Дисертації з теми "Reactor physic"
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Pope, Michael A. (Michael Alexander). "Reactor physics design of supercritical CO₂-cooled fast reactors." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/33633.
Повний текст джерелаIncludes bibliographical references (p. 109-113).
Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO₂ (S-CO₂) as a Brayton cycle working fluid in a direct cycle is evaluated. By using S- CO₂ at turbine inlet conditions of 20 MPa and 550⁰C - 700⁰C, efficiencies between 45% and 50% can be achieved with extremely compact components. Neutronic evaluation of candidate core materials was performed for potential use in block-type matrix fueled GFRs with particular concentration on lowering coolant void reactivity to less than $1. SiC cercer fuel was found to have relatively low coolant void worth (+22 cents upon complete depressurization of S-CO₂ coolant) and tolerable reactivity- limited burnup at matrix volume fractions of 60% or less in a 600 MWth core having H/D of 0.85 and titanium reflectors. Pin-type cores were also evaluated and demonstrated higher kff versus burnup, and higher coolant void reactivity than the SiC cercer cores (+$2.00 in ODS MA956-clad case having H/D of 1).
(cont.) It was shown, however, that S-CO₂ coolant void reactivity could be lowered significantly - to less than $1 - in pin cores by increasing neutron leakage (e.g. lowering the core H/D ratio to 0.625 in a pin core with ODS MA956 cladding), an effect not observed in cores using helium coolant at 8 MPa and 500⁰C. An innovative "block"-geometry tube-in-duct fuel consisting of canisters of vibrationally compacted (VIPAC) oxide fuel was introduced and some preliminary calculations were performed. A reference tube-in-duct core was shown to exhibit favorable neutron economy with a conversion ratio (CR) at beginning of life (BOL) of 1.37, but had a coolant void reactivity of +$ 1.4. The high CR should allow designers to lower coolant void worth by increasing leakage while preserving the ability of the core to reach high burnup. Titanium, vanadium and scandium were found to be excellent reflector materials from the standpoint of ... and coolant void reactivity due to their unique elastic scattering cross-section profiles: for example, the SiC cercer core having void reactivity of +$0.22 with titanium was shown to have +$0.57 with Zr₃Si₂.
(cont.) Overall, present work confirmed that the S-CO₂-cooled GFR concept has promising characteristics and a sufficiently broad opion space such that a safe and competitive design could be developed in future work with considerably less than $1 void reactivity and a controllable [delta]k due to burnup.
by Michael A. Pope.
S.M.
Sadeghi, Mohammad Mehdi 1959. "SYMBOLIC MANIPULATION IN REACTOR PHYSICS." Thesis, The University of Arizona, 1986. http://hdl.handle.net/10150/275520.
Повний текст джерелаBora, Pekicten Aziz. "Assembly homogenization of light water reactors by a monte carlo reactor physics method and verification by a deterministic method." Thesis, KTH, Kärnkraftsäkerhet, 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-34492.
Повний текст джерелаGottfridsson, Filip. "Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-148572.
Повний текст джерелаGonzalez, 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.
Повний текст джерелаBrinkmann, Torsten. "Use of catalytic membrane reactors for in situ reaction and separation." Thesis, University of Bath, 1999. https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.301546.
Повний текст джерелаEmmett, John Carter Alfred. "A standard neutron spectrum source of application to fast reactor physics." Thesis, Imperial College London, 2000. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.312126.
Повний текст джерелаChristensen, Eric Kurt. "Applications of Neutrino Physics." Diss., Virginia Tech, 2014. http://hdl.handle.net/10919/64864.
Повний текст джерелаPh. D.
MAEDA, REINALDO de M. "Determinação experimental de parâmetros de física de reatores utilizando refletor de água pesada no reator IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10128.
Повний текст джерелаMade available in DSpace on 2014-10-09T14:00:44Z (GMT). No. of bitstreams: 0
Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
Tuttelberg, Kaur. "STORM in Monte Carlo reactor physics calculations." Thesis, KTH, Fysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-146284.
Повний текст джерелаMovalo, Raisibe Shirley. "Fuel management study for a pebble bed modular reactor core." Thesis, Stellenbosch : Stellenbosch University, 2010. http://hdl.handle.net/10019.1/4234.
Повний текст джерелаENGLISH ABSTRACT: This dissertation reports on the impact of a set of selected nuclear fuel management parameters on reactor operations of the PBMR core. This is achieved by performing an assessment of the impact of nuclear fuel management parameter variations on the most important safety and economics issues for the PBMR core. These include the maximum fuel temperature at steady state and during Depressurized Loss of Forced Cooling (DLOFC) accident conditions. The reactivity worth of the Reactor Control System (RCS which determines the shutdown capability of the reactor core and the average discharge burn-up of fuel are also established. The fuel management parameters considered in this study include different enrichment levels, heavy metal loadings and fuel sphere circulation regimes. The impact and importance of these parameters on plant safety and economics is assessed. The dissertation will report the effects on the standard core physics parameters such as power peaking, multiplication factor, burn-up (safety and economics) and derive the benefits and drawbacks from the results. Based upon the findings from this study, and also experimental data, an optimum fuel management scheme is proposed for the PBMR core.
