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International audience; Fuel cycle simulator development started many years ago by several research and engineering institutions or consulting firms for a wide range of applications. To improve confidence in the results, institutions may be tempted to increase the complexity of their software even if this complexity might not be necessary. On the other hand, some simulators may be used outside their range of validity when used in very specific applications. The FIT (Functionality Isolation Test) project is an international effort devoted to improve the confidence in the data produced by fuel cycle simulation tools. The scientific goal is to determine the optimum level of detail a fuel cycle simulator needs according to the type of study and the required confidence level. The project relies on a wide variety of fuel cycle simulators with a large range of complexity levels. The FIT project consists of isolating the impact of one targeted functionality on fuel cycle simulations. The impact of the functionality is assessed using a set of simple basic exercises specifically designed for this purpose, called ”functionality isolation.” The present work focuses on the impact on simulation results of using a fuel loading model (a relation that links the stock isotopic composition with the fresh fuel fabrication according to the reactor requirements) or a fixed fraction approach (the fresh fuel fissile fraction is fixed and does not depend on the stock isotopic composition). The paper first presents the FIT project. The exercise design is described and results show that using a fuel loading model approach has an important impact on fuel cycle outputs under certain conditions that are described. This result is reinforced by the fact that all fuel cycle simulators used in this exercise provide similar conclusions.

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International audience; The aim of this paper is to analyze and comparenuclear scenarios including the Argentinian and theBrazilian nuclear fleets working either in a collaborativescenario or separately.Brazilian nuclear fleet is composed of PWRs (PressurizedWater Reactors). Fordecades,ArgentinahasbeenoperatingPressurizedHeavyWaterReactors(PHWR),inparticularaCanadianDeuteriumUranium(CANDU)typereactor.Thankstotheirveryefficientneutroneconomy,thosereactorsareknowntobeabletoacceptlargespectrumoffueltypes.TheycanbecriticalwhenfueledbyDUPIC(DirectUseofPWRfuelinCANDU),orwithreprocessedfuelfromirradiatedenricheduranium.Theycouldbeevenalmostbreedersifusedwiththorium-­‐basedfuels.Inthispaper,wepresenthowthespecificreactorphysicsofCANDUhavebeenimplementedinthescenariocodeCLASS(CoreLibraryforAdvancedScenarioSimulation).ThenwediscusstheadvantagesofacollaborativescenariobetweenArgentinaandBrazil.

International audience; In nuclear scenario studies, the use of simplified assembly models to simulate fuel irradiation in nuclear reactor core can lead to important biases. To calculate the spent fuel composition, the CLASS package (Core Library for Advanced Scenario Simulations) uses a physic model (usually, we define physics model as: 1- Fuel Loading Model, 2- Cross Section Predictor, 3- Bateman Solver), based on a large number of different fuel evolution simulations. For that, a simple and reproducible model has been chosen: the “assembly model” that is described in this paper. In this model, the whole core is represented by a single assembly, or a part of it, with reflecting boundary conditions.As a first step to quantify the errors related to this approximation, three real size assembly configurations have been simulated: a reflected one, an axial open one and a realistic one (with moderator and reflector at its top and bottom). An exhaustive comparison is presented in this paper. Hence this work focuses on the neutron axial leakage impact on burn-up and isotope inventories calculations.Evolutions up to 50 GWd/t have been simulated with the depletion code MURE for all the geometries and main isotope inventories, fission and capture cross sections and Keff have been reported for each time step. Axial leakage effects on neutron spectrum and inventory distributions have been observed. In this paper, a bias quantification of axial leakage and a comparison with homogeneous leakage models, available in the Serpent code, are analyzed.

ISBN 978-1-49-51-6286-2; International audience; Accelerator Driven Systems (ADS) are specifically studied for their capacity intransmuting Minor Actinides (MA). Electronuclear scenarios involving MA transmutation in ADSare widely researched. The dynamic fuel cycle simulation code CLASS (Core Library forAdvanced Scenarios Simulations) is used for predicting the inventory evolution induced by acomplex nuclear fleet. For managing reactors, the code CLASS is based on physic models. A FuelLoading Model (FLM) provides the fuel composition at Beginning Of Cycle (BOC) according tothe storages composition and the reactor requirements. A Cross Section Predictor (CSP) estimatesmean cross sections needed for solving evolution equations. Physic models are built from reactorscalculation set ahead of the scenario calculation. An ADS standard composition at BOC is amixture of plutonium and MA oxide. The high number of fissile isotopes present in the sub-criticalcore leads to an issue for building an ADS FLM. A high number of isotopic vectors at BOC isneeded to get an exhaustive simulation set. Also, ADS initial reactivity is adjusted with an inertmatrix that induces an additional degree of freedom. The building of an ADS FLM for CLASSrequires two steps. For any heavy nuclide composition at beginning of cycle, the core reactivitymust be imposed at a sub-critical level. Also, the reactivity coefficient evolution should bemaintained during the irradiation. For building a FLM, a simulation set has been built. Reactorsimulations are done with the transport code MCNP6 (Monte Carlo N particle transport code).The ADS geometry is based on the EFIT (European Facility for Industrial-Scale Transmutation)concept. The simulation set is composed of more than 8000 randomized. A complete neutronicstudy is presented that highlight the effect on MgO on neutronic parameters.

