Made available in DSpace on 2014-10-09T12:44:01Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:25Z (GMT). No. of bitstreams: 1 06864.pdf: 11106654 bytes, checksum: 851c7803db872d59fc1f49dc465fa8af (MD5) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2014-11-10T11:11:38Z No. of bitstreams: 0 / Made available in DSpace on 2014-11-10T11:11:38Z (GMT). No. of bitstreams: 0 / O objetivo deste trabalho é verificar a validade e a acurácia dos resultados fornecidos pelos programas computacionais FRAPCON-3.4a e FRAPTRAN-1.4, utilizados no processo de simulação do comportamento de varetas combustíveis de reatores a água leve pressurizada PWR (Pressurized Water Reactor), sob situações operacionais de regimes permanente e transiente, em condições de alta queima (high burnup). Para realizar a verificação, foi utilizada a base de dados FUMEX-III, que fornece dados relativos a experimentos realizados com diversos tipos de combustíveis nucleares, submetidos a diversas condições operacionais. Através dos resultados obtidos nas simulações computacionais com os programas FRAPCON-3.4a e FRAPTRAN-1.4 e da sua comparação com os dados experimentais da base FUMEX-III, foi possível constatar que os programas empregados possuem um boa capacidade de predizer o comportamento operacional de varetas combustíveis de PWR em regime permanente a altas queimas e sob condição de transiente inicializado por reatividade (Reactivity Initiated Accident RIA). / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

L’objectif de cette thèse est d’étudier et de reproduire les caractéristiques spécifiques de la microstructure du combustible nucléaire irradié à fort taux de combustion, appelée structure HBS (High Burnup Structure). Il s’agit d’étudier les différents paramètres pertinents impliqués dans la formation d’une telle structure, en évaluant leur importance, et en clarifiant leurs éventuelles synergies. Cet objectif a été réalisé en utilisant un système de modèle ultra simplifié, à savoir des monocristaux de dioxyde d’uranium (UO₂) irradiés par des ions de basse énergie (quelques centaines de keV) de Lanthane (La) ou de xénon (Xe) à une température de 773 K, correspondant à celle de la périphérie des véritables pastilles de combustible en réacteur. Les énergies et les masses des ions ont été choisies pour étudier la déstabilisation du solide en fonction de deux paramètres-clefs: (i) les collisions nucléaires élastiques et (ii) la contribution chimique de l'incorporation d'impuretés à forte concentration. Les deux espèces ont été choisies délibérément pour leurs solubilités très différentes dans le dioxyde d’uranium: les ions La sont solubles dans l'UO₂ jusqu’à de très fortes concentrations, tandis que les ions Xe sont insolubles. Les techniques de la Microscopie Électronique en Transmission (TEM) et de Spectrométrie de Rétrodiffusion Rutherford en canalisation RBS/C ont été conduites in situ couplée avec l’irradiation. Ces deux techniques utilisées pour visualiser, quantifier et fournir des informations concernant la fraction des défauts induits par l’irradiation et la formation des bulles, de cavités ou de précipités dans le solide. Les données de canalisation ont été analysées par simulation Monte-Carlo en supposant l’existence de deux catégories de défauts : (i) des atomes aléatoirement déplacés (RDA) et (ii) des distorsions des rangés atomiques (BC). L’évolution de la fraction de défaut de type RDA montre une forte augmentation entre 0.4 à 4.0 dpa (correspondant à une très faible concentration des ions implantés), indépendamment de la nature des ions. Elle est suivie par une saturation de la fraction de RDA pour les deux ions sur une large gamme d’irradiation quoi s’étend jusque 100 dpa. Une forte élévation de la fraction de RDA est observée en particulier pour les cristaux implantés avec des ions Xe pour une concentration élevée dépassant les 4%. En ce qui concerne l’évolution de BC, elle augmente fortement jusqu’à 4 dpa et sature ensuite deux ions La et Xe. Les résultats de microscopie électronique in situ montrent que des défauts identiques pour les deux ions induits par l’irradiation apparaissent, et présentent la même évolution en fonction de la fluence. Les différents défauts évoluent en fonction de la fluence : la première étape correspond à la formation de ‘black dots’ ; la deuxième étape est caractérisée par la formation de boucles puis de lignes de dislocations, qui