AFRIKAANSE OPSOMMING: Hierdie verhandeling beskryf die uitwerking van ‘n gekose stel kernbrandstofparameters op die bedryf van die PBMR reaktor. Die impak wat variasies in kernbrandstofparameters op belangrike veiligheids- en ekonomiese oorwegings het, is tydens hierdie studie ondersoek. Van die belangrikste oorwegings is die maksimum brandstoftemperatuur tydens normale, konstante bedryf, asook gedurende ‘n “Depressurized Loss of Forced Cooling (DLOFC)” insident waar alle verkoeling gestaak word. Ander belangrike fasette wat ondersoek is, is die reaktiwiteitwaarde van die beheerstelsel (RCS), wat die aanleg se vermoë om veilig af te sluit bepaal, asook die totale kernverbruik van die brandstof. Die kernbrandstofparameters wat in ag geneem is, sluit die brandstofverryking, swaarmetaalinhoud en die aantal brandstofsirkulasies deur die reaktorhart in. Die belangrikheid en impak van elk van hierdie parameters is ondersoek en word in die verhandeling beskryf . Daar word verslag gelewer oor die voor- en nadele, asook die uitwerking van hierdie variasies op standaard reaktorfisika-parameters soos drywingspieke in die brandstof, neutronvermenigvuldigingsfaktore en kernverbuik van die brandstof, vanaf ‘n veiligheids- en ekonomiese oogpunt. Gebaseer op die gevolgtrekkings van hierdie studie, tesame met eksperimentele data, word ‘n optimale kernbrandstofbestuurprogram voorgestel.
Hidalga, García-Bermejo Patricio. "Development and validation of a multi-scale and multi-physics methodology for the safety analysis of fast transients in Light Water Reactors." Doctoral thesis, Universitat Politècnica de València, 2021. http://hdl.handle.net/10251/160135.
Повний текст джерела[CA] La tecnologia nuclear per a l'ús civil genera més preocupació per la seguretat que moltes altres tecnologies d'ús quotidià. L'Autoritat Nuclear defineix les bases de com ha de realitzar-se l'operació segura d'una Central Nuclear. D'acord amb les directrius establertes per l'Autoritat Nuclear, una Central Nuclear ha d'analitzar una envoltant d'escenaris hipotètics I comprovar de manera determinista que els criteris d'acceptació per a l'esdeveniment seleccionat es compleixen. L'Anàlisi Determinista de Seguretat utilitza eines de simulació que apliquen la física coneguda sobre el comportament de la Central Nuclear per avaluar l'evolució d'una variable de seguretat i assegurar que els límits no es traspassen. El desenvolupament de la tecnologia informàtica, els mètodes matemàtics i de la física que envolta el comportament d'una Central Nuclear han proporcionat eines de simulació potents amb capacitat de predir el comportament de les variables de seguretat amb una precisió significativa. Això permet analitzar escenaris de manera realista evitant assumir condicions conservadores que fins al moment compensaven la mancança de coneixement. Les eines de simulació conegudes com De Millor Estimació son capaces d'analitzar esdeveniment transitoris a diferent escales. A més, utilitzen models analítics per a les diferents físiques amb més detall així com correlacions experimentals més actualitzades i realistes. Un pas més endavant en l'Anàlisi Determinista de Seguretat pretén combinar les diferents eines de Millor Estimació que se utilitzen per analitzar les distintes físiques d'una Central Nuclear, considerant inclús la interacció entre ells i l'anàlisi progressiu a diferents escales, amb la finalitat de poder analitzar fenòmens locals. Per a aquest fi, esta tesi presenta una metodologia d'anàlisi multi-física i multi-escala que utilitza diferents codis de simulació analitzant l'escenari proposat a diferents escales, és a dir, des d'un nivell de planta que inclou els distints components, fins al volum de control que suposa el refrigerant passant entre les varetes de combustible. Esta metodologia permet un flux de informació que va des de l'anàlisi d'una escala major a una menor. El desenvolupament d'aquesta metodologia ha sigut validada i verificada amb dades de planta i els resultats han sigut analitzats a fi d'avaluar la capacitat de la metodologia i les possibles línies de treball futur. A més s'han afegit els principals resultats de verificació i validació que han sorgit en les distintes etapes d'aquest treball.
[EN] The nuclear technology for civil use has generated more concerns for the safety than several other technologies applied to the daily life. The Nuclear Regulators define the basis of how the Safety Operation of Nuclear Power Plants is to be done. According to these guidelines, a Nuclear Power Plant must analyze an envelope of hypothetical events and deterministically define if the acceptance criteria for these events is met. The Deterministic Safety Analysis uses simulation tools that apply the physics known in the behavior of the Nuclear Power Plant to evaluate the evolution of a safety varia-ble and assure that the safety limits will not be exceeded. The development of the computer science, the numerical methods and the physics involved in the behavior of a Nuclear Power Plant have yield powerful simulation tools that are capable to predict the evolution of safety variables which significant accuracy. This allows to consider more realistic simulation scenarios instead of con-servative approaches in order to compensate the lack of knowledge in the applied prediction methods. The so called Best Estimate simulation tools are capable to analyze the transient events in different scales. Furthermore, they account more detailed analytical models and experimental correlations. A step forward in the Deterministic Safety Analysis intends to combine the Best Estimate simulation tools of the different physics considering the interaction among them and analyzing the different scales, considering more local approaches if necessary. For this purpose, this thesis work presents a multi-scale and multi-physics methodology that uses different physics codes and has the aim of modeling postulated scenarios in different scales, i.e. from system models representing the components of the plants to the subchannel models that analyze the behavior of the coolant between the fuel rods. This methodology allows a flow of information where the output of one scale is used as input in a more detailed scale to predict a more local analysis of parameters, such as the Critical Power Ratio, which are of great importance for the estimation of safety margins. The development of this methodology has been validated against plant data with the aim of evaluating the scope of this methodology and in order to provide future lines of development. In addition, different results of the validation and verifi-cation yielded in the development of the parts of this methodology are presented.
