Proposition d'un modele de coeur pour le RSF thorium minimisant la quantite de moderateur graphite en coeur

Rapport intern

La gestion des déchets nucléaires

La gestion des déchets nucléaires constitue pour le public un problème majeur et inquiétant bien que pour la majorité des acteurs du nucléaire, le stockage géologique soit une solution appropriée qui répond parfaitement aux légitimes exigences de sécurité. La loi de 1991 relative à la gestion des déchets nucléaires définit un cadre législatif qui a organisé pendant 15 ans les recherches sur ce sujet, et qui a pris soin d'y associer la société civile. Au vu des résultats obtenus, le parlement a voté en juin 2006 une nouvelle loi permettant de poursuivre et finaliser les recherches, avec pour objectif de proposer au parlement une solution industrielle pérenne à l'horizon 2015. Après avoir appréhendé la complexité du sujet, les principaux acquis de la loi de 1991 seront présentés. Ils permettront d'une part de mieux comprendre les nouveaux enjeux des recherches actuelles et d'autre part de situer les solutions de gestion possible par rapport aux différents scénarios concernant la poursuite de l'utilisation de l'énergie nucléaire

The Thorium Molten Salt Reactor : Moving on from the MSBR

A re-evaluation of the Molten Salt Breeder Reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the Thorium Molten Salt Reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the Molten Salt Reactor configurations that deserve further evaluation.Comment: 11 pages, 8 figures, 6 table

Impact of the MSBR concept technology on long-lived radio-toxicity and proliferation resistance

The MSBR (Molten Salt Breeder Reactor) was an industrial project designed at the beginning of the seventies at Oak Ridge National Laboratory and based on Thorium. Just before, the MSRE worked very well during four years with molten fuel. The MSBR system, where a maximum breeding was wanted, included a graphite moderated core with the circulation of a 71.7%LiF-16%BeF2-12%ThF4-0.3%UF4 salt and a pyro-chemical reprocessing unit. To obtain a maximum breeding ratio, Protactinium was extracted and stored allowing decay out of the neutron flux. This required the entire salt volume to be reprocessed in ten days, the gaseous fission products and Minor Actinides being extracted continuously by helium bubbling and pyro-chemical methods. The doubling time was evaluated to around 25 years. The project has since been re-evaluated especially within the frame of the EURATOM concerted Action MOST. To have an acceptable global reactivity feedback coefficient, studies have shown various possibilities based on core geometry, neutron moderation ratio and salt composition. When requiring only a breeding ratio of one, it is possible to avoid continuous reprocessing and to strongly simplify it. These various options will be discussed. The detailed inventory will be given showing clearly the interest of the Thorium Molten Salt Reactor where the production of Americium and Curium is a factor of one hundred lower that for the U-Pu RNR. The amount of Uranium 232 which is always produced in the Thorium cycle will be calculated as well as its decay rate since its decay chain eventually results in a 2.6 MeV γ –ray which may be used to detect and hence control the U233 fuel movements. As the U233 has to be produced in other reactors (PWR, RNR or other MSR), special cares have to be taken and will be discussed

Estimation of fiber diameters in the spinal dorsal columns from clinical data

Lack of human morphometric data regarding the largest nerve fibers in the dorsal columns (DCs) of the spinal cord has lead to the estimation of the diameters of these fibers from clinical data retrieved from patients with a new spinal cord stimulation (SCS) system. These patients indicated the perception threshold of stimulation induced paresthesia in various body segments, while the stimulation amplitude was increased. The fiber diameters were calculated with a computer model, developed to calculate the effects of SCS on spinal nerve fibers. This computer model consists of two parts: (1) a three-dimensional (3-D) volume conductor model of a spinal cord segment in which the potential distribution due to electrical stimulation is calculated and (2) an electrical equivalent cable model of myelinated nerve fiber, which uses the calculated potential field to determine the threshold stimulus needed for activation. It is shown that the largest fibers in the medial DCs are significantly smaller than the largest fibers in the lateral parts. This finding is in accordance with the fiber distribution in cat, derived from the corresponding propagation velocities. Moreover, it is shown that the mediolateral increase in fiber diameter is mainly confined to the lateral parts of the DCs. Implementation of this mediolateral fiber diameter distribution of the DCs in the computer model enables the prediction of the recruitment order of dermatomal paresthesias following increasing electrical stimulation amplitud

MURE : MCNP Utility for Reactor Evolution - Description of the methods, first applications and results

PACSInternational audienc

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