of uncertainty in the performance of safety systems as well as the economic implications of the passive safety systems. Two advanced pressurized water reactors and two advanced boiling water reactors, one representing passive reactors and the other active reactors for each type of coolant, are compared in terms of operation and responses to accidents as reported by the vendors. Considering a simplified decay heat removal system that utilizes an isolation condenser for decay heat removal, the uncertainty in the main parameters affecting the system performance upon a reactor isolation accident is characterized when the system is to rely on natural convection and when it is to rely on a pump to remove the core heat. It is found that the passive system is less certain in its performance if the pump of the active system is tested at least once every five months. In addition, a cost model is used to evaluate the economic differences and benefits between the active and passive reactors. It is found that while the passive systems could have the benefit of fewer components to inspect and maintain during operation, they do suffer from a larger uncertainty about the time that would be required for their licensing due to more limited data on the reliability of their operation. Finally, a survey among nuclear energy experts with a variety of affiliations was conducted to determine the current professional attitude towards these two competing nuclear design options. The results of the survey show that reactors with passive safety systems are more desirable among the surveyed expert groups. The perceived advantages of passive systems are an increase in plant safety with a decrease in cost.
by Ryan C. Eul.
S.M.
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Ford, Michael J. "Studies in Nuclear Energy: Low Risk and Low Carbon." Research Showcase @ CMU, 2017. http://repository.cmu.edu/dissertations/872.
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The amount of greenhouse gas emissions mitigation required to prevent the most dramatic climate change scenarios postulated in the 2014 IPCC Synthesis Report is substantial. Prior analyses have examined the potential for nuclear energy to play a role in decarbonizing the energy sector, one of the largest contributors to emissions worldwide. However, advanced, non-light water reactors, while often touted as a viable alternative for development, have languished. Large light water development projects have a repeated history of extended construction timelines, re-work delays, and significant capital risk. With few exceptions, large-scale nuclear projects have demonstrated neither affordability nor economic competitiveness, and are not well suited to nations with smaller energy grids, or to replace fossil generation in the industrial process heat sector. If nuclear power is to play a role in decarbonization, new policy and technical solutions will be needed. In this manuscript, we examine key aspects of past performance across the nuclear enterprise and explore the future potential of nuclear energy worldwide, focusing on policy and technical solutions that may be needed to move nuclear power forward as a part of a low-carbon energy future. We do so first at a high level, examining the history of nuclear power research and development in the United States, the nation that historically has led the way in the development of this generating technology. A significant portion of our analysis is focused on new developments in this technology – advanced non-light water reactors and small modular reactors. We find that while there are promising technical solutions available, improved funding and focus in research and new models of deployment may be needed if nuclear is to play a continuing or future role. We also find that in examining potential new markets for the technology, a continuing focus on institutional readiness is critical.
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Ulmer, Richard Marion. "Benchmark description of an advanced burner test reactor and verification of COMET for whole core criticality analysis in fast reactors." Thesis, Georgia Institute of Technology, 2014. http://hdl.handle.net/1853/52222.
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This work developed a stylized three dimensional benchmark problem based on Argonne National Laboratory's conceptual Advanced Burner Test Reactor design. This reactor is a sodium cooled fast reactor designed to burn recycled fuel to generate power while transmuting long term waste. The specification includes heterogeneity at both the assembly and core levels while the geometry and material compositions are both fully described. After developing the benchmark, 15 group cross sections were developed so that it could be used for transport code method verification. Using the aforementioned benchmark and 15 group cross sections, the Coarse-Mesh Transport Method (COMET) code was compared to Monte Carlo code MCNP5 (MCNP).
Results were generated for three separate core cases: control rods out, near critical, and control rods in. The cross section groups developed do not compare favorably to the continuous energy model; however, the primary goal of these cross sections is to provide a common set of approachable cross sections that are widely usable for numerical methods development benchmarking.
Eigenvalue comparison results for MCNP vs. COMET are strong, with two of the models within one standard deviation and the third model within one and a third standard deviation. The fission density results are highly accurate with a pin fission density average of less than 0.5% for each model.
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Sukjai, Yanin. "Silicon carbide performance as cladding for advanced uranium and thorium fuels for light water reactors." Thesis, Massachusetts Institute of Technology, 2014. http://hdl.handle.net/1721.1/87496.
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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2014.Cataloged from PDF version of thesis.
Includes bibliographical references (pages 285-288).