Hidalga García-Bermejo, P. (2020). Development and validation of a multi-scale and multi-physics methodology for the safety analysis of fast transients in Light Water Reactors [Tesis doctoral]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/160135
TESIS
Mullen, Christopher. "Radical-molecule reaction dynamics studied using a pulsed supersonic Laval nozzle flow reactor between 53 and 188 Kelvin." Diss., The University of Arizona, 2004. http://hdl.handle.net/10150/280633.
Повний текст джерелаOLIVEIRA, FERNANDO L. de. "Solução analítica da cinética espacial do modelo de difusão para sistemas homogêneos subcríticos acionados por fonte externa." reponame:Repositório Institucional do IPEN, 2008. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11671.
Повний текст джерелаMade available in DSpace on 2014-10-09T14:09:26Z (GMT). No. of bitstreams: 0
Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP
Yarsky, Peter. "Core design and reactor physics of a breed and burn gas-cooled fast reactor." Thesis, Massachusetts Institute of Technology, 2005. http://hdl.handle.net/1721.1/34650.
Повний текст джерелаIncludes bibliographical references (p. 245-248).
In order to fulfill the goals set forth by the Generation IV International Forum, the current NERI funded research has focused on the design of a Gas-cooled Fast Reactor (GFR) operating in a Breed and Burnm (B&B) fuel cycle mode. B&B refers to a once-through fuel cycle where low enriched uranium (less than 5 w/o 235U in U) subcritical assemblies are loaded into the core in equilibrium, yet in-situ plutonium breeding carries the fuel through a discharge burnup on the order of 150 MWD/kgHM. The B&B fuel cycle meets the GenIV goals of sustainability, economics, and proliferation resistance by increasing fuel burnup without the need for spent fuel reprocessing, recycle, or reuse of any kind. The neutronic requirements for B&B are strict and require an ultra-hard neutron spectrum. Therefore, the GFR is ideally suited for this fuel cycle. In the present work the B&B GFR concept evolved into two practical reactor designs, both of which build on extensive previous gas-cooled reactor design experience. The first version is the "demonstration" concept using highly neutronically reactive U15N fuel in a hexagonal pin fuel array that is nearly 50 v/o fuel. The core is helium cooled, with an outlet temperature of 570 °C.
The helium primary circuit is coupled to a steam Rankine power conversion system essentially identical to that for the British Advanced Gas-cooled Reactors. One advantage of the low coolant temperature compared to other GenIV GFR concepts is that it allows for the use of oxide dispersion strengthened stainless steels (ODS) in core. The fuel is manufactured using advanced vibration compaction techniques, clad in ODS, and vented in order to achieve the high burnup goal. The second version, the "advanced" concept builds on the experience of the demonstration concept to develop a B&B GFR without the need for expensive U'5N fuel. In order to substitute the nitride fuel with carbide, significantly higher heavy metal loadings are required (60 v/o fuel for UC versus 50 v/o fuel for U'5N) which are not practically achievable with a conventional pin fuel array. Therefore, an innovative tube-in-duct assembly design was proposed to achieve B&B operation with the less neutronically reactive carbide fuel. The advanced core offers significantly reduced natural uranium requirements and lower equilibrium fuel cycle costs (5 mills/kWhre) compared with conventional light water reactors (7 mills/kWhre), as the burnup is tripled for the same reload enrichment.
(cont.) The B&B GFR designs, though requiring active decay heat removal, are semi-self-regulating from a reactivity feedback standpoint and are designed to withstand all plausible accident scenarios, including loss of flow, loss of heat sink, and transient overpower all without scram. Reactor pressure vessel blowdown (LOCA) was investigated and while the B&B GFR has a low positive coolant void reactivity (less than 1$), the added reactivity during blowdown is compensated through other strong negative reactivity feedback mechanisms, thereby allowing for the safe operation of the B&B GFR.
by Peter Yarsky.
Ph.D.
Serra, André da Silva. "Determinação experimental da reatividade subcrítica utilizando correlação de terceira ordem." Universidade de São Paulo, 2012. http://www.teses.usp.br/teses/disponiveis/43/43134/tde-26032013-121339/.
Повний текст джерелаThe present work aims to contribute to the systematic development of new experimental methods of measuring the reactivity of any subcritical fissile arrangements using: high-order statistics of neutron counts from neutron detectors working in pulse mode, the recent concept general reactivity, and the IPEN/MB-01 facility. This thesis brought together in a single text various aspects concerning the proper implementation of these types of measures. Unlike other techniques used in measurements of subcritical reactivity, the methodologies presented in this thesis has the potential to allow the experimental measurement of subcritical reactivity without the prior estimate of any other kinetic parameters, obtained from experiments or from theoretical considerations, external sources calibrations or detectors e ciency measurements. At first, the high-order statistical methods of neutron counts allow to obtain directly the value of the subcriticality (or multiplication factor) from an fissile arrangement regardless the type of subcritical physical theory, and also without the use of unusual infrastructure (such as a pulsed neutron source). These methods are a natural extension of those that use lower order statistics - for example, Feymann-. The greater information content in high order statistics of neutron counting is the main reason for the implementation of this work. Despite its potential, the experimental implementation of the method found huge problems concerning acquisition time and rate of data acquisition. This difficulty overcome any effort in order to obtain a useful measurement inside the IPEN/MB-01 nuclear reactor (a critical facility). However, there are other ways to exploit higher order statistics. For example, an extension of the Rossi- method suggested in this thesis used self bicorrelations. Though the high variance values of obtained results, the fundamental statistical requirements of a measurement were preserved, once the proposed methodologies are observed. It was proposed a methodology to handle dead time issues, in order to allow one to carry out measurement at higher detection rates. Throughout its execution, this thesis aimed to fulfill some gaps in the experimental procedures apparently not addressed by other authors, allowing the establishment of more rigorous statistical procedures. Regarding those contributions, dead time corrections stands out together with the concerning for correlation treatment between the statistical parameters. From the theoretical point of view, this thesis suggests two new ways to address the same problem of using high order statistics of neutron detections in pulse mode: (1) self-bicorrelations, and (2) self-bispectra (power spectral density in two axis). The first was experimentally tested and exhaustively detailed, the second one was only suggested as a theoretical speculation to be confronted against experimental evidence
Svanström, Sebastian. "Load following with a passive reactor core using the SPARC design." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-296803.
