1,083 research outputs found
Development and Application of Data Assimilation Methods in Reactor Physics
No full textSimulations of nuclear reactor physics can disagree significantly from experimental evidence, even when the most accurate models are used. An important part of this bias from experiment is caused by nuclear data. The nuclear data have inherent uncertainties due to the way they are evaluated, which then propagate to nuclear reactor simulations. This creates a bias and an uncertainty in a predicted reactor parameter like \keff~or the composition of spent fuel. This thesis focuses on data assimilation techniques to ameliorate the effects of nuclear data. Data assimilation takes integral experiments and assimilates them in a Bayesian way to improve simulations. It can also be used to find trends and areas needing improvement in evaluated nuclear data. The research focuses on advancing the data assimilation theory and knowledge used in reactor physics, especially on techniques that require stochastic sampling of the nuclear data. Furthermore, the research takes advantage of rich experimental data available from the Proteus research reactor at the Paul Scherrer Institute.
The thesis showed, for the first time, that two methods based on stochastic sampling (called MOCABA and BMC) gave equivalent results to each other and to the traditional method called GLLS. This was corroborated with two independent studies that used different experiments, neutron transport codes, nuclear data, and processing codes. The first study used the JEZEBEL-Pu239 benchmark, the Serpent2 neutron transport code, and NUSS. The second study used reactivity experiments from the LWR-Phase II experiments at Proteus, CASMO-5 for neutron transport, and SHARK-X. While using Serpent2, several questions pertaining to the stochastic uncertainty of its sensitivity coefficients arose. To address these, a new method called eXtended GLLS, or xGLLS, was proposed and tested in the thesis. xGLLS showed that the uncertainties associated with sensitivity coefficients have a negligible effect on the data assimilation as long as the calculated integral parameters themselves were converged. The final study focused on adjusting the fission yields and covariances made by the GEF code with post-irradiation examination experiments from Proteus. The adjustment improved the accuracy of predicted nuclide concentrations in spent fuel and improved the agreement between the GEF fission yields and those of ENDF/B-VIII.0 and JEFF3.3.LR
Multi-physics modeling of VVERs with high fidelity high-resolution codes
No full textThe calculations performed for the design and operation of a Nuclear Power Plant (NPP) are a key factor for their safety analyses. The standard for the computational analysis of NPPs is the so called conventional approach, which relies on coarse mesh diffusion for the neutronic solver and 1D channels for the T/H solver. The recent evolution of computing clusters allows the use of a novel approach, with codes performing first-principle based multi-physics simulations with high-resolution of the calculated parameters. Due to their computational cost, these methods can only be used as an audit tool of the conventional approach. In recent years, the nuclear industry has moved towards the establishment of Best Estimate Plus Uncertainty safety limits. The novel approach is a step forward to that direction. At the same time, VVER technology is expanding, with new reactors being built worldwide. However, there is currently no high-resolution tool available for VVER steady state and cycle analysis. There is a clear need for more refined simulation tools capable of handling hexagonal geometries.
The goal of this PhD is the development of a novel computational scheme based on the 3D sub-pin neutron transport code nTRACER-FAST (nTF), coupled to the sub-channel code COBRA-TF (CTF) for VVER full core steady state and cycle analysis. The coupling of a neutronic code with sub-pin resolution to a sub-channel T/H solver, as well as the use of CTF for full core VVER sub-channel analysis are one of the main novelties of this work, to the extent of the writer's knowledge. In this thesis, nTF is verified & validated for VVER 3D core standalone neutronic calculations against data published in international benchmarks. The internal sub-pin coupling of nTF to CTF for hexagonal geometries is fully developed on the course of this study. A double domain decomposition scheme that allows both codes to be executed in parallel is designed. The multi-physics core solver is verified for steady state simulations. nTF/CTF is also compared with nTF using a simplified 1D T/H solver, proving that to achieve accurate predictions, computational tools of different fidelity should not be mixed. Finally, nTF/CTF is developed for 3D full core cycle analysis. The coupled code system is validated with experimental data, achieving the target accuracy set for industrial use. The capability of the solver for sub-pin predictions throughout the depletion cycle is demonstrated.