There has been an ongoing interest in replacing the fuel cladding zirconium-based alloys by other materials to reduce if not eliminate the autocatalytic and exothermic chemical reaction with water and steam at above 1,200 °C. The search for an accident tolerant cladding intensified after the Fukushima events of 2011. Silicon carbide (SiC) possesses several desirable characteristics as fuel cladding in light water reactors (LWRs). Compared to zirconium, SiC has higher melting point, higher strength at elevated temperature, and better dimensional stability when exposed to radiation, as well as lower thermal expansion, creep rate, and neutron absorption cross-section. However, under irradiation, the thermal conductivity of SiC is degraded considerably. Furthermore, lack of creep down towards the fuel causes the fuel-cladding gap and gap thermal resistance to stay relatively large during in-core service. This leads to higher fuel temperature during irradiation. In order to reduce the high fuel temperature during operation, the following fuel design options were investigated in this study: using beryllium oxide (BeO) additive to enhance fuel thermal conductivity, changing the gap bond material from helium to lead-bismuth eutectic (LBE) and adding a central void in the fuel pellet. In addition, the consequences of using thorium oxide (ThO₂) as host matrix for plutonium oxide (PuO₂) were covered. The effects of cladding thickness on fuel performance were also analyzed. A steady-state fuel performance modeling code, FRAPCON 3.4, was used as a primary tool in this study. Since the official version of the code does not include the options mentioned above, modifications of the source code were necessary. All of these options have been modeled and integrated into a single version of the code called FRAPCON 3.4-MIT. Moreover, material properties including thermal conductivity, swelling rate, and helium production/release rate of BeO have been updated. Material properties of ThO₂ have been added to study performance of ThO₂-PuO₂ . This modified code was used to study the thermo-mechanical behavior of the most limiting fuel rod with SiC cladding, and explore the possibility to improve the fuel performance with various design options. The fuel rod designs and operating conditions of a 4-loop Westinghouse pressurized water reactors (PWR) and Babcock and Wilcox (B&W) mPower small modular reactors (SMR) were reactors (PWR) and Babcock and Wilcox (B&W) mPower small modular reactors (SMR) were chosen as representatives of conventional PWRs and upcoming SMRs, respectively. Sensitivity analyses on initial helium gap pressure, linear heat generation rate (LHGR) history, and peak rod assumptions have been performed. The results suggest that, because of its lower thermal conductivity, SiC is more sensitive to changes in these parameters than zirconium alloys. For a low-conducting material like SiC, an increase in cladding thickness plays a significant role in fuel performance. With a thicker cladding (from 0.57 to 0.89 mm), the temperature drop across the cladding increases, which makes the fuel temperature higher than that with the thin cladding. Reduction of fuel volume to accommodate the thicker cladding also causes negative impact on fuel performance. However, if the extra volume of the cladding replaces some coolant, the reduced coolant fraction design (RCF) has superior performance to the decreased fuel volume fraction design. In general, the most effective fuel temperature improvement option appears to be the option of mixing beryllium oxide into the fuel. This method outperforms others because it improves the overall thermal conductivity and reduces the overall temperature of the fuel. With lower fuel temperature, fission gas release and eventually plenum pressure -- one of the most life-limiting factor for SiC -- can be lowered.
by Yanin Sukjai.
S.M.
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PARSONS, MICHAEL E. "A STUDY OF AEROBIC METHANOL ADDITION IN DENITRIFYING SEQUENCING BATCH REACTORS." University of Cincinnati / OhioLINK, 2007. http://rave.ohiolink.edu/etdc/view?acc_num=ucin1172761209.
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Resende, Jose Wilson. "Interaction between controlled reactors and converters : a harmonic analysis." Thesis, University of Aberdeen, 1986. http://digitool.abdn.ac.uk/R?func=search-advanced-go&find_code1=WSN&request1=AAIU367868.
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This thesis presents the development of a generalised computer program to calculate harmonic currents and voltages in six and twelve-pulse thyristor controlled reactors under non-ideal conditions. Thyristor controlled reactors are a relatively new source of harmonic distortion in power systems. The steady state characteristic harmonics are well known. Other non-characteristic harmonics can, however, be generated. A detailed representation is therefore necessary. Apart from the most common non-ideal conditions, such as voltage, impedance and firing pulse unbalances, this work allows voltage harmonic distortions, two firing pulse control methods, the effect of the feedback control in the equally spaced firing pulse control and the effect of the step-down transformer saturation. The effect of the a.c. system impedance, filters and capacitor banks is also included. Four different models of filters were implemented. With non-infinite a.c. systems, the harmonic currents generated are not totally absorbed by the filters. The remaining distortion may affect the main busbar voltages. Therefore, an iterative method was adopted in which the distorted voltages calculated at the end of one iteration are used to calculate the new currents and voltages. The process is repeated until convergence is reached. Several cases were then studied using this program which was then joined to an existing steady-state converter harmonic program. For instance, the need for a more complete representation of controlled reactors, converters and a.c. system network is illustrated. This study begins considering an hvdc station under ideal conditions which are then gradually moved towards more real conditions. The influence of the a.c. system representation in harmonic studies is also discussed. This analysis also compares the performance of two filter designs, namely the tuned and the damped filters. A study of harmonic magnification in the presence of a.c. and d.c. resonances is also included. The harmonic calculations program presented in this thesis is able to study so many conditions of operation of converters and/or thyristor controlled reactors that it is impractical to show all the possible cases. For instance, filters and capacitor banks can be installed at the converter busbar or at any controlled reactor busbar. Furthermore, the three-phase calculation approach allows studies in which some abnormal operation, such as the absence of a filter branch or capacitor bank at one phase, can be observed.
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White, Robert Patrick. "Pathways and frameworks for the licensing and regulation of advanced nuclear reactors in the United States." Thesis, Massachusetts Institute of Technology, 2019. https://hdl.handle.net/1721.1/121714.
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This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2019
Cataloged student-submitted from PDF version of thesis.
Includes bibliographical references (pages 93-96).
Nuclear regulation lies at the nexus of public policy, business, and engineering. While current regulatory processes have become fairly effective for the regulation of existing nuclear power plants, the use of prescriptive technology-specific requirements may present a challenge for the licensing and deployment of advanced nuclear power plants. These advanced nuclear power plants can utilize passive systems, new fuel forms or coolants, or other new design features to accomplish their safety and security functions. Advanced reactors may not comply with existing requirements for nuclear power plant licensing due to their departure from the design philosophies and reactor technologies used in existing nuclear power plants. The challenge of licensing advanced nuclear power plants using existing regulatory requirements could increase the time and costs associated with licensing new plants, and jeopardize the commercial viability of the industry.
