Neutronic study of slightly modified water reactors and application to transition scenarios

International audienceIn this paper we have studied slightly modified water reactors and their applications to transition scenarios. The PWR and CANDU reactors have been considered. New fuels based on Thorium have been tested : Thorium/Plutonium and Thorium/Uranium- 233, with different fissile isotope contents. Changes in the geometry of the assemblies were also explored to modify the moderation ratio, and consequently the neutron flux spectrum. A core equivalent assembly methodology was introduced as an exploratory approach and to reduce the computation time. Several basic safety analyses were also performed. We have finally developed a new scenario code, named OSCAR (Optimized Scenario Code for Advanced Reactors), to study the efficiency of these modified reactors in transition to GenIV reactors or in symbiotic fleet

Neutron-induced fission cross sections of short-lived actinides with the surrogate reaction method

Neutron-induced fission cross sections for 242,243Cm and 241Am have been obtained with the surrogate reaction method. Recent results for the neutron-induced cross section of 243Cm are questioned by the present data. For the first time, the 242Cm cross section has been determined up to the onset of second-chance fission. The good agreement at the lowest excitation energies between the present results and the existing neutron-induced data indicates that the distributions in spin and parity of states populated with both techniques are similar

Timing of treatment initiation for mild gestational diabetes mellitus and perinatal outcomes

To examine the association between gestational age (GA) at the time of treatment initiation for gestational diabetes (GDM) and maternal and perinatal outcomes

Couplage 3D neutronique thermohydraulique. Développement d'outils pour les études de sûreté des réacteurs innovants

Nuclear reactors are complex systems and modelling of their behaviour involves several sub-disciplines of physics. The most important are the neutronics, which governs the neutron transport and chain reaction in the core, and thermo-hydraulics, which treats the fluid flow of the coolant and the heat transfer from the fuel. These two different physical phenomena are coupled in reactor cores in a complex way : The fission chain reaction affects the heat produced and hence fuel and coolant temperatures and densities, and in turn, these affect the cross sections for the nuclear reactions. Thanks to the massive growth in computer power over the last few decades it is only now that it is possible to imagine to simulate this phenomenological complexity in a reasonable time. For this reason stochastic neutronics codes of the ”Monte-Carlo” type are used much more widely than in the past. They offer the great advantage of the ability of this type of probabilistic code resides in their ability to reproduce to ”faithfully” re-produce reality without recourse to modeling approximations. It is in this context that the following thesis work has been performed : a generic coupling of a Monte-Carlo based neutronics code to a thermo-hydraulics code to ensure the most accurate 3-dimensional description possible of operating conditions in a reactor core. This work is driven by the new demands for future reactor generations of increased security, the optimization of natural resources and the minimization of nuclear waste production. This manuscript presents the methodology for the development of an automated external coupling between the Monte Carlo based neutron transport code, MCNP, and the thermo-hydraulique/thermique code, COBRA-EN. The development these new and high precision simulation tools was accompanied with new physical-numeric problems which had to be solved. The problems encountered are highlighted in the manuscript. Finally, the validation of the coupled scheme was carried out on a complex, heterogenous benchmark in order to prove the robustness of the code developments undertaken and the feasibility of such a couplingLes études relatives aux réacteurs nucléaires font appel `a plusieurs disciplines dont les principales sont la neutronique et la thermo-hydraulique. Les phénomènes physiques qui se déroulent dans le coeur 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 `a reproduire ”fidèlement” la réalit´e sans recours à des approximations de modélisation. C'est dans ce contexte que cette thèse a être initiée : coupler un code Monte Carlo de neutronique `a un code de thermo-hydraulique coeur afin d'assurer une description la plus précise possible des conditions de fonctionnement d'un coeur de réacteur nucléaire. Ces travaux s'inscrivent dans une démarche évolutionnaire motivée par les exigences accrues de la sˆuret´e, d'optimisation des ressources et de minimisation des déchets pour les systémes 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 `a résoudre, dont les principaux sont exposés dans ce mémoire. La validation du schéma couplé a été réalisée sur un cas très complexe de coeur de réacteur et a permis de prouver la robustesse des développements entrepris et la faisabilité d'un tel couplage

Enhancements to the Nodal Drift Method for a Rod Ejection Accident in a PWR-like mini-core with lumped thermal model

International audiencePrevious burnup calculations of High Conversion Water Reactor (HCWR) options have given promising fuel cycle results. Now safety must be assessed. This requires the development of academic computing tools, coupling as simply as possible neutronics and thermal-hydraulics. The diffusion-based Nodal Drift Method (NDM) for spatial core analysis has been developed towards a broad and coarse safety review of the most promising candidates. NDM has been already validated on a simplified CANDU LOCA benchmark, without thermal coupling. Dedicated to the calculation of a Rod Ejection Accident (REA) as defined by a recent PWR-like mini-core benchmark, the present work describes and validates all the necessary enhancements to NDM. A first 2D geometry with coupling to a lumped thermal model has been validated. Then, 3D features including a refined model for the ejected control rod have been added. Particular attention has been paid on the analysis of discrepancies with reference results. Limitations due to the oversimplified lumped thermal model are finally discussed

Design of a Fast Molten Salt Reactor for Space Nuclear Electric Propulsion

International audienceThe French National Center for Scientific Research (CNRS) is carrying out design studies on a nuclear electric propulsion (NEP) engine based on a molten salt reactor (MSR). A NEP engine based on liquid nuclear fuel could allow developing a core design with relatively high power densities and temperatures while using simple reactivity control systems and keeping low pressure and temperature gradients in the fuel. Nevertheless, the design work of such an engine poses significant technical challenges and requires the use of advanced numerical simulation tools. Different MSRs for space are currently being studied. In this work, a MSR concept using a fast neutron spectrum is investigated using a multiphysics tool based on a numerical coupling between the OpenFOAM (computational fluid dynamics) and SERPENT 2 (Monte Carlo neutronics) codes. The analysis of this paper is focused on the reactor core coupled neutronic and thermal-hydraulic phenomena. Steady state full-power conditions are calculated for two different fast MSR designs using low-enriched uranium (LEU) and highly enriched uranium. The results show that the proposed core layout and materials allow obtaining a satisfactory temperature distribution in the core (maximal values and gradients) without significant penalization of the reactor operating conditions. A reactivity control strategy excluding the use of control rods is studied for the LEU concept. Transient and safety studies are also performed and show acceptable performance