Dissertations / Theses on the topic 'Reactor physic'

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.
Includes 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.

The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillating behavior forced an automatic emergency shutdown of the reactor. This phenomenon lead to a lot of downtime of the reactor and is still unsolved. However, the most probable cause of the transients is radial movements of the core, referred to as core-flowering. This study has investigated the available documentation of the A.U.R.N. events. A simplified model of core-flowering was also created in order to simulate how radial expansion affects the reactivity of a sodium-cooled core. Serpent, which is a Monte-Carlo based simulation code, was chosen as calculation tool. Furthermore, a model of the Phénix core was successfully created and partly validated. The model of the core has a k_eff = 1.00298 and a neutron flux of (8.43+-0.02)!10^15 neutrons/cm^2 at normal state. The result obtained from the simulations shows that an expansion of the core radius decreases the reactivity. A linear approximation of the result gave the relation: change in k_eff/core extension = - 60 pcm/mm. This value corresponds remarkably well to the around - 60 pcm/mm that was obtained from the dedicated core-flowering experiments in Phénix made by the CEA. Core-flowering can recreate similar signals to those registered during the A.U.R.N. events, though the absence of trace of core movements in Phénix speaks against this. However, if core-flowering is the sought answer, it can be avoided by design. The equipment that registered the A.U.R.N. events have proved to be insensitive to noise. Though, the high amplitude of the transients and their rapidness have made some researcher believe that the events are a combination of interference in the equipment of Phénix and a mechanical phenomenon. Regardless, the origin of A.U.R.N. seems to be bound to some specific parameter of Phénix due to the fact that the transients only have occurred in this reactor. A safety analysis made by an expert committee, appointed by CEA, showed that the A.U.R.N. events are not a threat to the safety of Phénix. However, the origin of these negative transients has to be found before any construction of a commercial size sodium-cooled fast reactor can begin. Thus, further research is needed.

Neutrino physics has entered a precision era in which understanding backgrounds and systematic uncertainties is particularly important. With a precise understanding of neutrino physics, we can better understand neutrino sources. In this work, we demonstrate dependency of single detector oscillation experiments on reactor neutrino flux model. We fit the largest reactor neutrino flux model error, weak magnetism, using data from experiments. We use reactor burn-up simulations in combination with a reactor neutrino flux model to demonstrate the capability of a neutrino detector to measure the power, burn-up, and plutonium content of a nuclear reactor. In particular, North Korean reactors are examined prior to the 1994 nuclear crisis and waste removal detection is examined at the Iranian reactor. The strength of a neutrino detector is that it can acquire data without the need to shut the reactor down. We also simulate tau neutrino interactions to determine backgrounds to muon neutrino and electron neutrino measurements in neutrino factory experiments.
Ph. D.

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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Thesis (MSc (Physics))--Stellenbosch University, 2010.
ENGLISH 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.

[ES] La tecnología nuclear para el uso civil genera más preocupación por la seguridad que muchas otras tecnologías que se usan a diario. La Autoridad Nuclear define las bases de cómo debe realizarse la operación segura de una Central Nuclear. De acuerdo a las directrices establecidas por la Autoridad Nuclear, una Central Nuclear debe analizar una envolvente de escenarios hipotéticos y comprobar de manera determinista que los criterios de aceptación para dicho evento se cumplen. El Análisis Determinista de Seguridad utiliza herramientas de simulación que aplican la física conocida sobre el comportamiento de la Central Nuclear para evaluar la evolución de una variable de seguridad y asegurar que los límites no se sobrepasan. El desarrollo de la tecnología informática, de los métodos matemáticos y de la física que envuelve el comportamiento de una Central Nuclear han proporcionado herra-mientas de simulación potentes que son capaces de predecir el comportamiento de las variables de seguridad con una importante precisión. Esto permite analizar escenarios de manera más realista evitando asumir condiciones conservadoras que hasta la fecha compensaban la falta de conocimiento modelado en las herramientas de simulación. Las herramientas conocidas como De Mejor Estimación son capaces de analizar even-tos transitorios en diferentes escalas. Además, emplean modelos analíticos de las dife-rentes físicas más detallados, así como correlaciones experimentales más realistas y actuales. Un paso adelante en el Análisis Determinista de Seguridad pretende combinar las diferentes herramientas de Mejor Estimación que se emplean para analizar las dis-tintas físicas de una Central Nuclear, considerando incluso la interacción entre ellas y el análisis progresivo a diferentes escalas, llegando a analizar fenómenos más locales si es necesario. Para este fin, esta tesis presenta una metodología de análisis multi-físico y multi-escala que emplea diferentes códigos de simulación analizando el escenario propuesto a dife-rentes escalas, es decir, desde un nivel de planta que incluye los distintos componentes, hasta el volumen de control que supone el refrigerante pasando entre las varillas de combustible. Esta metodología permite un flujo de información que va desde el análi-sis a mayor escala hasta el de menor escala. El desarrollo de esta metodología ha sido validado con datos de planta para poder evaluar el alcance de esta metodología y pro-porcionar nuevas líneas de trabajo futuro. Además, se han añadido los resultados de los distintos procesos de validación y verificación que han surgido a lo largo de este trabajo.
[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