Повний текст джерелаKnight, M. P. "The application of modern nodal methods to PWR reactor physics analysis." Thesis, Open University, 1988. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.382929.
Повний текст джерелаBloore, David A. "Reactor physics assessment of thick silicon carbide clad PWR fuels." Thesis, Cambridge, Massachussetts, Massachussetts Institute of Technology, 2013. http://hdl.handle.net/10945/40219.
Повний текст джерелаHigh temparature tolerance, chemical stability and low neutron affinity make silicon carbide (SiC) a potential fuel cladding material that may improve the economics and safety of light water reactors (LWRs). "Thick" SiC cladding (0.089 cm.) is easier (and thus more economical) to manufacture than SiC of conventional Zircaloy (Zr) classing thickness (0.05 cm.) Five fuels and clad combinations are analyzed: Zr with solid UO2 pellets, reduced fuel fraction "thick" SiC (Thick SiC) with annular UO2 pellets, Thick Sic with solid UO2/BeO pellets, reduced coolant fraction annular fuel with "Thick" SiC (Thick SiC RCF), and Thick Sic with solid PuO2/ThO2 pellets. CASMO-4E and SIMULATE-3 have been utilized to model the above in a 193 assembly, 4-loop Westinghouse pressurized water reactor (PWR). A new program, CSpy, has been written to use CASMO/SIMULATE to conduct optimization searches of burnable poison layouts and core reload patterns. All fuel/clad combinations have been modeled using 84 assembly reloads, and Thick SiC clad annular UO2 has been modeled using both 84 and 64 assembly reloads. Dual Binary Swap (DBS) optimization via three Objective Functions (OFs) has been applied to each clad/fuel/reload # case to produce a single reload enrichment equilibrium core reload map. The OFs have the goals of minimal peaking, balancing lower peaking with longer cycle length, or maximal cycle length. Results display the tradeoff betwween minimized peaking and maximized cycle length for each clad/fuel/reload # case. The presented Zr reference cases and Thick SiC RCF cases operate for an 18 month cycle at 3587 MWth using 4/3% and 4/8% enrichment, respectively. A 90% capacity factor was applied to all SiC cladding cases to reflect the challenge to introduction of a new fuel. The Thick SiC clad annular UO2 (84 reload cores) and Think SiC UO2/BeO exhibit similar reactor physics performance but require higher enrichments that 5%. The Thick SiC RCF annular UO2 fuel cases provide the required cycle length with less than 5% enrichment. The Thick SiC clad PuO/2/ThO2 cores can operate with a Pu% of heavy metal of about 12%, however they may have unacceptable shutdown margins without altering the control rod materials.
Bloore, David A. (David Allan). "Reactor physics assessment of thick silicon carbide clad PWR fuels." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82454.
Повний текст джерелаCataloged from PDF version of thesis.
Includes bibliographical references (pages 84-86).
High temperature tolerance, chemical stability and low neutron affinity make silicon carbide (SiC) a potential fuel cladding material that may improve the economics and safety of light water reactors (LWRs). "Thick" SiC cladding (0.089 cm) is easier (and thus more economical) to manufacture than SiC of conventional Zircaloy (Zr) cladding thickness (0.057 cm). Five fuel and clad combinations are analyzed: Zr with solid U0 2 pellets, reduced fuel fraction "thick" SiC (Thick SiC) with annular U0 2 pellets, Thick SiC with solid U0 2/BeO pellets, reduced coolant fraction annular fuel with "thick" SiC (Thick SiC RCF), and Thick SiC with solid PuO2/ThO2 pellets. CASMO-4E and SIMULATE-3 have been utilized to model the above in a 193 assembly, 4-loop Westinghouse pressurized water reactor (PWR). A new program, CSpy, has been written to use CASMO/SIMULATE to conduct optimization searches of burnable poison layouts and core reload patterns. All fuel/clad combinations have been modeled using 84 assembly reloads, and Thick SiC clad annular U0 2 has been modeled using both 84 and 64 assembly reloads. Dual Binary Swap (DBS) optimization via three Objective Functions (OFs) has been applied to each clad/fuel/reload # case to produce a single reload enrichment equilibrium core reload map. The OFs have the goals of: minimal peaking, balancing lower peaking with longer cycle length, or maximal cycle length. Results display the tradeoff between minimized peaking and maximized cycle length for each clad/fuel/reload # case. The presented Zr reference cases and Thick SiC RCF cases operate for an 18 month cycle at 3587 MWth using 4.3% and 4.8% enrichment, respectively. A 90% capacity factor was applied to all SiC cladding cases to reflect the challenge to introduction of a new fuel. The Thick SiC clad annular U0 2 (84 reload cores) and Thick SiC U0 2/BeO exhibit similar reactor physics performance but require higher enrichments than 5%. The Thick SiC RCF annular U0 2 fuel cases provide the required cycle length with less than 5% enrichment. The Thick SiC clad PuO2/ThO 2 cores can operate with a Pu% of heavy metal of about 12%, however they may have unacceptable shutdown margins without altering the control rod materials.
by David A. Bloore.
