Développement et validation de schémas de calcul dédiés à l'interprétation des mesures par oscillation pour l'amélioration des données nucléaires

Reactivity measurements by the oscillation technique, as those performed in the Minerve reactor, enable to access various neutronic parameters on materials, fuels or specific isotopes. Usually, expected reactivity effects are small, about ten pcm at maximum. Then, the modeling of these experiments should be very precise, to obtain reliable feedback on the pointed parameters. Especially, calculation biases should be precisely identified, quantified and reduced to get precise information on nuclear data. The goal of this thesis is to develop a reference calculation scheme, with well quantified uncertainties, for in-pile oscillation experiments. In this work are presented several small reactivity calculation methods, based on deterministic and/or stochastic calculation codes. Those method are compared thanks to a numerical benchmark, against a reference calculation. Three applications of these methods are presented here: a purely deterministic calculation with exact perturbation theory formalism is used for the experimental validation of fission product cross sections, in the frame of reactivity loss studies for irradiated fuel; an hybrid method, based on a stochastic calculation and the exact perturbation theory is used for the readjustment of nuclear data, here 241Am; and a third method, based on a perturbative Monte Carlo calculation, is used in a conception study.Les mesures de réactivité par la technique d'oscillation, comme celles effectuées dans le réacteur Minerve, permettent de tester de nombreux paramètres neutroniques sur des matériaux, des combustibles ou des isotopes spécifiques. Généralement, les effets attendus sont très faibles, tout au plus de l'ordre de la dizaine de pcm. La modélisation de ces expériences doit donc être particulièrement précise, afin d'obtenir un retour fiable et précis sur les paramètres ciblés. En particulier, les biais de calcul doivent être clairement identifiés, quantifiés et maîtrisés afin d'obtenir des informations pertinentes sur les données nucléaires de base. L'enjeu de cette thèse est le développement d'un schéma de calcul de référence, dont les incertitudes sont clairement identifiées et quantifiées, permettant l'interprétation des mesures par oscillation. Dans ce document plusieurs méthodes de calcul de ces faibles effets en réactivité sont présentées, basées sur des codes de calculs neutroniques déterministes et/ou stochastiques. Ces méthodes sont comparées sur un benchmark numérique, permettant leur validation par rapport à un calcul de référence. Trois applications sont ici présentées dans le détail : une méthode purement déterministe utilisant la théorie des perturbations exacte pour la qualification des sections efficaces des principaux produits de fission en REP, dans le cadre d'études sur l'estimation de la perte du réactivité du combustible au cours du cycle ; une méthode hybride, basée sur un calcul stochastique et la théorie des perturbations exacte, permet d'obtenir un retour précis sur les données nucléaires de bases d'isotopes, dans notre cas l'241Am; et enfin, une troisième méthode, reposant sur un calcul perturbatif Monte Carlo, est utilisée pour une étude de conception

Development and validation of calculation schemes dedicated to the interpretation of small reactivity effects for nuclear data improvement

Les mesures de réactivité par la technique d'oscillation, comme celles effectuées dans le réacteur Minerve, permettent de tester de nombreux paramètres neutroniques sur des matériaux, des combustibles ou des isotopes spécifiques. Généralement, les effets attendus sont très faibles, tout au plus de l'ordre de la dizaine de pcm. La modélisation de ces expériences doit donc être particulièrement précise, afin d'obtenir un retour fiable et précis sur les paramètres ciblés. En particulier, les biais de calcul doivent être clairement identifiés, quantifiés et maîtrisés afin d'obtenir des informations pertinentes sur les données nucléaires de base. L'enjeu de cette thèse est le développement d'un schéma de calcul de référence, dont les incertitudes sont clairement identifiées et quantifiées, permettant l'interprétation des mesures par oscillation. Dans ce document plusieurs méthodes de calcul de ces faibles effets en réactivité sont présentées, basées sur des codes de calculs neutroniques déterministes et/ou stochastiques. Ces méthodes sont comparées sur un benchmark numérique, permettant leur validation par rapport à un calcul de référence. Trois applications sont ici présentées dans le détail : une méthode purement déterministe utilisant la théorie des perturbations exacte pour la qualification des sections efficaces des principaux produits de fission en REP, dans le cadre d'études sur l'estimation de la perte du réactivité du combustible au cours du cycle ; une méthode hybride, basée sur un calcul stochastique et la théorie des perturbations exacte, permet d'obtenir un retour précis sur les données nucléaires de bases d'isotopes, dans notre cas l'241Am; et enfin, une troisième méthode, reposant sur un calcul perturbatif Monte Carlo, est utilisée pour une étude de conception.Reactivity measurements by the oscillation technique, as those performed in the Minerve reactor, enable to access various neutronic parameters on materials, fuels or specific isotopes. Usually, expected reactivity effects are small, about ten pcm at maximum. Then, the modeling of these experiments should be very precise, to obtain reliable feedback on the pointed parameters. Especially, calculation biases should be precisely identified, quantified and reduced to get precise information on nuclear data. The goal of this thesis is to develop a reference calculation scheme, with well quantified uncertainties, for in-pile oscillation experiments. In this work are presented several small reactivity calculation methods, based on deterministic and/or stochastic calculation codes. Those method are compared thanks to a numerical benchmark, against a reference calculation. Three applications of these methods are presented here: a purely deterministic calculation with exact perturbation theory formalism is used for the experimental validation of fission product cross sections, in the frame of reactivity loss studies for irradiated fuel; an hybrid method, based on a stochastic calculation and the exact perturbation theory is used for the readjustment of nuclear data, here 241Am; and a third method, based on a perturbative Monte Carlo calculation, is used in a conception study

