<p>The fuel assemblies used at the Swedish nuclear power plants contain typically between 100 and 300 fuel rods. An experimental technique has been demanded for determining the relative activities of specific isotopes in individual fuel rods without dismantling the assemblies. The purpose is to validate production codes, which requires an experimental relative accuracy of <2 % (1 σ).</p><p>Therefore, a new, non-destructive tomographic measurement technique for irradiated nuclear fuel assemblies has been developed. The technique includes two main steps: (1) the gamma-ray flux distribution around the assembly is recorded, and (2) the interior gamma-ray source distribution in the assembly is reconstructed. The use of detailed gamma-ray transport calculations in the reconstruction procedure enables accurate determination of the relative rod-by-rod source distribution.</p><p>To investigate the accuracy achievable, laboratory equipment has been constructed, including a fuel model with a well-known distribution of ¹³⁷Cs. Furthermore, an instrument has been constructed and built for in-pool measurements on irradiated fuel assemblies at nuclear power plants.</p><p>Using the laboratory equipment, a relative accuracy of 1.2 % was obtained (1 σ). The measurements on irradiated fuel resulted in a repeatability of 0.8 %, showing the accuracy that can be achieved using this instrument. The agreement between rod-by-rod data obtained in calculations using the POLCA–7 production code and measured data was 3.1 % (1 σ).</p><p>Additionally, there is a safeguards interest in the tomographic technique for verifying that no fissile material has been diverted from fuel assemblies, i.e. that no fuel rods have been removed or replaced. The applicability has been demonstrated in a measurement on a spent fuel assembly. Furthermore, detection of both the removal of a rod as well as the replacement with a non-active rod has been investigated in detail and quantitatively established using the laboratory equipment.</p>

The fuel assemblies used at the Swedish nuclear power plants contain typically between 100 and 300 fuel rods. An experimental technique has been demanded for determining the relative activities of specific isotopes in individual fuel rods without dismantling the assemblies. The purpose is to validate production codes, which requires an experimental relative accuracy of &lt;2 % (1 σ). Therefore, a new, non-destructive tomographic measurement technique for irradiated nuclear fuel assemblies has been developed. The technique includes two main steps: (1) the gamma-ray flux distribution around the assembly is recorded, and (2) the interior gamma-ray source distribution in the assembly is reconstructed. The use of detailed gamma-ray transport calculations in the reconstruction procedure enables accurate determination of the relative rod-by-rod source distribution. To investigate the accuracy achievable, laboratory equipment has been constructed, including a fuel model with a well-known distribution of 137Cs. Furthermore, an instrument has been constructed and built for in-pool measurements on irradiated fuel assemblies at nuclear power plants. Using the laboratory equipment, a relative accuracy of 1.2 % was obtained (1 σ). The measurements on irradiated fuel resulted in a repeatability of 0.8 %, showing the accuracy that can be achieved using this instrument. The agreement between rod-by-rod data obtained in calculations using the POLCA–7 production code and measured data was 3.1 % (1 σ). Additionally, there is a safeguards interest in the tomographic technique for verifying that no fissile material has been diverted from fuel assemblies, i.e. that no fuel rods have been removed or replaced. The applicability has been demonstrated in a measurement on a spent fuel assembly. Furthermore, detection of both the removal of a rod as well as the replacement with a non-active rod has been investigated in detail and quantitatively established using the laboratory equipment.