In this work, the principles of nuclear regulation are presented and discussed in the historical context and evolution of licensing and regulating nuclear power plants in the United States and abroad. The current licensing system for commercial nuclear power plants in the United States is then discussed in detail. Existing and proposed processes for advanced nuclear reactor licensing are presented, and challenges of advanced reactor licensing are discussed. Finally, a methodology is developed and presented for selecting an appropriate licensing pathway for a proposed advanced reactor. Answers to ten characterization questions are used to recommend which existing regulatory tools and pathways available in the United States could enable the most effective licensing of an advanced reactor. The proposed methodology could prove a valuable tool for companies seeking to develop new reactor technologies while minimizing licensing costs, schedules, and related uncertainties.
The methodology is accessible for users with limited experience with (or knowledge of) existing nuclear regulations. The recommendations for policy changes and the advanced reactor pathway selection methodology presented in this work could enable the more efficient licensing and deployment of advanced nuclear reactors.
by Robert Patrick White.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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Yamamoto, Akio. "Study on Advanced In-Core Fuel Management for Pressurized Water Reactors Using Loading Pattern Optimization Methods." Kyoto University, 1998. http://hdl.handle.net/2433/156982.
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本文データは平成22年度国立国会図書館の学位論文(博士)のデジタル化実施により作成された画像ファイルを基にpdf変換したものであるKyoto University (京都大学)
0048
新制・課程博士
博士(エネルギー科学)
甲第7440号
博第3号
新制||エネ||1(附属図書館)
UT51-98-G369
京都大学大学院エネルギー科学研究科エネルギー社会・環境科学専攻
(主査)教授 神田 啓治, 教授 吉川 榮和, 教授 代谷 誠治
学位規則第4条第1項該当
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Kingsbury, Christopher W. "Fuel cycle cost and fabrication model for fluoride-salt high-temperature reactor (FHR) "Plank" fuel design optimization." Thesis, Georgia Institute of Technology, 2015. http://hdl.handle.net/1853/54337.
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The fluoride-salt-cooled high-temperature reactor (FHR) is a novel reactor design benefitting from passive safety features, high operating temperatures with corresponding high conversion efficiency, to name a few key features. The fuel is a layered graphite plank configuration containing enriched uranium oxycarbide (UCO) tri-structural isotropic (TRISO) fuel particles. Fuel cycle cost (FCC) models have been used to analyze and optimize fuel plate thicknesses, enrichment, and packing fraction as well as to gauge the economic competitiveness of this reactor design.
Since the development of the initial FCC model, many corrections and modifications have been identified that will make the model more accurate. These modifications relate to corrections made to the neutronic simulations and the need for a more accurate fabrication costs estimate. The former pertains to a MC Dancoff factor that corrects for fuel particle neutron shadowing that occurs for double-heterogeneous fuels in multi-group calculations. The latter involves a detailed look at the fuel fabrication process to properly account for material, manufacturing, and quality assurance cost components and how they relate to the heavy metal loading in a FHR fuel plank.
It was found that the fabrication cost may be a more significant portion of the total FCC than was initially attributed. TRISO manufacturing cost and heavy metal loading via packing fraction were key factors in total fabrication cost. This study evaluated how much neutronic and fabrication cost corrections can change the FCC model, optimum fuel element parameters, and the economic feasibility of the reactor design.
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Winter, Dominik [Verfasser]. "Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts / Dominik Winter." Aachen : Hochschulbibliothek der Rheinisch-Westfälischen Technischen Hochschule Aachen, 2014. http://d-nb.info/1052303331/34.
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Salgado, Ricardo Manuel Nunes. "The removal of xenobiotic compounds from wastewater through the use of biological processes and advanced oxidation technologies." Doctoral thesis, Faculdade de Ciências e Tecnologia, 2011. http://hdl.handle.net/10362/6918.
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Dissertação apresentada para obtenção do Grau de Doutor em Engenharia Química e Bioquímica pela Universidade Nova de Lisboa, Faculdade de Ciências e TecnologiaFCT/MCTES projects PTDC/AMB/65702/2006 and SFRH/PROTEC/49449/2009 and SFRH/BPD/30800/2006 ; COST Action 636
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Aziz, Norashid. "Dynamic optimisation and control of batch reactors : development of a general model for batch reactors, dynamic optimisation of batch reactors under a variety of objectives and constraints and on-line tracking of optimal policies using different types of advanced control strategies." Thesis, University of Bradford, 2001. http://hdl.handle.net/10454/4402.
