Couplage 3D neutronique thermohydraulique (développement d outils pour les études de sûreté des réacteurs innovants)

Les études relatives aux réacteurs nucléaires font appel à plusieurs disciplines dont les principales sont la neutronique et la thermo-hydraulique. Les phénomènes physiques qui se déroulent dans le cœur d une centrale nucléaire comme la réaction en chaîne des fissions nucléaires, le mouvement des fluides et les transferts de chaleur se couplent de manière forte et complexe. De part l avancement des connaissances dans ces disciplines et la croissance massive de la puissance des ordinateurs, cette complexité phénoménologique peut aujourd hui être simulée en des temps raisonnables. C est pour cette raison que les codes de neutronique stochastiques, dits Monte Carlo, sont bien plus utilisés de nos jours que par le passé. Un grand intérêt de ce type de code probabiliste réside dans leur aptitude à reproduire fidèlement la réalité sans recours à des approximations de modélisation. C est dans ce contexte que cette thèse a été initiée : coupler un code Monte Carlo de neutronique à un code de thermo-hydraulique cœur afin d assurer une description la plus précise possible des conditions de fonctionnement d un cœur de réacteur nucléaire. Ces travaux s inscrivent dans une démarche évolutionnaire motivée par les exigences accrues de la sûreté, d optimisation des ressources et de minimisation des déchets pour les systèmes nucléaires du futur. Ce manuscrit présente la méthodologie employée pour le développement d un couplage externe automatisé entre le code Monte Carlo MCNP et le code de thermo-hydraulique/thermique COBRA-EN. Cette recherche d une meilleure performance et précision des outils de calcul s accompagne de nouveaux types de problèmes physico-numériques à résoudre, dont les principaux sont exposés dans ce mémoire. La validation du schéma couplé a été réalisée sur un cas très complexe de cœur de réacteur et a permis de prouver la robustesse des développements entrepris et la faisabilité d un tel couplage.ORSAY-PARIS 11-BU Sciences (914712101) / SudocSudocFranceF

Bold application of sNDM to REA in a SSCR core with azimuthal mesh

Previously applied to REA (Rod Ejection Accident) in several PWR-like cores, the minimalistic Nodal Drift Method (NDM) has recently been generalized to sNDM (super NDM). Both developed and validated on a heat-up transient of the KRUSTY experiment made of a few homogeneous parts, sNDM basically feeds the one-group diffusion approximation with so-called corrective Surface Factors (SF) for internodal currents from MCNP F1 tallies. In order to specify its practical usefulness for exploratory design studies, sNDM at its turn is put to the demanding test of REA in a PWR-like core. The chosen test case is a 600 MWth D2O/H2O-cooled thorium-fueled Spectral Shift Control Reactor (SSCR) core retrieved from previous studies, whose main results on conversion (by our MC-based tool SMURE) and safety (by NDM) are first summed up. Enhanced MCNP core models at HFP, CZP and HZP (respectively Hot Full, Cold Zero and Hot Zero Power) are detailed that have been specially adapted to a 2D radial-azimuthal mesh of few large nodes towards an even simpler REA calculation by sNDM. Other settings, necessary at BOT (Beginning Of Transient from HZP), include fuel and coolant thermal feedbacks as well as the global conductance of a lumped thermal model. Last but not least, the special cases of a few SF found variable between BOT and EOL (End Of Launch at t = 0.05 s) are addressed by an iterative transient calculation based on their linear interpolation. This method is proven effective at the cost of accepting an irreducible discrepancy for the radial exchange rate of the ejected node, provided that a proper so-called global way of computing all SF is used. Finally, main transient results are given until EOT equilibrium (End Of Transient at t = 300 s), with various sanity checks (including a partial safety one)

Study of D2O/H2O-cooled thorium-fueled PWR-like SMR cores using the KNACK toolbox: conversion and safety assessment