S.M.
Ignas, Mickus. "Response Matrix Reloaded : for Monte Carlo Simulations in Reactor Physics." Licentiate thesis, KTH, Kärnenergiteknik, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-263412.
Повний текст джерелаDenna avhandling undersöker Monte Carlo-metoder som används för kritikalitets- och tidsberoende problem i reaktorfysik. På grund av deras noggrannhet och flexibilitet betraktas Monte Carlo-metoder som en ‘gyllene standard’ i reaktorfysikberäkningar. Fördelarna kommer dock till priset av betydande datorkostnad. Trots den kontinuerliga ökningen av lättillgänglig datorkraft är en råstyrka Monte Carlo-beräkningar av vissa problem fortfarande utanför räckvidden för reaktorfysikaliska rutinanalyser. De två artiklarna som denna avhandling bygger på försöker ta itu med beräkningskostnadsproblemet genom att föreslå metoder för att utföra Monte Carlo-reaktorfysikberäkningar mer effektivt. Den första metoden behandlar effektiviteten för de vitt använda beräkningarna av k-egenvärdet med Monte Carlo. Den antyder att beräkningseffektiviteten kan ökas genom en gradvis ökning av neutronpopulationens storlek som simuleras under varje kritikalitetscykel, och föreslår ett sätt att bestämma den optimala neutronpopulationens storlek. Den andra metoden behandlar tillämpningen av Monte Carlo-beräkningar för reaktortransienter. Medan beräkningar av reaktortransienter i princip kan utföras uteslutande med Monte Carlo-metoder, tar sådana beräkningar flera tusentals CPU-timmar för att beräkna flera sekunder av en transient. Den föreslagna metoden erbjuder en medelväg, med användning av ett stokastiskt-deterministiskt hybridschema baserat på responsmatrisformalismen. Tidigare har responsmatrisformalismen huvudsakligen beaktats för tidsoberoende problem, med begränsad tillämpning på tidsberoende problem. Denna avhandling föreslår ett nytt sätt att använda information från Monte Carlo-kritikalitetsberäkningar för att lösa tidsberoende problem via responsmatrisen.
Examinator: Professor Pär Olsson
Persson, Carl-Magnus. "Reactivity Assessment in Subcritical Systems." Licentiate thesis, Stockholm : Fysiska institutionen, Kungliga Tekniska högskolan, 2007. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-4363.
Повний текст джерелаEllis, Matthew Shawn. "Methods for including multiphysics feedback in Monte Carlo reactor physics calculations." Thesis, Massachusetts Institute of Technology, 2017. http://hdl.handle.net/1721.1/112381.
Повний текст джерелаThis electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 314-321).
The ability to model and simulate nuclear reactors during steady state and transient conditions is important for designing efficient and safe nuclear power systems. The accurate simulation of a nuclear reactor is particularly challenging because the multiple physical processes within the reactor are tightly coupled, which requires that the numerical methods used to resolve each physical process can accurately and efficiently transfer and utilize data from other applications. Monte Carlo methods are desirable for solving the neutron transport equation required in reactor analysis because of the inherent accuracy of the method, but the Computational Solid Geometry (CSG) representation of the physical geometry makes it difficult to accurately and efficiently perform multiphysics reactor analyses with other applications that utilize finite element or finite volume representations. To address this limitation, a multiphysics coupling framework that minimizes the need for spatial discretization in the Monte Carlo geometry is presented in this thesis. The coupling framework uses Functional Expansion Tallies to transfer multiphysics information from the Monte Carlo application to other multiphysics tools. Additionally, the coupling framework uses a modified method for transporting neutrons through spatially continuous total macroscopic cross section distributions in order to incorporate continuous multiphysics feedback fields such as fuel temperature and coolant density into the Monte Carlo simulation. It has been shown that separable Zernike and Legendre Function Expansion Tallies can effectively reconstruct a continuous distribution of fission power density. Additionally, using a prototypical three-dimensional Light Water Reactor pin cell, the method used to transport neutrons through a continuously varying fuel temperature and coolant density distribution was shown to be 1.7 times faster than a comparable discretized simulation with volume-averaged properties, while still providing a high level of accuracy. Finally, in order to make the overall multiphysics coupling scheme useful for reactor analyses, a novel spatially continuous depletion methodology was developed and investigated. With the spatially continuous depletion methodology, number densities can be represented as a linear combination of polynomials, and those polynomial representations can be integrated through time to predict reactor operation. The spatially continuous depletion methodology was able to accurately predict the eigenvalue and number density distributions in a two-dimensional LWR pin cell depletion containing Gd-157 from a 2 weight percent GdO2 and seven other nuclides in the depletion matrix. Analyses of the spatially continuous depletion methodology showed that significant reductions in the number of tallied values could be achieved if polynomial representations were optimized for each nuclide reaction rate. From the depletion simulations in this thesis, a 23% reduction in the required number of reaction rate tallies compared to a lower-fidelity, 10 radial ring pin discretization was shown to be achievable with nuclide polynomial optimization. In addition to showing potential for reductions in tally memory and computational requirements, the spatially continuous depletion simulation was shown to be equal in computational performance to a discrete simulation with 10 radial rings and 8 azimuthal cuts, while providing a much higher level of spatial fidelity in number density concentrations.
by Matthew Shawn Ellis.
Ph. D.
Gale, Micah D. (Micah David). "Developing modern graphite exponential pile experiments to augment reactor physics education." Thesis, Massachusetts Institute of Technology, 2018. http://hdl.handle.net/1721.1/119041.
Повний текст джерелаThis electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 39-40).