Développement et validation de schémas de calcul dédiés à l'interprétation des mesures par oscillation pour l'amélioration des données nucléaires

Les mesures de réactivité par la technique d'oscillation, comme celles effectuées dans le réacteur Minerve, permettent de tester de nombreux paramètres neutroniques sur des matériaux, des combustibles ou des isotopes spécifiques. Généralement, les effets attendus sont très faibles, tout au plus de l'ordre de la dizaine de pcm. La modélisation de ces expériences doit donc être particulièrement précise, afin d'obtenir un retour fiable et précis sur les paramètres ciblés. En particulier, les biais de calcul doivent être clairement identifiés, quantifiés et maîtrisés afin d'obtenir des informations pertinentes sur les données nucléaires de base. L'enjeu de cette thèse est le développement d'un schéma de calcul de référence, dont les incertitudes sont clairement identifiées et quantifiées, permettant l'interprétation des mesures par oscillation. Dans ce document plusieurs méthodes de calcul de ces faibles effets en réactivité sont présentées, basées sur des codes de calculs neutroniques déterministes et/ou stochastiques. Ces méthodes sont comparées sur un benchmark numérique, permettant leur validation par rapport à un calcul de référence. Trois applications sont ici présentées dans le détail : une méthode purement déterministe utilisant la théorie des perturbations exacte pour la qualification des sections efficaces des principaux produits de fission en REP, dans le cadre d'études sur l'estimation de la perte du réactivité du combustible au cours du cycle ; une méthode hybride, basée sur un calcul stochastique et la théorie des perturbations exacte, permet d'obtenir un retour précis sur les données nucléaires de bases d'isotopes, dans notre cas l'241Am; et enfin, une troisième méthode, reposant sur un calcul perturbatif Monte Carlo, est utilisée pour une étude de conception.Reactivity measurements by the oscillation technique, as those performed in the Minerve reactor, enable to access various neutronic parameters on materials, fuels or specific isotopes. Usually, expected reactivity effects are small, about ten pcm at maximum. Then, the modeling of these experiments should be very precise, to obtain reliable feedback on the pointed parameters. Especially, calculation biases should be precisely identified, quantified and reduced to get precise information on nuclear data. The goal of this thesis is to develop a reference calculation scheme, with well quantified uncertainties, for in-pile oscillation experiments. In this work are presented several small reactivity calculation methods, based on deterministic and/or stochastic calculation codes. Those method are compared thanks to a numerical benchmark, against a reference calculation. Three applications of these methods are presented here: a purely deterministic calculation with exact perturbation theory formalism is used for the experimental validation of fission product cross sections, in the frame of reactivity loss studies for irradiated fuel; an hybrid method, based on a stochastic calculation and the exact perturbation theory is used for the readjustment of nuclear data, here 241Am; and a third method, based on a perturbative Monte Carlo calculation, is used in a conception study.SAVOIE-SCD - Bib.électronique (730659901) / SudocGRENOBLE1/INP-Bib.électronique (384210012) / SudocGRENOBLE2/3-Bib.électronique (384219901) / SudocSudocFranceF

