A theoretical study on the formation of iodine oxide aggregates and monohydrates

Biotic and abiotic emissions of molecular iodine and iodocarbons from the sea or the ice surface and the intertidal zone to the coastal/polar marine boundary layer lead to the formation of iodine oxides, which subsequently nucleate forming iodine oxide particles (IOPs). Although the link between coastal iodine emissions and ultrafine aerosol bursts is well established, the details of the nucleation mechanism have not yet been elucidated. In this paper, results of a theoretical study of a range of potentially relevant aggregation reactions of different iodine oxides, as well as complexation with water molecules, are reported. Thermochemical properties of these reactions are obtained from high level ab initio correlated calculations including spin–orbit corrections. The results show that the nucleation path most likely proceeds through dimerisation of I2O4. It is also shown that water can hinder gas-to-particle conversion to some extent, although complexation with key iodine oxides does not remove enough of these to stop IOP formation. A consistent picture of this process emerges from the theoretical study presented here and the findings of a new laboratory study reported in the accompanying paper (Gomez Martin et al., 2013).

Theoretical investigation on iodine oxides formation in the production of atmospheric aerosols

In this contribution, results of a theoretical study on different reactions that odine oxides, in the presence of water, can undergo to form iodine oxides particles in the atmosphere. Thermodynamic and kinetic properties of these reactions have been obtained at high level ab initio correlated calculations.

Highly correlated ab initio study of the far infrared spectra of methyl acetate

Highly correlated ab initio calculations (CCSD(T)) are used to compute gas phase spectroscopic parameters of three isotopologues of the methyl acetate (CH3COOCH3, CD3COOCH3, and CH3COOCD3), searching to help experimental assignments and astrophysical detections. The molecule shows two conformers cis and trans separated by a barrier of 4457 cm−1. The potential energy surface presents 18 minima that intertransform through three internal rotation motions. To analyze the far infrared spectrum at low temperatures, a three-dimensional Hamiltonian is solved variationally. The two methyl torsion barriers are calculated to be 99.2 cm−1 (C–CH3) and 413.1 cm−1 (O–CH3), for the cis-conformer. The three fundamental torsional band centers of CH3COOCH3 are predicted to lie at 63.7 cm−1 (C–CH3), 136.1 cm−1 (O–CH3), and 175.8 cm−1 (C–O torsion) providing torsional state separations. For the 27 vibrational modes, anharmonic fundamentals and rovibrational parameters are provided. Computed parameters are compared with those fitted using experimental data.

Development of a methodology for nuclear data uncertainty propagation on isotopic evolution calculations for advanced nuclear systems