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Batch reactor is an essential unit operation in almost all batch-processing industries. Different types of reaction schemes (such as series, parallel and complex) and different order of model complexity (short-cut, detailed, etc. ) result in different sets of model equations and computer coding of all possible sets of model equations is cumbersome and time consuming. In this work, therefore, a general computer program (GBRM - General Batch Reactor Model) is developed to generate all possible sets of equations automatically and as required. GBRM is tested for different types of reaction schemes and for different order of model complexity and its flexibility is demonstrated. The above GBRM computer program is lodged with Dr. I. M. Mujtaba. One of the challenges in batch reactors is to ensure desired performance of individual batch reactor operations. Depending on the requirement and the objective of the process, optimisation in batch reactors leads to different types of optimisation problems such as maximum conversion, minimum time and maximum profit problem. The reactor temperature, jacket temperature and jacket flow rate are the main control variables governing the process and these are optimised to ensure maximum benefit. In this work, an extensive study on mainly conventional batch reactor optimisation is carried out using GBRM coupled with efficient DAEs (Differential and Algebraic Equations) solver, CVP (Control Vector Parameterisation) technique and SQP (Successive Quadratic Programming) based optimisation technique. The safety, environment and product quality issues are embedded in the optimisation problem formulations in terms of constraints. A new approach for solving optimisation problem with safety constraint is introduced. All types of optimisation problems mentioned above are solved off-line, which results to optimal operating policies. The off-line optimal operating policies obtained above are then implemented as set points to be tracked on-line and various types of advanced controllers are designed for this purpose. Both constant and dynamic set points tracking are considered in designing the controllers. Here, neural networks are used in designing Direct Inverse and Inverse-Model-Based Control (IMBC) strategies. In addition, the Generic Model Control (GMC) coupled with on-line neural network heat release estimator (GMC-NN) is also designed to track the optimal set points. For comparison purpose, conventional Dual Mode (DM) strategy with PI and PID controllers is also designed. Robustness tests for all types of controllers are carried out to find the best controller. The results demonstrate the robustness of GMC-NN controller and promise neural controllers as potential robust controllers for future. Finally, an integrated framework (BATCH REACT) for modelling, simulation, optimisation and control of batch reactors is proposed.
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Tingle, Christopher P. "Optimal fuel management of CANDU reactors at approach to refuelling equilibrium, an investigation into the optimal use of advanced fuels in the CANDU-6 reactor." Thesis, National Library of Canada = Bibliothèque nationale du Canada, 1998. http://www.collectionscanada.ca/obj/s4/f2/dsk1/tape11/PQDD_0001/MQ44922.pdf.
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Lapins, Janis [Verfasser], and Eckart [Akademischer Betreuer] Laurien. "An advanced three-dimensional simulation system for safety analysis of gas cooled reactors / Janis Lapins. Betreuer: Eckart Laurien." Stuttgart : Universitätsbibliothek der Universität Stuttgart, 2016. http://d-nb.info/1084003309/34.
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Hannink, Ryan Christopher. "Investigation of the use of nanofluids to enhance the In-Vessel Retention capabilities of Advanced Light Water Reactors." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/41314.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering; and, (S.M.)--Massachusetts Institute of Technology, Engineering Systems Division, Technology and Policy Program, 2007.Includes bibliographical references (p. 126-130).
Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, employed in Advanced Light Water Reactors, was investigated. A model simulating the two-phase flow and heat transfer on the reactor vessel outer surface quantified the increase in decay power that can be removed using a nanofluid, predicting that the use of a nanofluid will allow a stable operating power ~40% greater than the power allowable using water to be achieved, while holding the Minimum Departure from Nucleate Boiling Ratio (MDNBR) constant. A nanofluid injection system that would take advantage of the enhanced CHF properties of the nanofluid in order to provide a higher safety margin than the current IVR strategy or, for given margin, enable IVR at higher core power, is proposed. A risk-informed analysis has revealed that this injection system has a reasonably high success probability of 0.99, comparable to the success probability without the injection system. Potential regulatory, environmental, and health risk issues were analyzed, and it was concluded that the current regulatory regimes are adequate for ensuring that the implementation of nanofluids in IVR will not endanger public health and safety. However, experimental verification of nanofluid CHF enhancement at prototypical IVR conditions and periodic nanofluid property testing as a surveillance requirement are needed to reduce the key uncertainties related to nanofluid performance. Finally, a periodic review of the health and environmental risks of nanofluids and, if necessary,follow-up research are ecommended to ensure the health of the public and environment.
by Ryan Christopher Hannink.
S.M.
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Mezohegyi, Gergo. "Catalytic azo dye reduction in advanced anaerobic bioreactors." Doctoral thesis, Universitat Rovira i Virgili, 2010. http://hdl.handle.net/10803/8593.
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En un reactor anaeróbico de lecho empacado y de flujo ascendente con carbón activado (AC) biológico se obtuvieron altas velocidades de conversión del colorante azoico Acid Orange 7 a tiempos espaciales muy cortos, hasta 99% en 2.0 min. Tanto el área superficial específica como la conductividad electrónica del AC contribuyeron a las mayores velocidades de reducción. La agitación en el lecho de carbón produjo un incremento de la bioconversión del colorante. Se estableció un modelo cinético de decoloración que implica catálisis heterogénea y bioreducción. La biodegradabilidad anaeróbica de un colorante azoico en el sistema reactivo agitado pudo ser predicha a partir de su potencial de reducción. Las velocidades de decoloración fueron significativamente influenciadas por las propiedades texturales del AC y moderadamente afectadas por su química superficial. Este bioreactor catalítico parece ser una alternativa atractiva para la mejora económica de las tecnologías de tratamiento de aguas residuales textiles y de colorantes.In an anaerobic upflow packed-bed reactor with biological activated carbon (AC), high azo dye Acid Orange 7 conversion rates were achieved during very short space times up to 99% in 2.0 min. Both electron conductivity and specific surface area of AC contributed to higher reduction rates. The application of stirring in the carbon bed resulted in an increase of dye bioconversion. A decolourisation model was developed involving both heterogeneous catalysis and bioreduction. The anaerobic biodegradability of an azo dye could be predicted by its reduction potential in the stirred reactor system. The decolourisation rates were found to be significantly influenced by the textural properties of AC and moderately affected by its surface chemistry. This catalytic bioreactor system seems to be an attractive alternative for economically improving textile/dye wastewater technologies.