A conventional computational route for VVER analysis, built with CASMO5-VVER as a lattice code for the nodal neutronic solver PARCS, is also established. The use of PARCS with CASMO5-VVER is another novelty of the present work. CASMO5-VVER/PARCS is verified and validated for VVER standalone neutronic and multi-physics steady state simulations and cycle analysis. The modeling options that optimize the use of the code system and the generation of CASMO5-VVER cross-sections for PARCS are described.
The comparison of the novel and conventional computational routes demonstrates the accuracy enhancement enabled by the high-resolution core solver, during steady state and depletion calculations. The work performed during this thesis also showed that the nTF/CTF computational requirements are manageable for VVER cycle analysis, when a state-of-the-art computing cluster is used. This illustrates that such high fidelity, high-resolution computational scheme can be used as an audit tool for the conventional route.LR
Performance assessment of a 3-D steady-state and spatial kinetics model for the CROCUS reactor
No full textThis dissertation covers both experimental and numerical neutronics studies to evaluate
the adequacy of the Serpent/PARCS code sequence for modeling the steady-state and
kinetics behavior of the CROCUS reactor. The reactor presents design characteristics
that raise questions about the acceptability of diffusion theory for its modeling. The
PARCS model of CROCUS has been developed considering several potential sources of
biases. More precisely, albedo boundary conditions were used to limit the axial geometry
to the grid plates where diffusion theory may lead to inaccuracies due to the presence of
Cadmium layers. Proper treatment of scattering anisotropies through in-scatter correction
of diffusion coefficients were also fundamental for producing accurate eigenvalues in the
CROCUS reactor. A parametric study has been conducted to evaluate transport effects
and the impact of energy discretization on eigenvalue and pin power distribution.
Steady-state and time-dependent experimental data has been obtained from CROCUS
with the purpose of validating the computational scheme. A comprehensive evaluation of
experimental uncertainties provided support for the generation of reliable experimental
data. Particular focus was placed upon the development of transient experiments that
involve local perturbations of the flux. Delayed neutron effects were not captured in these
transients because of the tightly coupled nature of the reactor.
The comparison of PARCS simulations against experimental data indicated that control
rod reactivity worth is predicted within (43)%. PARCS radial fission rate distributions
are in considerable disagreement with experimental data for the outer core region, where
differences are as large as 15%. This was attributed to the fact that PARCS does not
allow using adaptable mesh sizes in the radial plane, which results in a mismatch between
the mesh and explicit pins of the outer core region. However, from a safety viewpoint,
these biases are conservative and are located in the outer core region where the power is
low. PARCS axial fission rate profiles agree within 1% with experimental data for the
bottom and mid regions for the core. On the other hand, larger deviations of about 20%
were encountered for the top region, which are attributed to transport effects near the
water/air interface. Finally, the investigation on neutron kinetic effects verified that the
PARCS code is capable of modeling the transient experiments with spatial effects in the
CROCUS reactor, where maximum differences are in the order of 5%.