Una apropiada evaluación de los márgenes de seguridad de una instalación nuclear, por ejemplo, una central nuclear, tiene en cuenta todas las incertidumbres que afectan a los cálculos de diseño, funcionanmiento y respuesta ante accidentes de dicha instalación. Una fuente de incertidumbre son los datos nucleares, que afectan a los cálculos neutrónicos, de quemado de combustible o activación de materiales. Estos cálculos permiten la evaluación de las funciones respuesta esenciales para el funcionamiento correcto durante operación, y también durante accidente. Ejemplos de esas respuestas son el factor de multiplicación neutrónica o el calor residual después del disparo del reactor. Por tanto, es necesario evaluar el impacto de dichas incertidumbres en estos cálculos. Para poder realizar los cálculos de propagación de incertidumbres, es necesario implementar metodologías que sean capaces de evaluar el impacto de las incertidumbres de estos datos nucleares. Pero también es necesario conocer los datos de incertidumbres disponibles para ser capaces de manejarlos. Actualmente, se están invirtiendo grandes esfuerzos en mejorar la capacidad de analizar, manejar y producir datos de incertidumbres, en especial para isótopos importantes en reactores avanzados. A su vez, nuevos programas/códigos están siendo desarrollados e implementados para poder usar dichos datos y analizar su impacto. Todos estos puntos son parte de los objetivos del proyecto europeo ANDES, el cual ha dado el marco de trabajo para el desarrollo de esta tesis doctoral. Por tanto, primero se ha llevado a cabo una revisión del estado del arte de los datos nucleares y sus incertidumbres, centrándose en los tres tipos de datos: de decaimiento, de rendimientos de fisión y de secciones eficaces. A su vez, se ha realizado una revisión del estado del arte de las metodologías para la propagación de incertidumbre de estos datos nucleares. Dentro del Departamento de Ingeniería Nuclear (DIN) se propuso una metodología para la propagación de incertidumbres en cálculos de evolución isotópica, el Método Híbrido. Esta metodología se ha tomado como punto de partida para esta tesis, implementando y desarrollando dicha metodología, así como extendiendo sus capacidades. Se han analizado sus ventajas, inconvenientes y limitaciones. El Método Híbrido se utiliza en conjunto con el código de evolución isotópica ACAB, y se basa en el muestreo por Monte Carlo de los datos nucleares con incertidumbre. En esta metodología, se presentan diferentes aproximaciones según la estructura de grupos de energía de las secciones eficaces: en un grupo, en un grupo con muestreo correlacionado y en multigrupos. Se han desarrollado diferentes secuencias para usar distintas librerías de datos nucleares almacenadas en diferentes formatos: ENDF-6 (para las librerías evaluadas), COVERX (para las librerías en multigrupos de SCALE) y EAF (para las librerías de activación). Gracias a la revisión del estado del arte de los datos nucleares de los rendimientos de fisión se ha identificado la falta de una información sobre sus incertidumbres, en concreto, de matrices de covarianza completas. Además, visto el renovado interés por parte de la comunidad internacional, a través del grupo de trabajo internacional de cooperación para evaluación de datos nucleares (WPEC) dedicado a la evaluación de las necesidades de mejora de datos nucleares mediante el subgrupo 37 (SG37), se ha llevado a cabo una revisión de las metodologías para generar datos de covarianza. Se ha seleccionando la actualización Bayesiana/GLS para su implementación, y de esta forma, dar una respuesta a dicha falta de matrices completas para rendimientos de fisión. Una vez que el Método Híbrido ha sido implementado, desarrollado y extendido, junto con la capacidad de generar matrices de covarianza completas para los rendimientos de fisión, se han estudiado diferentes aplicaciones nucleares. Primero, se estudia el calor residual tras un pulso de fisión, debido a su importancia para cualquier evento después de la parada/disparo del reactor. Además, se trata de un ejercicio claro para ver la importancia de las incertidumbres de datos de decaimiento y de rendimientos de fisión junto con las nuevas matrices completas de covarianza. Se han estudiado dos ciclos de combustible de reactores avanzados: el de la instalación europea para transmutación industrial (EFIT) y el del reactor rápido de sodio europeo (ESFR), en los cuales se han analizado el impacto de las incertidumbres de los datos nucleares en la composición isotópica, calor residual y radiotoxicidad. Se han utilizado diferentes librerías de datos nucleares en los estudios antreriores, comparando de esta forma el impacto de sus incertidumbres. A su vez, mediante dichos estudios, se han comparando las distintas