Reactor Physics is not always an intuitive subject for students to understand. When nuclear engineering was beginning as a field it was common for students to complete measurements on sub-critical reactors, which could not sustain a fission chain reaction, in order to develop student intuition. The Massachusetts Institute of Technology has one such reactor, a graphite exponential pile, which went unused for decades. In this thesis the MIT Graphite Exponential Pile was returned to experimental operation, and a prototypic student experiment was completed. The material buckling was found by indium foil activations completed with a plutonium-beryllium source in the pile. From the experimental results it was calculated the pile would have to be a cube with sides that are 5.42m long to become a critical reactor. This proof of concept experiment makes it possible for mens et manus based education at MIT for reactor physics.
by Micah D. Gale.
S.B.
Jones, Christopher LaDon. "Prediction of the reactor antineutrino flux for the Double Chooz experiment." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/79519.
Повний текст джерелаCataloged from PDF version of thesis.
Includes bibliographical references (p. 183-191).
This thesis benchmarks the deterministic lattice code, DRAGON, against data, and then applies this code to make a prediction for the antineutrino flux from the Chooz BI and B2 reactors. Data from the destructive assay of rods from the Takahama-3 reactor and from the SONGS antineutrino detector are used for comparisons. The resulting prediction from the tuned DRAGON code is then compared to the first antineutrino event spectra from Double Chooz. Use of this simulation in nuclear nonproliferation studies is discussed.
by Christopher LaDon Jones.
Ph.D.
Keller, Steven Ede. "Flux-limited Diffusion Coefficient Applied to Reactor Analysis." Diss., Georgia Institute of Technology, 2007. http://hdl.handle.net/1853/16126.
Повний текст джерелаAmmar, Karim. "Conception multi-physique et multi-objectif des cœurs de RNR-Na hétérogènes : développement d’une méthode d’optimisation sous incertitudes." Thesis, Paris 11, 2014. http://www.theses.fr/2014PA112390/document.
Повний текст джерелаSince Phenix shutting down in 2010, CEA does not have Sodium Fast Reactor (SFR) in operating condition. According to global energetic challenge and fast reactor abilities, CEA launched a program of industrial demonstrator called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a reactor with electric power capacity equal to 600MW. Objective of the prototype is, in first to be a response to environmental constraints, in second demonstrates the industrial viability of:• SFR reactor. The goal is to have a safety level at least equal to 3rd generation reactors. ASTRID design integrates Fukushima feedback;• Waste reprocessing (with minor actinide transmutation) and it linked industry.Installation safety is the priority. In all cases, no radionuclide should be released into environment. To achieve this objective, it is imperative to predict the impact of uncertainty sources on reactor behaviour. In this context, this thesis aims to develop new optimization methods for SFR cores. The goal is to improve the robustness and reliability of reactors in response to existing uncertainties. We will use ASTRID core as reference to estimate interest of new methods and tools developed.The impact of multi-Physics uncertainties in the calculation of the core performance and the use of optimization methods introduce new problems:• How to optimize “complex” cores (i.e. associated with design spaces of high dimensions with more than 20 variable parameters), taking into account the uncertainties?• What is uncertainties behaviour for optimization core compare to reference core?• Taking into account uncertainties, optimization core are they still competitive? Optimizations improvements are higher than uncertainty margins?The thesis helps to develop and implement methods necessary to take into account uncertainties in the new generation of simulation tools. Statistical methods to ensure consistency of complex multi-Physics simulation results are also detailed.By providing first images of innovative SFR core, this thesis presents methods and tools to reduce the uncertainties on some performance while optimizing them. These gains are achieved through the use of multi-Objective optimization algorithms. These methods provide all possible compromise between the different optimization criteria, such as the balance between economic performance and safety
Kennedy, William B. (William Blake) 1979. "Analysis of the MIT research reactor fission product and actinide radioactivity inventories." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/32723.
Повний текст джерелаMIT Institute Archives copy: leaves 92-111 bound in reverse order.
Includes bibliographical references (leaf 57).
The current analysis of the MITR core radioactivity inventory eliminates unnecessary assumptions made in previous estimates of the inventory, and revises the list of contributory isotopes to include all actinide and fission product isotopes necessary for a proper accident source term calculation. The result is a power-history-dependent inventory that increases with bum-up, and comprises 41 actinide isotopes and 596 fission product isotopes. The analysis uses the ORIGEN2 depletion code to calculate the activity of actinide and fission product isotopes for eight MITR input models at 32 intervals over a period of 5376MWD. The input models simulate a MITR core loaded with high- enrichment, U-Alx cermet fuel or low-enrichment, monolithic U-Mo fuel, and operated at 6MW with a continuous-burn-up or cyclic-burn-up-and-decay power history. Reorganization of the ORIGEN2 output file, and application of an element reduction criterion creates the condensed matrix file for each MITR input model. This file lists the contribution of each isotope to the core radioactivity inventory at each output interval, and is the basis for all inventory analysis. The inventory analysis yields three important conclusions. First, the assumption of an equilibrium inventory of isotopes in the fuel is accurate to within 3% for all time after 10% fuel bum-up, and conservative over the entire fuel cycle. The equilibrium fuel assumption is invalid for the actinides due to a slow rate of inventory growth. Second, the cyclic-bum-up-and-decay power history yields a lower core inventory than the continuous-burn-up power history for both fuel enrichments. The difference is minimized by increasing the ratio of irradiation time to decay time.