Overall, the Serpent/PARCS scheme shows satisfactory performance for modeling the
CROCUS reactor, except for the estimation of radial reaction rate profiles, where biases
were attributed to the impossibility of adapting the mesh size to match the fuel pitch of
both fuel zones.LR
Validation of Intrapin Reaction Rate Distributions from Deterministic Transport Codes
No full textEfforts are underway worldwide to develop advanced simulation methods of nuclear power plants, with an improved resolution in space, angle and energy. These tools aim to enhance operational efficiency through the predictions of local phenomena critical for optimizing the plant power output without degrading safety. Verification and validation procedures are essential to demonstrate the accuracy of the high-resolution information produced by these novel simulation methods; yet access to experimental data remains limited, especially regarding localized quantities such as the neutron spatial distribution within a fuel rod. The NECTAR experiments, conducted on the CROCUS zero-power reactor operated at EPFL, aim to bridge this gap by providing precise intra-pin reaction rate experimental data. The CROCUS reactor serves as an ideal testbed for validating high-fidelity simulation codes due to its relative ease of access and available space; as well as challenging core geometry. Therefore, an instrumented fuel rod for intra-pin reaction rate measurements has been designed, licensed, built and installed in CROCUS. The rod can accommodate dosimeters which can be divided post-irradiation into either 6 radial or 8 azimuthal sections, allowing to precisely measure the reaction rate spatial variations within the dosimeter. Relative radial and azimuthal intra-pin 197Au(n,g) reaction rate distributions, were measured in 3 positions within the CROCUS core lattice. Each measured reaction rate is provided with a maximum experimental uncertainty of 0.17%, which is one order of magnitude lower than existing datasets available publicly. Repeatability within the quoted experimental uncertainty was achieved for both radial and azimuthal profiles. With respect to the modeling aspects of this work, we began by developing a low-fidelity model of CROCUS using the GeN-Foam SP3 solver and an unstructured mesh description. This model's predictions were compared to existing reaction rate measurements in the core moderator, reducing discrepancies in the outer core lattice compared to an existing structured model using the nodal code PARCS. To obtain pin-resolved predictions, we developed 2 high-fidelity models: one using GeN-Foam with the discrete ordinates method and the other using MPACT with the Method Of Characteristics. Both models were verified against Serpent Monte Carlo predictions in terms of keff values, pin power maps, and inter- and intra-pin thermal neutron flux profiles. Despite discrepancies in pin power predictions for both models in the range of ±6%, attributed respectively to the definition of the cross-section homogenization regions and library issues, predictions for thermal neutron flux in the fuel were within 1% of Serpent's. Finally, predictions of the high-fidelity models were compared to the NECTAR's reaction rate measurements. GeN-Foam showed minor yet statistically significant discrepancies in terms of intra-pin radial distributions with a maximum difference of 0.8%. With respect to the azimuthal reaction rate distributions, GeN-Foam produced results with a 2% deviation from the measurements. In contrast, MPACT simulations led to large inaccuracies due to the absence of gold self-shielding data within its library. Observed deviations in the prediction of azimuthal distribution primarily originate from global effects at the reactor core scale rather than shortcomings in the deterministic solvers' ability to account for intra-pin variations.LR
Analysis of Mathieu Equation Stable Solutions in the First Zone of Stability
No full textAbstractThe paper presents the results of a homogeneous Mathieu equation studies. Mathieu equation solutions are oscillations, modulated in amplitude and frequency. In the computational experiments we found dependences of the given oscillations on the ratio of the coefficients. These dependences are shown in graphs that can be used for an approximate estimation of the Mathieu equation solutions without integration
Pensar las escalas para pensar las luchas: Autor: Mathieu UHEL
No full textA través de un título sugerente, “pensar las escalas para pensar las luchas”, Mathieu Uhel entreteje la construcción teórico-crítica del concepto escala, generada por la geografía radical anglosajona de finales del siglo XX, con la necesidad/utilidad práctica de la escala para concienciar las luchas sociales. El artículo cumple un doble propósito: por un lado, delinear los elementos de lectura sobre el concepto escala; y, con ello, promover la atención de esta problemática en las luchas contemporáneas. En un primer apartado, Uhel ubica las discusiones académicas en torno a la escala, como herramienta metodológica útil para comprender la complejidad de las sociedades capitalistas; en el segundo apartado, el autor avanza la exposición en torno al contexto de la dimensión escalar del imperialismo capitalista; finalmente, el autor se centra en el rol de la actividad política a escala nacional en la tensa relación entre las imposiciones del capital y la lucha social.Por meio de um título sugestivo, “pensando escalas para pensar lutas”, Mathieu Uhel entrelaça a construção teórico-crítica do conceito