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Wutzler, Whitney A. "Analysis of Accidents in Sodium-Cooled Fast Reactors." The Ohio State University, 2011. http://rave.ohiolink.edu/etdc/view?acc_num=osu1306791253.
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Gottfridsson, 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.
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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.
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Umbel, Marissa. "Containment Source Terms for Sodium-Cooled Fast Reactor Accidents." The Ohio State University, 2011. http://rave.ohiolink.edu/etdc/view?acc_num=osu1306506275.
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Sumner, Tyler. "Effects of fuel type on the safety characteristics of a sodium cooled fast reactor." Diss., Georgia Institute of Technology, 2010. http://hdl.handle.net/1853/37217.
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A series of accident simulations were performed using INL's thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric's S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. Once key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters were calculated, S-PRISM was redesigned for a Nitride fueled core to take advantage of the Nitride fuel's high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.
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Johnson, Brian C. (Brian Carl). "Application of the technology neutral framework to sodium cooled fast reactors." Thesis, Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/76499.
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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 115-116).
Sodium cooled fast reactors (SFRs) are considered as a novel example to exercise the Technology Neutral Framework (TNF) proposed in NUREG- 1860. One reason for considering SFRs is that they have historically had a licensing problem due to postulated core disruptive accidents. Two SFR designs are considered, and both meet the goals of the TNF that LWRs typically would not. Considering these goals have been met, a method for improving economics is proposed where systems of low risk-importance are identified as candidates for removal, simplification, or removal from safety grade. Seismic risk dominates these designs and is found to be a limiting factor when applying the TNF. The contributions of this thesis are the following: 1. Functional event trees are developed as a tool to allow different designs to be compared on an equal basis. Functional event trees are useful within the TNF as a method for the selection of Licensing Basis Events (LBEs) which take the place of traditional Design Basis Accidents. 2. A new importance measure, Limit Exceedance Factor (LEF), is introduced that measures the margin in system failure probability. It can be used directly with the TNF where standard importance measures cannot. It also reveals that some systems that appear to be of high risk-importance with standard importance measures may have significant margin. 3. The seismic risk dominates these designs. It is shown that even under optimistic assumptions, the goals of the TNF cannot be met by a typical reactor. The effect of seismic isolation to reduce the frequency of seismically initiated large releases is also analyzed and found to be insufficient to reach the goals of the TNF. A limit on initiating event frequency that is consistent with current practices is proposed.
by Brian C. Johnson.
Ph.D.
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Verma, Vasudha. "Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors." Doctoral thesis, Uppsala universitet, Tillämpad kärnfysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-319945.
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Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility study of self-powered neutron detectors (SPNDs) with platinum emitters as in-core power profile monitors for SFRs at full power is performed. The study shows that an SPND with a platinum emitter generates a prompt current signal induced by neutrons and gammas of the order of 600 nA/m, which is large enough to be measurable. Therefore, it is possible for the SPND to follow local power fluctuations at full power operation. Ex-core and in-core detector locations are investigated with two types of detectors, fission chambers and self-powered neutron detectors (SPNDs) respectively, to study the possibility of detection of the spatial changes in the power profile during two different transient conditions, i.e. inadvertent withdrawal of control rods (IRW) and one stuck rod during reactor shutdown (OSR). It is shown that it is possible to detect the two simulated transients with this set of ex-core and in-core detectors before any melting of the fuel takes place. The detector signal can tolerate a noise level up to 5% during an IRW and up to 1% during an OSR.
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Zhang, Youpeng. "Transmutation of Am in sodium fast reactors and accelerator driven systems." Doctoral thesis, KTH, Reaktorfysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-72427.
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In this thesis, the feasibility to use sodium cooled fast reactors loaded with MOX, metallic and nitride fuels for efficient transmutation of americium is investigated by performing transient analysis for cases with different americium contents in fuels, using safety parameters obtained with the SERPENT Monte Carlo code. It was then demonstrated that there is no solid limit for the Am introduction into oxide, metallic and nitride fuels that were loaded into sodium fast reactors. Instead, higher Am contents could be permitted if specific levels of power penalty were accepted. Transient analysis of a new Accelerator Driven System design with higher neutron source efficiency than the reference EFIT-400 design, was also performed. Based on simulation results, the suggested ADS design was proved to survive the full set of transients, preserving 130 K margin to cladding rupture during the most limiting transient. After comparing Am transmutation performances in SFRs and the suggested ADS, it can be concluded that: 1. Nitride fuel could provide the highest Am transmutation efficiency, when loaded into SFRs; 2. One SFR loaded with nitride fuel is sufficient to transmute Am inventory produced by more than 15 commercial LWRs within the same time period, which is three times higher than the supporting ratio reported for the suggested ADS; 3. The total fraction of ADS power in the power park is half of cases for critical reactors.QC 20120201
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Nitta, Christopher C. "Applying risk informed methodologies to improve the economics of sodium-cooled fast reactors." Thesis, Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/58086.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2010.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 94-97).
In order to support the increasing demand for clean sustainable electricity production and for nuclear waste management, the Sodium-Cooled Fast Reactor (SFR) is being developed. The main drawback has been its high capital and operating costs in comparison with traditional light water reactors. In order to compete, the SFR must be shown to be economically competitive. This study makes use of the proposed Technology Neutral Framework (TNF) being developed by the U.S. NRC. By applying this risk-based approach to safety, rather than the traditional approach of applying deterministic requirements, it will be shown that significant savings can be realized without compromising fundamental safety. A methodology was developed using the Technology Neutral Framework to judge design alternatives based on risk significance that provide acceptable safety within the framework at less cost. The key probabilistic metrics of Risk Achievement Worth and Limit Exceedence Factor will be used to assess whether a system or component plays an important safety function. If not the system, structure or component either can be eliminated, modified or its safety grade can be reduced resulting in cost savings. In addition, assessments were made to determine how to improve thermal efficiency by raising reactor exit temperature and by applying other design alternatives to reduce costs as evaluated on a safety, reliability and economic basis.
(cont.) This methodology was applied in a series of case studies demonstrating the value of the approach in design. The probabilistic risk assessment, the reference economic model and the Technology Neutral Framework tools required for this methodology are described. A reference economic model for a pool-type SFR was developed using the G4-ECONS model since it is an acceptable standard model for economic analysis. Since cost predictions for sodium cooled fast reactors are highly uncertain, the results of the economic analysis are used to estimate the relative improvement in cost as a function of the design alternatives proposed by the TNF methodology approach. This study used generic and comparative numbers for the ALMR and SPRISM reactors for cost of components of the SFR, to identify capital cost drivers for further study and cost reduction. For comparative purposes, the light water reactor (LWR) economic model in the G4-ECONS model was used and benchmarked to current LWR data. As a result of the case studies in which the methodology was applied, it was shown that the capital cost of the SFR could be reduced by almost 18% ($336 million) over the reference design and the levelized generating costs could be reduced by over 10% (almost 1 cent/kw-hr). These savings come largely from improvements in thermal efficiency, elimination of the energetic core disruptive accident as a design basis event and simplification of the reactor shutdown system based on risk analysis and safety significance. Should this methodology be applied to the entire plant design, it is expected that significant additional savings could be identified.
by Christopher C. Nitta.
S.M.
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Brunett, Acacia Joann. "A Methodology for Analyzing the Consequences of Accidents in Sodium-Cooled Fast Reactors." The Ohio State University, 2010. http://rave.ohiolink.edu/etdc/view?acc_num=osu1280956877.
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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.
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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.
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Lindström, Tobias. "SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-263506.
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In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristics of a metal fuelled, sodium-cooled, pool-type reactor system. Whilst mimicking the passive safety features of the IFR, the vision of the SPARC design is a battery type reactor, which can operate with minimum interference from human actors. In this thesis, two reactor examples have been developed which operate using different fuel compositions. One reactor operates on recycled nuclear waste from today's nuclear power plants, and the other reactor operates on enriched uranium. Both reactors have a thermal power of 150 MW, and are meant to operate for 30 years without refuelling. The design was developed using the ADOPT software, and was simulated in Serpent. Using Serpent, criticality analyses were carried out which show that the ARC system is able to control the long term reactivity changes of the reactors.
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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.
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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.
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Kauric, Guilhem. "Contribution to the investigation of the chemical interaction between sodium and irradiated MOX fuel for the safety of Sodium-cooled Fast Reactors." Thesis, université Paris-Saclay, 2020. http://www.theses.fr/2020UPASF027.
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Dans le cas d’un accident grave dans un réacteur refroidi au sodium, des interactions entre le combustible irradié et le sodium pourraient se produire à très haute température. Pour pouvoir prédire la dégradation des aiguilles combustibles et les phases produites en tenant compte de tous les systèmes présents dans le combustible irradié, une étude des systèmes Na-PF-Pu-U-O avec PF= Ba, Cs, I, Mo, Te est primordiale. Pour un système avec de nombreux éléments à étudier sur une large échelle de température et de composition, la méthode Calphad est particulièrement adaptée. Celle-ci permet de prédire les phases formées en fonction de la température, pression et composition du système. Les composés dans le système Na-O-Pu-U ont été synthétisé par synthèse solide avec quatre ratios Pu/(U+Pu) différents dans les nanoparticules de MOX utilisés comme précurseurs. Les structures obtenues ont été caractérisés par DRX, ²³Na-RMN et HR-XANES. Le degré d’oxydation des échantillons a systématiquement été déterminé pour le plutonium et l’uranium et des mécanismes de compensations de charge avec insertion de sodium ou de lacunes d’oxygène ont été suggéré lorsque le degré d’oxydation du plutonium mesuré était différent du théorique. Les structures des composés quaternaires dans les systèmes Ba-Mo-Na-O et Cs-Mo-Na-O ont ensuite ont été étudiés par DRX, diffraction neutronique, DRX-HT, spectroscopie Raman-HT, XAS. L’enthalpie standard de formation et l’enthalpie de décomposition ont aussi été déterminées. Finalement, la section Cs₂MoO₄-Na₂MoO₄ a été ré-étudiée par DSC et un model Calphad pour ce système a été developpéIn case of a severe accident in Sodium-cooled Fast Reactors, interactions between partly molten fuel and sodium could happen at high temperature. Therefore, to predict the degradation evolution of fuel pins and phase formation in the different systems existing in the irradiated fuel, a thorough study of the Na-FP-Pu-U-O with FP= Ba, Cs, I, Mo, Te has to be performed. For such multicomponent system and large temperature and composition range, the Calphad method is a suitable way for developing a thermodynamic database to predict the phase formation depending on the temperature, pressure and composition of the system. Compositions with four Pu/(U+Pu) ratio in the Na-O-Pu-U system were synthesised by solid state synthesis method using nanoparticules of MOX fuel and characterised by XRD, ²³Na-NMR and HR-XANES techniques. The oxidation state of plutonium and uranium in the different structures was systematiquely investigated. When the measured oxidation state of actinides was different from the theoretical one, charge compensation mechanisms were suggested either by adding sodium in the structure or oxygen vacancies. Then, the structure of quaternary compounds in the Ba-Mo-Na-O and Cs-Mo-Na-O systems were also investigated by several structural techniques (XRD, neutron diffraction, HT-XRD, HT-Raman spectroscopy, XAS). Thermodynamic properties such as standard enthalpy of formation or enthalpy of decomposition were also determined. Finally, the Cs₂MoO₄-Na₂MoO₄ pseudo-binary section was re-investigated experimentally by DSC and a Calphad model for this system was developed
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Saxena, Aakanksha. "Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor." Thesis, Aix-Marseille, 2014. http://www.theses.fr/2014AIXM4341.