évoluent finalement jusqu’à commencer à devenir moins différenciables; le processus de restructuration se poursuit et forme un réseau de dislocations enchevêtrées. Une forte densité de bulles de gaz, de taille nanométrique et avec un diamètre moyen de 2 nm est observée pour le cristal Xe implanté à une dose seuil de 4 dpa. Le couplage des deux techniques conduites in situ montre que la différence entre les valeurs à saturation des fractions RDA des deux ions, d’une part, et l'augmentation drastique de RDA à très forte concentration d'ions Xe implantés d’autre part peuvent être attribuées à : (i) la solubilité des ions La vis-à-vis des ions Xe, conduisant à la formation des bulles de gaz de taille nanométrique et (ii) la taille des espèces implantées dans la matrice UO₂, pour laquelle les atomes Xe insolubles ont un rayon atomique beaucoup plus grand que le rayon cationique des atomes U⁴⁺(les atomes La³⁺ ont un rayon atomique similaire à celui des atomes U⁴⁺), responsable de plus de contraintes supplémentaires dans la structure cristalline. / The aim of this thesis is to investigate and reproduce the specific features of the microstructure of the high burnup structure of the irradiated nuclear fuel and to explore the various relevant parameters involved in the formation of such a structure, in evaluating their importance, and in clarifying the synergies between them. This have been performed by using a very simplified model system – namely uranium dioxide single crystals- irradiated with low energy La and Xe ions at 773 K, corresponding to the temperature at the periphery of the genuine fuel. The energies and masses of bombarding ions were chosen to investigate the destabilization of the solid due to: (i) the elastic nuclear collisions and by (ii) the chemical contribution of implanting impurities at high concentrations by implanting different ions in UO₂, namely Xe and La, having very different solubility: La species are soluble in UO₂ while Xe ions are insoluble. In situ Transmission electron Microscopy (TEM) and in situ Rutherford Backscattering Spectrometry in the channeling mode (RBS/C), both techniques coupled to ion irradiation, were performed to visualize, quantify and provide information with respect to the fraction of radiation-induced defects and the formation of bubbles, cavities, or precipitates. The channeling data were analyzed afterwards by Monte Carlo simulations assuming two class of defects comprising (i) randomly displaced atoms (RDA) and (ii) bent channels (BC) defects. Regarding the RDA evolution, a sharp increase step appears from 0.4 to 4.0 dpa (corresponding to a low concentration of implanted ions) regardless of nature of ions followed by a saturation of the fraction of RDA for both ions over a wide range of irradiation. A sharp increase of RDA fraction is observed specifically for crystals implanted with Xe ions at a high concentration exceeding 1.5 % (corresponding to the dose of more than 125 dpa). Regarding the BC evolution, for both ions, the evolution shows an increase in the fraction of BC up to 4.0 dpa then the fraction of BC almost saturates for Xe and La ions. In situ TEM results show that a similar radiation-induced defects appear for both ions and the same evolution of defects as a function of the fluence is observed. The various defects evolved as a function of the fluence: starting from the black dot defects formation that were observed as a first type of defects created, then dislocation lines and loops appeared and evolved until they started to be become less distinguishable, the restructuring process continued by forming a tangled dislocation network. A high density of nanometer-sized gas bubbles with a mean diameter 2 nm were observed at room temperature for the Xe-implanted crystal at a threshold dose of 4 dpa. The coupling between both techniques (in situ RBS/C and TEM) demonstrates that the difference between the two plateaus of saturation between the two ions and the dramatic increase of RDA at high concentration of implanted Xe ions can be ascribed to: (i) the solubility of La compared to Xe ions leading to the formation of nanometer-sized gas bubbles and (ii) the size of implanted species in UO₂ matrix where insoluble Xe atoms have an atomic radius much larger than the cationic radius of U⁴⁺ atoms, (La³⁺ atoms have a similar atomic radius as U⁴⁺ atoms) responsible for more stress in UO₂ crystal.