International audienceBased on SMURE (Serpent2/MCNP Utility) and NDM (Nodal Drift Method for time-dependent diffusion), a full set of academic methods named KNACK (Knack of Nodal Approach to Core Kinetics) has been used for the design of 600 MWth D2O/H2O-cooled thorium-fueled SMR (Small Modular Reactor) cores. Three types of lattice, with 17x17, 19x19 or 21x21 PWR-like FAs (Fuel Assemblies), have been considered. After initial fissile zoning for power flattening, full core burnup calculations with D2O/H2O Spectral Shift Control have been performed at HFP (Hot Full Power) for the comparison of conversion performance. Temperature dependences of diffusion data have been implemented within a thermal lumped model for safety. A simple criterion, on coolant temperatures only, has finally been used for the comparative analysis of Rod Ejection Accidents (REA) from HZP (Hot Zero Power)

Study of D2O/H2O-cooled thorium-fueled PWR-like SMR cores using the KNACK toolbox: conversion and safety assessment

International audienceBased on SMURE (Serpent2/MCNP Utility) and NDM (Nodal Drift Method for time-dependent diffusion), a full set of academic methods named KNACK (Knack of Nodal Approach to Core Kinetics) has been used for the design of 600 MWth D2O/H2O-cooled thorium-fueled SMR (Small Modular Reactor) cores. Three types of lattice, with 17x17, 19x19 or 21x21 PWR-like FAs (Fuel Assemblies), have been considered. After initial fissile zoning for power flattening, full core burnup calculations with D2O/H2O Spectral Shift Control have been performed at HFP (Hot Full Power) for the comparison of conversion performance. Temperature dependences of diffusion data have been implemented within a thermal lumped model for safety. A simple criterion, on coolant temperatures only, has finally been used for the comparative analysis of Rod Ejection Accidents (REA) from HZP (Hot Zero Power)

Fundamentals of reactor physics with a view to the (possible) futures of nuclear energy

International audienceThis paper present basic nuclear reactor physics that may help to understand next challenges that nuclear industry have to face in the future. The ones considered in this work are waste production and natural resources consumption. This paper shows that waste and resources are linked by the plutonium status that could be considered as the principal waste or a valuable material that should be saved for a future transition to breeder reactors that could be for instance Sodium-cooled Fast Reactors (SFRs). This kind of reactors does not rely on natural resource supply, as it produces its own fissile material, plutonium-239, after a neutron capture on uranium-238. However, the operation of SFRs needs a huge amount of plutonium that should be produced in current Pressurized Water Reactors (PWRs).Natural uranium available on earth is expected to allow the operation of the global current fleet until approximately 2100 without major issues. The transition from PWRs to SFRs is then needed if and only if the nuclear industry will face an increase at a global scale during this century. If not, plutonium, the most radiotoxic element produced in PWRs, should be considered as a waste. Consequently, the plutonium status depends on the future evolution of nuclear energy. This paper shows that a status-quo means that plutonium should be considered as the most problematic element that should be handled on a long-term basis. On the other hand, a strong increase in nuclear energy on a global scale would imply that plutonium is a valuable material that would make a transition to SFRs possible

Advanced plutonium management in PWR, complementarity of thorium and uranium

International audienc

Americium mono-recycling in PWR: A step towards transmutation

International audienceIn contrast to the straight final disposal solution, countries like France have opted to reprocess their nuclear reactors spent fuel and defined another way to take care of sensitive elements such as the plutonium or minor actinides. Even in countries which have chosen to reprocess their spent fuel, americium is still considered as a final disposal waste. Among the minor actinides, americium will remain the main contributor to the toxicity and the decay heat of the spent fuel for thousand of years. Therefore it is important to reduce its quantity. At this time, only fast neutron future reactors are accepted to be efficient enough to transmute the americium from the thermal reactors spent fuel. As we can presume these future reactors will not be available before many decades, a new strategy which consists in recycling americium together with plutonium in pressurize water reactors mixed oxide fuel is proposed. In this paper the benefit and after-effect of this waiting strategy is analyzed. It demonstrates that the americium is indeed transmuted in a PWR quite efficiently (transmutation rate of around 43%) however the spent fuel is, as expected, more concentrated in curium of heavier nuclei. The impact on the fuel cycle (transportation, cooling time) is investigated showing that the key point would be the fabrication of the MOx-Am fuel