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La thèse porte sur la simulation de la thermohydraulique et des transferts thermiques dans un faisceau d'aiguilles d'assemblage combustible de réacteur à neutrons rapides à caloporteur sodium.Des premiers calculs ont été réalisés par une approche moyennée de type RANS à l'aide du code industriel STAR-CCM+. De cette modélisation, il ressort une meilleure compréhension des transferts de chaleur opérés entre les aiguilles et le sodium. Les principales grandeurs macroscopiques de l'écoulement sont en accord avec les corrélations. Cependant, afin d'obtenir une description détaillée des fluctuations de température au niveau des fils espaceur, une approche plus détaillée de type LES et DNS est apparue indispensable. Pour la partie LES, le code TRIO_U a été utilisé. Concernant la partie DNS, un code de recherche a été utilisé. Ces approches requièrent des temps de calculs considérables qui ont nécessité des géométries représentatives mais simplifiées.L'approche DNS permet d'étudier l'écoulement à bas nombre de Prandtl, qui induit un comportement très différent du champ thermique relativement au champ hydraulique. Le calcul LES de l'assemblage montre que la présence du fil espaceur génère l'apparition de points chauds locaux (~20°C) en aval de celui-ci par rapport à l'écoulement sodium, au niveau de son contact avec l'aiguille. Les fluctuations de température au niveau des fils espaceur sont faibles (~1°C-2°C). En régime nominal, l'analyse spectrale montre l'absence de grande amplitude d'oscillations de température à basse fréquence (2-10 Hz); les conséquences sur la tenue mécanique des structures devront être analyséesThe thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR).First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO_U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry.The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (~20°C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (~1-2°C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures
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Russo, Genevieve V. (Genevieve Virgina). "The economics of alternative fuel cycles on sodium-cooled fast reactors and uncertainty and sensitivity analysis of cost estimates." Thesis, Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/62708.
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Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 78-79).
Previous work was done to create a baseline capital cost model for the SFR in which case studies were performed to identify ways to decrease the capital costs while maintaining safety and performance. This thesis expands on the capital cost model to include uncertainty analysis, as well as performing qualitative cost evaluations for the fuel cycle and O&M costs. An uncertainty model was developed from the available data for capital account costs, the point estimate and the contingency. After evaluating several alternative uncertainty models, a reasonable model was created by assuming the capital account costs to be lognormal distributions and assigning the point estimate as the 10th percentile and the contingency amount as the standard deviation for each of the accounts. The resulting Total Capital Cost distribution fits the mental model of cost estimating practices for Engineering, Procurement and Constructor (EPC) contractors. The Total Cost Distribution determined in this thesis has the best estimate without contingency below the 10th percentile, the best estimate plus contingency near the 50 th percentile, and the 90 th percentile equal to 110% of the best estimate plus contingency. Statistical analysis was performed on the SFR capital cost distribution and an assumed LWR capital cost distribution. The purpose of this exercise was to determine the statistical significance of the difference between the expected LWR and SFR capital costs for given mean costs and standard deviations. The result is that the estimated cost of the SFR is greater than the expected LWR cost and the difference is statistically significant. While the capital cost distributions for the two reactors overlap, the methodology provides a way to quantify the differences. While fuel cycle costs for the reprocessing alternatives, aqueous and pyroprocessing, are largely unknown, a comparison was made between the two based on several factors that could influence the cost: capital costs, waste streams, proliferation concerns, and technology readiness to inform the fuel choices for SFRs - namely oxide or metallic from a fuel cycle perspective. As a result, the aqueous process is thought to be more economical based on proven technology, reduced and lower cost waste streams, and the process costs of reprocessing and fuel fabrication. Other factors, such as the symbiotic relationship between LWRs and SFRs and the location of the facility, could have significant effects on the fuel cycle cost component of the total cost of electricity for an SFR. Centralized facilities have economic advantages in fuel cycle cost savings and depending on how spent fuel from light water reactors is treated - namely as a cost to the utility for removal, it could be a fuel cycle savings for SFRs. Based on SFR operating experience, there have been several O&M issues unique to the SFR design which have the potential to increase O&M costs. Past operating experience is not encouraging but if the lessons learned are applied in design, such as ease of inspection, monitoring, and ease of replacement of known troublesome equipment, the expectation is that there should be no inherent O&M cost differentials between LWRs and SFRs if best practices are followed. The choice of fuel type for SFRs, namely metal or oxide, is left to other operating plant capital cost decisions based on risk informed safety analyses.
by Genevieve V. Russo.
S.M.and S.B.
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Hartig, Maximilian [Verfasser], and T. [Akademischer Betreuer] Schulenberg. "Numerical Simulation of Fluid-Structure Interaction during the Expansion Phase in Sodium Cooled Fast Reactors / Maximilian Hartig ; Betreuer: T. Schulenberg." Karlsruhe : KIT-Bibliothek, 2019. http://d-nb.info/120047130X/34.
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Lázaro, Chueca Aurelio. "Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors." Doctoral thesis, Universitat Politècnica de València, 2014. http://hdl.handle.net/10251/39353.
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El Generation IV International Forum (GIF) [1] es un programa internacional dedicado a apoyar, coordinar y dirigir las iniciativas de investigación y desarrollo encaminados a implementar las soluciones tecnológicas que caracterizarán a la siguiente generación de reactores nucleares. Estos reactores se caracterizaran por una gestión más eficiente del combustible nuclear, un incremento en las exigencias de seguridad y una alta competitividad económica.
Con tales objetivos, GIF propuso una serie de diseños potencialmente capaces de alcanzarlos. Estos diseños son tecnológicamente muy distintos a las plantas nucleares comerciales actuales al utilizar neutrones de espectro rápido y consecuentemente refrigeración por metales líquidos.
Estos nuevos diseños requieren el desarrollo y validación de herramientas computacionales capaces de simular el comportamiento de la planta tanto en fase estacionaria como en transitoria y por tanto sean aplicables en los procesos de diseño y licitación de dichas plantas.
El objetivo de esta tesis es el de adaptar los códigos computacionales actuales aplicados a la simulación de reactores refrigerados por agua a reactores rápidos refrigerados por metales líquidos, tales como el sodio o el plomo y el desarrollo de modelos capaces de simular de una manera consistente el comportamientos de los sistemas ante determinados eventos que constituyen la base de diseño de la planta
Para ello se adaptaran dichos códigos a la fenomenología específica de estos reactores, se desarrollaran modelos termo-hidráulicos y neutrónicos tanto unidimensionales como tridimensionales de los diseños propuestos y se validarán los resultados para demostrar su aplicabilidad.
El trabajo incluye la implementación de correlaciones específicas para habilitar los códigos para el cálculo de la condiciones termo-hidráulicas de los refrigerantes así como la adaptación de los esquemas de acoplamiento termo-hidráulico-neutrónicos existentes a esta nueva tecnología.Lázaro Chueca, A. (2014). Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/39353
TESIS
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Richard, Joshua (Joshua Glenn). "A strategy for transition from a uranium fueled, open cycle SFR to a transuranic fueled, closed cycle sodium cooled fast reactor." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/76972.
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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 110-111).
Reactors utilizing a highly energetic neutron spectrum, often termed fast reactors, offer large fuel utilization improvements over the thermal reactors currently used for nuclear energy generation. Conventional fast reactor deployment has been hindered by the perceived need to use plutonium as fuel, coupling the commercial introduction of fast reactors to the deployment of large-scale thermal reactor used fuel reprocessing. However, the future of used fuel treatment in the United States is highly uncertain, creating a bottleneck for the introduction of fast reactor technology. A strategy centered around using uranium-fueled fast reactor cores in a once-through mode-a uranium startup fast reactor (USFR)-decouples fast reactor commercialization from fuel reprocessing and enables transition to a recycle mode once the technology becomes available and economic. The present work investigates the optimal strategy for recycling spent fuel from once-through sodium cooled fast reactors (SFRs), by analyzing the performance of various designs. A range of acceptable transitions are described and their economic, breeding, nonproliferation, and safety performance are characterized. A key finding is that the burnups of all cores were limited by the allowable fluence to the cladding rather than by the core reactivity. The carbide cores achieve fluence-limited burnups 15-25% greater than the comparable metal cores, though the metal cores can be optimized via decrementing the fuel volume fraction to reach fluence-limited burnups within 10% of the carbide cores. The removal of minor actinides from the recycled fuel has a minimal impact on the achievable burnups of both types of fuels, decreasing the fluence-limited burnup by less than half a percent in all cases. Similarly, long-term storage of the USFR fuel had minimal impact on the achievable burnups of all cores, decreasing the fluencelimited burnup by no more than 2% in all cases. Levelized fuel costs were in the range of 5.98 mills/kWh to 7.27 mills/kWh for the carbide cores, and 6.81 mills/kWh to 7.57 mills/kWh for the optimized metal cores, which is competitive with fuel costs of current LWRs and once-through SFRs. The metal and carbide multicore cores, made using slightly more than one once-through SFR core, functioned as slight fissile burners with fissile inventory ratios (FIRs) near 0.9. The uranium+ cores, made using one oncethrough SFR core plus natural uranium makeup, functioned in a fissile self-sustaining mode with FIRs near unity. All cores discharged fuel that was less attractive for weapon use than that of an LWR. The carbide cores had maximum sodium void worths in the range of $2.81-$2.86, approximately half the worth of the metal cores, which were in the range of $4.97-$5.14. Carbide and metal multicore cores possessed initial reactivities in the range of 15,000 pcm, requiring either multi-batch staggered reloading or control system modifications to achieve acceptable shutdown margins. The uranium+ carbide and metal cores achieved acceptable shutdown margin with the nominal control configuration and the single-batch reloading scheme. The overall conclusion is that USFR spent fuel is readily usable for recycle.
by Joshua Richard.
S.M.
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Romedenne, Marie Michelle. "Étude de la carburation et de la boruration d'aciers inoxydables en milieu sodium : interaction entre la gaine et le carbure de bore." Thesis, Toulouse, INPT, 2018. http://www.theses.fr/2018INPT0098/document.