<p>Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel.</p><p>This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebäck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed.</p><p>In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies.</p><p>Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA-7) and the in-pool measurements of thermal power indicates that the nodal power can generally be predicted with an accuracy within 4% and the bundle power with an accuracy better than 2%, expressed as rms errors.</p><p>In-pile experiments have successfully simulated the conditions that occur in a fuel rod following a primary debris failure, being secondary fuel degradation. It was concluded that massive hydrogen pick-up takes place during the first few days following the primary failure and that a pre-oxidized layer does not function as a barrier towards hydriding in an environment with a very high partial pressure of hydrogen. Another series of in-pile experiments clearly indicate that increased UO₂ grain size is an effective way of suppressing fission gas release in LWR fuel up to the burnup level covered (55 MWd/kgUO₂).</p>

Presently there is a clear trend of increasing demands on in-pile performance of nuclear fuel. Higher target burnups, part length rods and various fuel additives are some examples of this trend. Together with an increasing demand from the public for even safer nuclear power utilisation, this implies an increased focus on various experimental, preferably non-destructive, methods to characterise the fuel. This thesis focuses on the development and experimental evaluation of such methods. In its first part, the thesis presents a method based on gamma-ray spectroscopy with germanium detectors that have been used at various power reactors in Europe. The aim with these measurements is to provide information about the thermal power distribution within fuel assemblies in order to validate core physics production codes. The early closure of the Barsebäck 1 BWR offered a unique opportunity to perform such validations before complete depletion of burnable absorbers in Gd-rods had taken place. To facilitate the measurements, a completely submersible measuring system, LOKET, was developed allowing for convenient in-pool measurements to be performed. In its second part, the thesis describes methods that utilise in-pile measurements. These methods have been used in the Halden test-reactor for determination of fission gas release, pellet-cladding interaction studies and fuel development studies. Apart from the power measurements, the LOKET device has been used for fission gas release (FGR) measurements on single fuel rods. The significant reduction in fission gas release in the modern fuel designs, in comparison with older designs, has been demonstrated in a series of experiments. A FGR database covering a wide range of burnup, power histories and fuel designs has been compiled and used for fuel performance analysis. The fission gas release has been measured on fuel rods with average burnups well above 60 MWd/kgU. The comparison between core physics calculations (PHOENIX-4/POLCA-7) and the in-pool measurements of thermal power indicates that the nodal power can generally be predicted with an accuracy within 4% and the bundle power with an accuracy better than 2%, expressed as rms errors. In-pile experiments have successfully simulated the conditions that occur in a fuel rod following a primary debris failure, being secondary fuel degradation. It was concluded that massive hydrogen pick-up takes place during the first few days following the primary failure and that a pre-oxidized layer does not function as a barrier towards hydriding in an environment with a very high partial pressure of hydrogen. Another series of in-pile experiments clearly indicate that increased UO2 grain size is an effective way of suppressing fission gas release in LWR fuel up to the burnup level covered (55 MWd/kgUO2).

This PhD study, developed at Universitat Politècnica de València (UPV), aims to cover the first phase of the benchmark released by the expert group on Uncertainty Analysis in Modeling (UAM-LWR). The main contribution to the benchmark, made by the thesis' author, is the development of a MATLAB program requested by the benchmark organizers. This is used to generate neutronic libraries to distribute among the benchmark participants. The UAM benchmark pretends to determine the uncertainty introduced by coupled multi-physics and multi-scale LWR analysis codes. The benchmark is subdivided into three phases: 1. Neutronic phase: obtain collapsed and homogenized problem-dependent cross sections and criticality analyses. 2. Core phase: standalone thermohydraulic and neutronic codes. 