(cont.) Finally, the analysis indicates that conversion to a U-Mo fuel will produce an actinide inventory 18 times greater than that of the current U-Alx fuel, with no significant change in the fission product inventory. However, the actinide inventory is a small fraction of the fission product inventory. The worst-case core inventory available for release is 2.91 E+7Ci for the high-enrichment fuel, and 2.94E+7Ci for the low-enrichment fuel, with a core loading of 24 elements in each case. The best-estimate core inventory available for release is 2.83E+7Ci, and 2.82E+7Ci respectively, and accounts for typical cyclic operation of the MITR.
by William B. Kennedy.
S.B.
Palfelt, Alexander, Wilhelm Thunberg, and Anders Winka. "Determining the Sensitivity of Reactor Parameters in a Sodium Cooled Fast Reactor." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2020. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-413073.
Повний текст джерелаBrown, Craig J. "Characterization of a parallel plate electrochemical reactor." Thesis, University of Southampton, 1992. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.358040.
Повний текст джерелаDufek, Jan. "Advanced Monte Carlo methods in reactor physics : eigenvalue and steady-state problems /." Stockholm : Fysiska institutionen, Kungliga Tekniska högskolan, 2007. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-4458.
Повний текст джерелаClifford, Ivor David. "Object-oriented multi-physics applied to spatial reactor dynamics / Ivor David Clifford." Thesis, North-West University, 2007. http://hdl.handle.net/10394/1656.
Повний текст джерелаThesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2008.
Dobisesky, Jacob P. (Jacob Paul) 1987. "Reactor physics considerations for implementing silicon carbide cladding into a PWR environment." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76525.
Повний текст джерелаCataloged from PDF version of thesis.
Includes bibliographical references (p. 110-112).
Silicon carbide (SiC) offers several advantages over zirconium (Zr)-based alloys as a potential cladding material for Pressurized Water Reactors: very slow corrosion rate, ability to withstand much higher temperature with little reaction with steam, and more favorable neutron absorption. To evaluate the feasibility of longer fuel cycles and higher power density in SiC clad fuel, a core design study was completed with uranium dioxide fuel and SiC cladding in a standard, Westinghouse 4-loop PWR. NRC-limited values for hot channel and hot spot values were taken into account as well as acceptable values for the reactivity feedback and control mechanisms and shutdown margin. The Studsvik Core Management System, which consisted of CASMO-4E, CMS-Link, and SIMULATE-3, provided an accurate tool to design the new core loading patterns that would satisfy current nuclear industry standards. Libraries of Westinghouse robust fuel assemblies (RFAs) were modeled in CASMO-4E with varying enrichments, burnable poison layouts, and power conditions. Using these assemblies, full core, three-dimensional analyses were performed in SIMULATE-3 for operating conditions similar to the Seabrook Nuclear Power Station. In this study, SiC-clad fuel rods held 10% less heavy metal to allow for central holes in the U0 2 pellets, limiting peak fuel temperature during anticipated operational transients but raising the average enrichment per fuel batch. The cladding dimensions remained similar to the current Zircaloy 4 cladding. Three approaches were followed in creating the PWR core designs: 1) constant core power density (or total reactor power) and cycle length, but fewer fresh assemblies loaded, 2) constant cycle length, but increased core power density to the maximum feasible level, staying within the capability of the reactor etc., and 3) constant power density, but extended fuel cycle length from 18 to 24 months. Sixteen core designs were completed with three different types of burnable poison (IFBA, WABA, and gadolinium) that achieved the desired operating cycle lengths and target values for reactor physics parameters limited by the NRC. Batch average discharge burnups ranged from ~41 to ~80 MWd/kgU, reinforcing SiC's advantage and potential appeal to power utilities. Additionally, a power uprate of 10% was found to be feasible, but beyond this value would require a redesign of the control rod material and/or layout to allow for an acceptable shutdown margin by end of cycle (EOC). Nevertheless, all other reactivity coefficients and safety margins were met, confirming the feasibility of operating to higher burnups beyond the current limits of Zr cladding.
by Jacob P. Dobisesky.
S.M.
ROSSI, LUBIANKA F. R. "Acoplamento entre os métodos diferencial e da teoria da perturbação para o cálculo dos coeficientes de sensibilidade em problemas de transmutação nuclear." reponame:Repositório Institucional do IPEN, 2014. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23594.
Повний текст джерелаMade available in DSpace on 2015-03-17T10:41:16Z (GMT). No. of bitstreams: 0
Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
Reed, Mark W. (Mark Wilbert). "A steady-state L-mode tokamak fusion reactor : large scale and minimum scale." Thesis, Massachusetts Institute of Technology, 2009. http://hdl.handle.net/1721.1/58088.
Повний текст джерелаCataloged from PDF version of thesis.
Includes bibliographical references (p. 69-70).
We perform extensive analysis on the physics of L-mode tokamak fusion reactors to identify (1) a favorable parameter space for a large scale steady-state reactor and (2) an operating point for a minimum scale steady-state reactor. The identification of the large scale parameter space is part of the 2008 MIT Nuclear Systems Design Project, which also includes sustainability and economic optimizations to identify a plausible operating point for a large scale (a 14 m major radius) hydrogen production reactor dubbed HYPERION. Due to the potentially prohibitive capital cost (a $50 billion) and exorbitant thermal power (a 35 GWth) of HYPERION, we identify a conservative estimate for the minimum scale of a similar steady-state L-mode reactor of approximately 7.5 meters, half the size of HYPERION and only 20% larger than ITER. This minimum scale reactor would require an on-coil magnetic field of a 16 T and a blanket power density of ~ 5 MW/m 2 . It would produce 7 GWth of power with a power gain of 30, and it would operate far from all stability and confinement limits. To confirm the viability of this operating point, we perform various 1-D calculations. The crucial advantage of a steady-state (or fully non-inductive) reactor is that it is not limited by flux swing and can operate continuously, recharging its solenoid during operation. The crucial advantages of L-mode are that it avoids instabilities associated with edge localized modes (ELMs) and that it allows volumetric heating in the mantle due to the absence of a pedestal. Steady-state L-mode tokamak reactors could be the future of controlled fusion research and even play an important role in meeting the world's clean energy needs.
by Mark Reed.