Full textAbstract:
Les barres de commande du futur démonstrateur de réacteur à neutrons rapides refroidi au sodium (RNR – Na) nommé ASTRID sont constituées de pastilles de B4C enfermées dans une gaine en acier inoxydable AIM1 (15Cr-15Ni-0,4Ti). En service, les pastilles de B4C sont plongées dans le sodium liquide à une température allant de 500 à 600 °C. Les retours d’expérience des RNR - Na ont mis en évidence que la durée de vie des barres de commande était limitée par leur cinétique de carburation. Cependant, un phénomène de boruration des gaines a été observé lors d’essais réalisés « hors réacteur / hors irradiation ». Afin de maîtriser la durabilité des barres de commandes, il est donc nécessaire d’évaluer précisément la nature de l’interaction entre les gaines en acier et le B4C dans le sodium liquide. Ainsi, deux campagnes d’essai ont été menées : 1. Trois aciers inoxydables (AIM1, 316L et EM10) ont été exposés dans du sodium liquide fortement carburant (ac > 1) à 500, 600 et 650 °C. 2. Les mêmes nuances d’aciers ont été exposées dans du sodium liquide contenant de la poudre de B4C en excès à 500 et 600 °C. La première campagne a été réalisée pour avoir une meilleure compréhension des mécanismes et des cinétiques de carburation des barres de commande. Tout d’abord, l’état de carburation a été caractérisé finement au moyen de différentes techniques d’analyse (microsonde de Castaing, diffraction des rayons X du rayonnement synchrotron, microscopie électronique en transmission). Ensuite, la cinétique de carburation a été simulée à l’aide d’un modèle analytique simplifié de la carburation puis grâce à un outil commercial plus complet de simulation numérique de la diffusion à l’équilibre thermodynamique (DICTRA). Des écarts ont été observés entre les simulations des états de carburation réalisées avec DICTRA et les mesures expérimentales (profil de concentration en carbone et population de carbures). Afin de prédire au mieux l’état de carburation des aciers rencontré à 500 et 600 °C, il a notamment été démontré qu’il est probablement nécessaire de prendre en compte la diffusion du carbone dans les joints de grains et un écart à l’équilibre thermodynamique entre le carbone piégé dans les carbures et le carbone dissout dans la matrice. La deuxième campagne expérimentale a concerné l’étude du système : acier – B4C – Na. Des caractérisations couplées à des études thermodynamique et cinétique ont permis de proposer un mécanisme de carburation et de boruration des aciers. Après la dissolution du B4C dans le sodium, deux phénomènes ont été observés. Le bore réagit avec les aciers pour former une couche duplexe de borures à la surface (MB, M2B) et des borures dans les joints de grains du substrat. La cinétique de formation de la couche de borures dans les aciers suit une loi parabolique. Le carbone entraine une légère carburation des aciers plus en profondeur et le degré de carburation des aciers s’est avéré constant entre 250 et 3000 h d’exposition, ce qui suggère que le phénomène de carburation s’opère probablement avant la formation d’une couche continue de boruresPellets of boron carbide, B4C, enclosed in AIM1 (15Cr-15Ni-0.4Ti) stainless steel tubes are constitutive materials of the control rods in the future French Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID). During reactor operation, the B4C pellets are immersed in liquid sodium in the temperature range 773-873 K. Based on the feedback from operation of former Sodium Fast Reactors (SFR), the lifetime of the control rods has been shown to be limited by their carburization kinetics. Although, boriding of the steels was observed in out-ofpile studies. In order to increase the lifetime prediction of the aforementioned components in service, detailed information on the chemical interaction between the steel and B4C in liquid sodium is required. As a result, two sets of out-of-pile experiments were conducted: 1. Three stainless steels (AIM1, 316L, EM10) were exposed to highly carburizing sodium (ac > 1) at 773, 873 and 923 K. 2. The same grades were exposed to high purity B4C powder in liquid sodium at 773 and 873 K. The first campaign was performed in order to have a better understanding of the carburization phenomenology and kinetics of the control rods. The extent of carburization was evaluated. A good description of the carburization kinetics was obtained by means of two models and a simulation tool (DICTRA). The limits of the simulation tools were exposed. It was shown that the grain boundary diffusion of carbon had to be taken into account. The second set of experiments was carried out in order to study the system: steel – B4C – Na. A thorough examination of the nature of the chemical interaction was performed. The characterizations were combined with a thermodynamic and kinetic study to propose a carburization and boriding mechanism. The B4C powder dissolved in liquid sodium and reacted with the steels to form a boride layer (MB and M2B) at the surface, borides in the grain boundaries and a carburized zone underneath. The growth kinetics of the boron affected zone was shown to be parabolic. The carburization depth did not evolve between 250 and 3000 h and suggested that this phenomenon occurred during a transient stage
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Venturi, Claudia. "Development and application of a thermo-hydraulic model supporting the design and the deterministic safety analysis in a sodium cooled fast reactor." Master's thesis, Alma Mater Studiorum - Università di Bologna, 2014. http://amslaurea.unibo.it/7636/.
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Il CP-ESFR è un progetto integrato di cooperazione europeo sui reattori a sodio SFR realizzato sotto il programma quadro EURATOM 7, che unisce il contributo di venticinque partner europei.
Il CP-ESFR ha l'ambizione di contribuire all'istituzione di una "solida base scientifica e tecnica per il reattore veloce refrigerato a sodio, al fine di accelerare gli sviluppi pratici per la gestione sicura dei rifiuti radioattivi a lunga vita, per migliorare le prestazioni di sicurezza, l'efficienza delle risorse e il costo-efficacia di energia nucleare al fine di garantire un sistema solido e socialmente accettabile di protezione della popolazione e dell'ambiente contro gli effetti delle radiazioni ionizzanti. "
La presente tesi di laurea è un contributo allo sviluppo di modelli e metodi, basati sull’uso di codici termo-idraulici di sistema, per l’ analisi di sicurezza di reattori di IV Generazione refrigerati a metallo liquido.
L'attività è stata svolta nell'ambito del progetto FP-7 PELGRIMM ed in sinergia con l’Accordo di Programma MSE-ENEA(PAR-2013). Il progetto FP7 PELGRIMM ha come obbiettivo lo sviluppo di combustibili contenenti attinidi minori
1. attraverso lo studio di due diverse forme: pellet (oggetto della presente tesi) e spherepac
2. valutandone l’impatto sul progetto del reattore CP-ESFR.
La tesi propone lo sviluppo di un modello termoidraulico di sistema dei circuiti primario e intermedio del reattore con il codice RELAP5-3D© (INL, US). Tale codice, qualificato per il licenziamento dei reattori nucleari ad acqua, è stato utilizzato per valutare come variano i parametri del core del reattore rilevanti per la sicurezza (es. temperatura di camicia e di centro combustibile, temperatura del fluido refrigerante, etc.), quando il combustibile venga impiegato per “bruciare” gli attinidi minori (isotopi radioattivi a lunga vita contenuti nelle scorie nucleari). Questo ha comportato, una fase di training sul codice, sui suoi modelli e sulle sue capacità. Successivamente, lo sviluppo della nodalizzazione dell’impianto CP-ESFR, la sua qualifica, e l’analisi dei risultati ottenuti al variare della configurazione del core, del bruciamento e del tipo di combustibile impiegato (i.e. diverso arricchimento di attinidi minori).
Il testo è suddiviso in sei sezioni. La prima fornisce un’introduzione allo sviluppo tecnologico dei reattori veloci, evidenzia l’ambito in cui è stata svolta questa tesi e ne definisce obbiettivi e struttura. Nella seconda sezione, viene descritto l’impianto del CP-ESFR con attenzione alla configurazione del nocciolo e al sistema primario. La terza sezione introduce il codice di sistema termico-idraulico utilizzato per le analisi e il modello sviluppato per riprodurre l’impianto. Nella sezione quattro vengono descritti: i test e le verifiche effettuate per valutare le prestazioni del modello, la qualifica della nodalizzazione, i principali modelli e le correlazioni più rilevanti per la simulazione e le configurazioni del core considerate per l’analisi dei risultati.
I risultati ottenuti relativamente ai parametri di sicurezza del nocciolo in condizioni di normale funzionamento e per un transitorio selezionato sono descritti nella quinta sezione. Infine, sono riportate le conclusioni dell’attività.
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Garcia, Cervantes Elias Yammir. "Modeling and analysis of Superphénix fast reactor tests with the neutronic calculation scheme APOLLO-3®-SFR for application to the ASTRID reactor." Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0274.
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Le développement de RNR-Na vise à concevoir des conceptions avancées avec une sûreté améliorée. Cela représente une complexité dans la modélisation, raison pour laquelle le développement d'une nouvelle plate-forme neutronique est en cours: APOLLO-3. Dans le cadre du processus VVQI de APOLLO-3 et en soutien à ASTRID, cette thèse analyse plusieurs essais de Superphénix portant sur différentes thématiques. Les essais de démarrage neutroniques de Superphénix sont effectués grâce à des schémas innovants de calcul APOLLO-3 dans le cadre de sa validation. À partir de ce modèle, la production des données pour une analyse thermo hydraulique avec CATHARE-3, chaînée au code de mécanique du combustible, GERMINAL, permet d'évaluer un transitoire de réactivité et les coefficients de sûreté kgh. Enfin, l'application à ASTRID des schémas validés avec APOLLO-3 est effectuée. En conclusion, même si le réacteur Superphénix n'est pas entièrement représentatif d'ASTRID, différentes techniques de modélisation avec APOLLO-3® donnent des évaluations appropriéesThe SFR development aims to conceive advanced designs with improved safety. This represents a complexity on its modeling, reason for which the development of a new neutronic platform is undergoing: APOLLO-3. Under the framework of the VVUQ process of APOLLO-3 and in support to ASTRID, this thesis analyses a set of the Superphénix tests concerning multiple topics. The Superphénix start-up tests are performed for the calculation schemes validation with APOLLO-3, followed by the analysis of certain parameters at power conditions. From this model the production of the data for a thermal-hydraulic analysis with CATHARE-3, chained to the fuel performance code GERMINAL code is done to assess a transient over power and the stationary kgh coefficients. Finally, the application to ASTRID of the validated schemes with APOLLO-3 is performed. In conclusion, even if the Superphénix is not entirely representative to the ASTRID, different modeling techniques with APOLLO-3® show appropriate evaluations
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Singh, Shifali. "Radioscopie X pour les interactions corium-sodium lors d'un scénario d'accident grave." Thesis, Université Paris-Saclay (ComUE), 2019. http://www.theses.fr/2019SACLS114.
