Properties of low-lying states in some high-nuclearity Mn, Fe and V clusters: Exact studies of Heisenberg models

Using an efficient numerical scheme that exploits spatial symmetries and spin parity, we have obtained the exact low-lying eigenstates of exchange Hamiltonians for the high nuclearity spin clusters, Mn_{12}, Fe_8 and V_{15}. The largest calculation involves the Mn_{12} cluster which spans a Fock space of a hundred million. Our results show that the earlier estimates of the exchange constants need to be revised for the Mn_{12} cluster to explain the level ordering of low-lying eigenstates. In the case of the Fe_8 cluster, correct level ordering can be obtained which is consistent with the exchange constants for the already known clusters with butterfly structure. In the V_{15} cluster, we obtain an effective Hamiltonian that reproduces exactly, the eight low-lying eigenvalues of the full Hamiltonian.Comment: Revtex, 12 pages, 16 eps figures; this is the final published versio

Oxide phase characterization in simulated high burn-up UO2 fuels in the early stages of a nuclear severe accident

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Characterization of the S = 9 excited state in Fe8Br8 by Electron Paramagnetic Resonance

High Frequency electron paramagnetic resonance has been used to observe the magnetic dipole, $\Delta$ M$_s$ = $\pm$ 1, transitions in the $S = 9$ excited state of the single molecule magnet Fe$_8$Br$_8$. A Boltzmann analysis of the measured intensities locates it at 24 $\pm$ 2 K above the $S = 10$ ground state, while the line positions yield its magnetic parameters D = -0.27 K, E = $\pm$0.05 K, and B$_4^0$ = -1.3$\times$ 10$^{-6}$ K. D is thus smaller by 8% and E larger by 7% than for $S = 10$. The anisotropy barrier for $S = 9$ is estimated as 22 K,which is 25% smaller than that for $S = 10$ (29 K). These data also help assign the spin exchange constants(J's) and thus provide a basis for improved electronic structure calculations of Fe$_8$Br$_8$.Comment: 7 pages, Figs included in text, submitted to PR

Modelling nuclear fuel behaviour with TAF-ID: Calculations on the VERDON-1 experiment, representative of a nuclear severe accident

International audienceThe chemical behaviour of the main elements present in a PWR irradiated fuel sample (UO$_2$ fuel, Zr from the Zircaloy-cladding, Pu, Np, and fission products such as Ba, Ce, Cs, Gd, I, La, Mo, Nd, Pd, Rh, Ru, Sr, Tc, Te, etc.) submitted to a nuclear severe accident type sequence (VERDON-1 experiment) has been investigated by thermodynamic calculations using the TAF-ID database. Particular emphasis has been placed on the chemical behaviour of the fission products Ba, Cs, Mo, and Zr, and the interactions between the Zircaloy-cladding and the UO$_2$ matrix. Calculation results have been compared to experimental observations completed during the VERDON-1 experiment (instantaneous release of fission products and final chemical state of elements in the sample). Results presented in this manuscript do not account for non-equilibrium phenomena that have a major impact on fission product behaviour (e.g., chemical diffusion, mass transport induced by temperature gradients in the fuel pellet, etc.). None the less, these TAF-ID calculations have described accurately the melting of the sample following the formation of a mixed (U,Zr)O$_2$-x phase because of the interaction between the fuel and the Zircaloy-cladding. Furthermore, calculations have assisted in the explanation of the instantaneous released fractions observed for Ba and Mo during the experiment (as a function of the experiment atmosphere and temperature)

A laser-based system to heat nuclear fuel pellets at high temperature

International audienceAnnealing tests are of utmost importance in nuclear fuel research, particularly to study the thermophysical properties of the material, microstructure evolution, or the released gas as a function of temperature. As an alternative to conventional furnace or induction annealing, we report on a laser-heating experiment allowing one to heat a nuclear fuel pellet made of uranium dioxide, UO2, or potentially other nuclear fuel pellets in an isothermal and controlled manner. For that purpose, we propose to use an indirect heating method based on a two compartment tungsten crucible, one containing the sample and the other acting as a laser susceptor for efficient and homogeneous heating of the assembly. With this concept, we demonstrate the heating of UO2 samples up to 1500 °C at a maximum heating rate of 30 °C/s with the use of two 500 W lasers. The system is, however, scalable to higher heating rates or higher temperatures by increasing the laser power up to few kW. The experiment has been designed to heat a pressurized water reactor fuel pellet, but the concept could be easily applied to other sample geometries or materials

Implementation of a new gamma spectrometer on the MERARG loop Application to the volatile fission products release measurement.

International audienceThe MERARG facility initially aims at the annealing of irradiated fuel samples to study the gaseous fission products release kinetics. In order to complete the evaluation of the source term potentially released during accidental situation, the MERARG experimental circuit has been enhanced with a new gamma spectrometer. This one is directly sighting the fuel and is devoted to the fission products release kinetics. Because of the specificities of the fuel measurements, it has been dimensioned and designed to match the specific requirements. The acquisition chain and the collimation system have been optimized for this purpose and a first set of two experiments have shown the good functioning of this new spectrometry facility

Coupled modeling of irradiated fuel thermochemistry and gas diffusion during severe accidents

In this paper, a novel approach where irradiated fuel thermochemistry and gas release are coupled is presented in details and illustrated by the simulations of some tests of the VERCORS program characterized by increasing temperatures and varying gas composition in the furnace (oxidizing or reducing conditions). At each step of the tests, the oxidation/reduction of the nuclear fuel and the fission product chemical speciation are precisely assessed thanks to a thermochemical equilibrium calculation relying on the OpenCalphad thermochemical solver and on a built-in thermochemical database derived from the SGTE database and completed by a solid solution model for the U-O-fission product system. Fission product releases are estimated from the chemically reactive gases that form in the fuel (according to the thermochemical calculation) and from a gas diffusion model based on the equivalent sphere model. The gas diffusion model takes into account not only the noble gases available in the fuel prior to the test but also the chemically reactive gases that form during the test. It is shown that the proposed coupled approach provides a consistent estimation of fission product release (I, Te, Cs, Mo, Ba) during the VERCORS tests in spite of the simple gas diffusion mechanism considered in the simulations (no distinction between the fission products). The proposed coupled approach is used to test some thermochemical hypotheses to improve the calculated release of some fission products (Ba, Mo)

Modeling high burnup fuel thermochemistry, fission product release and fuel melting during the VERDON 1 and RT6 tests

This paper presents simulations of the VERDON 1 and RT6 tests (temperature increase up to fuel-clad melting, oxidizing and/or reducing conditions within the furnace) performed with high burnup UO 2 fuel (i.e., up to 72 GWd/tU) and considering a coupling between irradiated fuel thermochemistry and a fission gas release model. The thermochemical calculations rely on the Thermodynamics of Advanced Fuels-International Database (TAF-ID) for the description of the phases likely to form from the 15 fission product considered in the fuel (Ba