Modelling the long-term dynamics of the energy transition accounting for socioeconomic behaviour and biophysical constraints: overview of the Wiliam Energy Module

WILIAM (Within Limit Integrated Assessment Model) is a global multiregional IAM that combines economic, social, demographic, environmental, energy and material related aspects into one system dynamics model. It aims to provide stakeholders with an open source, welldocumented model to assess the feasibility, effectiveness, costs and impacts of different sustainability policy options. The adequate representation of energy production is key to assess future sustainability pathways. The main function of the developed energy module is to estimate the primary energy requirements and related GHG emissions for satisfying the economic demand. This goal was achieved by 7 major sub-modules: (1) End-use: translates the economic demand into final energy demand through a hybrid approach combining bottom-up with energy intensities for different sectors. (2) Energy transformation: maps the entire energy conversion chain from final to primary energy, including intermediary energy commodities and an allocation function for power plant utilization. (3) Energy capacity: keeps track of the current power plant capacity stock, decommissioning of expired capacities, as well as the build-up of new capacities. An allocation function for choosing the suitable technology types for new capacities stands at the core of this sub-module. (4) Computation of the EROI of green technologies (5) Variability and storage: keeps track of sub-annual time scale effects on annual energy balances depending on the current power system setup (DSM, Storage, sector coupling). (6) Consideration of techno-sustainable potentials of RES considering geographical, resource and Energy Return on Energy Investment (EROI) constraints. (7) Computation of the energy-related GHG emissions

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

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

Direct sensitivity analysis to nuclear data of thorium molten salt reactors at equilibrium using MURE

Proc on lineInternational audienceSustainability and deployment potential of GEN IV reactors is linked to their breeding capacity. Therefore, nuclear data uncertainties should be assessed carefully so as to demonstrate the confidence that we have on the simulation of this key parameter. Breeding fissile material requires the reprocessing of the fuel. Then the actinide content after refueling depends on previous fuel evolutions and so does the sensitivity of any observable linked to these nuclides. Furthermore, keeping the reactor critical at all time can imply strong feedback on the sensitivity calculations too. First results using analytical tools and the equilibrium approach have shown that very strong error compensation may occur because of reprocessing. This means that breeding sensitivities cannot be based on static calculations or one stage evolution calculations. In this paper we propose to use the MURE package to confirm our previous results. Our test case is a Thorium Molten Salt Reactor. It is particularly important to do uncertainty calculations for this reactor concept as the breeding is more difficult to obtain with thorium when nuclear data is known not to be as much validated as for uranium. We will recalculate the fuel equilibrium after modifications of the ACE files for the reactions that are expected to have the strongest contribution to the uncertainty: 233U capture cross section and neutron fission yield

Direct sensitivity analysis to nuclear data of thorium molten salt reactors at equilibrium using MURE

In this paper we propose to use the MURE (MCNP Utility for Reactor Evolution) package to calculate directly nuclear data sensitivities of the breeding gain of a thorium molten salt reactor. The continuous fuel reprocessing is used to control reactivity. This control coupled with the fact that the fuel has reached its equilibrium, induces feedback effects on nuclear data sensitivities. That is why sensitivities are calculated directly by recalculating the fuel equilibrium of a simplified model after modifications of the ACE files for the reactions that are expected to have the strongest contribution to the uncertainty: 233U capture cross section and neutron fission yield

Measurement of mass yields from the 241Am(2nth,f) reaction at the Lohengrin spectrometer

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Measurement of fission yields from the

The study of fission yields has a major impact on the characterization and understanding of the fission process and is mandatory for reactor applications. While the yields are known for the major actinides (235U, 239Pu) in the thermal neutron-induced fission, only few measurements have been performed on 242Am. This paper presents the results of a measurement at the Lohengrin mass spectrometer (ILL, France) on the reaction 241Am(2nth,f): a total of 41 mass yields in the light and the heavy peaks have been measured and compared with the fission process simulation code GEF. Modus operandi and first results of a second experiment performed in May 2013 on the same reaction but with the goal of extracting the isotopic yields are presented as well: 8 mass yields were re-measured and 18 isotopic yields have been investigated and are being analyzed. Results concerning the kinetic energy and its comparison with the GEF Code are also presented in this paper

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

Measurements of the mass and isotopic yields of the 233U(nth,f) reaction at the Lohengrin spectrometer

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