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Dans les réacteurs à neutrons rapides refroidis au sodium (RNR-Na), une défaillance dans le système de refroidissement du coeur ou de protection du réacteur peut conduire à un accident grave. Dans un tel scenario, les matériaux du coeur (combustible et acier) fondent pour former un mélange appelé corium susceptible d'interagir avec le réfrigérant (sodium). Une telle interaction (Fuel Coolant Interaction, FCI) peut générer des évènements énergétiques aptes à mettre en danger les structures du réacteur. La violence de l'explosion dépend essentiellement de l'état du mélange corium-sodium. Connaître les caractéristiques des trois phases en présence dans la zone d'interaction (corium, sodium liquide, vapeur de sodium) est donc crucial. Le manque de connaissances de la phénoménologie de l'interaction conduit à l'étudier au moyen de dispositifs expérimentaux. PLINIUS-2, la future plateforme expérimentale du CEA Cadarache consacrée aux grandes masses, sera dédiée à l'étude de la phénoménologie de l'interaction corium-réfrigérant (eau et sodium). L'objectif du présent travail est de développer, pour cette installation, un système d'imagerie basée sur des radiographies X afin de visualiser et de mieux comprendre l'interaction corium-sodium. Un algorithme de post-traitement de ces images a aussi été écrit afin de les analyser, d'en déduire la répartition des trois phases dans le système et de contribuer ainsi à l'amélioration de la modélisation numérique. Ce travail de thèse a été conduit en trois étapes. La première était consacrée à une étude bibliographique des expériences passées d'interaction corium-sodium et des mécanismes de fragmentation mis en jeu. Cette étude, complétée d'une analyse statistique des distributions de particules obtenue alors, a montré que les particules générées lors de ces expériences sont extrêmement fines avec des diamètres caractéristiques inférieurs au mm. Ces très petites tailles de particules combinées aux limites de détection de notre système d'imagerie nous ont amenés à travailler plutôt sur la détection de nuages de particules que de particules individuelles. Dans une deuxième étape, la simulation du nuage de particules de corium suivie de la conception de fantômes (maquettes 3D) représentant les trois phases du milieu ont été réalisées. Les simulations de nuages de fragments dans le sodium liquide et la vapeur ont été réalisées avec le logiciel CEA MODHERATO. Sur la base des résultats de ces calculs, certains fantômes ont été fabriqués afin de générer des images "réelles" à l'aide d'une chaîne d'imagerie existant au CEA. Ces fantômes représentatifs de la zone d'interaction corium-sodium ont été fabriqués pour faciliter le développement et la calibration du logiciel de traitement d'images. La troisième étape de cette thèse était dédiée à la conduite des expériences avec les fantômes 3D et au développement du logiciel de traitement des images ainsi obtenues. Les radiographies X des fantômes ont été générées à l'aide du système d'imagerie de l'installation KROTOS du CEA Cadarache. Les images ont ensuite été traitées avec le logiciel appelé PICSEL que nous avons développé afin de localiser les trois phases en présence dans la zone d'interaction. PICSEL a ensuite été appliqué à une expérience d'interaction corium-eau réalisée dans l'installation KROTOS dans le cadre du projet ALISA de collaboration Europe-Chine; ceci a permis une première validation du logiciel. Ainsi, lors de cette thèse, un système de radioscopie X a été défini pour visualiser l'interaction corium-sodium dans les expériences de la future installation PLINIUS-2 du CEA Cadarache. Une analyse qualitative des images produites par ce système (expériences menées avec des fantômes) est aussi réalisée avec le logiciel PICSEL développé lors de cette thèse afin de mieux caractériser le mélange, de mieux comprendre les mécanismes de l'interaction et de contribuer ainsi à l'amélioration de la sûreté des réacteurs à neutrons rapidesIn Sodium-cooled Fast Reactors (SFR), hypothetical failure of the core cooling system or the plant protection system may lead to a severe accident scenario. In such a scenario, core materials (fuel and cladding) melt down generating a hot molten mixture called corium. This corium may interact with the coolant (liquid sodium) leading to Fuel Coolant Interaction (FCI) which can generate energetic events and hence jeopardize the reactor structures. The yield of these energetic events strongly depends on the state of the corium-sodium mixture prior to the energetic event. Therefore, the knowledge of the features of the mixture composed of three-phases (i.e., corium, liquid sodium, and sodium vapor) is crucial. The lack of knowledge on the phenomenology of the interaction emphasizes the need to study it with the help of experiments. PLINIUS-2, the future large-mass experimental platform of CEA Cadarache, will be dedicated to experiments aiming at understanding the interaction phenomenology of prototypic corium with coolant (sodium and water). The present research aims to develop a high-energy X-Ray imaging system for this facility, to visualize and better understand the corium-sodium interaction. An image-processing algorithm to analyze the three-phase repartition is also developed to contribute to the improvement of numerical modeling. This Ph.D. research has been executed in three steps. In the first step, a bibliographic study of the past experiments was carried out to better understand the physics of the interaction and the mechanism of fragmentation during corium-sodium interaction. This bibliographic study, along with a statistical analysis of the particle size distribution data of various experiments conducted in the past, revealed that the particles formed in these tests are extremely fine fragments with characteristic diameters smaller than 1 mm. Due to the small particle size and the detection limitations of corium fragments in sodium with our X-Ray system, clouds of particles were detected instead of individual particles. In the second phase, the simulation of clouds of corium particles followed by the designing of phantoms (3D mock-ups) representing the 3-phase medium was carried out. Simulations of clouds of corium fragments in liquid sodium and vapor were performed using the CEA Cadarache in-house tool MODHERATO. Based on the results obtained from the simulations, certain phantoms were designed to conduct some physical experiments. These phantoms representative of the FCI interaction zone were manufactured to experimentally evaluate the performance of the radioscopy system and to facilitate the development and calibration of the image processing software. The third step of this work was dedicated to performing experiments with the phantoms and analyzing the radiographic images by developing an image processing algorithm. Experiments were carried out with phantoms in several configurations with the X-Ray radiography system at the CEA Cadarache KROTOS facility. The radioscopic images obtained were treated by developing a new comprehensive image processing and analysis code called PICSEL to identify the three phases composing the medium. Further verification and validation of the PICSEL software were carried out on a test conducted between corium and water at the KROTOS facility under the Euro-Chinese project “ALISA”. Thus, in this Ph.D. research, an X-Ray imaging system was qualified to visualize the corium-sodium interaction in the future PLINIUS-2-FR facility. A qualitative analysis of the images produced by this system was also performed using the PICSEL software to better characterize the evolution of the three-phase mixture and understand the FCI phenomenon, knowledge of which is deemed essential to improve the safety and designs of future sodium-cooled fast reactors
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Andriolo, Lena. "Impact des combustibles sphere-pac innovants sur les performances de sûreté des réacteurs à neutrons rapides refroidis au sodium." Thesis, Université Grenoble Alpes (ComUE), 2015. http://www.theses.fr/2015GREAI067/document.
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Les futurs réacteurs à neutrons rapides refroidis au sodium (RNR-Na) doivent remplir les critères GEN-IV à savoir présenter des qualités d'économie, de sûreté améliorée, de résistance à la prolifération et de minimisation des déchets. Ce projet de thèse est dédié à l'étude de l'impact des combustibles innovants (spécialement le combustible oxyde sphere-pac chargé en actinides mineurs) sur les performances de sûreté des RNR-Na dédiés à la transmutation.Le code de calcul SIMMER-III, développé à l'origine pour les phases avancées d'un accident grave, est utilisé pour les simulations. Ce code a été étendu dans le cadre de cette thèse afin d'améliorer la simulation de la phase primaire de l'accident, en introduisant le traitement des effets en réactivité liés à la dilatation du cœur et les spécificités du combustible sphere-pac (conductivité thermique, gap). Les transitoires complets (de la phase d'initiation aux phases avancées) sont simulés avec cette version étendue du code. Dans le cadre de cette thèse, les propriétés thermiques du combustible sphere-pac ont été modélisées et adaptées à SIMMER. Une méthodologie innovante tenant compte des effets en réactivité liés à la dilation thermique du cœur dans un maillage Eulérien et dans le cadre de la cinétique spatiale a ensuite été développée. A chaque pas de temps, les dimensions et densités dilatées sont calculées pour chaque cellule suite aux variations de températures. Des facteurs correctifs sont appliqués aux densités dilatées pour obtenir une configuration équivalente (en réactivité) ayant les dimensions non-dilatées et des densités modifiées. De nouvelles sections efficaces sont calculées à partir de ces densités et l'effet en réactivité lié à la dilatation est calculé. Les résultats sont prometteurs pour des dilatations uniformes et non-uniformes. Des limitations dans le cas de dilatations non-uniformes ont été identifiées et des calculs neutroniques ont été effectués en vue de futurs développements SIMMER. Les résultats préliminaires sont encourageants. Enfin, deux cœurs RNR-Na, issus du précédent projet CP-ESFR, ont été modélisés avec des combustibles sphere-pac : le Working Horse et le CONF2 (présentant un plénum sodium élargi pour une diminution de l'effet de vide sodium). Des analyses de sûreté ont été effectuées afin de fournir une première évaluation du comportement du combustible sphere-pac comparé au combustible pastille. Les deux options sont analysées en situation nominale et accidentelle (accident de perte de débit primaire) en début de vie du cœur et après irradiation. Les analyses révèlent deux phases à considérer en début de vie pour le combustible sphere-pac. Au démarrage du réacteur, ce combustible n'est pas restructuré et sa conductivité thermique est très inférieure à celle du combustible pastille. Après quelques heures sous irradiation, il se restructure suite aux importants gradients de température, ce qui améliore sa conductivité. Il se comporte alors de façon similaire au combustible pastille. Ce travail a également permis d'évaluer le comportement accidentel du cœur CONF2 qui subit un transitoire doux, prouvant que le large plénum sodium prévient efficacement de larges insertions de réactivité positive. Cependant, avec l'ajout d'américium ou suite à l'irradiation, des excursions de puissance et de réactivité plus prononcées sont observées. Ce travail a permis de démontrer que le combustible sphere-pac ne semble pas causer de problèmes de sûreté spécifiques comparé au combustible pastille, dans les conditions de simulations actuelles. La prise en compte des effets en réactivité liés à la dilatation du cœur avec cette version étendue de SIMMER retarde et réduit le potentiel énergétique lors d'un accident. Les analyses confirment également l'action atténuante du plénum sodium sur les transitoires conduisant à la vidange du sodium du coeur. Le comportement du combustible sphere-pac dans ces conditions ouvre une perspective à son utilisation en RNR-NaFuture sodium cooled fast reactors (SFRs) have to fulfill the GEN-IV requirements of enhanced safety, minimal waste production, increased proliferation resistance and high economical potential. This PhD project is dedicated to the evaluation of the impact of innovative fuels (especially minor actinides bearing oxide sphere-pac fuels) on the safety performance of advanced SFRs with transmutation option. The SIMMER-III code, originally tailored to mechanistically analyze later phases of core disruptive accidents, is employed for accident simulations. During the PhD project, the code has been extended for a better simulation of the early accident phase introducing the treatment of thermal expansion reactivity effects and for taking into account the specifics of sphere-pac fuels (thermal conductivity and gap conditions). The entire transients (from the initiating event to later accident phases) have been modeled with this extended SIMMER version. Within this PhD work, first the thermo-physical properties of sphere-pac fuel have been modeled and casted into SIMMER-III. Then, a new computational method to account for thermal expansion feedbacks has been developed to improve the initiation phase modeling of the code. The technique has the potential to evaluate these reactivity feedbacks for a fixed Eulerian mesh and in a spatial kinetics framework. At each time step, cell-wise expanded dimensions and densities are calculated based on temperature variations. Density factors are applied to the expanded densities to get an equivalent configuration (in reactivity) with original dimensions and modified densities. New cross sections are calculated with these densities and the reactivity of the equivalent configuration is computed. The developed methods show promising results for uniform and non-uniform expansions. For non-uniform expansions, model improvement needs have been identified and neutronics simulations have been carried out to support future SIMMER extensions. Preliminary results are encouraging. In the third part of the PhD, two core designs with conventional and sphere pac fuels are compared with respect to their transient behavior. These designs were established in the former CP-ESFR project: the working horse core and the optimized CONF2 core (with a large sodium plenum above the core for coolant void worth reduction). The two fuel design options are compared for steady state and transient conditions (Unprotected Loss of Flow accident, ULOF) either at beginning of life (BOL) or under irradiated conditions. Analyses for sphere-pac fuel reveal two main phases to consider at BOL. At start-up, the non-restructured sphere-pac fuel shows a low thermal conductivity compared to pellet fuel of same density. However, the fuel restructures quickly (in a few hours) due to the high thermal gradients and its thermal conductivity recovers. The fuel then shows a behavior close to the pellet one. The study also shows that the CONF2 core leads to a very mild transient for a ULOF accident at BOL. The large upper sodium plenum seems to effectively prevent large positive reactivity insertions. However, stronger reactivity and power peaks are observed under irradiated conditions or when americium is loaded in the core and lower axial blanket. This PhD work demonstrates, under current simulation conditions, that sphere-pac fuels do not seem to cause specific safety issues compared to standard pellet fuels, when loaded in SFRs. The accurate simulation of core thermal expansion reactivity feedbacks by means of the extended SIMMER version plays an important role in the accident timing (simulations confirm the expected delay in the first power peak) and on the energetic potential compared to the case where these feedbacks are omitted. The analyses also confirm the mitigating impact of a large sodium plenum on transients with voiding potential. The behavior of sphere-pac fuel in these conditions opens a perspective to its practical application in SFRs
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Faure, Bastien. "Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype." Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0289.