de escala, gerado pela geografia radical anglo-saxônica do final do século XX, com a necessidade / utilidade prática escala para aumentar a consciência das lutas sociais. O artigo tem um duplo propósito: por um lado, delinear os elementos de leitura sobre o conceito de escala; e, com isso, promover atenção a esse problema nas lutas contemporâneas. Na primeira seção, Uhel localiza as discussões acadêmicas em torno da escala, como uma ferramenta metodológica útil para compreender a complexidade das sociedades capitalistas; na segunda seção, o autor avança a exposição em torno do contexto da dimensão escalar do imperialismo capitalista; por fim, o autor enfoca o papel da atividade política em escala nacional na tensa relação entre as imposições do capital e a luta social.Mathieu Uhel\u27s suggestive title, “Thinking about scales to think about struggles”, he interweaves the theoretical-critical construction of concept scale, generated by radical Anglo-Saxon geography in the late 20th century, with it´s practical utility to social struggles. The article serves two purposes: on the one hand, Uhel locates academic discussion around scale; and, with this, he promotes attention to this problem in contemporary struggles. In the first section, Uhel locates academic discussions around scale, as a useful methodological tool to understand the complexity of capitalist societies; in the second section, the author advances the argument around the context of the scalar dimension of capitalist imperialism; finally, the author focuses on the role of political activity on a national scale in the tense relationship between the impositions of capital and the social movement
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Full Core, Heterogeneous, Time Dependent Neutron Transport Calculations with the 3D Code DeCART
No full textThe current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, burnup, control rod position, etc...) anticipated during the transient. The core calculation is then performed using the few group cross sections in a core simulator which uses some type of coarse mesh nodal method. The multi-step approach was identified as inadequate for several applications such as the design of MOX cores and other highly hetereogeneous, high leakage core designs. Because of the considerable advances in computing power over the last several years, there has been interest in high-fidelity solutions of the Boltzmann Transport equation. A practical approach developed for high-fidelity solutions of the 3D transport equation is the 2D-1D methodology in which the method of characteristics (MOC) is applied to the heterogeneous 2D planar problem and a lower order solution is applied to the axial problem which is, generally, more uniform. This approach was implemented in the DeCART code. Recently, there has been interest in extending such approach to the simulations of design basis accidents, such as control rod ejection accidents also known as reactivity initiated accidents (RIA). The current 2D-1D algorithm available in DeCART only provide 1D axial solution based on the diffusion theory whose accuracy deteriorates in case of strong flux gradient that can potentially be observed during RIA simulations. The primary ojective of the dissertation is to improve the accuracy and range of applicability of the DeCART code and to investigate its ability to perform a full core transient analysis of a realistic RIA. The specific research accomplishments of this work include:* The addition of more accurate 2D-1D coupling and transverse leakage splitting options to avoid the occurrence of negative source terms in the 2D MOC equations and the subsequent failure of the DeCART calculation and the improvement of the convergence of the 2D-1D method.* The implementation of a higher order transport axial solver based on NEM-Sn derivation of the Boltzmann equation. * Improved handling of thermal hydraulic feedbacks by DeCART during transient calculations.* A consistent comparison of the DeCART transient methodology with the current multistep approach (PARCS) for a realistic full core RIA.An efficient direct whole core transport calculation method involving the NEM-Sn formulation for the axial solution and the MOC for the 2-D radial solution was developed. In this solution method, the Sn neutron transport equations were developed within the framework of the Nodal Expansion Method. A RIA analysis was performed and the DeCART results were compared to the current generation of LWR core analysis methods represented by the PARCS code. In general there is good overall agreement in terms of global information from DeCART and PARCS for the RIA considered. However, the higher fidelity solution in DeCART provides a better spatial resolution that is expected to improve the accuracy of fuel performance calculations and to enable reducing the margin in several important reactor safety analysis events such as the RIA
Mathieu Ichou, Les Enfants d’immigrés à l’école
No full textIt is common to hear in the fields of educational and immigration sociology that on average, the children of immigrants do not perform as well in school as children of native-born parents. Mathieu Ichou offers an innovative sociological analysis on a topic that is heavily exploited by political and media discourse, and subject to much scientific controversy. The author takes distance from the homogenized vision of a “second generation” of students who have totally failed academically, and rep..
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