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Macedo, Luiz Alberto. "Controle de Sistemas Passivos de Resfriamento de Emergência de Reatores Nucleares por Meio de Linhas de Desvio." Universidade de São Paulo, 2001. http://www.teses.usp.br/teses/disponiveis/85/85133/tde-29052003-092313/.
Full textAbstract:
Neste trabalho são apresentados resultados experimentais, de um circuito operando em circulação natural, que permitem analisar o comportamento de um sistema de resfriamento de emergência quando é aberta uma linha de desvio entre a fonte quente e a fonte fria. O trabalho tem ainda a importância de documentar os testes de caracterização hidráulica do circuito experimental, fornecendo inclusive os fatores de perda de pressão específicos para o circuito. Observou-se que, para uma mesma potência, quando é aberta a linha de desvio, a temperatura na saída da fonte quente aumenta substancialmente. Esse aumento ocorre porque a vazão através do aquecedor diminui. A vazão através do trocador de calor (fonte fria) aumenta ligeiramente, sendo sempre a soma das vazões na linha de desvio e no aquecedor. O trabalho mostra ainda que a posição de conexão da linha de desvio com a perna quente determina o sentido de escoamento, podendo ocorrer a inversão a partir de uma determinada cota. Para comprovar a possibilidade de simulação precisa dos experimentos foi ainda desenvolvido um modelo numérico das equações de conservação, utilizando o programa Engineering Equation Solver (EES). Esse modelo foi utilizado para reproduzir os experimentos de circulação natural pelo circuito externo.This work presents experimental results of a circuit when operating in natural circulation. These results allow to analyze the behavior of an emergency core cooling system when a bypass line that connects the hot source with the cold source is opened. This work also reports the hydraulic characterization of the experimental loop, given geometric and hydraulic data including experimental friction factors specific to this circuit. It was observed that, to a fixed thermal power, when the bypass line is opened, the heater outlet temperature increases. This temperature increase is due to the decrease in the flow rate through the heater. The heat exchanger's flow rate is subjected to a small increase. This flow rate is the sum of the bypass line and heater mass flow rates. This work also shows that the vertical position of the connection of the bypass line in the hot-leg determines the flow direction in the bypass line. If the bypass line connection is in the lowest position, the flow is from the cold to the hot-leg. If the bypass connection is in the highest position, the flow is from the hot to the cold-leg. A numerical model used to evaluate friction factors and heat transfer coefficients influence was developed. It was used to confirm the possibility of precise experiments simulation. The conservation equations are solved using Engineering Equation Solver (EES), a thermal hydraulics analysis tool. The model was adjusted with natural circulation experimental data and was tested with results of natural circulation without bypass lines.
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Lange, Carsten [Verfasser], Antonio [Akademischer Betreuer] Hurtado, and Rizwan-uddin [Akademischer Betreuer]. "Advanced nonlinear stability analysis of boiling water nuclear reactors / Carsten Lange. Gutachter: Antonio Hurtado ; Rizwan-uddin. Betreuer: Antonio Hurtado." Dresden : Saechsische Landesbibliothek- Staats- und Universitaetsbibliothek Dresden, 2009. http://d-nb.info/1063279836/34.
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Bremer, Jens Verfasser], and Kai [Gutachter] [Sundmacher. "Advanced operating strategies for non-isothermal fixed-bed reactors exemplified for CO 2 methanation / Jens Bremer ; Gutachter: Kai Sundmacher." Magdeburg : Universitätsbibliothek Otto-von-Guericke-Universität, 2020. http://d-nb.info/1232911909/34.
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Hiezl, Zoltan. "Processing and microstructural characterisation of UO2-based simulated spent nuclear fuel ceramics for the UK's advanced gas-cooled reactors." Thesis, Imperial College London, 2015. http://hdl.handle.net/10044/1/28254.
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Processing and characterisation of depleted UO2-based Simulated Spent Nuclear Fuel (SIMFuel), which aims to replicate both chemistry and microstructure of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR) after a prolonged cooling time is described in this thesis. Thirteen fission product surrogates were blended with depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/t U. Pure depleted UO2 pellets were also investigated as a reference. The fission product (FP) inventory was calculated using the FISPIN code provided by the UK National Nuclear Laboratory. Experiments were conducted in two phases, during which SIMFuel pellets were sintered for 5 and 12 h at 1730 °C in reducing atmosphere. Some pellets were also heat-treated to simulate microstructural changes in SNF while in the reactor. SIMFuel pellets were up to 92% dense, with grain sizes between 1.5 μm and 5 μm and porosity 4% and 10%. Undoped reference pellet density was ~96.5%, with grain size of 10.3 ± 3.0 μm and ~4.5 area % porosity. Heat treatment of the UO2 samples increased grain size by ~50%, while little change occurred in the doped samples. The chemistry of the various FPs was reproduced with limitations. Notably, during the sintering process oxide precipitates ((Ba,Sr)ZrO3 perovskite phase) and Pd-Ru-Rh-Mo metallic precipitates formed within the UO2 matrix, as originally sought. Spherical oxide precipitates measured up to 30 μm in diameter, while the metallic precipitates were 0.8 ± 0.7 μm. FPs with high solubility in UO2, such as La, Nd and Y, dissolved into the UO2 matrix. ICP-MS analysis showed that some dopants, e.g. Cs and Te, evaporated from the pellets, while the concentration of other elements had also changed during sample fabrication. Very scarce information on real PWR and AGR SNF are reviewed and compared to AGR SIMFuel fabricated in this project. Thorough analysis reveal the severe limitations of the SIMFuel technique in general and call for more experimental work and accessible publications on SNF.