3. System phase: coupled thermohydraulic and neutronic code. In this thesis the objectives of the first phase are covered. Specifically, a methodology is developed to propagate the uncertainty of cross sections and other neutronic parameters through a lattice physics code and core simulator. An Uncertainty and Sensitivity (U&S) analysis is performed over the cross sections contained in the ENDF/B-VII nuclear library. Their uncertainty is propagated through the lattice physics code SCALE6.2.1, including the collapse and homogenization phase, up to the generation of problem-dependent neutronic libraries. Afterward, the uncertainty contained in these libraries can be further propagated through a core simulator, in this study PARCSv3.2. The module SAMPLER -available in the latest release of SCALE- and DAKOTA 6.3 statistical tool are used for the U&S analysis. As a part of this process, a methodology to obtain neutronic libraries in NEMTAB format -to be used in a core simulator- is also developed. A code-to-code comparison with CASMO-4 is used as a verification. The whole methodology is tested using a Boiling Water Reactor (BWR) reactor type. Nevertheless, there is not any concern or limitation regarding its use in any other type of nuclear reactor. The Gesellschaft für Anlagen und Reaktorsicherheit (GRS) stochastic methodology for uncertainty quantification is used. This methodology makes use of the high-fidelity model and nonparametric sampling to propagate the uncertainty. As a result, the number of samples (determined using the revised Wilks' formula) does not depend on the number of input parameters but only on the desired confidence and uncertainty of output parameters. Moreover, the output Probability Distribution Functions (PDFs) are not subject to normality. The main disadvantage is that each input parameter must have a pre-defined PDF. If possible, input PDFs are defined using information found in the related literature. Otherwise, the uncertainty definition is based on expert judgment. A second scenario is used to propagate the uncertainty of different thermohydraulic parameters through the coupled code TRACE5.0p3/PARCSv3.0. In this case, a PWR reactor type is used and a transient control rod drop occurrence is simulated. As a new feature, the core is modeled chan-by-chan following a fully 3D discretization. No other study is found using a detailed 3D core. This U&S analysis also makes use of the GRS methodology and DAKOTA 6.3. / Este trabajo de doctorado, desarrollado en la Universitat Politècnica de València (UPV), tiene como objetivo cubrir la primera fase del benchmark presentado por el grupo de expertos Uncertainty Analysis in Modeling (UAM-LWR). La principal contribución al benchmark, por parte del autor de esta tesis, es el desarrollo de un programa de MATLAB solicitado por los organizadores del benchmark, el cual se usa para generar librerías neutrónicas a distribuir entre los participantes del benchmark. El benchmark del UAM pretende determinar la incertidumbre introducida por los códigos multifísicos y multiescala acoplados de análisis de reactores de agua ligera. El citado benchmark se divide en tres fases: 1. Fase neutrónica: obtener los parámetros neutrónicos y secciones eficaces del problema específico colapsados y homogenizados, además del análisis de criticidad. 2. Fase de núcleo: análisis termo-hidráulico y neutrónico por separado. 3. Fase de sistema: análisis termo-hidráulico y neutrónico acoplados. En esta tesis se completan los principales objetivos de la primera fase. Concretamente, se desarrolla una metodología para propagar la incertidumbre de secciones eficaces y otros parámetros neutrónicos a través de un código lattice y un simulador de núcleo. Se lleva a cabo un análisis de incertidumbre y sensibilidad para las secciones eficaces contenidas en la librería neutrónica ENDF/B-VII. Su incertidumbre se propaga a través del código lattice SCALE6.2.1, incluyendo las fases de colapsación y homogenización, hasta llegar a la generación de una librería neutrónica específica del problema. Luego, la incertidumbre contenida en dicha librería puede continuar propagándose a través de un simulador de núcleo, para este estudio PARCSv3.2. Para el análisis de incertidumbre y sensibilidad se ha usado el módulo SAMPLER -disponible en la última versión de SCALE- y la herramienta estadística DAKOTA 6.3. Como parte de este proceso, también se ha desarrollado una metodología para obtener librerías neutrónicas en formato NEMTAB para ser usadas en simuladores de núcleo. Se ha realizado una comparación con el código CASMO-4 para obtener una verificación de la metodología completa. Esta se ha probado usando un reactor de agua en ebullición del tipo BWR. Sin embargo, no hay ninguna preocupación o limitación respecto a su uso con otro tipo de reactor nuclear. Para la cuantificación de la incertidumbre se usa la metodología estocástica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Esta metodología hace uso del modelo de alta fidelidad y un muestreo no paramétrico para propagar la incertidumbre. Como resultado, el número de muestras (determinado con la fórmula revisada de Wilks) no depende del número de parámetros de entrada, sólo depende del nivel de confianza e incertidumbre deseados de los parámetros de salida. Además, las funciones de distribución de probabilidad no están limitadas a normalidad. El principal inconveniente es que se ha de disponer de las distribuciones de probabilidad de cada parámetro de entrada. Si es posible, las distribuciones de probabilidad de entrada se definen usando información encontrada en la literatura relacionada. En caso