Validation of real time dispersion of Tritium clouds over the Western Mediterranean Basin in different assesments

Real time Tritium concentrations in air coming from an ITER-like reactor as source were coupled the European Centre Medium Range Weather Forecast (ECMWF) numerical model with the lagrangian atmospheric dispersion model FLEXPART. This tool ECMWF/FLEXPART was analyzed in normal operating conditions in the Western Mediterranean Basin during 45 days at summer 2010. From comparison with NORMTRI plumes over Western Mediterranean Basin the real time results have demonstrated an overestimation of the corresponding climatologically sequence Tritium concentrations in air outputs, at several distances from the reactor. For these purpose two clouds development patterns were established. The first one was following a cyclonic circulation over the Mediterranean Sea and the second one was based in the cloud delivered over the Interior of the Iberian Peninsula by another stabilized circulation corresponding to a High. One of the important remaining activities defined then, was the tool qualification. The aim of this paper is to present the ECMWF/FLEXPART products confronted with Tritium concentration in air data. For this purpose a database to develop and validate ECMWF/FLEXPART tritium in both assessments has been selected from a NORMTRI run. Similarities and differences, underestimation and overestimation with NORMTRI will allowfor refinement in some features of ECMWF/FLEXPART

Neighborhood-corrected interface discontinuity factors for multi-group pin-by-pin diffusion calculations for LWR

Performing three-dimensional pin-by-pin full core calculations based on an improved solution of the multi-group diffusion equation is an affordable option nowadays to compute accurate local safety parameters for light water reactors. Since a transport approximation is solved, appropriate correction factors, such as interface discontinuity factors, are required to nearly reproduce the fully heterogeneous transport solution. Calculating exact pin-by-pin discontinuity factors requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration; however, inaccurate correction factors are one major source of error in core analysis when using multi-group diffusion theory. An alternative to generate accurate pin-by-pin interface discontinuity factors is to build a functional-fitting that allows incorporating the environment dependence in the computed values. This paper suggests a methodology to consider the neighborhood effect based on the Analytic Coarse-Mesh Finite Difference method for the multi-group diffusion equation. It has been applied to both definitions of interface discontinuity factors, the one based on the Generalized Equivalence Theory and the one based on Black-Box Homogenization, and for different few energy groups structures. Conclusions are drawn over the optimal functional-fitting and demonstrative results are obtained with the multi-group pin-by-pin diffusion code COBAYA3 for representative PWR configurations.

A comparative study of Monte Carlo-coupled depletion codes applied to a Sodium Fast Reactor design loaded with minor actinides

inor actinides (MAs) transmutation is a main design objective of advanced nuclear systems such as generation IV Sodium Fast Reactors (SFRs). In advanced fuel cycles, MA contents in final high level waste packages are main contributors to short term heat production as well as to long-term radiotoxicity. Therefore, MA transmutation would have an impact on repository designs and would reduce the environment burden of nuclear energy. In order to predict such consequences Monte Carlo (MC) transport codes are used in reactor design tasks and they are important complements and references for routinely used deterministic computational tools. In this paper two promising Monte Carlo transport-coupled depletion codes, EVOLCODE and SERPENT, are used to examine the impact of MA burning strategies in a SFR core, 3600 MWth. The core concept proposal for MA loading in two configurations is the result of an optimization effort upon a preliminary reference design to reduce the reactivity insertion as a consequence of sodium voiding, one of the main concerns of this technology. The objective of this paper is double. Firstly, efficiencies of the two core configurations for MA transmutation are addressed and evaluated in terms of actinides mass changes and reactivity coefficients. Results are compared with those without MA loading. Secondly, a comparison of the two codes is provided. The discrepancies in the results are quantified and discussed.