ISBN 978-1-49-51-6286-2; International audience; Fuel depletion calculation codes require one-group mean cross sections of manynuclides to solve Bateman's equations. In such codes, the mean cross sections are assed by themean of iterative calls of Boltzmann equation solver thanks to neutron transport codes. This is atime consuming task, especially with Monte Carlo codes such as MCNP. This paper presents amethodology based on neural network for building a cross section predictor for a PWR reactorloaded with any MOX fuel. This approach allows performing fuel depletion calculation in lessthan one minute with an excellent accuracy. A maximum deviation of 3% on actinides is obtainedat the end of cycle between inventories calculated from neural networks and from the referencecoupled neutron transport / fuel depletion calculation.

International audience; The global warming, the increase of world population and the depletion of fossil resourceshave lead us in a major energy crisis. Using electronuclear energy could be one of the meansto solve a part of these issues. The way out of this crisis may be enlightened by the study oftransitional scenarios, guiding the political decisions. The reliability of those studies passesthrough the wide variety of the simulation tools and the comparison between them.From this perspective and in order to perform complex electronuclear scenario simulation,the open source Core Library for Advance Scenario Simulation (CLASS) is being developed.CLASS main asset is its ability to include any kind of reactor, whether the system is innovativeor standard. A reactor is fully described by its evolution database that must contain a set ofdifferent fuel compositions in order to simulate transitional scenarios. CLASS aims at being auseful tool to study scenarios involving Generation IV reactors as well as innovative fuel cycles,like the Thorium cycle.The following contribution will present in detail the CLASS software. Starting with the workingprinciple of this tool, one will explain the working process of the different modules suchas the evolution module. It will be followed by an exhaustive presentation of the UOX-MOXbases generation procedure. Finally a brief analysis of the error made by the CLASS evolutionmodule will be presented.

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In the framework of the EUROTRANS project an experimental accelerator driven system “XT-ADS” is being designed. This document is the final deliverable of work package 1.4, the project in which the design of the spallation target for the XT-ADS is developed. The present status of the design of a windowless spallation target for the XT-ADS is discussed with particular focus on the modifications made as compared to MYRRHA draft 2. The document is composed of two major parts. After the definition of the design boundary conditions, a first major part deals with the general layout of the spallation target loop with a discussion of the current lay-out options. In the second part the different design support studies that have been performed are described. These include the spallation target zone and spallation loop thermal-hydraulics, system dynamics, beam impact studies, neutronics, mechanical integration and safety analyses. In the last section some conclusions are formulated.

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Proc. CD Series IAEA-13-CN-173 ISBN 9789201504104; International audience; EURISOL DS (Europe Isotope Separation On-Line Design Study) project is the European common effort in planning a next generation RIB factory able to deliver secondary beams up to 1013 pps at energies up to 150 MeV u-1. The proposed schematic layout of the facility is based on four target stations, three direct targets of 100 kW of beam power and one multi-MW target two stages assembly. Being produced via spallation the RIBs produced in the direct targets are mainly proton rich. While in the multi-MW target high intensity RIBs of neutron rich isotopes are produced by fission in actinide targets placed in the fast neutron spectrum given by a liquid metal spallation source. The purpose of this paper is to summarize the work carried out within Task 11 “Beam Intensity Calculations” with special emphasis to the estimation of the in-target yield intensities produced in the various target configurations. Benchmark studies were performed initially in order to verify the accurate description of the spallation models used by the MCNPX2.5.0 code and to choose the best options to be used for the present work requirements. The predictions of the code tested against measured data are presented. Several calculations using MCNPX2.5.0 combined with the evolution code CINDER'90 were carried out to assess the performance of the direct targets. A complex analysis was performed to study the in-target production RIB intensities varying with various parameters: target dimensions, materials and incident proton beam energies. The optimized configurations for the targets together with the corresponding quantitative estimates of the production rates for all interested nuclei resulted from this investigation are discussed.

