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

Le Thorium Molten Salt Reactor : Au-delà du MSBR

La re-évaluation du concept de Molten Salt Breeder Reactor a fait apparaître des problèmes liés à la sûreté et à la complexité du retraitement. Une nouvelle réflexion est menée afin de trouver des solutions et ainsi d'aboutir au concept du Thorium Molten Salt Reactor. Plusieurs contraintes principales sont établies et vont servir de guides aux études paramétriques. Celles-ci permettent alors de comprendre l'influence de paramètres importants du coeur sur le comportement du réacteur. Le but de cet article est de présenter ce vaste domaine de recherche et d'indiquer quelles configurations intéressantes de Réacteurs à Sels Fondus peuvent être étudiées plus avant

Molten salt reactors and possible scenarios for future nuclear power deployment

PAC

DEPHI : un analyseur de fonctionnement pour Superphénix

Intermédiaire entre le code de calcul et le simulateur, l'analyseur est un outil d'utilisation simple pour l'étude du fonctionnement normal, incidentel ou accidentel d'une installation. Elaboré par EDF à la demande de la NERSA, société responsable de la réalisation du surgénérateur Superphénix, l'analyseur DEPHI permet d'explorer le fonctionnement de cette centrale. Les auteurs en décrivent les caractéristiques et présentent un exemple concret d'utilisation

Physique des réacteurs et transmutation

La transmutation des transuraniens par recyclage en réacteur permet de ramener en quelques siècles de refroidissement la radio-toxicité des déchets de haute activité à vie longue mis au stockage au niveau de celle de l'uranium naturel ayant servi à produire la même quantité d'énergie en cycle ouvert. Les principales difficultés pour sa mise en oeuvre découlent des propriétés neutroniques et physiques des actinides mineurs, aboutissant à privilégier les réacteurs à spectre rapide, avec toutefois de gros problèmes de faisabilité du combustible pour les options d'incinération "hétérogène" où les actinides mineurs sont chargés à part dans des cibles spécifiques placées dans les réacteurs électrogènes, ou dans une double strate de réacteurs sous-critiques dédiés. Quant aux produits de fission à vie longue, leur transmutation ne semble pas, en l'état actuel des études, pouvoir présenter une option de gestion suffisamment efficace

Analysis of uncertainty propagation in scenario studies Surrogate models application to the French historical PWR fleet

International audienceNuclear scenario studies simulate the whole fuel cycle over a period of time, from extraction of natural resources to geological storage. They enable the comparison of different strategies related to the reactor fleet evolution, fuel cycle materials management, etc. based on criteria such as the installed capacity per reactor technology, mass inventories and flows, in the fuel cycle as well as in the waste. Several sources of uncertainty have an impact on the scenario results, such as nuclear data and industrial parameters. Nuclear data uncertainties propagate in the scenario along the isotopic chains through depletion, cooling and fuel equivalence models, while industrial parameters impact directly the fuel cycle facilities, such as the plutonium and minor actinides recovery rates at the reprocessing plant or the spent fuels burnup. A method dedicated to uncertainty propagation in scenario studies based on a Monte-Carlo approach and surrogate models was developed. In the present study, the uncertainty propagation methodology is applied to the French historical PWR fleet, up to 2010

HELIOS: Irradidation of U-Free Fuels and Targets for Americium Transmutation

Americium is one of the radioactive elements that contributes to a large part of the radiotoxicity of nuclear spent fuel. Transmutation by irradiation in nuclear reactors of long-lived nuclides like 241Am is, therefore, an option for the reduction of the mass and radiotoxicity of nuclear waste. The analysis of previous irradiation experiments which were carried out with targets of MgAl2O4+11wt 241Am showed that the release/trapping of helium is the key issue for target design. In fact, in those experiments a significant volume swelling was observed which was partly attributed to the production of helium, which is characteristic for 241Am transmutation. These findings led to the conclusion, that a new experiment should be designed in such a way that helium shall be released from the target already during irradiation. Such release of helium might be achieved either with a fuel temperature kept sufficiently high during the whole irradiation or with release paths for helium created by inclusion of tailored open porosity in the targets. A new irradiation experiment called HELIOS is currently being designed and will be carried out in the High Flux Reactor (HFR) in Petten (The Netherlands) in the frame of the 4-year project EUROTRANS of the EURATOM 6th Framework Programme (FP6). The main objective of the HELIOS irradiation is to study the in-pile behaviour of U-free fuels and targets such as CerCer (Pu, Am, Zr)O2 and Am2Zr2O7 MgO or CerMet (Pu, Am)O2 Mo in order to gain knowledge on the role of the microstructure and of the temperature on the gas release and on fuel swelling. The irradiation temperature will be high enough to be able to tune the release of a significant fraction of helium produced so that the material swelling can be minimised as much as reasonably possible. Besides, the irradiation duration has been chosen as a compromise such to ensure that the central temperature in the (Pu,Am,Zr,Y)O2 pellets be always higher than that of the (Am,Zr,Y)O2 pellets in order to be able to investigate, during the Post-Irradiation Examinations (PIE), the influence of the higher irradiation temperature on the helium release. The HELIOS irradiation experiment is planned to be carried out in the HFR core and shall last 300 full power days starting in the first quarter of 2007. The proposed irradiation position is a high flux position of the HFR core, which has a thermal flux of about 1 1018 m-2s-1 and a total neutron flux of about 6 1018 m-2s-1. The use of a high flux position is required in order to transmute a substantial fraction of the 241Am, within the planned duration of the HFR irradiation. In the present paper the fabrication procedure and the development of the HELIOS irradiation, its rationales and objectives are described and discussed.JRC.F.3-High Flux and Future Reactor

Radioactive waste inventories in the case of different nuclearoptions for the French reactor fleet

International audienc