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Hardie, Christopher David. "Micro-mechanics of irradiated Fe-Cr alloys for fusion reactors." Thesis, University of Oxford, 2013. http://ora.ox.ac.uk/objects/uuid:a3ac36ba-ca6f-4129-8f37-f1278ef8a559.
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In the absence of a fusion neutron source, research on the structural integrity of materials in the fusion environment relies on current fission data and simulation methods. Through investigation of the Fe-Cr system, this detailed study explores the challenges and limitations in the use of currently available radiation sources for fusion materials research. An investigation of ion-irradiated Fe12%Cr using nanoindentation with a cube corner, Berkovich and spherical tip, and micro-cantilever testing with two different geometries, highlighted that the measurement of irradiation hardening was largely dependent on the type of test used. Selected methods were used for the comparison of Fe6%Cr irradiated by ions and neutrons to a dose of 1.7dpa at a temperature of 288°C. Micro-cantilever tests of the Fe6%Cr alloy with beam depths of 400 to 7000nm, identified that size effects may significantly obscure irradiation hardening and that these effects are dependent on radiation conditions. Irradiation hardening in the neutron-irradiated alloy was approximately double that of the ion-irradiated alloy and exhibited increased work hardening. Similar differences in hardening were observed in an Fe5%Cr alloy after ion-irradiation to a dose of 0.6dpa at 400°C and doses rates of 6 x 10-4dpa/s and 3 x 10-5dpa/s. Identified by APT, it was shown that increased irradiation hardening was likely to be caused by the enhanced segregation of Cr observed in the alloy irradiated with the lower dose rate. These observations have significant implications for future fusion materials research in terms of the simulation of fusion relevant radiation conditions and micro-mechanical testing.
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Blake, Bryan P. "Initial Testing of Single-Mode Optical Fibers Interrogated with an Optical Backscatter Reflectometer at High Temperatures and in Radiation Environments for Advanced Instrumentation in Nuclear Reactors." The Ohio State University, 2012. http://rave.ohiolink.edu/etdc/view?acc_num=osu1344873713.
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Wood, Thomas W. Jr. "Evaluation of Single-Mode and Bragg Grating Optical Fibers Interrogated with an Optical Backscatter Reflectometer (OBR) in High Temperature Environments for Advanced Instrumentation in Nuclear Reactors." The Ohio State University, 2013. http://rave.ohiolink.edu/etdc/view?acc_num=osu1373364266.
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Vidal, Carlos Magno de Sousa. "Avaliação da microfiltração tangencial como alternativa de tratamento avançado de efluente gerado em sistema de tratamento de esgoto sanitário constituído de reator UASB (Upflow Anaerobic Sludge Blanket) seguido de tanque de aeração." Universidade de São Paulo, 2006. http://www.teses.usp.br/teses/disponiveis/18/18138/tde-06112006-233334/.
Full textAbstract:
A proposta desta pesquisa consistiu na avaliação da microfiltração tangencial para o tratamento avançado de efluente gerado em sistema de tratamento de esgoto sanitário constituído de reator UASB (Upflow Anaerobic Sludge Blanket) seguido de tanque de aeração. Neste trabalho foram avaliados métodos de limpeza física (retrolavagem com ar comprimido) e química (ácida e básica) das membranas, bem como análise comparativa da colmatação e do desempenho de membranas de tamanhos de poro de 0,2 e 1,0 'mü'm, quando aplicadas ao pós-tratamento de efluentes do tanque de aeração em diferentes concentrações de SST. Foram contemplados ainda estudos de desinfecção de efluentes da microfiltração por radiação UV e aplicação do processo de coagulação antecedendo a microfiltração tangencial. Os experimentos foram realizados em unidade piloto com membranas tubulares de polipropileno com área efetiva de filtração de 0,036 'M POT.2'. Constatou-se que a limpeza física foi essencial para obtenção de maiores valores de fluxo de permeado na unidade de microfiltração. A limpeza química das membranas por solução básica foi mais eficiente quando comparada a limpeza ácida. Foram obtidos melhores resultados com a membrana de 0,2 'mü'm, quando comparada a membrana de 1 'mü'm, a qual apresentou intensa colmatação interna de seus poros. Para membrana de 0,2 'mü'm obteve-se permeado de excelente qualidade microbiológica (E.Coli < 1 UFC/100 mL e Colifagos < 16 UFP/100 mL), bem como valores de turbidez inferiores a 1,46 uT e remoção praticamente completa de SST. A prévia coagulação das amostras de licor misto do tanque de aeração contribuiu para obtenção de maiores taxas e melhor remoção de P-'PO IND.4'POT.3-' e DQOt na unidade de microfiltração. Para dosagem de 40 mg/L de cloreto férrico obteve-se a maior taxa média (139,7 L/'M POT.2'.h), concentrações residuais de P-'PO IND.4'POT.3-' inferiores a 1,4 mgP/L e DQOt menor que 33 mg/L. A radiação UV permitiu inativação completa de E.Coli e Colifagos das amostras de permeado. Concluiu-se que a microfiltração tangencial apresentou grande potencialidade para ser aplicada no tratamento avançado de efluente gerado em sistema de tratamento de esgoto sanitário constituído de reator UASB (Upflow Anerobic Sludge Blanket) seguido de tanque de aeração.The proposal of this research was to evaluate the crossflow microfiltration as an alternative for an advanced treatment of effluent generated in a system of sewage treatment composed by a UASB (Upflow Anaerobic Sludge Blanket) reactor followed by an aeration tank. This work aimed to evaluate the membranes physical cleaning methods (backwashing with compressed air) and the chemical ones (acid and basic), as well as the comparative analysis between the fouling event and the 0,2 and 1,0 'mü'm pore size membranes performance, when applied to the aeration tank effluents in a post-treatment stage at TSS different concentrations. Studies for microfiltration effluents disinfection by UV radiation and the application of the coagulation process preceding the crossflow microfiltration were also developed. The experiments were performed in a pilot unit with polypropylene tubular membranes with 0.036 'M POT.2' of effective filtration