Review of the C-nat(n,gamma) cross section and criticality calculations of the graphite moderated reactor BR1

A review of the experimental data for natC(n,c) and 12C(n,c) was made to identify the origin of the natC capture cross sections included in evaluated data libraries and to clarify differences observed in neutronic calculations for graphite moderated reactors using different libraries. The performance of the JEFF-3.1.2 and ENDF/B-VII.1 libraries was verified by comparing results of criticality calculations with experimental results obtained for the BR1 reactor. This reactor is an air-cooled reactor with graphite as moderator and is located at the Belgian Nuclear Research Centre SCK-CEN in Mol (Belgium). The results of this study confirm conclusions drawn from neutronic calculations of the High Temperature Engineering Test Reactor (HTTR) in Japan. Furthermore, both BR1 and HTTR calculations support the capture cross section of 12C at thermal energy which is recommended by Firestone and Révay. Additional criticality calculations were carried out in order to illustrate that the natC thermal capture cross section is important for systems with a large amount of graphite. The present study shows that only the evaluation carried out for JENDL-4.0 reflects the current status of the experimental data.

Methodologies for an improved prediction of the isotopic content in high burnup samples. Application to Vandellos-II reactor core

Fuel cycles are designed with the aim of obtaining the highest amount of energy possible. Since higher burnup values are reached, it is necessary to improve our disposal designs, traditionally based on the conservative assumption that they contain fresh fuel. The criticality calculations involved must consider burnup by making the most of the experimental and computational capabilities developed, respectively, to measure and predict the isotopic content of the spent nuclear fuel. These high burnup scenarios encourage a review of the computational tools to find out possible weaknesses in the nuclear data libraries, in the methodologies applied and their applicability range. Experimental measurements of the spent nuclear fuel provide the perfect framework to benchmark the most well-known and established codes, both in the industry and academic research activity. For the present paper, SCALE 6.0/TRITON and MONTEBURNS 2.0 have been chosen to follow the isotopic content of four samples irradiated in the Spanish Vandellós-II pressurized water reactor up to burnup values ranging from 40 GWd/MTU to 75 GWd/MTU. By comparison with the experimental data reported for these samples, we can probe the applicability of these codes to deal with high burnup problems. We have developed new computational tools within MONTENBURNS 2.0. They make possible to handle an irradiation history that includes geometrical and positional changes of the samples within the reactor core. This paper describes the irradiation scenario against which the mentioned codes and our capabilities are to be benchmarked.

Monte Carlo uncertainty propagation approaches in ADS burn-up calculations

In activation calculations, there are several approaches to quantify uncertainties: deterministic by means of sensitivity analysis, and stochastic by means of Monte Carlo. Here, two different Monte Carlo approaches for nuclear data uncertainty are presented: the first one is the Total Monte Carlo (TMC). The second one is by means of a Monte Carlo sampling of the covariance information included in the nuclear data libraries to propagate these uncertainties throughout the activation calculations. This last approach is what we named Covariance Uncertainty Propagation, CUP. This work presents both approaches and their differences. Also, they are compared by means of an activation calculation, where the cross-section uncertainties of 239Pu and 241Pu are propagated in an ADS activation calculation.

Analysis of the operator action and the single failure criteria in a SGTR sequence using best estimate assumptions with TRACE 5.0

Steam Generator Tube Rupture (SGTR) sequences in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are a special kind of transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path from the reactor coolant system to the environment. The first methodology used to perform the Deterministic Safety Analysis (DSA) of a SGTR did not credit the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that period of time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that the operators usually take more than 30 min to stop the leakage in actual sequences. Some methodologies were raised to overcome that fact, considering operator actions from the beginning of the transient, as it is done in Probabilistic Safety Analysis. This paper presents the results of comparing different assumptions regarding the single failure criteria and the operator action taken from the most common methodologies included in the different Deterministic Safety Analysis. One single failure criteria that has not been analysed previously in the literature is proposed and analysed in this paper too. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP) with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The behaviour of the reactor is quite diverse depending on the different assumptions made regarding the operator actions. On the other hand, although there are high conservatisms included in the hypothesis, as the single failure criteria, all the results are quite far from the regulatory limits. In addition, some improvements to the Emergency Operating Procedures to minimize the offsite release from the damaged SG in case of a SGTR are outlined taking into account the offsite dose sensitivity results.