contrario, la incertidumbre se define en base a la opinión de un experto. Se usa un segundo escenario para propagar la incertidumbre de diferentes parámetros termo-hidráulicos a través del código acoplado TRACE5.0p3/PARCSv3.0. En este caso, se utiliza un reactor tipo PWR para simular un transitorio de una caída de barra. Como nueva característica, el núcleo se modela elemento a elemento siguiendo una discretización totalmente en 3D. No se ha encontrado ningún otro estudio que use un núcleo tan detallado en 3D. También se usa la metodología GRS y el DAKOTA 6.3 para este análisis de incertidumbre y sensibilidad. / Aquest treball de doctorat, desenvolupat a la Universitat Politècnica de València (UPV), té com a objectiu cobrir la primera fase del benchmark presentat pel grup d'experts Uncertainty Analysis in Modeling (UAM-LWR). La principal contribució al benchmark, per part de l'autor d'aquesta tesi, es el desenvolupament d'un programa de MATLAB sol¿licitat pels organitzadors del benchmark, el qual s'utilitza per a generar llibreries neutròniques a distribuir entre els participants del benchmark. El benchmark del UAM pretén determinar la incertesa introduïda pels codis multifísics i multiescala acoblats d'anàlisi de reactors d'aigua lleugera. El citat benchmark es divideix en tres fases: 1. Fase neutrònica: obtenir els paràmetres neutrònics i seccions eficaces del problema específic, col¿lapsats i homogeneïtzats, a més de la anàlisi de criticitat. 2. Fase de nucli: anàlisi termo-hidràulica i neutrònica per separat. 3. Fase de sistema: anàlisi termo-hidràulica i neutrònica acoblats. En aquesta tesi es completen els principals objectius de la primera fase. Concretament, es desenvolupa una metodologia per propagar la incertesa de les seccions eficaces i altres paràmetres neutrònics a través d'un codi lattice i un simulador de nucli. Es porta a terme una anàlisi d'incertesa i sensibilitat per a les seccions eficaces contingudes en la llibreria neutrònica ENDF/B-VII. La seua incertesa es propaga a través del codi lattice SCALE6.2.1, incloent les fases per col¿lapsar i homogeneïtzar, fins aplegar a la generació d'una llibreria neutrònica específica del problema. Després, la incertesa continguda en la esmentada llibreria pot continuar propagant-se a través d'un simulador de nucli, per a aquest estudi PARCSv3.2. Per a l'anàlisi d'incertesa i sensibilitat s'ha utilitzat el mòdul SAMPLER -disponible a l'última versió de SCALE- i la ferramenta estadística DAKOTA 6.3. Com a part d'aquest procés, també es desenvolupa una metodologia per a obtenir llibreries neutròniques en format NEMTAB per ser utilitzades en simuladors de nucli. S'ha realitzat una comparació amb el codi CASMO-4 per obtenir una verificació de la metodologia completa. Aquesta s'ha provat utilitzant un reactor d'aigua en ebullició del tipus BWR. Tanmateix, no hi ha cap preocupació o limitació respecte del seu ús amb un altre tipus de reactor nuclear. Per a la quantificació de la incertesa s'utilitza la metodologia estocàstica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Aquesta metodologia fa ús del model d'alta fidelitat i un mostreig no paramètric per propagar la incertesa. Com a resultat, el nombre de mostres (determinat amb la fórmula revisada de Wilks) no depèn del nombre de paràmetres d'entrada, sols depèn del nivell de confiança i incertesa desitjats dels paràmetres d'eixida. A més, las funcions de distribució de probabilitat no estan limitades a la normalitat. El principal inconvenient és que s'ha de disposar de les distribucions de probabilitat de cada paràmetre d'entrada. Si és possible, les distribucions de probabilitat d'entrada es defineixen utilitzant informació trobada a la literatura relacionada. En cas contrari, la incertesa es defineix en base a l'opinió d'un expert. S'utilitza un segon escenari per propagar la incertesa de diferents paràmetres termo-hidràulics a través del codi acoblat TRACE5.0p3/PARCSv3.0. En aquest cas, s'utilitza un reactor tipus PWR per simular un transitori d'una caiguda de barra. Com a nova característica, cal assenyalar que el nucli es modela element a element seguint una discretizació totalment 3D. No s'ha trobat cap altre estudi que utilitze un nucli tan detallat en 3D. També s'utilitza la metodologia GRS i el DAKOTA 6.3 per a aquesta anàlisi d'incertesa i sensibilitat.¿ / Mesado Melia, C. (2017). Uncertainty Quantification and Sensitivity Analysis for Cross Sections and Thermohydraulic Parameters in Lattice and Core Physics Codes. Methodology for Cross Section Library Generation and Application to PWR and BWR [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/86167 / TESIS