La nécessité de simulations précises d’un réacteur nucléaire et spécialement dans des cas de cœurs et de configurations de combustible complexes, a imposé un usage accru de Codes Neutroniques Stochastiques (CNS). De plus, une demande a émergé pour des CNS à capacité inhérente d’estimation en continu de la variation de la composition isotopique du cœur ainsi qu’à couplage thermo-hydraulique optimisé. Des capacités supplémentaires sont exigées pour ces codes au vu de leur utilisation pour l’étude de nouveaux concepts de réacteur comme les Réacteurs Conduits par Accélérateur (RCA). Plus précisément, le réacteur hybride comprenant un réacteur nucléaire conventionnel et un accélérateur, nécessite l’analyse des deux composantes (réacteur – accélérateur) par un outil capable de couvrir le spectre énergétique neutronique extrêmement étendu qui caractérise ce système hybride.Ce travail présente les principales caractéristiques et capacités du nouveau CNS ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) développé en collaboration du NCSR Demokritos (Grèce) avec CNRS/IDRIS et UPMC (France) et couvrant autant que possible les exigences exposées ci-dessus. ANET est basé sur la version ouverte du code PHE GEANT3.21 et est destiné à effectuer des analyses de cœurs de réacteurs conventionnels de génération II et III ainsi que des RCA. ANET est construit avec la capacité inhérentea) d’effectuer des calculs d’évolution du combustibleb) de simuler le processus de spallation dans le cas des RCAtout en tenant compte de la thermo-hydraulique du système.La version actuelle d’ANET utilise les trois estimateurs standard Monte Carlo pour le calcul du facteur de multiplication neutronique effectif (keff), soit l’estimateur de collision, celui d’absorption et celui de longueur de trace. Pour ce qui est du calcul du débit de fluence neutronique et des taux de réaction, les estimateurs de collision et de longueur de trace sont implémentés dans ANET suivant la procédure standard Monte Carlo. Pour ce qui concerne les calculs d’évolution (par exemple la consommation du combustible), une approche purement stochastique est implémentée dans ANET. A noter que la procédure usuelle consiste à coupler le code neutronique stochastique avec un code déterministe qui calcule la consommation du combustible. Pour les besoins d’analyse des RCA, le module INCL/ABLA a été incorporé dans ANET de façon à ce que le processus de spallation soit simulé par le code. La capacité d’ANET de simuler des configurations classiques a été démontrée en utilisant des résultats de mesures et des simulations de vérification effectuées en utilisant d’autres codes bien établis, ainsi qu’il est montré par la suite.Des données provenant de plusieurs installations et des analyses de problèmes-type internationaux ont été utilisés pour vérifier et valider les capacités d’ANET.Pour conclure, les résultats obtenus lors des comparaisons avec des mesures ou avec des simulations effectuées en utilisant d’autres codes neutroniques stochastiques ou déterministes, montrent qu’ANET possède la capacité de calculer correctement d’importants paramètres de systèmes critiques ou sous-critiques. Par ailleurs, l’application préliminaire d’ANET à des problèmes dépendant du temps fournit des résultats encourageants. ANET produit des estimations de consommation de combustible raisonnables, compte tenu que des incertitudes dans ce domaine sont souvent de l’ordre de 20% ou plus. Finalement, les performances du code dans le cas de KUCA montrent qu’ANET peut analyser des RCA de façon satisfaisante. / The necessity for precise simulations of a nuclear reactor especially in case of complex core and fuel configurations has imposed the increasing use of Monte Carlo (MC) neutronics codes. Besides, a demand of additional stochastic codes’ inherent capabilities has emerged regarding mainly the simulation of the temporal variations in the core isotopic composition as well as the incorporation of the T-H feedback. In addition to the above, the design of innovative nuclear reactor concepts, such as the Accelerator Driven System (ADSs), imposed extra requirements of simulation capabilities. More specifically, the combination of an accelerator and a nuclear reactor in the ADS requires the simulation