Concrete constitutive model, calibration and applications

A methodology has been developed for characterising the mechanical behaviour of concrete, based on the damaged plasticity model, enriched with a user subroutine (V)USDFLD in order to capture better the ductility of the material under moderate confining pressures. The model has been applied in the context of the international benchmark IRIS_2012, organised by the OECD/NEA/CSNI Nuclear Energy Agency, dealing with impacts of rigid and deformable missiles against reinforced concrete targets. A slightly modified version of the concrete damaged plasticity model was used to represent the concrete. The simulation results matched very well the observations made during the actual tests. Particularly successful predictions involved the energy spent by the rigid missile in perforating the target, the crushed length of the deformable missile, the crushed and cracked areas of the concrete target, and the values of the strains recorded at a number of locations in the concrete slab.

Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

Comportamiento mecánico de nuevas aleaciones de wolframio en función de la temperatura

La presente memoria de tesis tiene como objetivo principal la caracterización mecánica en función de la temperatura de nueve aleaciones de wolframio con contenidos diferentes en titanio, vanadio, itria y lantana. Las aleaciones estudiadas son las siguientes: W-0.5%Y2O3, W-2%Ti, W-2% Ti-0.5% Y2O3, W-4% Ti-0.5% Y2O3, W-2%V, W- 2%Vmix, W-4%V, W-1%La2O3 and W-4%V-1%La2O3. Todos ellos, además del wolframio puro se fabrican mediante compresión isostática en caliente (HIP) y son suministradas por la Universidad Carlos III de Madrid. La investigación se desarrolla a través de un estudio sistemático basado en ensayos físicos y mecánicos, así como el análisis post mortem de las muestras ensayadas. Para realizar dicha caracterización mecánica se aplican diferentes ensayos mecánicos, la mayoría de ellos realizados en el intervalo de temperatura de 25 a 1000 º C. Los ensayos de caracterización que se llevan a cabo son: • Densidad • Dureza Vicker • Módulo de elasticidad y su evolución con la temperatura • Límite elástico o resistencia a la flexión máxima, y su evolución con la temperatura • Resistencia a la fractura y su comportamiento con la temperatura. • Análisis microestructural • Análisis fractográfico • Análisis de la relación microestructura-comportamiento macroscópico. El estudio comienza con una introducción acerca de los sistemas en los que estos materiales son candidatos para su aplicación, para comprender las condiciones a las que los materiales serán expuestos. En este caso, el componente que determina las condiciones es el Divertor del reactor de energía de fusión por confinamiento magnético. Parece obvio que su uso en los componentes del reactor de fusión, más exactamente como materiales de cara al plasma (Plasma Facing Components o PFC), hace que estas aleaciones trabajen bajo condiciones de irradiación de neutrones. Además, el hecho de que sean materiales nuevos hace necesario un estudio previo de las características básicas que garantice los requisitos mínimos antes de realizar un estudio más complejo. Esto constituye la principal motivación de la presente investigación. La actual crisis energética ha llevado a aunar esfuerzos en el desarrollo de nuevos materiales, técnicas y dispositivos para la aplicación en la industria de la energía nuclear. El desarrollo de las técnicas de producción de aleaciones de wolframio, con un punto de fusión muy alto, requiere el uso de precursores de sinterizado para lograr densificaciones más altas y por lo tanto mejores propiedades mecánicas. Este es el propósito de la adición de titanio y vanadio en estas aleaciones. Sin embargo, uno de los principales problemas de la utilización de wolframio como material estructural es su alta temperatura de transición dúctil-frágil. Esta temperatura es característica de materiales metálicos con estructura cúbica centrada en el cuerpo y depende de varios factores metalúrgicos. El proceso de recristalización aumenta esta temperatura de transición. Los PFC tienen temperaturas muy altas de servicio, lo que facilita la recristalización del metal. Con el fin de retrasar este proceso, se dispersan partículas insolubles en el material permitiendo temperaturas de servicio más altas. Hasta ahora se ha utilizado óxidos de torio, lantano e itrio como partículas dispersas. Para entender cómo los contenidos en algunos elementos y partículas de óxido afectan a las propiedades de wolframio se estudian las aleaciones binarias de wolframio en comparación con el wolframio puro. A su vez estas aleaciones binarias se utilizan como material de referencia para entender el comportamiento de las aleaciones ternarias. Dada la estrecha relación entre las propiedades del material, la estructura y proceso de fabricación, el estudio se completa con un análisis fractográfico y micrográfico. El análisis fractográfico puede mostrar los mecanismos que están implicados en el proceso de fractura del material. Por otro lado, el estudio micrográfico ayudará