Dans l’étude de tout nouveau réacteur nucléaire, la conception de son cœur est une étape décisive. Or il s’agit d’un problème complexe, qui couple fortement la neutronique, la thermomécanique du combustible et la thermo-hydraulique. Actuellement cette conception se fait par longues itérations successives entre les différentes spécialités. Afin d’optimiser de façon plus globale et complète la conception d’un cœur, une nouvelle démarche appelée FARM (FAst Reactor Methodology) a été développée dans le cadre de la thèse. Elle consiste à établir des modèles simplifiés de neutronique, mécanique et thermo-hydraulique, sous forme analytique ou d’interpolation de calculs de codes de référence, puis à les coupler, de manière à pré-dimensionner automatiquement un cœur à partir de variables d’optimisation. Une fois ce modèle établi, on peut explorer et optimiser directement de nombreux cœurs, à partir d’algorithmes génétiques de façon à améliorer leurs performances (inventaire Plutonium en cycle, …) et leur sûreté (estimateurs de sûreté pour accidents protégés et non-protégés). Une réflexion a également due être menée pour déterminer les performances d’un cœur, ainsi que la façon de prendre en compte la sûreté. Cette nouvelle approche a été utilisée pour optimiser la conception de trois concepts de cœurs de Réacteur à Neutrons Rapides refroidi au Gaz (RNR-G). Tout d’abord, la conception du RNR-G à combustible carbure et à aiguilles en SiC a pu être optimisée. Les résultats ont permis d’une part de démontrer que le cœur de référence issu de la méthode itérative était optimal (c'est-à-dire sur le front de Pareto). D’autre part, l’optimisation a également permis de proposer de nombreux autres cœurs, où en dégradant un estimateur de sûreté ou une performance (sur lesquels des marges étaient disponibles), on améliore les autres performances. Une évolution de ce concept utilisant la nouvelle technologie du buffer, a également été modélisée dans FARM et optimisée. FARM a ainsi permis de proposer les premières images de cœur GFR carbure gainé en SiC utilisant la technologie buffer, et d’estimer leurs performances. Les résultats obtenus montrent que cette innovation permet d’atteindre des cœurs beaucoup plus performants et/ou beaucoup plus « sûrs » (plusieurs profils de cœurs étant proposés). Une troisième application de FARM a été réalisée sur un concept de GFR carbure gainé en Vanadium, où là aussi FARM a proposé les premières images de cœur. Toutefois les grandes incertitudes en jeu ne permettent pas véritablement de conclure sur les performances de ce concept, qui semble prometteur.Ainsi, la faisabilité d’une optimisation globale, couplant les différentes physiques d’un cœur de réacteur nucléaire a été démontrée. Si la méthode ainsi obtenue (FARM) est moins précise que la méthode classique, elle permet d’explorer et d’optimiser beaucoup plus rapidement (en quelques semaines au lieu de quelques mois) un grand nombre de cœurs et est parfaitement adaptée pour l’étape de préconception des cœurs de réacteurs ; d’autres études détaillées permettant ensuite d’affiner l’image de cœur retenue. / In the study of any new nuclear reactor, the design of the core is an important step. However designing and optimising a reactor core is quite complex as it involves neutronics, thermal-hydraulics and fuel thermomechanics and usually design of such a system is achieved through an iterative process, involving several different disciplines. In order to solve quickly such a multi-disciplinary system, while observing the appropriate constraints, a new approach has been developed to optimise both the core performance (in-cycle Pu inventory, fuel burn-up, etc…) and the core safety characteristics (safety estimators) of a Fast Neutron Reactor. This new approach, called FARM (FAst Reactor Methodology) uses analytical models and interpolations (Metamodels) from CEA reference codes for neutronics, thermal-hydraulics and fuel behaviour, which are coupled to automatically design a core based on several optimization variables. This global core model is then linked to a genetic algorithm and used to explore and optimise new core designs with improved performance. Consideration has also been given to which parameters can be best used to define the core performance and how safety can be taken into account.This new approach has been used to optimize the design of three concepts of Gas cooled Fast Reactor (GFR). For the first one, using a SiC/SiCf-cladded carbide-fuelled helium-bonded pin, the results demonstrate that the CEA reference core obtained with the traditional iterative method was an optimal core, but among many other possibilities (that is to say on the Pareto front). The optimization also found several other cores which exhibit some improved features at the expense of other safety or performance estimators. An evolution of this concept using a “buffer”, a new technology being developed at CEA, has hence been introduced in FARM. The FARM optimisation produced several core designs using this technology, and estimated their performance. The results obtained show that this innovative feature leads to much higher performing and/or safer cores. The FARM approach has also been applied to a GFR concept using a vanadium cladding. However the large uncertainties involved do not really enable one to evaluate the performance of this promising concept.In summary, the feasibility of a global multi-disciplinary optimization has been demonstrated. Although the resulting method (FARM) is less accurate than the conventional method, it allows fast optimization and permits a large number of cores to be explored quickly, and is ideally suited for the preliminary designs studies before further refinement of the core design.