Thermal neutron activation experiments on Ag, In, Cs, Eu, V, Mo, Zn, Sn and Zr in the MINERVE facility

The MAESTRO experimental program has been designed to improve nuclear data uncertainty on a large range of materials used for detection, absorption, moderation and structures in LWRs. It consists of pile-oscillation and neutron activation experiments, carried out in the MINERVE low power facility. For this program, the core configuration has been designed to be representative of HZP (Hot Zero Power) conditions of a typical PWR. Samples of high purity elements have been manufactured with severe technological constraints to reach a target accuracy of ±2% (1σ) on the measurement. This paper presents a preliminary analysis of activation experiments, based on TRIPOLI4 Monte-Carlo calculations and various nuclear data libraries

Representativity studies of PROTEUS UO2_2 fuel for new GEN-III+-type configurations in the EOLE Critical Facility

Extended abstract Within the framework of an international collaboration between VENUS, EOLE and PROTEUS experimental teams (VEP), project agreements in the fields of experimental reactor physics design and techniques are proposed to be shared between parties. In this context, the Experimental Physics Division (SPEx) of CEA is currently designing a new program devoted to core physics of GEN-III type lattices that requires higher enriched U5 PWR fuel pins. Those lattices are composed of a central 17x17 5wt% enriched $^{235}$U UO$_2$ assembly, surrounded by portions of 3,7% enriched $^{235}$U UO$_2$ PWR fuel pins. The EOLE stockpile requires manufacturing the central assembly, hence increasing the overall costs of the program. The PROTEUS situation is such that its 5% UO$_2$ pins stockpile could be used in those lattices, in place of new fuels to be manufactured. Nevertheless, due to their radial size, bigger than traditional PWR pins, a so-called representativity study must be started in order to verify that, under some constraints, those PROTEUS fuels could lead to adequate lattices characteristics, both in terms of critical mass, but also in terms of spectrum. To check this potential feasibility, several preliminary calculations were performed to determine adequate size and pitch of the mixed mock-up. Amongst them, representativity calculations were made, to optimize both overclad thickness and lattice pitch, compatible with the buffer zone. The targets being a 4.95% enriched UO$_2$ PWR cell, and then assembly, both at hot zero power condition, the process of representativity has been managed in two ways:-a traditional approach only based on conservation of the moderation ratio, and its analysis in terms of sensitivity profiles: the target moderation ratio being fixed once for all by adjusting the overclad thickness-a more innovative method based on optimization of the representativity itself, also based on similarities of the sensitivity profiles, and adding one or more degrees of freedom in technological parameter (clad thickness, pitch, etc…).The paper is split into 4 parts: after a summary of the experimental program needs and their design, the second chapter details the concept of representativity, as the tools required for this study. A section is devoted to the iterative method built to automatically optimize the representativity factor. The third part details the benchmark, the tools used by each party, as well as the covariance data. The last part makes a comparative analysis of results obtained by each participant, as well as the main feasibility conclusions and forthcoming studies

State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors

The paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs) and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n–γ field. A recently developed methodology based on the use of 7Li and 6Li-enriched TLDs, precalibrated both in photon and neutron fields, is a promising approach to deconvolute the two components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fibered setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1σ)

Some Considerations on the Energy Deposition During a RIA Transient Based On Monte Carlo Simulations

Specific research reactors are capable of reproducing reactivity injection accidents in order to study the behavior of the nuclear fuel pins in accidental situations. In the CABRI research reactor, the fuel pin to be examined (test pin) is placed in the center of the core in a dedicated test loop. It is then subjected to a power transient, obtained by the fast depressurization of the 3He neutron absorber gas from the transient rods located in the core. One of the central parameters of the experiment is the energy deposition in the test pin, which is currently not measured during a transient. Instead, it is assumed that the relative energy distribution between the core and the test pin is constant regardless the operational state of the reactor. Currently, this correlation is measured in steady state. As such, the impact of the variations in the neutron flux, fuel and moderator temperatures during the transient is assumed equivalent on the energy deposition in the core and in the test pin. The goal of this work is to improve our knowledge on the mechanisms involved in the transient energy deposition. The aim of this paper is to present a methodological approach for the energy deposition estimation during a CABRI transient, based on static Monte Carlo calculations. The results suggest that the transient energy deposition rate is mainly dependent on the helium pressure and the Doppler feedback, and the relative energy distribution between the core and test pin changes during the transient