S.B.
Kennedy, Ryanne Ariel. "Quantifying Uncertainty in Reactor Flux/Power Distributions." The Ohio State University, 2011. http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901.
Повний текст джерелаCampos, Claudio Milton Montenegro. "Physical aspects affecting granulation in UASB reactors." Thesis, University of Newcastle Upon Tyne, 1990. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.278700.
Повний текст джерелаAnnand, Kirsty June. "The nanoscale mechanisms of Zircaloy-4 corrosion in simulated nuclear reactor conditions." Thesis, University of Glasgow, 2018. http://theses.gla.ac.uk/8781/.
Повний текст джерелаJasim, Mahdi H. "Elastic and inelastic scattering of fast neutrons in fusion reactor materials." Thesis, Aston University, 1985. http://publications.aston.ac.uk/10594/.
Повний текст джерелаDieuaide, Manon. "SAMOFAR Molten Salt Fast Reactor reprocessing unit design." Thesis, KTH, Fysik, 2018. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-228095.
Повний текст джерелаZhu, Kaixin. "Nuclear Reactor Seismic Analysis Considering Soil-Structure Interaction." Thesis, KTH, Fysik, 2018. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-231328.
Повний текст джерелаGrund, Jessica [Verfasser]. "Online coupling of TRIGA-TRAP to the research reactor TRIGA Mainz / Jessica Grund." Mainz : Universitätsbibliothek Mainz, 2018. http://d-nb.info/1164715569/34.
Повний текст джерелаDepnering, Wilfried Walter [Verfasser]. "Scintillation Light Transport In The Large Reactor Antineutrino Detector JUNO / Wilfried Walter Depnering." Mainz : Universitätsbibliothek der Johannes Gutenberg-Universität Mainz, 2021. http://d-nb.info/1234655209/34.
Повний текст джерелаTalamo, Alberto. "Advanced In-Core Fuel Cycles for the Gas Turbine-Modular Helium Reactor." Doctoral thesis, Stockholm, 2006. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-3901.
Повний текст джерелаLuszczek, Karol. "Validation and Benchmarking of Westinghouse BWR lattice physics methods." Thesis, KTH, Reaktorteknologi, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563.
Повний текст джерелаTroville, Jonathan. "Multiscale Modeling of Carbon Nanotube Synthesis in a Catalytic Chemical Vapor Deposition Reactor." Wright State University / OhioLINK, 2017. http://rave.ohiolink.edu/etdc/view?acc_num=wright1495839218743389.
Повний текст джерелаAlthafiri, Faisal. "Treatment of endocrine disrupting chemicals using the downflow gas contractor reactor." Thesis, University of Birmingham, 2016. http://etheses.bham.ac.uk//id/eprint/6830/.
Повний текст джерелаIvanov, Aleksandar Stoyanov [Verfasser], and R. [Akademischer Betreuer] Stieglitz. "High Fidelity Monte Carlo Based Reactor Physics Calculations / Aleksandar Stoyanov Ivanov. Betreuer: R. Stieglitz." Karlsruhe : KIT-Bibliothek, 2015. http://d-nb.info/1079594868/34.
Повний текст джерелаOlsson, Pär. "Modelling of Formation and Evolution of Defects and Precipitates in Fe-Cr Alloys of Reactor Relevance." Doctoral thesis, Uppsala University, Department of Neutron Research, 2005. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-6014.
Повний текст джерелаFe-Cr alloys form the basis of many industrially important steels. Due to their excellent resistance to radiation induced swelling, ferritic steels are expected to be used for critical structural components in advanced nuclear systems, such as fast breeder reactors, accelerator driven systems and fusion reactors. In this thesis project, theoretical modelling of bulk properties of Fe-Cr alloys has been performed for a wide range of phenomena. Electronic structure calculations, based on density functional theory, have been used to determine equilibrium properties for different magnetic states of the alloy. Ferromagnetic alloys of low Cr concentration (<10% Cr) are anomalously stable, which is related to the variation in sign of the mixing enthalpy which was predicted for the first time in this work. This finding is in agreement with experimental evidence of long range ordering in Fe-Cr alloys with low Cr concentration, as well as the observed phase separation for compositions with higher Cr content.
The character of the interaction of point defects with solute Cr atoms in an iron matrix was investigated ab initio. It was found that due to magnetic interactions, interstitial defects are bound by Cr atoms in bulk iron. Vacancies, on the other hand, interact only weakly with Cr. These results may offer qualitative explanations to the observed concentration dependence of radiation induced swelling in Fe-Cr model alloys.
The ab initio predictions inspired an effort to develop an interatomic alloy potential capable of reproducing both the thermodynamic bulk behaviour of the alloy, such as the mixing enthalpy, and the point defect interactions, in order to perform large scale atomistic and stochastic simulations on scales out of reach for density functional theory. A two-band extension of the embedded atom method of interatomic potentials was developed in order to model ferromagnetic Fe-Cr alloys of arbitrary composition. Kinetic Monte-Carlo simulations of thermal aging, using this two-band potential, reproduce the experimentally measured formation and evolution of solute precipitation as a function of concentration for temperatures relevant to structural materials in nuclear reactors.
Markillie, Gavin A. J. "Reaction dynamics of small polyatomic molecules." Thesis, University of Oxford, 1997. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.363979.
Повний текст джерела