a entender este comportamiento a través de la identificación de las posibles fases presentes. La medida del tamaño de grano es una parte de la caracterización microestructural. En esta investigación, la medida del tamaño de grano se llevó a cabo por ataque químico selectivo para revelar el límite de grano en las muestras preparadas. Posteriormente las micrografías fueron sometidas a tratamiento y análisis de imágenes. El documento termina con una discusión de los resultados y la compilación de las conclusiones más importantes que se alcanzan después del estudio. Actualmente, el desarrollo de nuevos materiales para aplicación en los componentes de cara al plasma continúa. El estudio de estos materiales ayudará a completar una base de datos de características que permita hacer una selección de ellos más fiable. The main goal of this dissertation is the mechanical characterization as a function of temperature of nine tungsten alloys containing different amounts of titanium, vanadium and yttrium and lanthanum oxide. The alloys under study were the following ones: W-0.5%Y2O3, W-2%Ti, W-2% Ti-0.5% Y2O3, W-4% Ti-0.5% Y2O3, W-2%V, W- 2%Vmix, W-4%V, W-1%La2O3 and W-4%V-1%La2O3. All of them, besides pure tungsten, were manufactured using a Hot Isostatic Pressing (HIP) process and they were supplied by the Universidad Carlos III de Madrid. The research was carried out through a systematic study based on physical and mechanical tests as well as the post mortem analysis of tested samples. Diverse mechanical tests were applied to perform this characterization; most of them were conducted at temperatures in the range 25-1000 ºC. The following characterization tests were performed: • Density • Vickers hardness • Elastic modulus • Yield strength or ultimate bending strength, and their evolution with temperature • Fracture toughness and its temperature behavior • Microstructural analysis • Fractographical analysis • Microstructure-macroscopic relationship analysis This study begins with an introduction regarding the systems where these materials could be applied, in order to establish and understand their service conditions. In this case, the component that defines the conditions is the Divertor of magnetic-confinement fusion reactors. It seems obvious that their use as fusion reactor components, more exactly as plasma facing components (PFCs), makes these alloys work under conditions of neutron irradiation. In addition to this, the fact that they are novel materials demands a preliminary study of the basic characteristics which will guarantee their minimum requirements prior to a more complex study. This constitutes the motivation of the present research. The current energy crisis has driven to join forces so as to develop new materials, techniques and devices for their application in the nuclear energy industry. The development of production techniques for tungsten-based alloys, with a very high melting point, requires the use of precursors for sintering to achieve higher densifications and, accordingly, better mechanical properties. This is the purpose of the addition of titanium and vanadium to these alloys. Nevertheless, one of the main problems of using tungsten as structural material is its high ductile-brittle transition temperature. This temperature is characteristic of metallic materials with body centered cubic structure and depends on several metallurgical factors. The recrystallization process increases their transition temperature. Since PFCs have a very high service temperature, this facilitates the metal recrystallization. In order to inhibit this process, insoluble particles are dispersed in the material allowing higher service temperatures. So far, oxides of thorium, lanthanum and yttrium have been used as dispersed particles. Tungsten binary alloys are studied in comparison with pure tungsten to understand how the contents of some elements and oxide particles affect tungsten properties. In turn, these binary alloys are used as reference materials to understand the behavior of ternary alloys. Given the close relationship between the material properties, structure and manufacturing process, this research is completed with a fractographical and micrographic analysis. The fractographical analysis is aimed to show the mechanisms that are involved in the process of the material fracture. Besides, the micrographic study will help to understand this behavior through the identification of present phases. The grain size measurement is a crucial part of the microstructural characterization. In this work, the measurement of grain size was carried out by chemical selective etching to reveal the boundary grain on prepared samples. Afterwards, micrographs were subjected to both treatment and image analysis. The dissertation ends with a discussion of results and the compilation of the most important conclusions reached through this work. The development of new materials for plasma facing components application is still under study. The analysis of these materials will help to complete a database of the features that will allow a more reliable materials selection.