A pulsed supersonic Laval nozzle flow reactor has been employed to investigate a variety of neutral-radical reaction processes at temperatures between 53 and 188 Kelvin. These supersonic flows simulate the conditions found in the Earth's upper atmosphere as well as certain environments in the interstellar medium and outer planetary atmospheres and thus provide direct information on the chemistry and physical processes occurring in those environments. Studies of this type, in the limit of 0 Kelvin, coupled with modern astronomical observations of planetary atmospheres and dense molecular clouds provide for a global understanding of chemistry in cold environments. With this in mind, the flow reactor was used to conduct fundamental studies involving the reactivity of hydroxyl (OH) and imidogen (NH) radical species with a variety of partners. More specifically, the reactions of OH+HBr and all of the H/D isotopic variants were explored between 53 and 135 K, with the goal of elucidating the kinetic isotope effects, both primary and secondary, for a reaction system occurring over a potential energy surface without an appreciable barrier, that demonstrates inverse temperature dependence. While not of direct astronomical importance, the reaction of OH+HBr does affect the partitioning of Br in the Earth's atmosphere, and knowledge of kinetic isotope effects helps one understand the chemistry leading to H/D fractionation observed in a variety of interstellar environments. The reactions of NH radical with NO, saturated, and unsaturated hydrocarbons were also studied between 53 and 188 Kelvin in the Laval nozzle flow reactor. These species were chosen as most are important constituents in the atmosphere of Titan, which is known to possess a rich organic chemistry. The reactions of NH with the unsaturated hydrocarbons are found to display negative temperature dependence over the window investigated, and are thought to proceed through an addition mechanism. Finally, the flow reactor was also coupled to a tunable vacuum and extreme ultraviolet frequency source based on four wave frequency mixing to allow for studies of radical species with their first electronic transitions in this frequency range. A discussion of the development, implementation, and future directions is included.

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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN/CNEN-SP

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2005.
Includes 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.

O presente trabalho visa contribuir com o desenvolvimento sistemático de novas metodologias experimentais da medida da reatividade de arranjos físseis subcríticos, utilizando: estatísticas de alta ordem das contagens de nêutrons com detectores no modo pulso, o recente conceito de reatividade generalizada, e as instalações do reator IPEN/MB-01. Este trabalho reuniu em um só texto diversos aspectos da implementação destes tipos de medidas. Diferentemente das demais técnicas utilizadas nas medidas da reatividade subcrítica, as metodologias apresentadas neste trabalho tem o potencial para permitir a medida experimental da reatividade subcrítica sem a necessidade da estimativa prévia de quaisquer outros parâmetros cinéticos, obtidos de forma teórica ou experimental, calibração de fontes externas ou detectores.A princípio, os métodos estatísticos de alta ordem das contagens de nêutrons permitem obter diretamente o valor da subcriticalidade (ou o fator de multiplicação) de um arranjo físsil, independentemente do modelo de física subcrítica utilizado, sem a utilização de infra-estrutura diferenciada (como uma fonte pulsada de nêutrons), sendo uma extensão natural das metodologias que utilizam estatísticas de ordens inferiores - por exemplo, Feymann-. E este conteúdo estatístico diferenciado dos momentos de altas ordens das contagens de nêutrons, o principal motivador da implementação deste trabalho. Apesar de suas potencialidades, a implementação experimental do método esbarra no tempo e taxa de aquisição de dados; ou seja, na quantidade de conteúdo estatístico necessária para a obtenção de medida útil. Exatamente esta dificuldade impediu a obtenção de uma medida útil/prática nas instalações do reator IPEN/MB-01. Existem, entretanto, outras formas de explorar estatísticas ordem superior. Por exemplo, uma extensão do método de Rossi- sugerida neste trabalho pode utilizar auto bi-correlações (coincidências triplas não acidentais de contagens). A despeito do alto valor das incertezas, os aspectos estatísticos fundamentais de uma medida foram preservados nos métodos empregados neste trabalho. O método das auto bicorrelações é conceitualmente mais robusto contra as influências do tempo morto do sistema de aquisição de dados. Ao longo de sua execução, o presente trabalho visou preencher algumas lacunas de procedimentos experimentais aparentemente pouco abordadas por outros autores, permitindo estabelecer métodos estatisticamente mais rigorosos. Entre as contribuições neste sentido destacam-se, entre outras, as correções por tempo morto ou as geradas pela correlação entre os parâmetros estatísticos em tela. Do ponto de vista teórico, este trabalho sugere duas maneiras originais de abordar o mesmo problema da utilização de estatísticas de altas ordens: (a) auto bicorrelações; e (2) os biespectros de densidade de espectral de potência própria, sendo o primeiro explorado experimentalmente/estatisticamente em detalhes.
The 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