Full textAbstract:
Les réacteurs nucléaires refroidis au sodium offrent des perspectives intéressantes pour la filière nucléaire (utilisation optimale de l'uranium naturel, réduction de la radiotoxicité des déchets nucléaires). Cependant, la nécessité d’élever le niveau de sûreté de ces réacteurs aux standards du XXIe siècle a conduit à des designs de cœurs très hétérogènes.Ainsi, les objectifs de la thèse sont l’identification des principaux phénomènes physiques devant être pris en compte lors du calcul neutronique de cœurs hétérogènes en spectre rapide, ainsi que le développement de schémas de calcul adaptés dans le code APOLLO3 du CEA. Après quelques rappels théoriques et méthodologiques, ce document présente une analyse critique des schémas de calcul disponibles dans APOLLO3 pour les réacteurs refroidis au sodium. Cette analyse permet de mettre en évidence la nécessité de simuler, dès l’étape de préparation des sections efficaces, des modes angulaires du flux qui soient représentatifs de la configuration géométrique du cœur. Pour répondre à ce besoin dans le cadre de géométries présentant une forte hétérogénéité axiale, une approximation 2D/1D à l'équation du transport des neutrons 3D est développée. Cette dernière permet de représenter de manière cohérente, et à moindre coût, des effets d’anisotropie axiale dans des calculs 2D. Une nouvelle modélisation de type traverse de l’interface cœur / réflecteur est également proposée, ainsi qu’une méthode de calcul innovante des barres de contrôle. Ces méthodes permettent, in fine, de définir un schéma de calcul de référence unique et validé numériquement, adapté à la modélisation des cœurs de réacteurs refroidis au sodiumSodium-cooled nuclear reactors offer interesting perspectives in terms of uranium resources economy and radioactive waste management. In order to meet modern safety standards, though, increasingly complex core concepts have been proposed for this technology.Hence, the first objective of this thesis is the identification of the main physical phenomena that need to be taken into account when modeling the neutronic behavior of a heterogeneous nuclear core in a fast neutron spectrum. The second objective is the development of appropriate calculation schemes in the APOLLO3 code, developed at CEA.After a brief reminder of neutronic calculation theory and methods, this document presents a critical analysis of the neutronic calculation schemes available in APOLLO3 for sodium-cooled applications. This analysis highlights the necessity to model, during the cross section preparation phase, angular modes of the neutron flux that are representative of the core geometrical configuration. To meet this need in axially heterogeneous geometries, a 2D/1D approximation to the 3D neutron transport equation is derived and implemented in APOLLO3. In particular, it is shown that this approximation allows to consistently represent axial angular modes of the flux in 2D calculation domains. Besides, a new traverse model is proposed for the core/reflector radial interface, as well as an innovative control rod calculation method. The combination of these methods allows to define a unique, and numerically validated, reference calculation scheme in APOLLO3, suitable for the calculation of a wide range of complex sodium-cooled nuclear cores
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Zylbersztejn, Florian. "Diagnostic des vibrations coeur dans les réacteurs de 4 ieme génération par des techniques de bruit neutronique en puissance." Thesis, Grenoble, 2013. http://www.theses.fr/2013GRENI111.
Full textAbstract:
Les r'{e}acteurs rapides refroidis au sodium (RNR) font partie des six designs s'{e}lectionn'{e}s par le forum international sur les r'{e}acteurs de quatri`{e}me g'{e}n'{e}ration (GIF), et la construction d'un d'{e}monstrateur technologique est pr'{e}vue en France. Le niveau de s^{u}ret'{e} des futurs RNR doit ^{e}tre '{e}quivalent `{a} celui de l'EPR. L'am'{e}lioration de la s^{u}ret'{e} passe '{e}galement par l''{e}tude de tous les sc'{e}narios incidentels possibles et la surveillance de leurs signes pr'{e}curseurs.Les d'{e}formations m'{e}caniques du coe ur peuvent avoir un impact n'{e}faste dans les RNR, tel que des variations impr'{e}vues de la puissance due `{a} l'augmentation de r'{e}activit'{e} en cas de compaction du coeur, ou la d'{e}t'{e}rioration excessive des mat'{e}riaux de structure. La surveillance de tels ph'{e}nom`{e}nes et de leurs '{e}ventuels signes pr'{e}curseurs est recommand'{e}. Cette surveillance peut s'op'{e}rer par des d'{e}tecteurs neutroniques plac'{e}s en cuve et hors cuve.La th`{e}se concerne l''{e}tude du bruit neutronique g'{e}n'{e}r'{e} par des d'{e}formations p'{e}riodiques des coe urs RNR, restreinte aux d'{e}formations de type 'compaction coe ur', dans lesquelles l''{e}paisseur de sodium inter assemblages fluctue par flexion des assemblages (les assemblages des RNR sont tenus par leur base). La th`{e}se a '{e}t'{e} r'{e}alis'{e}e en co-tutelle au sein de la collaboration entre le CEA (France) et l'Universit'{e} Chalmers (Su`{e}de).La th`{e}se a '{e}t'{e} organis'{e}e autour de plusieurs travaux cl'{e}s:newlineUn '{e}tat de l'art sur la surveillance des d'{e}formations du coe ur par l'interpr'{e}tation des mesures de bruit dans les RNR a '{e}t'{e} r'{e}alis'{e}. La base de donn'{e}es des mesures multi-physiques constitu'{e}e sur le r'{e}acteur PHENIX a '{e}t'{e} exploit'{e}e pour permettre l'interpr'{e}tation du bruit neutronique comme '{e}tant r'{e}v'{e}lateur d'un ph'{e}nom`{e}ne de vibration m'{e}canique. Ce travail a donn'{e} lieu `{a} une publication dans Annal of Nuclear Energy. Une des conclusions importante est de souligner le manque de connaissance th'{e}orique sur le bruit neutronique ainsi que le positionnement non adapt'{e} des d'{e}tecteurs comme '{e}tant les facteurs limitants des capacit'{e}s d'interpr'{e}tation des mesures post exp'{e}riences.La collaboration avec une '{e}quipe de l'Universit'{e} Chalmers a permis l'am'{e}lioration d'un code de calcul sp'{e}cialis'{e} dans le calcul du bruit neutronique (CORESIM). Ce travail a '{e}t'{e} r'{e}alis'{e} en commenc{c}ant par une prise en main d'une version ant'{e}rieure du code CORESIM utilis'{e} pour les r'{e}acteurs thermiques et l''{e}tude du bruit neutronique induit par les fluctuations statistiques de la temp'{e}rature du mod'{e}rateur. Ce travail a donn'{e} lieu `{a} une publication dans Annal of Nuclear Energy. J'ai ensuite particip'{e} `{a} l'adaptation du code CORESIM aux sp'{e}cificit'{e}s des r'{e}acteurs rapides et `{a} sa v'{e}rification sur un mod`{e}le pratique de coeur. Ce travail a donn'{e} lieu `{a} une publication dans Annal of Nuclear Energy.Le ph'{e}nom`{e}ne de compaction coe ur a '{e}t'{e} mod'{e}lis'{e} avec deux codes de calculs (ERANOS et CORESIM) et l'application directe sur le mod`{e}le de coe ur CP-ESFR. L'impact en r'{e}activit'{e} sp'{e}cifique au coe ur CP-ESFR a '{e}t'{e} estim'{e} pour deux mod`{e}les diff'{e}rents de d'{e}formation. Le bruit neutronique g'{e}n'{e}r'{e} par ces mod`{e}les de d'{e}formations p'{e}riodiques a '{e}t'{e} '{e}galement estim'{e}. La d'{e}pendance spatiale, '{e}nerg'{e}tique et fr'{e}quentielle du bruit neutronique calcul'{e} a '{e}t'{e} analys'{e}e et contribuera aux '{e}tudes de positionnement de la future instrumentation neutronique. La soumission d'une publication bas'{e}e sur ce travail est en coursThe Sodium cooled Fast Reactor (SFR) is one of the six reactor types selected by the Generation-IV international forum (GIF), and the building of an industrial prototype is planned in France. The safety standard of the future SFR has to be equivalent to the EPR's. The general improvement of the safety of the new reactor goes through the examination of all the potentially harmful scenarios and both the study and monitoring of early signs.The mechanical deformations of the core can have harmful consequences in sodium fast reactors, such as unexpected power variations due to the reactivity increase in case of core compaction, or the excessive deterioration of the mechanical structures. The monitoring of such phenomena and of their potential early signs is then needed. The monitoring of such phenomena can be done with neutron detectors placed inside and outside the tank.This PhD thesis deals with the study of the neutron noise generated by the periodic deformation of the SFR core, restricted to the so-called core compaction or core flowering phenomenon, a deformation consisting in the variation of the inter-assembly sodium width by a radial bending the assemblies (the assemblies in SFR are held by the base). The PhD thesis has been performed within a collaboration between CEA (France) and Chalmers Institute of Technology (Sweden). The work realized during the thesis led to the publication of 3 articles as first author and another as second author.This work has embraced the following topics:newlineA state of the art of the monitoring of the core deformation phenomenon by interpretation of the noise measurements in SFR has been done. The PHENIX reactor multi physics measurements database has been scrutinized to provide an interpretation of the neutron noise bringing out a mechanical vibration phenomena. An important conclusion was that the lack of theoretical knowledge about the neutron noise induced by the vibration phenomenon and the ill positioning of the neutron detectors are the key points limiting the capacities of interpretation of noise measurements.The collaboration with the Chalmers team has allowed the improvement of a calculation code solving the neutron noise equations (CORESIM). The work has started with the use of an earlier version of CORESIM code for thermal reactors and the study of the noise induced by the statistical fluctuations of the coolant temperature. That work led to a publication in Annal of Nuclear Energy. I took part in the adaptation of the CORESIM code to the specificities of fast reactors and its application to a working version of an SFR.newlineThe modeling of the core flowering phenomenon and the direct application of the code on the CP-ESFR core case were carried out. The reactivity impact specific to the CP-ESFR core was calculated for two models of core deformations. The neutron noise induced by the modeled deformation has been then calculated. The energy, space and frequency dependence of the neutron noise has been analysed and will contribute to the instrumentation positioning question.It comes out that such a phenomena could be monitored by placing several detectors outside of the core along the same axial channel at several heights. It would also be doable to identify the noise signature by the axial noise profile. One can note that the relative noise is significantly higher at the top fuel height than in the lower fuel height. This work could be continued by designing a neutron instrumentation dedicated to the core monitoring using the proposed neutron noise technique
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Galli, Giacomo. "Etude des décharges partielles dans une chambre à fission haute température." Thesis, Université Paris-Saclay (ComUE), 2018. http://www.theses.fr/2018SACLC109/document.
Full textAbstract:
Le Commissariat à l'Energie Atomique et aux Energies Alternatives (CEA) a en charge la conception d'un réacteur à neutrons rapides de quatrième génération.L'instrumentation neutronique de ce futur réacteur s'appuiera sur des chambres à fission placées en cuve. Ces chambres à fission à haute température (CFHT) devront fonctionner à pleine puissance à une température comprise entre 400°C et 650°C.Un bilan récent de la technologie CFHT a révélé que certains points sont à améliorer afin d'en garantir une plus grande fiabilité.En particulier, on recherche une meilleure compréhension du phénomène de décharges partielles. Celles-ci engendrent des impulsions non discernables de celles produites par les fragments de fission du dépôt fissile.Par ailleurs, elles pourraient accélérer le vieillissement des isolants minéraux.En s'appuyant sur une démarche expérimentale et théorique, ce travail de thèse a apporté plusieurs résultats.Les tests sur les différentes chambres à fission ont permis de caractériser les signaux de DP, vis à vis des signaux neutroniques et de trouver une méthode efficace de discrimination DP-neutron. De la même manière, les signaux DP ont été localisés et une solution technologique a été proposée et mise en oeuvre avec succès pour les éliminer.Un outil de calcul pour la simulation des impulsions neutroniques a été conçu et testé avec succès.Une expérience sur l'effet de la température sur la courbe de Paschen, dans un volume de gaz fermé, a été conçue et réalisée en donnant les premiers résultats intéressantsThe Commission for Atomic and Alternative Energy (CEA) is in charge of the fourth generation fast neutron reactor design. The instrumentation for neutron flux measurement of this future reactor will be based on fission chambers placed in-core. These high temperature fission chambers (HTFC) will have to operate at full reactor power, and thus at a temperature between 400°C and 650°C.A recent review of HTFC technology has revealed that some points need improvement to ensure greater reliability.In particular, a better understanding of the phenomenon of partial discharges (PD), which are observed in the fission chambers at high temperature, is needed. These PD pulses are indistinguishable from those produced by the products of fission caused by collision with neutrons with the fissile deposit within thechambers.In addition, they could accelerate aging of the ceramic insulators used in the chambers.Based on both experimental and theoretical approaches, this PhD work found several results.Tests on different fission chambers made it possible to characterize the DP signals vis-a-vis the neutron signals and to find an operational DP-neutron discrimination method. The DP signals were localized and a technological solution was proposed and successfully implemented to eliminate them.A calculation tool for neutron pulse simulation was also designed and tested successfully.An experiment on the effect of temperature on the Paschen curve, in a closed gas volume, was designed and carried out giving initial interesting results
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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.