Monte Carlo Neutronics and Thermal-hydraulics coupling applied to Fast Reactors

Un escenario habitualmente considerado para el uso sostenible y prolongado de la energía nuclear contempla un parque de reactores rápidos refrigerados por metales líquidos (LMFR) dedicados al reciclado de Pu y la transmutación de actínidos minoritarios (MA). Otra opción es combinar dichos reactores con algunos sistemas subcríticos asistidos por acelerador (ADS), exclusivamente destinados a la eliminación de MA. El diseño y licenciamiento de estos reactores innovadores requiere herramientas computacionales prácticas y precisas, que incorporen el conocimiento obtenido en la investigación experimental de nuevas configuraciones de reactores, materiales y sistemas. A pesar de que se han construido y operado un cierto número de reactores rápidos a nivel mundial, la experiencia operacional es todavía reducida y no todos los transitorios se han podido entender completamente. Por tanto, los análisis de seguridad de nuevos LMFR están basados fundamentalmente en métodos deterministas, al contrario que las aproximaciones modernas para reactores de agua ligera (LWR), que se benefician también de los métodos probabilistas. La aproximación más usada en los estudios de seguridad de LMFR es utilizar una variedad de códigos, desarrollados a base de distintas teorías, en busca de soluciones integrales para los transitorios e incluyendo incertidumbres. En este marco, los nuevos códigos para cálculos de mejor estimación ("best estimate") que no incluyen aproximaciones conservadoras, son de una importancia primordial para analizar estacionarios y transitorios en reactores rápidos. Esta tesis se centra en el desarrollo de un código acoplado para realizar análisis realistas en reactores rápidos críticos aplicando el método de Monte Carlo. Hoy en día, dado el mayor potencial de recursos computacionales, los códigos de transporte neutrónico por Monte Carlo se pueden usar de manera práctica para realizar cálculos detallados de núcleos completos, incluso de elevada heterogeneidad material. Además, los códigos de Monte Carlo se toman normalmente como referencia para los códigos deterministas de difusión en multigrupos en aplicaciones con reactores rápidos, porque usan secciones eficaces punto a punto, un modelo geométrico exacto y tienen en cuenta intrínsecamente la dependencia angular de flujo. En esta tesis se presenta una metodología de acoplamiento entre el conocido código MCNP, que calcula la generación de potencia en el reactor, y el código de termohidráulica de subcanal COBRA-IV, que obtiene las distribuciones de temperatura y densidad en el sistema. COBRA-IV es un código apropiado para aplicaciones en reactores rápidos ya que ha sido validado con resultados experimentales en haces de barras con sodio, incluyendo las correlaciones más apropiadas para metales líquidos. En una primera fase de la tesis, ambos códigos se han acoplado en estado estacionario utilizando un método iterativo con intercambio de archivos externos. El principal problema en el acoplamiento neutrónico y termohidráulico en estacionario con códigos de Monte Carlo es la manipulación de las secciones eficaces para tener en cuenta el ensanchamiento Doppler cuando la temperatura del combustible aumenta. Entre todas las opciones disponibles, en esta tesis se ha escogido la aproximación de pseudo materiales, y se ha comprobado que proporciona resultados aceptables en su aplicación con reactores rápidos. Por otro lado, los cambios geométricos originados por grandes gradientes de temperatura en el núcleo de reactores rápidos resultan importantes para la neutrónica como consecuencia del elevado recorrido libre medio del neutrón en estos sistemas. Por tanto, se ha desarrollado un módulo adicional que simula la geometría del reactor en caliente y permite estimar la reactividad debido a la expansión del núcleo en un transitorio. éste módulo calcula automáticamente la longitud del combustible, el radio de la vaina, la separación de los elementos de combustible y el radio de la placa soporte en función de la temperatura. éste efecto es muy relevante en transitorios sin inserción de bancos de parada. También relacionado con los cambios geométricos, se ha implementado una herramienta que, automatiza el movimiento de las barras de control en busca d la criticidad del reactor, o bien calcula el valor de inserción axial las barras de control. Una segunda fase en la plataforma de cálculo que se ha desarrollado es la simulació dinámica. Puesto que MCNP sólo realiza cálculos estacionarios para sistemas críticos o supercríticos, la solución más directa que se propone sin modificar el código fuente de MCNP es usar la aproximación de factorización de flujo, que resuelve por separado la forma del flujo y la amplitud. En este caso se han estudiado en profundidad dos aproximaciones: adiabática y quasiestática. El método adiabático usa un esquema de acoplamiento que alterna en el tiempo los cálculos neutrónicos y termohidráulicos. MCNP calcula el modo fundamental de la distribución de neutrones y la reactividad al final de cada paso de tiempo, y COBRA-IV calcula las propiedades térmicas en el punto intermedio de los pasos de tiempo. La evolución de la amplitud de flujo se calcula resolviendo las ecuaciones de cinética puntual. Este método calcula la reactividad estática en cada paso de tiempo que, en general, difiere de la reactividad dinámica que se obtendría con la distribución de flujo exacta y dependiente de tiempo. No obstante, para entornos no excesivamente alejados de la criticidad ambas reactividades son similares y el método conduce a resultados prácticos aceptables. Siguiendo esta línea, se ha desarrollado después un método mejorado para intentar tener en cuenta el efecto de la fuente de neutrones retardados en la evolución de la forma del flujo durante el transitorio. El esquema consiste en realizar un