aproximaciones del Método Híbrido y otras metodologías para la porpagación de incertidumbres de datos nucleares: Total Monte Carlo (TMC), desarrollada en NRG por A.J. Koning y D. Rochman, y NUDUNA, desarrollada en AREVA GmbH por O. Buss y A. Hoefer. Estas comparaciones demostrarán las ventajas del Método Híbrido, además de revelar sus limitaciones y su rango de aplicación. ABSTRACT For an adequate assessment of safety margins of nuclear facilities, e.g. nuclear power plants, it is necessary to consider all possible uncertainties that affect their design, performance and possible accidents. Nuclear data are a source of uncertainty that are involved in neutronics, fuel depletion and activation calculations. These calculations can predict critical response functions during operation and in the event of accident, such as decay heat and neutron multiplication factor. Thus, the impact of nuclear data uncertainties on these response functions needs to be addressed for a proper evaluation of the safety margins. Methodologies for performing uncertainty propagation calculations need to be implemented in order to analyse the impact of nuclear data uncertainties. Nevertheless, it is necessary to understand the current status of nuclear data and their uncertainties, in order to be able to handle this type of data. Great eórts are underway to enhance the European capability to analyse/process/produce covariance data, especially for isotopes which are of importance for advanced reactors. At the same time, new methodologies/codes are being developed and implemented for using and evaluating the impact of uncertainty data. These were the objectives of the European ANDES (Accurate Nuclear Data for nuclear Energy Sustainability) project, which provided a framework for the development of this PhD Thesis. Accordingly, first a review of the state-of-the-art of nuclear data and their uncertainties is conducted, focusing on the three kinds of data: decay, fission yields and cross sections. A review of the current methodologies for propagating nuclear data uncertainties is also performed. The Nuclear Engineering Department of UPM has proposed a methodology for propagating uncertainties in depletion calculations, the Hybrid Method, which has been taken as the starting point of this thesis. This methodology has been implemented, developed and extended, and its advantages, drawbacks and limitations have been analysed. It is used in conjunction with the ACAB depletion code, and is based on Monte Carlo sampling of variables with uncertainties. Different approaches are presented depending on cross section energy-structure: one-group, one-group with correlated sampling and multi-group. Differences and applicability criteria are presented. Sequences have been developed for using different nuclear data libraries in different storing-formats: ENDF-6 (for evaluated libraries) and COVERX (for multi-group libraries of SCALE), as well as EAF format (for activation libraries). A revision of the state-of-the-art of fission yield data shows inconsistencies in uncertainty data, specifically with regard to complete covariance matrices. Furthermore, the international community has expressed a renewed interest in the issue through the Working Party on International Nuclear Data Evaluation Co-operation (WPEC) with the Subgroup (SG37), which is dedicated to assessing the need to have complete nuclear data. This gives rise to this review of the state-of-the-art of methodologies for generating covariance data for fission yields. Bayesian/generalised least square (GLS) updating sequence has been selected and implemented to answer to this need. Once the Hybrid Method has been implemented, developed and extended, along with fission yield covariance generation capability, different applications are studied. The Fission Pulse Decay Heat problem is tackled first because of its importance during events after shutdown and because it is a clean exercise for showing the impact and importance of decay and fission yield data uncertainties in conjunction with the new covariance data. Two fuel cycles of advanced reactors are studied: the European Facility for Industrial Transmutation (EFIT) and the European Sodium Fast Reactor (ESFR), and response function uncertainties such as isotopic composition, decay heat and radiotoxicity are addressed. Different nuclear data libraries are used and compared. These applications serve as frameworks for comparing the different approaches of the Hybrid Method, and also for comparing with other methodologies: Total Monte Carlo (TMC), developed at NRG by A.J. Koning and D. Rochman, and NUDUNA, developed at AREVA GmbH by O. Buss and A. Hoefer. These comparisons reveal the advantages, limitations and the range of application of the Hybrid Method.