area. It was verified that the physical cleaning was essential to the attainment of higher permeate flux values in the microfiltration unit. The chemical cleaning of the membranes through basic solution was more efficient when compared to their acid cleaning. Better results were attained when the 0,2 'mü'm membrane was employed in comparison with the 1 'mü'm membrane, which presented intense internal blocking of its pores. It was attained an excellent microbiological quality (E.Coli < 1 FCU/100 mL and Coliphages < 16 FPU/100 mL) for the 0,2 'mü'm membrane, as well as turbidity levels under 1,46 uT and almost total removal of TSS. The previous mixed liquor samples coagulation of the aeration tank contributed to the attainment of higher rates and better removal of P-'PO IND.4'POT.3-' and CODt in the microfiltration unit. It was attained, for the ferric chloride 40 mg/L dosage, the higher mean rate (139,7 L/'M POT.2'.h), P-'PO IND.4'POT.3-' remaining concentrations under 1,4 mgP/L and CODt lesser than 33 mg/L. The UV radiation allowed the complete inactivation of E.Coli and Coliphages from the permeate samples. It was concluded that the crossflow microfiltration presents great possibilities of application in the advanced treatment of effluent generated in a system of sewage treatment composed by a UASB (Upflow Anaerobic Sludge Blanket) reactor followed by an aeration tank.
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Hought, Julian L. "Advanced control of batch chemical reactions." Thesis, University of Huddersfield, 1992. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.314422.
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Hanson, John P. "Evaluation of fiber optic technology for advanced reactor instrumentation." Connect to resource, 2010. http://hdl.handle.net/1811/45425.
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Nabi, Magdi Mohammed. "Neural model-based advanced control of Chylla-Haase reactor." Thesis, Liverpool John Moores University, 2015. http://researchonline.ljmu.ac.uk/4332/.
Full textAbstract:
The objective of this thesis is to develop advanced control method and to design advanced control system for the polymerization reactor (Chylla-Haase) to maintain the high accurate reactor temperature. The first stage of this research start with the development of mathematical model of the process. The sub-models for monomer concentration, polymerization rate, reactor temperature and jacket outlet/inlet temperature are developed and implemented in Matlab/Simulink. Four conventional control methods were applied to the reactor: a Proportional –Integral-Derivative (PID), Cascade control (CCs), Linear-Quadratic-Regulator (LQR), and Linear model predictive control (LMPC). The simulation results show that the PID controller is unable to perform satisfactorily due to the change of physical properties unless constant re-tuning takes place. Also, Cascade Control the most common control method used in such processes cannot guarantee a robust performance under varying disturbance and system uncertainty. In addition, LQR and linear MPC methods lead to better results compared with the previous two methods. But it is still under an assumption of the linearized plant. Three advanced neural network based control schemes are also proposed in this thesis: radial basis function RBF neural network inverse model based feedforward-feedback control scheme, RBF based model predictive control and multi-layer perception (MLP) based model predictive control. The major objective of these control schemes is to maintain the reactor temperature within its tolerance range under disturbances and system uncertainty. Satisfactory control performance in terms of effective regulation and robustness to disturbance have been achieved. In the feedforward-feedback control scheme, a neural network model is used to predict reactor temperature. Then, a neural network inverse model is used to estimate the valve position of the reactor, the manipulated variable. This method can identify thecontrolled system with the RBF neural network identifier. A PID controller is used in the feedback control to regulate the actual temperature by compensating the neural network inverse model output. Simulation results show that the proposed control has strong adaptability, robustness and satisfactory control performance. These advanced methods achieved the much improved control performance compared with conventional control schemes. The main contribution of this research lies in the following aspects. The MPC theory is realised to control Chylla-Haase polymerization reactor. Two adaptive reactor models including the RBF network model and MLP model are developed to predict the multiple-step-ahead values of the reactor output. Their modelling ability is compared with that of the models with fixed parameters and proven to be better. The RBF neural network and the MLP is trained by the recursive Least Squares (RLS) algorithm and is used to model parameter uncertainty in nonlinear dynamics of the Chylla-Haase reactor. The predictive control strategy based on the RBF neural network is applied to achieve set-point tracking of the reactor output against disturbances. The result shows that the RBF based model predictive control gives reliable result in the presence of some disturbances and keeps the reactor temperature within a tight tolerance range around the specified reaction temperature. Moreover, RBF neural network based model predictive control strategy has also been used to reduce the batch time in order to shorten the reaction period. RBF neural network is considered as a prediction model for control purpose which is based to minimize a cost function in order to determine an optimal sequence of control moves. The result shows that the RBF based model predictive control gives reliable result in the presence of variation of monomer and presence of some disturbances for keeping the reactor temperature within a tight tolerance range around the specified reaction temperature without harming the quality of the temperature control.
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Preston, Stephen David. "The effect of material property variations on the failure probability of an AGR moderator brick subject to irradiation induced self stress." Thesis, University of Salford, 1989. http://usir.salford.ac.uk/43034/.