cálculo cuasiestacionario por cada paso de tiempo con MCNP. La simulación cuasiestacionaria se basa EN la aproximación de fuente constante de neutrones retardados, y consiste en dar un determinado peso o importancia a cada ciclo computacial del cálculo de criticidad con MCNP para la estimación del flujo final. Ambos métodos se han verificado tomando como referencia los resultados del código de difusión COBAYA3 frente a un ejercicio común y suficientemente significativo. Finalmente, con objeto de demostrar la posibilidad de uso práctico del código, se ha simulado un transitorio en el concepto de reactor crítico en fase de diseño MYRRHA/FASTEF, de 100 MW de potencia térmica y refrigerado por plomo-bismuto. ABSTRACT Long term sustainable nuclear energy scenarios envisage a fleet of Liquid Metal Fast Reactors (LMFR) for the Pu recycling and minor actinides (MAs) transmutation or combined with some accelerator driven systems (ADS) just for MAs elimination. Design and licensing of these innovative reactor concepts require accurate computational tools, implementing the knowledge obtained in experimental research for new reactor configurations, materials and associated systems. Although a number of fast reactor systems have already been built, the operational experience is still reduced, especially for lead reactors, and not all the transients are fully understood. The safety analysis approach for LMFR is therefore based only on deterministic methods, different from modern approach for Light Water Reactors (LWR) which also benefit from probabilistic methods. Usually, the approach adopted in LMFR safety assessments is to employ a variety of codes, somewhat different for the each other, to analyze transients looking for a comprehensive solution and including uncertainties. In this frame, new best estimate simulation codes are of prime importance in order to analyze fast reactors steady state and transients. This thesis is focused on the development of a coupled code system for best estimate analysis in fast critical reactor. Currently due to the increase in the computational resources, Monte Carlo methods for neutrons transport can be used for detailed full core calculations. Furthermore, Monte Carlo codes are usually taken as reference for deterministic diffusion multigroups codes in fast reactors applications because they employ point-wise cross sections in an exact geometry model and intrinsically account for directional dependence of the ux. The coupling methodology presented here uses MCNP to calculate the power deposition within the reactor. The subchannel code COBRA-IV calculates the temperature and density distribution within the reactor. COBRA-IV is suitable for fast reactors applications because it has been validated against experimental results in sodium rod bundles. The proper correlations for liquid metal applications have been added to the thermal-hydraulics program. Both codes are coupled at steady state using an iterative method and external files exchange. The main issue in the Monte Carlo/thermal-hydraulics steady state coupling is the cross section handling to take into account Doppler broadening when temperature rises. Among every available options, the pseudo materials approach has been chosen in this thesis. This approach obtains reasonable results in fast reactor applications. Furthermore, geometrical changes caused by large temperature gradients in the core, are of major importance in fast reactor due to the large neutron mean free path. An additional module has therefore been included in order to simulate the reactor geometry in hot state or to estimate the reactivity due to core expansion in a transient. The module automatically calculates the fuel length, cladding radius, fuel assembly pitch and diagrid radius with the temperature. This effect will be crucial in some unprotected transients. Also related to geometrical changes, an automatic control rod movement feature has been implemented in order to achieve a just critical reactor or to calculate control rod worth. A step forward in the coupling platform is the dynamic simulation. Since MCNP performs only steady state calculations for critical systems, the more straight forward option without modifying MCNP source code, is to use the flux factorization approach solving separately the flux shape and amplitude. In this thesis two options have been studied to tackle time dependent neutronic simulations using a Monte Carlo code: adiabatic and quasistatic methods. The adiabatic methods uses a staggered time coupling scheme for the time advance of neutronics and the thermal-hydraulics calculations. MCNP computes the fundamental mode of the neutron flux distribution and the reactivity at the end of each time step and COBRA-IV the thermal properties at half of the the time steps. To calculate the flux amplitude evolution a solver of the point kinetics equations is used. This method calculates the static reactivity in each time step that in general is different from the dynamic reactivity calculated with the exact flux distribution. Nevertheless, for close to critical situations, both reactivities are similar and the method leads to acceptable practical results. In this line, an improved method as an attempt to take into account the effect of delayed neutron source in the transient flux shape evolutions is developed. The scheme performs a quasistationary calculation per time step with MCNP. This quasistationary simulations is based con the constant delayed source approach, taking into account the importance of each criticality cycle in the final flux estimation. Both adiabatic and quasistatic methods have been verified against the diffusion code COBAYA3, using a theoretical kinetic exercise. Finally, a transient in a critical 100 MWth lead-bismuth-eutectic reactor concept is analyzed using the adiabatic method as an application example in a real system.