Object Kinetic Monte Carlo calculations of irradiated Fe-Cr dilute alloys: The effect of the interaction radius between substitutional Cr and self-interstitial Fe

ObjectKineticMonteCarlo models allow for the study of the evolution of the damage created by irradiation to time scales that are comparable to those achieved experimentally. Therefore, the essential ObjectKineticMonteCarlo parameters can be validated through comparison with experiments. However, this validation is not trivial since a large number of parameters is necessary, including migration energies of point defects and their clusters, binding energies of point defects in clusters, as well as the interactionradii. This is particularly cumbersome when describing an alloy, such as the Fe–Cr system, which is of interest for fusion energy applications. In this work we describe an ObjectKineticMonteCarlo model for Fe–Cr alloys in the dilute limit. The parameters used in the model come either from density functional theory calculations or from empirical interatomic potentials. This model is used to reproduce isochronal resistivity recovery experiments of electron irradiateddiluteFe–Cr alloys performed by Abe and Kuramoto. The comparison between the calculated results and the experiments reveal that an important parameter is the capture radius between substitutionalCr and self-interstitialFe atoms. A parametric study is presented on the effect of the capture radius on the simulated recovery curves.

Propagation of Nuclear Data Uncertainties in Transmutation Calculations Using ACAB Code

The assessment of the accuracy of parameters related to the reactor core performance (e.g., ke) and f el cycle (e.g., isotopic evolution/transmutation) due to the uncertainties in the basic nuclear data (ND) is a critical issue. Different error propagation techniques (adjoint/forward sensitivity analysis procedures and/or Monte Carlo technique) can be used to address by computational simulation the systematic propagation of uncertainties on the final parameters. To perform this uncertainty assessment, the ENDF covariance les (variance/correlation in energy and cross- reactions-isotopes correlations) are required. In this paper, we assess the impact of ND uncertainties on the isotopic prediction for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) for a discharge burnup of 150 GWd/tHM. The complete set of uncertainty data for cross sections (EAF2007/UN, SCALE6.0/COVA-44G), radioactive decay and fission yield data (JEFF-3.1.1) are processed and used in ACAB code.

Importance of Nuclear Data Uncertainties in Criticality Calculations

To study the propagation of the uncertainty from basic data across different scale and physics phenomena -> through complex coupled multi-physics and multi-scale simulations

Propagation of nuclear data uncertainties in fuel cycle calculations using MONTE-CARLO Technique

The uncertainty propagation in fuel cycle calculations due to Nuclear Data (ND) is a important important issue for : issue for : • Present fuel cycles (e.g. high burnup fuel programme) • New fuel cycles designs (e.g. fast breeder reactors and ADS) Different error propagation techniques can be used: • Sensitivity analysis • Response Response Surface Method Surface Method • Monte Carlo technique Then, p p , , in this paper, it is assessed the imp y pact of ND uncertainties on the decay heat and radiotoxicity in two applications: • Fission Pulse Decay ( y Heat calculation (FPDH) • Conceptual design of European Facility for Industrial Transmutation (EFIT)

Uncertainty methods inactivation and inventory calculations

For a number of important nuclides, complete activation data libraries with covariance data will be produced, so that uncertainty propagation in fuel cycle codes (in this case ACAB,FISPIN, ...) can be developed and tested. Eventually, fuel inventory codes should be able to handle the complete set of uncertainty data, i.e. those of nuclear reactions (cross sections, etc.), radioactive decay and fission yield data. For this, capabilities will be developed both to produce covariance data and to propagate the uncertainties through the inventory calculations.

Midday measurements of leaf water potential and stomatal conductance are highly correlated with daily water use of Thompson Seedless grapevines

A study was conducted to determine the relationship between midday measurements of vine water status and daily water use of grapevines measured with a weighing lysimeter. Water applications to the vines were terminated on August 24th for 9 days and again on September 14th for 22 days. Daily water use of the vines in the lysimeter (ETLYS) was approximately 40 L vine−1 (5.3 mm) prior to turning the pump off, and it decreased to 22.3 L vine−1 by September 2nd. Pre-dawn leaf water potential (ΨPD) and midday Ψl on August 24th were −0.075 and −0.76 MPa, respectively, with midday Ψl decreasing to −1.28 MPa on September 2nd. Leaf g s decreased from ~500 to ~200 mmol m−2 s−1 during the two dry-down periods. Midday measurements of g s and Ψl were significantly correlated with one another (r = 0.96) and both with ETLYS/ETo (r = ~0.9). The decreases in Ψl, g s, and ETLYS/ETo in this study were also a linear function of the decrease in volumetric soil water content. The results indicate that even modest water stress can greatly reduce grapevine water use and that short-term measures of vine water status taken at midday are a reflection of daily grapevine water use

Analysis of Hail and Temperature Series for Peninsular Spain

Hail is a serious concern for agriculture on the Iberian Peninsula. Hailstorms affect crop yield and/or quality to a degree that depends on the crop species and the phenological time. In Europe, Spain is one of the countries that experience relatively high agricultural losses related to hailstorms. It is of high interest to study models that can support calculations of the probabilities of economic losses due to hail damage and of the tendency over time for such losses. Some studies developed in France and the Netherdlands show that the summer mean temperature was highly correlated with a yearly hail severity index developed from hailrelated parameters obtained for insurance purposes. Meanwhile, other studies in the USA point out that a highly significant correlation between both is not possible to find due to high climatic variability. The aim of this work is to test the correlation between average minimum temperatures and hail damage intensity over the Spanish Iberian Peninsula. With this purpose, correlation analyses on both variables were performed for the 47 Spanish provinces (as individuals and single set) and for all crops and four individual crops: grapes, wheat, barley and winter grains. Suitable crop insurance data are available from 1981 until 2007 and based on this period, temperature data were obtained. This study does not confirm the results previously obtained for France and the Netherlands that relate observed hail damage to the average minimum temperature. The reason for this difference and the nature of the cases observed are discussed.