Full textAbstract:
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.
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Krentz-Wee, Rebecca (Rebecca Elizabeth). "Radial power flattening in sodium fast reactors." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/77067.
Full textAbstract:
Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012."June 2012." Cataloged from PDF version of thesis.
Includes bibliographical references (p. 13).
In order to improve a new design for a uranium startup sodium cooled fast reactor which was proposed at MIT, this thesis evaluated radial power flattening by varying the fuel volume fraction at a fixed U-235 enrichment of 18.5%. Of particular interest was how best to reduce the radial power peaking at the center of the reactor. Two cores were modeled: one with a uniform fuel volume fraction of 37% and one with fuel volume fractions which increased with the core radius. The ERANOS code was used to determine the power density, reactivity, and material values at the beginning and end of life. The varied fuel was shown to have a flatter radial power profile, but slightly lower reactivity and more mass. The power in each cell was normalized with respect to the average power; the peak power ratio in the uniform fuel volume core was 1.59, while the peak power ratio in the varied fuel volume core was 1.16, a significant improvement. The reactivity at beginning of life dropped from 12573.9 pcm to 11734.0 pcm, and stayed about 500pcm lower over the cycle, which is not a very large amount. The total mass of the heavy metals increased from the uniform core to the varied core by less than 0.9% and the mass of U-235 by 1.2%, so the varied fuel does not significantly impact the overall fuel cycle cost.
by Rebecca Krentz-Wee.
S.B.
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Kuznet︠s︡ov, I. A. Bergeron Andrea E. "Accidents and transients in fast breeder reactors /." Monterey, Calif. : Monterey Institute of International Studies, 2002. http://library.miis.edu/thesis/TI02%5F1.pdf.
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Romano, Antonino 1972. "Optimization of actinide transmutation in innovative lead-cooled fast reactors." Thesis, Massachusetts Institute of Technology, 2003. http://hdl.handle.net/1721.1/17643.
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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2003.Includes bibliographical references (leaves 228-232).
The thesis investigates the potential of fertile free fast lead-cooled modular reactors as efficient incinerators of plutonium and minor actinides (MAs) for application to dedicated fuel cycles for transmutation. A methodology is developed that assesses the actinide incineration performance of several transmutation fuel cycles, both single and multi-tiered, on the basis of performance indicators related to the reduction of actinide mass at the repository and economics. Application of the methodology indicates that fast fertile-free critical reactors that bum plutonium and MAs applied to both one-tier and two-tier fuel cycles with fuel reprocessing have the potential to be more economical than other dedicated systems like the ATW. Hence the thesis studies two lead-cooled critical reactors that employ fertile-free fuel as possible candidates. The two designs proposed are: an actinide burner reactor (ABR), designed to incinerate mostly plutonium and some MAs from the spent fuel of the conventional LWR fleet, and a minor a ctinide buumer reactor (MABR), devoted to bum mostly minor actinides and some plutonium both recycled from the burner and produced by the LWR fleet that should be integrated in a two-tier fuel cycle where most of the plutonium is fed back to advanced thermal reactors or conventional fast reactors (first tier) for incineration. These designs incorporate several advanced technical solutions such as fuel assemblies with streaming elements, double-entry control rods and special absorbing materials that establish favorable neutronics characteristics and excellent self-controllability features, comparable to those of the Integral Fast Reactor (IFR) core.
(cont.) Specifically, the designs achieve (a) negative reactivity feedbacks based on coolant voiding, Doppler and fuel and core structure thermal expansion effects; (b) satisfactory values for the effective delayed neutron fractions, comparable to those reported for the IFR; (c) proliferation resistance with dilution of the plutonium vector, which is practically unusable for nuclear weapons fabrication. The safety features of these designs is assessed by applying a thermal-hydraulic code that simulates the worst accident scenarios envisaged for the burners. Application of the fuel cycle methodology to both designs shows that a) deployment of the MABR as the burner in the two-tier fuel cycle is the most efficient strategy to manage actinides and that b) both designs yield lower fuel cycle costs compared to the subcritical reactors.
by Antonino Romano.
Ph.D.
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Gezelius, Knut 1978. "Design of compact intermediate heat exchangers for gas cooled fast reactors." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/26911.
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Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2004.Includes bibliographical references (p. 171-174).
Two aspects of an intermediate heat exchanger (IHX) for GFR service have been investigated: (1) the intrinsic characteristics of the proposed compact printed circuit heat exchanger (PCHE); and (2) a specific design optimizing economic and technical efficiency while coupling a supercritical CO2 Brayton power cycle to a helium cooled fast reactor core. In particular, the wavy channel friction factor and the effective conduction thickness between channels were evaluated by simulations using state of the art software (FluentTm). To support the competitiveness of the PCHE, it was directly compared to other potential IHX candidates with respect to performance and size for identical operating conditions. All PCHE modeling conservatively assumed straight channels and was carried out using an MIT in-house code. The PCHEs designed specifically for the He/S-CO2 cycle were designed to be deployed in a prestressed cast iron reactor vessel (PCIV) pod and to permit a net cycle efficiency of at least 40%. Optimization theory, sensitivity studies, and thermal-hydraulic constraints contributed to shaping the final design. The friction factor analysis showed that the correlations cited in the literature overestimate the value by approximately a factor of two. As regards the effective conduction thickness ratio, it was found to be around 0.6 for a 2.0 mm channel diameter. Since the value of the ratio employed in the MIT in-house code is 1.0, the results generated by the code should be conservative.
Comparing the competing IHX types clearly illustrated the advantages of using a compact design, thus favoring PCHEs and plate-fin designs. A maximum net cycle efficiency of 40.9% was achieved for the proposed cycle utilizing a low-pressure-drop reference core. The cost and core volume of this 600 MWt PCHE design were estimated to be $2.4M and 16.4 inm3, respectively. The largest uncertainty associated with the computations is whether the PCIV pod provides sufficient space for deployment of the PCHE, a blower, and other ancillary equipment. However, studies of PCHEs based on zig-zag channels indicate that the compactness can be further enhanced by a factor of 2 to 3 thanks to the increased heat transfer capability of the saw-tooth channel geometry. More research is needed to verify this projection.
by Knut Gezelius.
S.M.and S.B.
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DeWitte, Jacob D. (Jacob Dominic). "Reactor protection system design alternatives for sodium fast reactors." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76523.
Full textAbstract:
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011."January 2011." Cataloged from PDF version of thesis.
Includes bibliographical references (p. 110-112).
Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a sodium-cooled fast reactor include an energetic core disruptive accident, vessel failure, and a large early release. These consequences can be avoided if unprotected transients are properly defended against, potentially improving the economics of sodium fast reactors. One way to defend against such accidents is to include a highly reliable reactor protection system. The perceived undesirability of the consequences arising from an unprotected transient has led some sodium fast reactor designers to consider incorporating several design modifications to the reactor protection system, including: self-actuated shutdown systems, articulated control rods, and seismic anticipatory scram systems. This study investigates the performance of these systems in sodium fast reactors. To analyze the impact of these proposed design alternatives, a model to analyze plant performance that incorporates uncertainty analysis is developed using RELAP5-3D and the ABR-1000 as the reference design. The performance of the proposed alternatives is analyzed during unprotected loss of flow and unprotected transient overpower scenarios, each exacerbated by a loss of heat sink. The recently developed Technology Neutral Framework is used to contextually rate performance of the proposed alternatives. Ultimately, this thesis offers a methodology for a designer to analyze reactor protection system design efficacy. The principle results of this thesis suggest that when using the Technology Neutral Framework as a licensing framework for a sodium-cooled fast reactor, the two independent scram systems of the ABR- 1000's reactor protection system perform well enough to screen unprotected transients from the design basis. While a regulator may still require consideration of accidents involving the failure of the reactor protection system, these events will not drive the design of the system. However, self-actuated shutdown systems may be called for to diversify the reactor protection system. Of these, the Curie point latch marginally reduces the conditional cladding damage probability for metal cores because of their rapid inherent feedback effects, but is more effective for the more sluggish oxide cores given reasonably long pump coastdown times. Flow levitated absorbers are highly effective at mitigating unprotected loss of flow events for both fuel types, but are limited in response during unprotected transient overpower events. When considered from a risk-informed perspective, a clear rationale and objective is needed to justify the inclusion of an additional feature such as self-actuated shutdown systems. The use of articulated safety rods as one of the diverse means of reactivity insertion and the implementation of an anticipatory seismic scram system may be the most cost-effective alternatives to provide defense in depth in light of the sodium fast reactor's susceptibility to seismic events.
by Jacob D. DeWitte.
S.M.
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Grabaskas, David. "Analysis of Transient Overpower Scenarios in Sodium Fast Reactors." The Ohio State University, 2010. http://rave.ohiolink.edu/etdc/view?acc_num=osu1265726176.
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Can, Levent. "Analysis of coolant options for advanced metal cooled nuclear reactors." Thesis, Monterey, Calif. : Naval Postgraduate School, 2006. http://bosun.nps.edu/uhtbin/hyperion.exe/06Dec%5FCan%5FAP.pdf.
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Thesis (M.S. in Applied Physics)--Naval Postgraduate School, December 2006.Thesis Advisor(s): Craig F. Smith "December 2006." Includes bibliographical references (p. 69-70). Also available in print.
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Denman, Matthew R. "Probabilistic transient analysis of fuel choices for sodium fast reactors." Thesis, Massachusetts Institute of Technology, 2011. http://hdl.handle.net/1721.1/76578.
Full textAbstract:
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis. "June 2011."
Includes bibliographical references (p. 180-184).
This thesis presents the implications of using a risk-informed licensing framework to inform the design of Sodium Fast Reactors. NUREG-1860, more commonly known as the Technology Neutral Framework (TNF), is a risk-informed licensing process drafted by the Nuclear Regulatory Commission's (NRC) Office of Nuclear Regulatory Research. The TNF determines the acceptability of accident sequences by examining the 95th percentile estimate of both the frequency and quantity of radioactive material release and compares this value to predetermined limits on a Frequency-Consequence Curve. In order to apply this framework, two generic pool type sodium reactors, one using metal fuel and one using oxide fuel, were modeled in RELAP5-3D in order to determine the transient response of reactors to unprotected transient overpower and unprotected loss of flow events. Important transient characteristics, such as the reactivity coefficients, were treated as random variables which determine the success or failure of surviving the transient. In this context, success is defined as the cladding remaining intact and the avoidance of sodium boiling. In order to avoid running an excessive amount of simulations, the epistemic uncertainties around the random variables are sampled using importance sampling. For metallic fuel, the rate of fuel/cladding eutectic formation has typically been modeled as an Arrhenius process which depends only on the temperature of the fuel/cladding interface. Between the 1960s and the 1990s, numerous experiments have been conducted which indicate that the rate of fuel/cladding eutectic formation is more complex, depending upon fuel/cladding interfacial temperature, fuel constituents (uranium metal or uranium zirconium), cladding type (stainless steel 316, D9 or HT9), linear power, plutonium enrichment and burnup. This thesis improves the modeling accuracy of eutectic formation through the application of multivariable regression using a database of fuel/cladding eutectic experiments and determines that the remaining uncertainty governing the rate of eutectic formation should not significantly affect the frequency of cladding failure for tested cladding options. The general conclusion from this thesis is that when using NUREG-1860 to license metal or oxide fueled SFRs, it is steady state, not transient, cladding considerations which control optimal operating temperature, currently corresponding to an approximate core outlet temperature of 550°C. Metallic cores traditionally have been designed with core outlet temperatures of 510°C and increasing this temperature to 550°C may decrease the busbar cost by 19% when combined with the adoption of a Supercritical-CO₂ power conversion cycle, reduced containment requirements, and Printed Circuit Heat Exchangers. While both fuel types will be shown to meet the NUREG-1860 requirements, the frequency of radiation release for unprotected loss of flow and unprotected transient overpower events for metallic fuel has been shown to be orders of magnitude lower than for oxide fuel.
by Matthew R. Denman.