Validation of the burn-up code EVOLCODE 2.0 with PWR experimental data and with a Sensitivity/Uncertainty analysis

A validation of the burn-up simulation system EVOLCODE 2.0 is presented here, involving the experimental measurement of U and Pu isotopes and some fission fragments production ratios after a burn-up of around 30 GWd/tU in a Pressurized Light Water Reactor (PWR). This work provides an in-depth analysis of the validation results, including the possible sources of the uncertainties. An uncertainty analysis based on the sensitivity methodology has been also performed, providing the uncertainties in the isotopic content propagated from the cross sections uncertainties. An improvement of the classical Sensitivity/ Uncertainty (S/U) model has been developed to take into account the implicit dependence of the neutron flux normalization, that is, the effect of the constant power of the reactor. The improved S/U methodology, neglected in this kind of studies, has proven to be an important contribution to the explanation of some simulation-experiment discrepancies for which, in general, the cross section uncertainties are, for the most relevant actinides, an important contributor to the simulation uncertainties, of the same order of magnitude and sometimes even larger than the experimental uncertainties and the experiment- simulation differences. Additionally, some hints for the improvement of the JEFF3.1.1 fission yield library and for the correction of some errata in the experimental data are presented.

Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

The calculation of the effective delayed neutron fraction, beff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for beff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of beff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of beff .

Modelización e integración de sistemas de adquisición de datos inteligentes en sistemas de instrumentación para dispositivos de fusión