This thesis is a follow up on "SPARC fast reactor design: Design of two passively metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control" by Tobias Lindström (2015). In this thesis the two reactors designed by Lindström in said thesis were evaluated. The goal was to determine the reactors ability to load follow as well as the burnup of the neutron absorber used in the passive control system. To be able to determine the dynamic behaviour of the reactors the reactivity feedbacks of the cores were modelled using Serpent, a Monte Carlo simulation software for 3D neutron transport calculations. These feedbacks were then implemented into a dynamic simulation of the core, primary and secondary circulation and steam generator. The secondary circulation and feedwater flow were used to regulate steam temperature and turbine power. The core was left at constant coolant flow and no control rods were used. The simulations showed that the reactor was able to load follow between 100 % and 40 % of rated power at a speed of 6 % per minute. It was also shown that the reactor could safely adjust its power between 100 % and 10 % of rated power suggesting that load following is possible below 40 % of rated power but at a lower speed. Finally the reactors were allowed compensate for the variations in a week of the Latvian wind power production in order to show one possible application of the reactor.

CIVINS
High 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.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.
Cataloged 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.

This thesis investigates Monte Carlo methods applied to criticality and time-dependent problems in reactor physics. Due to their accuracy and flexibility, Monte Carlo methods are considered as a “gold standard” in reactor physics calculations. However, the benefits come at a significant computing cost. Despite the continuous rise in easily accessible computing power, a brute-force Monte Carlo calculation of some problems is still beyond the reach of routine reactor physics analyses. The two papers on which this thesis is based try to address the computing cost issue, by proposing methods for performing Monte Carlo reactor physics calculations more efficiently. The first method addresses the efficiency of the widely-used k-eigenvalue Monte Carlo criticality calculations. It suggests, that the calculation efficiency can be increased through a gradual increase of the neutron population size simulated during each criticality cycle, and proposes a way to determine the optimal neutron population size. The second method addresses the application of Monte Carlo calculations to reactor transient problems. While reactor transient calculations can, in principle, be performed using only Monte Carlo methods, such calculations take multiple thousands of CPU hours for calculating several seconds of a transient. The proposed method offers a middle-ground approach, using a hybrid stochastic-deterministic scheme based on the response matrix formalism. Previously, the response matrix formalism was mainly considered for steady-state problems, with limited application to time-dependent problems. This thesis proposes a novel way of using information from Monte Carlo criticality calculations for solving time-dependent problems via the response matrix.
Denna 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

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2017.
This 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.

Thesis: S.B., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2018.
This 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.

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Physics, 2012.
Cataloged 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.