of both subsystems for an integrated system analysis. Therefore a need arises for more advanced simulation tools, able to cover the broad neutrons energy spectrum involved in these systems.This work presents the main features and capabilities of the new MC neutronics code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback), being developed in NCSR Demokritos (Greece) in cooperation with CNRS/IDRIS and UPMC (France) and intending to meet as effectively as possible the above described modelling requirements. ANET is based on the open-source version of the HEP code GEANT3.21 and is targeting to the creation of an enhanced computational tool in the field of reactor analysis, capable of simulating both GEN II/III reactors and ADSs. ANET is structured with inherent capability of (a) performing burnup calculations and (b) simulating the spallation process in the ADS analysis, while taking T-H feedback into account.The current ANET version utilizes the three standard Monte Carlo estimators for the neutron multiplication factor (keff) calculation, i.e. the collision estimator, the absorption estimator and the track-length estimator. Regarding the simulation of neutron fluence and reaction rates, the collision and the track-length estimators are implemented in ANET following the standard Monte Carlo procedure. For the burnup calculations ANET attempts to apply a pure Monte Carlo approach, adopting the typical procedure followed in stochastic codes. With respect to code improvements for the ADS analysis, so far ANET has incorporated the INCL/ABLA code so that the spallation process can be inherently simulated. The ANET reliability in typical computations was tested using observational data and parallel simulations by different codes as described in the following chapters.Various installations and international benchmarks were considered suitable for the verification and validation of all the previously mentioned features incorporated in the new code ANET. The obtained results are compared with experimental data from the nuclear infrastructures and with computations performed by well-established stochastic or deterministic neutronics codes and show satisfactory agreement with both measurements and independent computations, verifying thus ANET’s ability to successfully simulate important parameters of critical and subcritical systems. Also, the preliminary ANET application for dynamic analysis is encouraging since it indicates the code capability to inherently provide a reasonable prediction for the core inventory evolution taking into account the uncertainties of the order of 20% and even higher that are traditionally expected in core inventory evolution calculations. Lastly, the code performance in the KUCA case was found satisfactory demonstrating thus inherent capability of analyzing ADSs.

Le code MURE est basé sur le couplage d’un code Monte Carlo statique et le calcul de l’évolution pendant l’irradiation et les différentes périodes du cycle (refroidissement, fabrication). Le code MURE est ici utilisé pour analyser deux différentes questions : le mono-recyclage de l’Am dans les réacteurs français de type REP et la conversion du coeur du MNSR (Miniature Neutron Source Reactor) au Ghana d’un combustible à uranium hautement enrichi (HEU) vers un combustible faiblement enrichi (LEU), dans le cadre de la lutte contre la prolifération. Dans les deux cas, une comparaison détaillée est menée sur les taux d’irradiation et les radiotoxicités induites (combustibles usés, déchets).Le combustible UOX envisagé est enrichi de telle sorte qu’il atteigne un taux d’irradiation de 46 GWj/t et 68 GWj/t. Le combustible UOX usé est retraité, et le retraitement standard consiste à séparer le plutonium afin de fabriquer un combustible MOX sur base d’uranium appauvri. La concentration du Pu dans le MOX est déterminée pour atteindre un taux d’irradiation du MOX de 46 et 68 GWj/t. L’impact du temps de refroidissement de l’UOX usé est étudié (5 à 30 ans), afin de quantifier l’impact de la disparition du 241PU (fissile) par décroissance radioactive (T=14,3 ans). Un refroidissement de 30 ans demande à augmenter la teneur en Pu dans le MOX. L’241Am, avec une durée de vie de 432 ans, jour un rôle important dans le dimensionnement