La obtención de energía a partir de la fusión nuclear por confinamiento magnético del plasma, es uno de los principales objetivos dentro de la comunidad científica dedicada a la energía nuclear. Desde la construcción del primer dispositivo de fusión, hasta la actualidad, se han llevado a cabo multitud de experimentos, que hoy en día, gran parte de ellos dan soporte al proyecto International Thermonuclear Experimental Reactor (ITER). El principal problema al que se enfrenta ITER, se basa en la monitorización y el control del plasma. Gracias a las nuevas tecnologías, los sistemas de instrumentación y control permiten acercarse más a la solución del problema, pero a su vez, es más complicado estandarizar los sistemas de adquisición de datos que se usan, no solo en ITER, sino en otros proyectos de igual complejidad. Desarrollar nuevas implementaciones hardware y software bajo los requisitos de los diagnósticos definidos por los científicos, supone una gran inversión de tiempo, retrasando la ejecución de nuevos experimentos. Por ello, la solución que plantea esta tesis, consiste en la definición de una metodología de diseño que permite implementar sistemas de adquisición de datos inteligentes y su fácil integración en entornos de fusión para la implementación de diagnósticos. Esta metodología requiere del uso de los dispositivos Reconfigurable Input/Output (RIO) y Flexible RIO (FlexRIO), que son sistemas embebidos basados en tecnología Field-Programmable Gate Array (FPGA). Para completar la metodología de diseño, estos dispositivos van a ser soportados por un software basado en EPICS Device Support utilizando la tecnología EPICS software asynDriver. Esta metodología se ha evaluado implementando prototipos para los controladores rápidos de planta de ITER, tanto para casos prácticos de ámbito general como adquisición de datos e imágenes, como para casos concretos como el diagnóstico del fission chamber, implementando pre-procesado en tiempo real. Además de casos prácticos, esta metodología se ha utilizado para implementar casos reales, como el Ion Source Hydrogen Positive (ISHP), desarrollada por el European Spallation Source (ESS Bilbao) y la Universidad del País Vasco. Finalmente, atendiendo a las necesidades que los experimentos en los entornos de fusión requieren, se ha diseñado un mecanismo mediante el cual los sistemas de adquisición de datos, que pueden ser implementados mediante la metodología de diseño propuesta, pueden integrar un reloj hardware capaz de sincronizarse con el protocolo IEEE1588-V2, permitiendo a estos, obtener los TimeStamps de las muestras adquiridas con una exactitud y precisión de decenas de nanosegundos y realizar streaming de datos con TimeStamps. ABSTRACT Fusion energy reaching by means of nuclear fusion plasma confinement is one of the main goals inside nuclear energy scientific community. Since the first fusion device was built, many experiments have been carried out and now, most of them give support to the International Thermonuclear Experimental Reactor (ITER) project. The main difficulty that ITER has to overcome is the plasma monitoring and control. Due to new technologies, the instrumentation and control systems allow an approaching to the solution, but in turn, the standardization of the used data acquisition systems, not only in ITER but also in other similar projects, is more complex. To develop new hardware and software implementations under scientific diagnostics requirements, entail time costs, delaying new experiments execution. Thus, this thesis presents a solution that consists in a design methodology definition, that permits the implementation of intelligent data acquisition systems and their easy integration into fusion environments for diagnostic purposes. This methodology requires the use of Reconfigurable Input/Output (RIO) and Flexible RIO (FlexRIO) devices, based on Field-Programmable Gate Array (FPGA) embedded technology. In order to complete the design methodology, these devices are going to be supported by an EPICS Device Support software, using asynDriver technology. This methodology has been evaluated implementing ITER PXIe fast controllers prototypes, as well as data and image acquisition, so as for concrete solutions like the fission chamber diagnostic use case, using real time preprocessing. Besides of these prototypes solutions, this methodology has been applied for the implementation of real experiments like the Ion Source Hydrogen Positive (ISHP), developed by the European Spallation Source and the Basque country University. Finally, a hardware mechanism has been designed to integrate a hardware clock into RIO/FlexRIO devices, to get synchronization with the IEEE1588-V2 precision time protocol. This implementation permits to data acquisition systems implemented under the defined methodology, to timestamp all data acquired with nanoseconds accuracy, permitting high throughput timestamped data streaming.

Thermal Scattering Libraries Processing (F4E, Task 6.2)

Presentación del trabajo realizado en el marco del proyecto F4E, sobre el procesamiento de librerías de dispersión térmica de neutrones en formato ACE para su uso con el código MCNP. Se presentan tanto los métodos y procedimientos empleados, como los resultados y diferencias entre las distintas fuentes de datos.