The γ-ray emission probability distribution of 234Pa* has been measured between 5.2 and 6.2 MeV excitation energy using the transfer reaction 232Th(3He, p)234Pa* at an incident 3He energy of 24 MeV. The experimental set-up is made of four liquid scintillator C6D6 detectors and four Si telescopes arranged to detect γ rays and charged particles in coincidence. In order to determine the total number of γ-ray cascades in a given nucleus, the raw experimental detected gamma-ray events have to be corrected by the weighting function technique. This weighting requires complete knowledge of the efficiency and response functions for the C6D6 detector set-up and surroundings. These have been determined experimentally using gamma-ray sources, inelastic scattering reactions and transfer reactions on light nuclei. Simulations which reproduce successfully the detector response and efficiency are presented. From these measurements, the (n, γ) capture cross section of 233Pa, in the neutron energy range 0 to 1 MeV, has been deduced via the product of the measured γ emission probability of 234Pa* with the calculated compound nucleus formation cross section of the reaction 233Pa + n. A comparison between existing data bases or recent predictions and the present results is also made.

LPSC 0594; Molten Salt Reactor (MSR) based on Th/U cycle is one of the new generation concepts for nuclear energy production. A typical MSR is a graphite-moderated core with liquid fuel (^{7}LiF+ThF_{4}+UF_{4}). Many numerical studies based on Monte-Carlo codes are currently carried out but the validity of these numerical result relies on the precise knowledge of neutron cross sections used such as elastic scattering on carbon (\sigma_{C}), fluorine (\sigma_{F}) and lithium 7 (\sigma_{Li}). The goal of this work is to obtain \sigma_{C} and \sigma_{F} between 1 eV and 100 keV. Such measurements have been performed at the Laboratoire de Physique Subatomique et de Cosmologie (LPSC) de Grenoble on the experimental platform PEREN using slowing-down time spectrometers (C and CF_{2}) associated to a pulsed neutron generator (GENEPI). Capture rates are obtained for reference materials (Au, Ag, Mo and In) using YAP scintillator coupled to a photo-multiplier.Very precise simulations (MCNP code) of the experimental setup have been performed and comparison with experiments has led to the determination of \sigma_{C} and \sigma_{F} with accuracies of 1% and 2% respectively. These results show a small discrepancy to evaluated nuclear data file (ENDF).Measures of total cross-sections \sigma_{C} and \sigma_{F} at higher energy (200-600\, keV) were also carried out at Centre des Etudes Nucléaires de Bordeaux using a transmission method. Mono-energetic neutrons were produced by protons accelerated by a Van de Graaff accelerator on a LiF target and transmitted neutrons are counted in a proportional hydrogen gaseous detector. Discrepancies of 5\% and 9\% for \sigma_{C} and \sigma_{F} respectively with ENDF have been shown.; Le Réacteur à Sels Fondus (RSF) en cycle Th/U est un des concepts de nouvelle génération pour la production d'énergie nucléaire. Un RSF type se compose d'une structure en graphite, servant de modérateur, percé de canaux dans lesquels circule le sel fondu (^{7}LiF+ThF_{4}+UF_{4}). Les données neutroniques fondamentales pour les simulations numériques sont les sections efficaces de diffusion élastique d'un neutron sur des noyaux de carbone (\sigma_{C}), de fluor (\sigma_{F}) et de lithium 7 (\sigma_{Li}) dans le domaine épithermique. Le but de ce travail est de déterminer précisément \sigma_{C} et \sigma_{F} entre 1 eV et 100 keV. Ces mesures ont été réalisées auprès de la plate-forme d'Etude et de Recherche sur l'Electro-Nucléaire (PEREN) du Laboratoire de Physique Subatomique et de Cosmologie de Grenoble. Elle comprend un GEnérateur de NEutrons Pulsé Intense (GENEPI) produisant des neutrons de 2.5 MeV au centre de massifs ralentisseurs (C et CF_{2}). Des taux de capture sont obtenus, pour des échantillons composés de matériaux de référence (Au, Ag, Mo et In) placés dans les massifs, à l'aide d'un scintillateur YAP couplé à un photo-multiplicateur.Des simulations précises des expériences, avec le code MCNP, ont été réalisées et la comparaison entre la simulation et l'expérience a permis de déterminer \sigma_{C} et \sigma_{F} avec des précisions de 1% et 2% respectivement, mettant en évidence des écarts avec les données nucléaires évaluées (ENDF).Des mesures complémentaires des sections efficaces totales \sigma_{C} et \sigma_{F} à plus haute énergie (200 - 600keV) ont été effectuées au Centre d'Etudes Nucléaires de Bordeaux par une méthode de transmission. Un faisceau de neutrons mono-énergétique est produit par des protons accélérés, dans un accélérateur Van De Graaff, sur une cible de LiF. Les neutrons transmis sont comptés par un détecteur proportionnel à hydrogène. Des écarts de 5% (C) et de 9% (F) par rapport à ENDF ont été mis en évidence.