A new definition of the diffusion coefficient for use in reactor physics calculations is evaluated in this thesis. It is based on naturally flux-limited diffusion theory (FDT), sometimes referred to as Levermore-Pomraning diffusion theory. Another diffusion coefficient more loosely based on FDT is also evaluated in this thesis. Flux-limited diffusion theory adheres to the physical principle of flux-limiting, which is that the magnitude of neutron current is not allowed to exceed the scalar flux. Because the diffusion coefficients currently used in the nuclear industry are not flux-limited they may violate this principle in regions of large spatial gradients, and because they encompass other assumptions, they are only accurate when used in the types of calculations for which they were intended. The evaluations were performed using fine-mesh diffusion theory. They are in one spatial dimension and in 47, 4, and 2 energy groups, and were compared against a transport theory benchmark using equivalent energy structures and spatial discretization. The results show that the flux-limited diffusion coefficient (FD) outperforms the standard diffusion coefficient in calculations of single assemblies with vacuum boundaries, according to flux- and eigenvalue-errors. In single assemblies with reflective boundary conditions, the FD yielded smaller improvements, and tended to improve only the fast-group results. The results also computationally confirm that the FD adheres to flux-limiting, while the standard diffusion coefficient does not.

Depuis la fermeture de Phénix en 2010 le CEA ne possède plus de réacteur au sodium. Vus les enjeux énergétiques et le potentiel de la filière, le CEA a lancé un programme de démonstrateur industriel appelé ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), réacteur d’une puissance de 600MW électriques (1500 MW thermiques). L’objectif du prototype est double, être une réponse aux contraintes environnementales et démontrer la viabilité industrielle :• De la filière RNR-Na, avec un niveau de sureté au moins équivalent aux réacteurs de 3ème génération, du type de l’EPR. ASTRID intégrera dès la conception le retour d’expérience de Fukushima ;• Du retraitement des déchets (transmutation d’actinide mineur) et de la filière qui lui serait liée.La sûreté de l’installation est prioritaire, aucun radioélément ne doit être rejeté dans l’environnement, et ce dans toutes les situations. Pour atteindre cet objectif, il est impératif d’anticiper l’impact des nombreuses sources d’incertitudes sur le comportement du réacteur et ce dès la phase de conception. C’est dans ce contexte que s’inscrit cette thèse dont l’ambition est le développement de nouvelles méthodes d’optimisation des cœurs des RNR-Na. L’objectif est d’améliorer la robustesse et la fiabilité des réacteurs en réponse à des incertitudes existantes. Une illustration sera proposée à partir des incertitudes associées à certains régimes transitoires dimensionnant. Nous utiliserons le modèle ASTRID comme référence pour évaluer l’intérêt des nouvelles méthodes et outils développés.L’impact des incertitudes multi-Physiques sur le calcul des performances d’un cœur de RNR-Na et l’utilisation de méthodes d’optimisation introduisent de nouvelles problématiques :• Comment optimiser des cœurs « complexes » (i.e associés à des espaces de conception de dimensions élevée avec plus de 20 paramètres variables) en prenant en compte les incertitudes ?• Comment se comportent les incertitudes sur les cœurs optimisés par rapport au cœur de référence ?• En prenant en compte les incertitudes, les réacteurs sont-Ils toujours considérés comme performants ?• Les gains des optimisations obtenus à l’issue d’optimisations complexes sont-Ils supérieurs aux marges d’incertitudes (qui elles-Mêmes dépendent de l’espace paramétrique) ?La thèse contribue au développement et à la mise en place des méthodes nécessaires à la prise en compte des incertitudes dans les outils de simulation de nouvelle génération. Des méthodes statistiques pour garantir la cohérence des schémas de calculs multi-Physiques complexes sont également détaillées.En proposant de premières images de cœur de RNR-Na innovants, cette thèse présente des méthodes et des outils permettant de réduire les incertitudes sur certaines performances des réacteurs tout en les optimisant. Ces gains sont obtenus grâce à l’utilisation d’algorithmes d’optimisation multi-Objectifs. Ces méthodes permettent d’obtenir tous les compromis possibles entre les différents critères d’optimisations comme, par exemple, les compromis entre performance économique et sûreté
Since 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

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Physics, 2004.
MIT 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.

The sensitivity of two operational output parameters, criticality and isotopic composition during burnup, to specific design and operational reactor parameters in a Sodium Cooled Fast Reactor, is investigated. The computational simulation tool Serpent is used. The parameters varied include Uranium enrichment, Plutonium content, rod thickness, fuel temperature, and sodium density. In burnup, the development of the fraction of fissile isotopes, isotopes used for measurements, the isotopic composition of Plutonium, and isotopes that complicate fuel reprocessing is displayed. A surrogate model, optimized for use in determining how criticality develops between data points, is used. The results are displayed as plots created in Matlab. The results are discussed, with a focus on how large an effect varying different parameters have on different outputs related to the reactor's operation. It is concluded that the Plutonium content has the largest effect on the isotopic composition and that, based on the performed simulations, MOX fuel is potentially safer than Zirconium alloy fuel in a practical setting.