du site de stockage des déchets vitrifiés et dans leur radiotoxicité à long terme. Il est le candidat principal à la transmutation, et nous envisageons donc son recyclage dans le MOX, avec le plutonium. Cette stratégie permet de minimiser la puissance résiduelle et la radiotoxicité des verres, en laissant l’Am disponible dans les MOX usés pour une transmutation éventuelle future dans les réacteurs rapides. Nous avons étudié l’impact neutronique d’un tel recyclage. Le temps de refroidissement de l’UOX est encore plus sensible ici car l’241Am recyclé est un fort poison neutronique qui dégrade les performances du combustible (taux d’irradiation, coefficients de vide et de température). Néanmoins, à l’exception de quelques configurations, le recyclage de l’Am ne dégrade pas les coefficients de sûreté de base. Le réacteur MNSR du Ghana fonctionne aujourd’hui avec de l’uranium enrichi à 90,2% (HEU), et nous étudions ici la possibilité de le faire fonctionner avec de l’uranium enrichi à 12,5%, en passant d’un combustible sur base d’aluminium à un oxyde. Les simulations ont été menées avec le code MURE, et montrent que le coeur LEU peut-être irradié plus longtemps, mais demande d’intervenir plus tôt sur le pilotage en jouant sur la quantité de béryllium en coeur. Les flux de neutrons dans les canaux d’irradiation sont similaires pour les coeurs HEU et LEU, de même pour les coefficients de vide. Le combustible LEU usé présente cependant une radiotoxicité et une chaleur résiduelle plus élevée, du fait de la production plus importante de transuraniens pendant l’irradiation. / The MURE code is based on the coupling of a Monte Carlo static code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to analyse two different questions, concerning the mono-recycling of Am in present French Pressurized Water Reactor, and the conversion of high enriched uranium (HEU) used in the Miniature Neutron Source Reactor in Ghana into low enriched uranium (LEU) due to proliferation resistance issues. In both cases, a detailed comparison is made on burnup and the induced radiotoxicity of waste or spent fuel. The UOX fuel assembly, as in the open cycle system, was designed to reach a burn-up of 46GWd/T and 68GWd/T. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of Plutonium and addition of depleted Uranium to reach burn-ups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. Spent UOX fuel, after 30 years of cooling in the repository required higher concentration of Pu to be reprocessed into a MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has high radiotoxic level, high mid-term residual heat and a precursor for other long lived isotope. An innovative strategy consists of reprocessing not only the plutonium from the UOX spent fuel but also the americium isotopes which dominate the radiotoxicity of present waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all the Am. The main objective is to propose a “waiting strategy” for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOXAm (MOX and Americium isotopes) fuel was fabricated to see the effect of americium in MOX fuel on the burn-up, neutronic behavior and on radiotoxicity. The MOXAm fuel showed relatively good indicators both on burnup and on radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared to the open cycle. All fuel types understudy in the PWR cycle showed good safety inherent feature with the exception of the some MOXAm assemblies which have a positive void coefficient in specific configurations, which could not be consistent with safety features. The core lifetimes of the current operating 90.2% HEU UAl fuel and the proposed 12.5% LEU UOX fuel of the MNSR were investigated using MURE code. Even though LEU core has a longer core life due to its higher core loading and low rate of uranium consumption, the LEU core will have it first beryllium top up to compensate for reactivity at earlier time than the HEU core. The HEU and LEU cores of the MNSR exhibited similar neutron fluxes in irradiation channels, negative feedback of temperature and void coefficients, but the LEU is more radiotoxic after fission product decay due to higher actinides presence at the end of its core lifetime.

DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations.