Frequency-dependent opacity calculations for radiation transport simulations

We will present recent developments in the calculation of opacity tables suitable for including in the radiation hydrodynamic code ARWEN [1] to study processes like ICF or X-ray secondary sources. For these calculations we use the code BiG BART in LTE conditions, with self-consistent data generated with the Flexible Atomic Code (FAC) [2]. Non-LTE effects are approximately taken into account by means of the improved RADIOM model [3], which makes use of existing LTE data tables.

Estudio de la tolerancia al daño de los materiales y sistemas de pretensado

La mayoría de las estructuras de hormigón pretensadas construidas en los últimos 50 años han demostrado una excelente durabilidad cuando su construcción se realiza atendiendo las recomendaciones de un buen diseño así como una buena ejecución y puesta en obra de la estructura. Este hecho se debe en gran parte al temor que despierta el fenómeno de la corrosión bajo tensión típico de las armaduras de acero de alta resistencia. Menos atención se ha prestado a la susceptibilidad a la corrosión bajo tensión de los anclajes de postensado, posiblemente debido a que se han reportado pocos casos de fallos catastróficos. El concepto de Tolerancia al Daño y la Mecánica de la Fractura en estructuras de Ingeniería Civil ha empezado a incorporarse recientemente en algunas normas de diseño y cálculo de estructuras metálicas, sin embargo, aún está lejos de ser asimilado y empleado habitualmente por los ingenieros en sus cálculos cuando la ocasión lo requiere. Este desconocimiento de los aspectos relacionados con la Tolerancia al Daño genera importantes gastos de mantenimiento y reparación. En este trabajo se ha estudiado la aplicabilidad de los conceptos de la Mecánica de la Fractura a los componentes de los sistemas de postensado empleados en ingeniería civil, empleándolo para analizar la susceptibilidad de las armaduras activas frente a la corrosión bajo tensiones y a la pérdida de capacidad portante de las cabezas de anclajes de postensado debido a la presencia de defectos. Con este objeto se han combinado tanto técnicas experimentales como numéricas. Los defectos superficiales en los alambres de pretensado no se presentan de manera aislada si no que existe una cierta continuidad en la dirección axial así como un elevado número de defectos. Por este motivo se ha optado por un enfoque estadístico, que es más apropiado que el determinístico. El empleo de modelos estadísticos basados en la teoría de valores extremos ha permitido caracterizar el estado superficial en alambres de 5,2 mm de diámetro. Por otro lado la susceptibilidad del alambre frente a la corrosión bajo tensión ha sido evaluada mediante la realización de una campaña de ensayos de acuerdo con la actual normativa que ha permitido caracterizar estadísticamente su comportamiento. A la vista de los resultados ha sido posible evaluar como los parámetros que definen el estado superficial del alambre pueden determinar la durabilidad de la armadura atendiendo a su resistencia frente a la corrosión bajo tensión, evaluada mediante los ensayos que especifica la normativa. En el caso de las cabezas de anclaje de tendones de pretensado, los defectos se presentan de manera aislada y tienen su origen en marcas, arañazos o picaduras de corrosión que pueden producirse durante el proceso de fabricación, transporte, manipulación o puesta en obra. Dada la naturaleza de los defectos, el enfoque determinístico es más apropiado que el estadístico. La evaluación de la importancia de un defecto en un elemento estructural requiere la estimación de la solicitación local que genera el defecto, que permite conocer si el defecto es crítico o si puede llegar a serlo, si es que progresa con el tiempo (por fatiga, corrosión, una combinación de ambas, etc.). En este trabajo los defectos han sido idealizados como grietas, de manera que el análisis quedara del lado de la seguridad. La evaluación de la solicitación local del defecto ha sido calculada mediante el empleo de modelos de elementos finitos de la cabeza de anclaje que simulan las condiciones de trabajo reales de la cabeza de anclaje durante su vida útil. A partir de estos modelos numéricos se ha analizado la influencia en la carga de rotura del anclaje de diversos factores como la geometría del anclaje, las condiciones del apoyo, el material del anclaje, el tamaño del defecto su forma y su posición. Los resultados del análisis numérico han sido contrastados satisfactoriamente mediante la realización de una campaña experimental de modelos a escala de cabezas de anclaje de Polimetil-metacrilato