Ph.D.
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Rossi, Fabiana <1982>. "Safety, Security And safeguards In GEN IV sodium fast reactors." Doctoral thesis, Alma Mater Studiorum - Università di Bologna, 2015. http://amsdottorato.unibo.it/6836/.
Full textAbstract:
This work presents first a study of the national and international laws in the fields of safety, security and safeguards. The international treaties and the recommendations issued by the IAEA as well as the national regulations in force in France, the United States and Italy are analyzed. As a result of this, a comparison among them is presented. Given the interest of the Japan Atomic Energy Agency for the aspects of criminal penalties and monetary, also the Japanese case is analyzed.
The main part of this work was held at the JAEA in the field of proliferation resistance (PR) and physical protection (PP) of a GEN IV sodium fast reactor. For this purpose the design of the system is completed and the PR & PP methodology is applied to obtain data usable by designers for the improvement of the system itself. Due to the presence of sensitive data, not all the details can be disclosed. The reactor site of a hypothetical and commercial sodium-cooled fast neutron nuclear reactor system (SFR) is used as the target NES for the application of the methodology. The methodology is applied to all the PR and PP scenarios: diversion, misuse and breakout; theft and sabotage. The methodology is applied to the SFR to check if this system meets the target of PR and PP as described in the GIF goal; secondly, a comparison between the SFR and a LWR is performed to evaluate if and how it would be possible to improve the PR&PP of the SFR. The comparison is implemented according to the example development target: achieving PR&PP similar or superior to domestic and international ALWR. Three main actions were performed: implement the evaluation methodology; characterize the PR&PP for the nuclear energy system; identify recommendations for system designers through the comparison.
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Vivaldi, Daniele. "Modeling of underexpanded reactive CO2-into-sodium jets, in the frame of sodium fast reactors." Thesis, Saint-Etienne, EMSE, 2013. http://www.theses.fr/2013EMSE0707/document.
Full textAbstract:
Ce travail de thèse s’est inscrit dans le contexte d’utilisation d’un cycle de conversion de l’énergie de type Brayton au CO2 supercritique, pour les réacteurs à neutrons rapides refroidis au sodium (RNRNa). Dans le cas d’une fuite accidentelle dans l’échangeur de chaleur Na − CO2 d’un RNRNa, le CO2, avec une pression opérative d’environ 200 bars, serait injecté dans le sodium liquide qui se trouve à basse pression,provoquant un jet sous-détendu et réactif deCO2 dans le sodium. L’objectif principal de ce travail de thèse était le développement d’un modèle numérique du jet réactif diphasique de CO2 dans du sodium.Un modèle numérique d’un jet sous-détendu non-réactif de gaz dans du liquide, utilisant une approche3D non-stationnaire de type multi-fluide CFD, a été développé. Les résultats numériques ont été validés à travers la comparaison avec résultats expérimentaux obtenus avec mesures optiques. Un modèle décrivant la réaction chimique entre le sodium et le CO2 a été ensuite développé et intégré dans le modèle 3D multi-fluide. Le modèle résultant permet de calculer les profils de température obtenus au sein du jet et sur les parois des tubes de l’échangeur de chaleurThis PhD work was motivated by the investigations in the frame of supercritical CO2 Brayton cycles as possible energy conversion cycles for the Sodium-cooled Fast nuclear Reactors (SFRs). Following an accidental leakage inside the sodium-CO2 heat exchanger of a SFR, the CO2, having an operating pressure of about 200 bars, would be injected into the low-operatingpressure liquid sodium, creating an underexpanded reactive CO2-into-sodium jet. The goal of this PhD work is the development of a numerical model of the two-phase reactive CO2-into-sodium jet.A numerical model of an underexpanded non-reactive gas-into-liquid jet was developed, adopting a 3D unsteady multi-fluid CFD approach. The numerical results have been validated through the experimental results obtained with a facility employing optical probe technique. A numerical model for the chemical reaction between sodium and CO2 was then developed and integrated into the 3D two-fluid model. The resulting model allows to determine the temperature profiles inside the reactive jet and on the heat exchanger tubes
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Riber, Marklund Anders. "Passive acoustic leak detection in energy conversion systems of sodium fast reactors." Doctoral thesis, KTH, Reaktorteknologi, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-199415.
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Reaching the goals of Generation IV nuclear power is challenging. However, no less than six reactor concepts have been identified as capable of fulfilling the demands. Among these, the Sodium Fast Reactor (SFR), probably represents the most mature technology as about 20 SFR plants have been operated to this day. One design-specific issue of the SFR is the risk of leak and sodium-water reaction inside a steam generator. Standard monitoring is based on hydrogen detection, resulting in high sensitivity but slow response. The alternative of acoustic leak detection methods has been studied since the 1970s since they are able to respond much faster. Demonstrating low false alarm rate while detecting the fairly weak and possibly unknown acoustic signals of leaks has however proven to be difficult. Today, the CEA performs R&D, notably within the scope of the ASTRID project, with the aim of eliminating the sodium-water reaction risk. This is achieved by a Brayton cycle, using a nitrogen turbine and compact sodium-nitrogen heat exchangers. In case of a leak in this system, the low solubility of nitrogen in sodium and the high pressure in the tertiary circuit would increase the secondary pressure, locally deteriorate performance and possibly result in harmful hydrodynamic effects. Together with the risks of a potential gas leak over to the reactor, this motivates the use of leak detection also for this design. This thesis concerns passive acoustic leak detection, primarily for a SFR sodium-nitrogen heat exchanger, arguing that this method is suitable based on experiments, numerical simulations and studies on algorithms. The word “passive” here refers to a system that does not send out any signals, but rather records the noise of the plant and detects leaks as changes in this signal. The thesis covers experiments on normal operation and leak-simulating setups as well as machine-learning based detection methods intended to be of interest also for change detection in general.QC 20170109
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Garcia, Ruben. "A Paranetric Study on Core Performance of Sodium Fast Reactors Using SERPENT Code." Thesis, KTH, Fysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-101995.
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Kumar, Mithlesh. "Magnetic flux distorsion in two-phase liquid metal flow." Phd thesis, Toulouse, INPT, 2016. http://oatao.univ-toulouse.fr/15854/1/Kumar_1.pdf.
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A Generation IV Sodium cooled Fast Reactor (SFR) is being researched and developed at CEA, Cadarache France under the project named ASTRID. Monitoring gas presence in SFR is important with respect to its safe operation. In accordance with the principles of diversity, techniques based on different measurement principles have been proposed. This thesis concerns the detection and characterization of void using magnetic flux perturbation principle. An Eddy Current Flow Meter (ECFM) device is used for this purpose. From the technological point of view, the objective is to evaluate the feasibility of ECFM as a flow and/or void monitoring/characterizing device; and to determine which parameters are of interest and what are the precision of these measurements; and whether it is possible to measure the flow rate and void fraction simultaneously with the same ECFM device. From the physics point of view, the ECFM system involves the magnetic flux perturbation due to voids in the presence of Faraday induction and Lorentz force effects. Therefore ECFM integrated signal contains informations about the void, Faraday induction and Lorentz force effects based perturbation in magnetic flux and their couplings. Our objective is to understand the nature and extent of these couplings. Specific experiments have been developed to study the effects of flow velocity, void fraction and magnetic flux pulsations on the response of an ECFM. It consists in modeling the two-phase flow by a moving aluminium rod (plug flow) with holes and grooves to simulate voids. Flow velocity range of variation was 0
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Langhans, Robert Florian. "Liquid Sodium Stratication Prediction and Simulation in a Two-Dimensional Slice." Thesis, Virginia Tech, 2017. http://hdl.handle.net/10919/76725.
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In light of rising global temperatures and energy needs, nuclear power is uniquely positioned to offer carbon-free and reliable electricity. In many markets, nuclear power faces strong headwinds due to competition with other fuel sources and prohibitively high capital costs. Small Modular Reactors (SMRs), such as the proposed Advanced Fast Reactor (AFR) 100, have gained popularity in recent years as they promise economies of scale, reduced capital costs, and flexibility of deployment. Fast sodium reactors commonly feature an upper plenum with a large inventory of sodium. When temperatures change due to transients, stratification can occur. It is important to understand the stratification behavior of these large volumes because stratification can counteract natural circulation and fatigue materials.
This work features steady-state and transient simulations of thermal stratification and natural circulation of liquid sodium in a simple rectangular slice using a commercial CFD code (ANSYS FLUENT). Different inlet velocities and their effect on stratification are investigated by changing the inlet geometry. Stratification was observed in the two cases with the lowest inlet velocities. An approach for tracking the stratification interface was developed that focuses on temperature gradients rather than differences. Other authors have developed correlations to predict stratification in three dimensional enclosures. However, these correlations predict stratified conditions for all simulations even the ones that did not stratify. The previous models are modified to reflect the two-dimensional nature of the flow in the enclosure. The results align more closely with the simulations and correctly predict stratification in the investigated
cases.Master of Science
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Fricano, Joseph William. "Integrated fuel performance and thermal-hydraulic sub-channel models for analysis of sodium fast reactors." Thesis, Massachusetts Institute of Technology, 2012. http://hdl.handle.net/1721.1/76919.
Full textAbstract:
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 191-197).
Sodium Fast Reactors (SFR) show promise as an effective way to produce clean safe nuclear power while properly managing the fuel cycle. Accurate computer modeling is an important step in the design and eventual licensing of SFRs. The objective of this work was to couple a model for metal fuel performance to a sub-channel analysis code to more precisely predict critical phenomena that could lead to pin failure for steady-state and transient scenarios. The fuel code that was used is the recently developed and benchmarked FEAST-METAL code. The sub-channel analysis code that was selected is COBRA-IV-I. This code was updated with current correlations for sodium for pressure drop, mixing, and heat transfer. The new code, COBRA-IV-I-MIT was then validated with experimental data from the Oak Ridge National Laboratory (ORNL) 19-Pin Bundle, the Toshiba 37-Pin Bundle, and the Westinghouse Advanced Reactors Division (WARD) 61-Pin Bundle. Important topics that were addressed for coupling the codes include the following. The importance of azimuthal effects in the fuel pin: FEAST only evaluates the fuel in two-dimensions, assuming azimuthal symmetry; however, coupling to COBRA produces an azimuthal temperature distribution. The acceptability of assuming a two-dimensional fuel rod with an average temperature was examined. Furthermore, how the fuel pin evolves over time affects the assembly geometry. How well a two-dimensional fuel rod allows for an accurate description of the changing assembly geometry was also considered. Related to this was how the evolution of the assembly geometry affects its thermal hydraulic behavior, which determined the exact form of coupling between the codes. Ultimately one-way coupling was selected with azimuthal temperature averaging around the fuel pin. The codes were coupled using a wrapper, the COBRA And FEAST Executer (CAFE), written in the Python programming language. Data from EBR-II was used to confirm and verify CAFE. It was found that the number of axial nodes used in FEAST can have a large effect on the result. Finally FEAST was used to parametrically study three different pin designs: driver fuel, radial blanket, and tight pitch breed and bum fuel. This study provides data for pin expected life in assembly design.
by Joseph William Fricano.