The MURE code is based on the coupling of a Monte Carlo static code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to analyse two different questions, concerning the mono-recycling of Am in present French Pressurized Water Reactor, and the conversion of high enriched uranium (HEU) used in the Miniature Neutron Source Reactor in Ghana into low enriched uranium (LEU) due to proliferation resistance issues. In both cases, a detailed comparison is made on burnup and the induced radiotoxicity of waste or spent fuel. The UOX fuel assembly, as in the open cycle system, was designed to reach a burn-up of 46GWd/T and 68GWd/T. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of Plutonium and addition of depleted Uranium to reach burn-ups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. Spent UOX fuel, after 30 years of cooling in the repository required higher concentration of Pu to be reprocessed into a MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has high radiotoxic level, high mid-term residual heat and a precursor for other long lived isotope. An innovative strategy consists of reprocessing not only the plutonium from the UOX spent fuel but also the americium isotopes which dominate the radiotoxicity of present waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all the Am. The main objective is to propose a "waiting strategy" for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOXAm (MOX and Americium isotopes) fuel was fabricated to see the effect of americium in MOX fuel on the burn-up, neutronic behavior and on radiotoxicity. The MOXAm fuel showed relatively good indicators both on burnup and on radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared to the open cycle. All fuel types understudy in the PWR cycle showed good safety inherent feature with the exception of the some MOXAm assemblies which have a positive void coefficient in specific configurations, which could not be consistent with safety features. The core lifetimes of the current operating 90.2% HEU UAl fuel and the proposed 12.5% LEU UOX fuel of the MNSR were investigated using MURE code. Even though LEU core has a longer core life due to its higher core loading and low rate of uranium consumption, the LEU core will have it first beryllium top up to compensate for reactivity at earlier time than the HEU core. The HEU and LEU cores of the MNSR exhibited similar neutron fluxes in irradiation channels, negative feedback of temperature and void coefficients, but the LEU is more radiotoxic after fission product decay due to higher actinides presence at the end of its core lifetime.

A detailed reactor core modeling of the LOVIISA-2 PWR and FORSMARK-3BWR was performed in the Serpent 2 Continuous Energy Monte-Carlocode.Both models of the reactors were completed but the approximations ofthe atomic densities of nuclides present in the core differedsignificantly.In the LOVIISA-2 PWR, the predicted atomic density for the nuclidesapproximated by Chebyshev Rational Approximation method (CRAM)coincided with the corrected atomic density simulated by the Serpent2 program. In the case of FORSMARK-3 BWR, the atomic density fromCRAM poorly approximated the data returned by the simulation inSerpent 2. Due to boiling of the moderator in the core of FORSMARK-3,the model seemed to encounter problems of fission density, whichyielded unusable results.The results based on the models of the reactor cores are significantto the FANCSEE Nuclear fuel cycle simulator, which will be used as adataset for the nuclear fuel cycle burnup in the reactors. / FANCSEE

DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations.

<p>Accelerator-driven systems for transmutation of nuclear waste have been suggested as a means for dealing with spent fuel components that pose potential radiological hazard for long periods of time. While not entirely removing the need for underground waste repositories, this nuclear waste incineration technology provides a viable method for reducing both waste volumes and storage times. Potentially, the time spans could be diminished from hundreds of thousand years to merely 1.000 years or even less. A central aspect for accelerator-driven systems design is the prediction of safety parameters and fuel economy. The simulations performed rely heavily on nuclear data and especially on the precision of the neutron cross section representations of essential nuclides over a wide energy range, from the thermal to the fast energy regime. In combination with a more demanding neutron flux distribution as compared with ordinary light-water reactors, the expanded nuclear data energy regime makes exploration of the cross section sensitivity for simulations of accelerator-driven systems a necessity. This fact was observed throughout the work and a significant portion of the study is devoted to investigations of nuclear data related effects. The computer code package EA-MC, based on 3-D Monte Carlo techniques, is the main computational tool employed for the analyses presented. Directly related to the development of the code is the extensive IAEA ADS Benchmark 3.2, and an account of the results of the benchmark exercises as implemented with EA-MC is given. CERN's Energy Amplifier prototype is studied from the perspectives of neutron source types, nuclear data sensitivity and transmutation. The commissioning of the n_TOF experiment, which is a neutron cross section measurement project at CERN, is also described.</p>