Molten Salt Reactor (MSR) based on Th/U cycle is one of the new generation concepts for nuclear energy production. A typical MSR is a graphite-moderated core with liquid fuel (^{7}LiF+ThF_{4}+UF_{4}). Many numerical studies based on Monte-Carlo codes are currently carried out but the validity of these numerical result relies on the precise knowledge of neutron cross sections used such as elastic scattering on carbon (\sigma_{C}), fluorine (\sigma_{F}) and lithium 7 (\sigma_{Li}). The goal of this work is to obtain \sigma_{C} and \sigma_{F} between 1 eV and 100 keV. Such measurements have been performed at the Laboratoire de Physique Subatomique et de Cosmologie (LPSC) de Grenoble on the experimental platform PEREN using slowing-down time spectrometers (C and CF_{2}) associated to a pulsed neutron generator (GENEPI). Capture rates are obtained for reference materials (Au, Ag, Mo and In) using YAP scintillator coupled to a photo-multiplier.Very precise simulations (MCNP code) of the experimental setup have been performed and comparison with experiments has led to the determination of \sigma_{C} and \sigma_{F} with accuracies of 1% and 2% respectively. These results show a small discrepancy to evaluated nuclear data file (ENDF).Measures of total cross-sections \sigma_{C} and \sigma_{F} at higher energy (200-600\, keV) were also carried out at Centre des Etudes Nucléaires de Bordeaux using a transmission method. Mono-energetic neutrons were produced by protons accelerated by a Van de Graaff accelerator on a LiF target and transmitted neutrons are counted in a proportional hydrogen gaseous detector. Discrepancies of 5\% and 9\% for \sigma_{C} and \sigma_{F} respectively with ENDF have been shown., Le Réacteur à Sels Fondus (RSF) en cycle Th/U est un des concepts de nouvelle génération pour la production d'énergie nucléaire. Un RSF type se compose d'une structure en graphite, servant de modérateur, percé de canaux dans lesquels circule le sel fondu (^{7}LiF+ThF_{4}+UF_{4}). Les données neutroniques fondamentales pour les simulations numériques sont les sections efficaces de diffusion élastique d'un neutron sur des noyaux de carbone (\sigma_{C}), de fluor (\sigma_{F}) et de lithium 7 (\sigma_{Li}) dans le domaine épithermique. Le but de ce travail est de déterminer précisément \sigma_{C} et \sigma_{F} entre 1 eV et 100 keV. Ces mesures ont été réalisées auprès de la plate-forme d'Etude et de Recherche sur l'Electro-Nucléaire (PEREN) du Laboratoire de Physique Subatomique et de Cosmologie de Grenoble. Elle comprend un GEnérateur de NEutrons Pulsé Intense (GENEPI) produisant des neutrons de 2.5 MeV au centre de massifs ralentisseurs (C et CF_{2}). Des taux de capture sont obtenus, pour des échantillons composés de matériaux de référence (Au, Ag, Mo et In) placés dans les massifs, à l'aide d'un scintillateur YAP couplé à un photo-multiplicateur.Des simulations précises des expériences, avec le code MCNP, ont été réalisées et la comparaison entre la simulation et l'expérience a permis de déterminer \sigma_{C} et \sigma_{F} avec des précisions de 1% et 2% respectivement, mettant en évidence des écarts avec les données nucléaires évaluées (ENDF).Des mesures complémentaires des sections efficaces totales \sigma_{C} et \sigma_{F} à plus haute énergie (200 - 600keV) ont été effectuées au Centre d'Etudes Nucléaires de Bordeaux par une méthode de transmission. Un faisceau de neutrons mono-énergétique est produit par des protons accélérés, dans un accélérateur Van De Graaff, sur une cible de LiF. Les neutrons transmis sont comptés par un détecteur proportionnel à hydrogène. Des écarts de 5% (C) et de 9% (F) par rapport à ENDF ont été mis en évidence.