Ph.D.
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Ludington, Alexander R. (Alexander Rockwell). "Tools for supercritical carbon dioxide cycle analysis and the cycle's applicability to sodium fast reactors." Thesis, Massachusetts Institute of Technology, 2009. http://hdl.handle.net/1721.1/52798.
Full textAbstract:
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009.This electronic version was submitted by the student author. The certified thesis is available in the Institute Archives and Special Collections.
"June 2009." Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (p. 127-131).
The Sodium-Cooled Fast Reactor (SFR) and the Supercritical Carbon Dioxide (S-C0₂) Recompression cycle are two technologies that have the potential to impact the power generation landscape of the future. In order for their implementation to be successful, they must compete economically with existing light water reactors and the conventional Rankine cycle. Improvements in efficiency, while maintaining safety and proliferation goals, will allow the SFR to better compete in the electricity generation market. These improvements will depend on core design as well as the balance of plant, including the choice of steam or C0₂ as the working fluid. This work has developed some of the tools necessary for evaluating different design core and balance of plant options. Much of it has concentrated on the S-C0₂ Recompression cycle. S-C0₂ promises to be useful as a working fluid in high-efficiency power conversion systems for SFRs because it achieves higher efficiencies at the high temperatures associated with SFRs. The recompression cycle is capable of operating with very high efficiencies due to the low compressor work needed when C0₂ approaches its critical point at the compressor inlet. The potential of this cycle to meet the needs of next-generation plants must be investigated across the entire range of operations and within each component of the system. A steady-state code for analysis of the recompression cycle was previously developed at MIT in the form of CYCLES II, but the present work has made significant improvements to this code that make the new version, CYCLES III, more versatile.
(cont.) This code can help to size components of the system and predict the costs and performance of the system at steady -state. Coupling of the primary and secondary loops is a major concern, the construction of the intermediate loop and associated heat exchangers (IHX) being critical to cost, efficiency, and safety. Furthermore, there is little experience in industry with large-scale compressors for S-C0₂. The experience that has been gained is typically proprietary. Most existing C0₂ compressors do not operate near the critical point and therefore, perform much like any other semi-ideal gas compressor. Accordingly, consistent, usable models of non-ideal gas compressors have been developed in the present work to produce preliminary designs and performance maps for the compressors in S-C0₂ recompression cycles. Compressor designs were developed for a 500 MWth S-C0₂ recompression cycle. The main compressor achieves an operating point total-to-static efficiency of 90.4 % and the recompressing compressor achieves 91.4 %. Further work can continue once these areas have been developed, including transient analysis, the effects of impurities on the system, and investigation of cycles which operate on other working fluids. Additionally, changes in the intermediate loop, the arrangement of the reactor vessel, and in-core changes will affect the efficiency of the SFR. These include the option of diluent grading in the fuel, flattening of the core outlet temperature profile, choosing Rankine or S-C0₂ for the balance of plant, and heat exchanger design. All these have been evaluated for their impact on plant efficiency.
(cont.) It has been determined that the S-C0₂ recompression cycle can provide efficiency benefits over conventional Rankine cycles for SFRs with core outlet temperatures at or above 510 oC. With the S-C0₂ cycle, SFRs can achieve thermal efficiencies of ~42 %.
by Alexander R. Ludington.
S.M.
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Wolniewicz, Peter. "Development of a Methodology for Detecting Coolant Void in Lead-cooled Fast Reactors by Means of Neutron Measurements." Doctoral thesis, Uppsala universitet, Tillämpad kärnfysik, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-232252.
Full textAbstract:
In a lead-cooled fast reactor (LFR), small bubbles (in the order of one mm or less) may enter the coolant from a leaking steam generator. If such a leakage is undetected the small bubbles may eventually coalesce into a larger bubble in local stagnation zones under the active core. If such a bubble or void releases and passes through the core, it could drive the reactor into prompt criticality. It is therefore desirable to be able to detect the initial stages of such void formation. In this thesis, a methodology to detect such leaks is presented together with a study on void-induced reactivity effects in various LFR's. The methodology developed is based on information from two fission chambers positioned radially outside the core. The fissile content of the fission chambers consist either of 235U or 242Pu making them sensitive to different parts of the neutron spectrum. It is shown that the information from the fission chambers can be used to obtain an early indication of the presence of a small leak within typically a month. Furthermore, it is shown that for all but the smallest LFR’s, prompt criticality due to voids passing the core cannot be excluded. One conclusion is that the methodology may form an attractive complement to the general monitoring system of future LFR’s but, as is noted, it has potential for further developments.
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Karahan, Aydin. "Modelling of thermo-mechanical and irradiation behavior of metallic and oxide fuels for sodium fast reactors." Thesis, Massachusetts Institute of Technology, 2009. http://hdl.handle.net/1721.1/57693.
Full textAbstract:
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009.Cataloged from PDF version of thesis.
Includes bibliographical references (p. 292-301).
A robust and reliable code to model the irradiation behavior of metal and oxide fuels in sodium cooled fast reactors is developed. Modeling capability was enhanced by adopting a non-empirical mechanistic approach to the extent possible, so that to increase the ability to extrapolate the existing database with a reasonable accuracy. Computational models to analyze in-reactor behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and U0 2-PuO 2 mixed oxide fuel pins have been developed and implemented into a new code, the Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe (1) Fission Gas Release and Swelling, (2) Fuel Chemistry and Restructuring, (3) Temperature Distribution, (4) Fuel Clad Chemical Interaction, (5) Fuel and Clad Mechanical Analysis and (6) Transient Creep- Fracture Model for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and cladding thermo-mechanical behavior at both steady state and design-basis accident scenarios. FEAST was written in FORTRAN-90 program language. The FEAST-METAL code mechanical analysis module implements the old Argonne National Laboratory (ANL)'s LIFE code algorithm.
(cont.) Fission gas release and swelling are modeled with the Korean GRSIS algorithm, which is based on detailed tracking of the fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of the thermo-transport theory. Fuel Clad Chemical Interaction (FCCI) models were developed for steady-state and transient situations, based on precipitation kinetics. A transient creep fracture model for the clad, based on the constrained diffusional cavity growth model, was adopted. FEAST-METAL has been benchmarked against available EBR-II database for (steady state) and furnace tests (transients). The results show that the code is able to predict important phenomena such as cladding strain, fission gas release, clad wastage, clad failure time and axial fuel slug deformation, satisfactorily. A similar code for oxide fuels, FEAST-OXIDE, was also developed. It adopts the OGRES model to describe fission gas release and swelling. However, the original OGRES model has been extended to include the effects of Joint Oxide Gain (JOG) formation on fission gas release and swelling. The fuel chemistry model includes diffusion models for radial actinide migration, cesium axial and radial migration, formation of the JOG, and variation of the oxygen to metal ratio. Fuel restructuring is also modeled, and includes the effects of porosity migration, irradiation-induced fuel densification and grain growth.
(cont.) The FEAST-OXIDE predictions has been compared to the available FFTF, EBR-II and JOYO databases, and the agreement between the code and data was found to be satisfactory. Both metal and oxide versions of FEAST are rather superior compared to many other fuel codes in the literature. Comparing metal and oxide versions, FEAST-OXIDE has a more sophisticated fission gas release and swelling model, which is based on vacancy flow. In addition, modeling of the chemistry module of the oxide fuel requires a much more detailed analysis to estimate its impact on the thermo-mechanical behavior with a reasonable accuracy. Finally, the melting of the oxide fuel and its effect on the thermo-mechanical performance have been modeled in case of transient scenarios.
by Aydin Karahan.
Ph.D.
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Ponomarev, Alexander [Verfasser], and R. [Akademischer Betreuer] Stieglitz. "Improved methodologies for evaluation of severe transient conditions of sodium-cooled fast systems / Alexander Ponomarev ; Betreuer: R. Stieglitz." Karlsruhe : KIT-Bibliothek, 2017. http://d-nb.info/1132996775/34.
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Sommer, Christopher. "Fuel cycle design and analysis of SABR subrcritical advanced burner reactor /." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2008. http://hdl.handle.net/1853/24720.
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Maddox, James Warren. "Fuel Cycle Optimization of a Helium-Cooled, Sub-Critical, Fast Transmutation of Waste Reactor with a Fusion Neutron Source." Thesis, Available online, Georgia Institute of Technology, 2006, 2006. http://etd.gatech.edu/theses/available/etd-03222006-174421/.
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Sumner, Tyler Scott. "A safety and dynamics analysis of the subcritical advanced burner reactor: SABR." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2008. http://hdl.handle.net/1853/24636.
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Guo, Hui. "Design of innovative systems for the optimized control of reactivity in Gen-IV fast neutron reactors." Thesis, Aix-Marseille, 2019. http://www.theses.fr/2019AIXM0245.
Full textAbstract:
Les réacteurs de quatrième génération pourraient tirer parti du spectre des neutrons rapides pour optimiser l'utilisation des ressources en uranium, faciliter la gestion des matières fissiles et améliorer la transmutation des déchets nucléaires. La barre de commande est habituellement combinée un réseau des aguilles non étanchées en carbure de bore (B4C) comme élément absorbant. Différents critères technologiques limitent la durée d’utilisation du B4C en irradiation. Par conséquent, des éléments absorbants alternatifs tels que l'oxyde de gadolinium (Gd2O3), l'oxyde d'europium (Eu2O3) et le diborure d'hafnium (HfB2) sont utilisés dans des configurations mixtes spécifiques contenant du modérateur pour optimiser la conception des barres de commande dans les réacteurs rapides refroidi au sodium (RNR-Na).Dans les réacteurs rapides « classiques », la barre de commande est très souvent le seul système de contrôle. Il est donc très difficile de démontrer la non-fusion du combustible lors d'accidents de remontée intempestive des barres (RIB). Par conséquent, deux conceptions de poisons consommables (PC) sont proposées et étudiées pour réduire les effets de RIB. La première conception de PC charge des actinides mineurs en mode homogène ou hybride. La deuxième conception combine B4C et modérateurs dans les assemblages dédiés.Ces conceptions sont étudiées à l'aide de schémas de calcul avancé dans APOLLO3® qui sont développés et validés dans cette thèse. Ces conceptions sont utilisées dans un grand RNR-Na industriel et dans un petit RNR-Na modulaire, et montrent leur grande flexibilité pour optimiser le contrôle de la réactivité dans les réacteurs rapidesThe Generation-IV reactors could benefit from the fast neutron spectrum to maximize the utilization of uranium resources, improve the management of fissile materials, and help the transmutation of nuclear waste. As the absorption cross-sections decrease with incident neutron energy, the fast spectrum challenges its reactivity control.The conventional control rod is a cluster of open pins with boron carbide (B4C) as the absorber. ^10B enrichment can be adjusted to satisfy the requirements of different cores. However, the operating lifetime of B4C is limited due to its characteristics under irradiation. Alternative absorbers such as gadolinium oxide (Gd2O3), europium oxide (Eu2O3) and hafnium diboride (HfB2) may present some advantages and be used with local addition of moderators to optimize the design of control rods in sodium fast reactors (SFRs).In the conventional fast reactors, the control rod is usually the only reactivity control system, which would lead to fuel melting in control rod withdrawal (CRW) accidents. Therefore, two burnable poison (BP) designs are investigated to reduce core excess reactivity and thus improve the inherent safety performance of reactors. The first BP design load minor actinides in homogenous or hybrid mode. The second BP design combines depleted B4C and moderators in dedicated assemblies.These designs are investigated using the advanced calculation scheme in APOLLO3® that is developed and validated in this thesis. These designs are applied in a large industrial SFR and a small modular SFR, which proves their excellent flexibility to optimize reactivity control in a wide range of fast reactors