en los que artificialmente se han introducido defectos de diversos tamaños y en distintas posiciones. ABSTRACT Most of the prestressed concrete structures built in the last 50 years have demonstrated an excellent durability when they are constructed in accordance with the rules of good design, detailing and execution. This is particularly true with respect to the feared stress corrosion cracking, which is typical of high strength prestressing steel wires. Less attention, however, has been paid to the stress corrosion cracking susceptibility of anchorages for steel tendons for prestressing concrete, probably due to the low number of reported failure cases. Damage tolerance and fracture mechanics concepts in civil engineering structures have recently started to be incorporated in some design and calculation rules for metallic structures, however it is still far from being assimilated and used by civil engineers in their calculations on a regular basis. This limited knowledge of the damage tolerance basis could lead to significant repair and maintenance costs. This work deals with the applicability of fracture mechanics and damage tolerance concepts to the components of prestressed systems, which are used in civil engineering. Such concepts have been applied to assess the susceptibility of the prestressing steel wires to stress corrosion cracking and the reduction of load bearing capability of anchorage devices due to the presence of defects. For this purpose a combination of experimental work and numerical techniques have been performed. Surface defects in prestressing steel wires are not shown alone, though a certain degree of continuity in the axial direction exist. A significant number of such defects is also observed. Hence a statistical approach was used, which is assumed to be more appropriate than the deterministic approach. The use of statistical methods based in extreme value theories has allowed the characterising of the surface condition of 5.2 mm-diameter wires. On the other hand the stress corrosion cracking susceptibility of the wire has been assessed by means of an experimental testing program in line with the current regulations, which has allowed statistical characterisasion of their performances against stress corrosion cracking. In the light of the test results, it has been possible to evaluate how the surface condition parameters could determine the durability of the active metal armour regarding to its resistance against stress corrosion cracking assessed by means of the current testing regulations. In the case of anchorage devices for steel tendons for prestressing concrete, the damage is presented as point defects originating from dents, scratches or corrosion pits that could be produced during the manufacturing proccess, transport, handling, assembly or use. Due to the nature of these defects, in this case the deterministic approach is more appropriate than the statistical approach. The assessment of the relevancy of defect in a structural component requires the computation of the stress intensity factors, which in turn allow the evaluation of whether the size defect is critical or could become critical with the progress of time (due to fatigue, corrosion or a combination of both effects). In this work the damage is idealised as tiny cracks, a conservative hypothesis. The stress intensity factors have been calculated by means of finite element models of the anchorage representing the real working conditions during its service life. These numeric models were used to assess the impact of some factors on the rupture load of the anchorage, such the anchorage geometry, material, support conditions, defect size, shape and its location. The results from the numerical analysis have been succesfully correlated against the results of the experimental testing program of scaled models of the anchorages in poly-methil methacrylate in which artificial damage in several sizes and locations were introduced.

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.

Importance of nuclear data uncertainties in criticality calculations

The aim of this paper is to study the importance of nuclear data uncertainties in the prediction of the uncertainties in keff for LWR (Light Water Reactor) unit-cells. The first part of this work is focused on the comparison of different sensitivity/uncertainty propagation methodologies based on TSUNAMI and MCNP codes; this study is undertaken for a fresh-fuel at different operational conditions. The second part of this work studies the burnup effect where the indirect contribution due to the uncertainty of the isotopic evolution is also analyzed.