Traditionally coupled field reactor analysis has been carried out using several loosely coupled solvers, each having been developed independently from the others. In the field of multi-physics, the current generation of object-oriented toolkits provides robust close coupling of multiple fields on a single framework. This research investigates the suitability of such frameworks, in particular the Open-source Field Operation and Manipulation (OpenFOAM) framework, for the solution of spatial reactor dynamics problems. For this a subset of the theory of the Time-dependent Neutronics and Temperatures (TINTE) code, a time-dependent two-group diffusion solver, was implemented in the OpenFOAM framework. This newly created code, called diffusionFOAM, was tested for a number of steady-state and transient cases. The solver was found to perform satisfactorily, despite a number of numerical issues. The object-oriented structure of the framework allowed for rapid and efficient development of the solver. Further investigations suggest that more advanced transport methods and higher order spatial discretization schemes can potentially be implemented using such a framework as well.
Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2008.

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.
Cataloged 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.

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-03-17T10:41:16Z No. of bitstreams: 0
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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Physics, June 2010.
Cataloged 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.

Worldwide, zirconium alloys have long been utilised for fuel cladding elements and other structural components within several commercial designs of nuclear reactor owing to their high creep resistance, superior corrosion resistance in highly aggressive environments, and low cross section for neutron absorption. The purpose of this cladding material is to separate the uranium dioxide (UO2) fuel and the coolant water in order to prevent the escape of fission products, whilst also maintaining heat transfer to the coolant. Consequently, water corrosion of the fuel containments has become a key factor in the limitation of the lifetime of fuel rods within nuclear reactors, and maintaining containment integrity under corrosion is critical to ensuring safe operation and preventing accidental release of radionuclides into the cooling water. Therefore, developing an understanding of the mechanisms which govern the corrosion of zirconium-based alloys is vital, and is the motivation behind this study. The analysis of the corrosion of an unirradiated zirconium based alloy - Zircaloy-4 - is performed in a three-fold manner in this thesis. Firstly, an investigation of the metal:oxide interface is carried out, the results of which are set out in Chapter 4 of this thesis. Providing a clear understanding of the nanoscale structure and chemistry of this interface, alongside a thorough investigation of the morphology of any suboxide phases generated during the process of corrosion is fundamental to understanding the overall corrosion of this alloy. Secondly, systematic analysis of the corrosion and incorporation of SPPs into the oxide layer is performed in Chapter 5 of this thesis, in order to help inform the role of SPPs on the corrosion process for both autoclave, and more importantly, on irradiated oxides. In addition, spatially resolved chemical mapping, and correlation to the crystallographic structure provides an understanding not previously shown in the literature on the complex corrosion that takes place. Finally, Chapter 6 presents findings from studying oxygen content through the oxide scale, performed in order to quantitatively elucidate the details of the oxygen content from the outer porous oxide, through the stoichiometric ZrO2, and into any metastable suboxide layers present. This analysis highlights the significance of understanding such microstructural behaviour, in order to interpret the overall macro structural corrosion behaviour of zirconium alloys.

In this 'WOrk , the angular distributions for eTastic ·and. iBela:~ii.tc scattering of fast neutrons in fusion .reactor materials l'ia:~te<~ studied Lithium and wad material are- -likely' ';;i;"be ~n CCIUfX)nents of fusion reactor wall con£igut'atiQn qesign .. .We m=asurements were perfonnedusing an associated part;icl~,~~-; flight technique • The 14 and 14 .. 44 Mev neutrons were p~u¢ed 1;Jy. ;tli.$ T(d,n} 4He reaction with deut.erons Peinga<;eelerated in a 150kev SAME..S type Jaccelerator at ASTON and in.the 3. Mev ~~ at the Jo.i;nt Radiation. Centre I Birmingham I. res~vely; .. The q,ss.Qcj.a.~~ alpha-particles and fast. neu.tJ;qri$ Were; deteeteCl!. ~;¥.'~l :o£'·~·:p~a;§~¢; scintillator rrpunted on. a fa:st£GC1.Jsed photoroillmplj;er