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The failure of graphite moderator bricks in an Advanced Gas cooled Reactor (AGR) is potentially a serious problem. This thesis describes the generation of self stress in the moderator brick during irradiation and the derivation of a simple analytical model to predict the magnitude of this stress. The magnitude of the self stress in the brick is affected by the variations in the material properties of the graphite used for the brick and this is also examined, developing a statistical approach to the analysis. Property variations between bricks are considered but no allowance has been made for material property variations within a brick. Finally, the thesis compares the self stress in one of the critical peak rated moderator bricks to the strength of the irradiated oxidised material on a statistical basis and predicts the failure probability of a brick due to self stress to be extremely low at 25.5 full power years (FPY). However, the failure probability rises steeply and for the peak rated bricks at 29 FPY it approaches 100%.
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Moore, Eugene James Thomas. "Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application." Texas A&M University, 2003. http://hdl.handle.net/1969.1/3996.
Full textAbstract:
High-temperature gas-cooled reactors (HTGR) are passively safe, efficient, and
economical solutions to the worldÂs energy crisis. HTGRs are capable of generating high
temperatures during normal operation, introducing design challenges related to material
selection and reactor safety. Understanding heat transfer and fluid flow phenomena
during normal and transient operation of HTGRs is essential to ensure the adequacy of
safety features, such as the reactor cavity cooling system (RCCS). Modeling abilities of
system analysis codes, used to develop an understanding of light water reactor
phenomenology, need to be proven for HTGRs. RELAP5-3D v2.3.6 is used to generate
two reactor plant models for a code-to-code and a code-to-experiment benchmark
problem.
The code-to-code benchmark problem models the Russian VGM reactor for
pressurized and depressurized pressure vessel conditions. Temperature profiles
corresponding to each condition are assigned to the pressure vessel heat structure.
Experiment objectives are to calculate total thermal energy transferred to the RCCS for
both cases. Qualitatively, RELAP5-3DÂs predictions agree closely with those of other
system codes such as MORECA and Thermix. RELAP5-3D predicts that 80% of thermal energy transferred to the RCCS is radiant. Quantitatively, RELAP5-3D computes
slightly higher radiant and convective heat transfer rates than other system analysis
codes. Differences in convective heat transfer rate arise from the type and usage of
convection models. Differences in radiant heat transfer stem from the calculation of
radiation shape factors, also known as view or configuration factors. A MATLAB script
employs a set of radiation shape factor correlations and applies them to the RELAP5-3D
model.
This same script is used to generate radiation shape factors for the code-toexperiment
benchmark problem, which uses the Japanese HTTR reactor to determine
temperature along the outside of the pressure vessel. Despite lacking information on
material properties, emissivities, and initial conditions, RELAP5-3D temperature trend
predictions closely match those of other system codes. Compared to experimental
measurements, however, RELAP5-3D cannot capture fluid behavior above the pressure
vessel. While qualitatively agreeing over the pressure vessel body, RELAP5-3D
predictions diverge from experimental measurements elsewhere. This difference reflects
the limitations of using a system analysis code where computational fluid dynamics codes
are better suited.
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Human, Gerhardus. "Model based predictive control for load following of a pressurised water reactor / Gerhardus Human." Thesis, North-West University, 2009. http://hdl.handle.net/10394/4017.
Full textAbstract:
By September 2009 the International Atomic Energy Agency reported that the
number of commercially operated nuclear reactors in 30 countries across the world
is 436, around 50 reactors are currently being constructed, 137 reactors have been
ordered or is already planned, and there are around 295 proposed reactors.
Pressurised water reactors (PWRs) make up the majority of these numbers. The
growing number of carbon emissions and the ongoing fight against fossil fuel power
stations might see the number of planned nuclear reactors increase even more to be
able to satisfy the world’s need for cleaner energy. To ensure that technology keeps
pace with this growing demand, ongoing research is essential. Not only is the
research of new reactor technologies (i.e. High Temperature Reactors) important,
but improving the current technologies (i.e. PWRs) is critical. With the increased
contribution of nuclear generated electricity to our grids, it is becoming more
common for nuclear reactors to be operated as load following units, and not base
load units as they are more commonly being operated. Therefore a need exists to
study and develop new strategies and technologies to improve the automatic load
following capabilities of reactors.
PWR power plants are multivariable systems. In this study a multivariable, more
specifically, a model predictive controller (MPC) is developed for controlling the load
following of a nuclear power plant, more specifically a PWR plant. In developing this
controller system identification is employed to develop a model of the PWR plant.
For the identification of the model, measured data from a computer based PWR
simulator is used as the input. The identified plant model is used to develop the MPC
controller. The controller is developed and tested on the plant model. The MPC
controller is also evaluated against another set of measured data from the simulator.
To compare the performance of the MPC controller to that of the conventional
controller the ITAE performance index is employed. During the process Matlab
®
, the
System Identification Toolbox™, the MPC Toolbox™ and Simulink
®
are used.
The results reveal that MPC is practicable to be used in the control of non-linear
systems such as PWR plants. The MPC controller showed good results for
controlling the system and also outperformed the conventional controllers. A further result from the dissertation is that system identification can successfully be used to
develop models for use in model based controllers like MPC controllers. The results
of the research show that a need exists for future research to improve the methods
to eventually have a controller that can be applied on a commercial plant.Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2010.
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Galvin, Mark Robert. "Maintenance cycle extension in advanced light water reactor plant design." Thesis, Springfield, Va. : Available from National Technical Information Service, 2001. http://handle.dtic.mil/100.2/ADA393174.
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