Search results for "MOX fuel"
showing 6 items of 6 documents
New measurement of the 242Pu(n,γ) cross section at n_TOF
2016
The use of MOX fuel (mixed-oxide fuel made of UO2 and PuO2 ) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. With the use of such new fuel composition rich in Pu, a better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United States (ENDF) nuclear data agencies. For the case of 242 Pu, the two only neutron capture time-of-flight measurements available, from 1973 and 1976, are not consistent with each other, which calls for a new time-of flight captu…
Stable anchoring of dispersed gold nanoparticles on hierarchic porous silica-based materials
2010
The nanometric organization of MOx (M = Co, Zn, Ni) domains partially embedded inside the mesoporous silica walls but accessible to the pore voids, which is achieved through a simple one-pot surfactant-assisted procedure, define optimal anchors for the nucleation and growth of gold nanoparticles, which in turn favours an exceptional thermal stability for the final Au-supported materials. As silica support we have selected a UVM-7 silica having a highly accessible architecture defined by two hierarchic pore systems. The combination of nanometric pore length, tortuous mesopores and MOx inorganic anchors favours the stability of the final Au/CoOx-UVM-7 nanocomposites.
Helium Behavior in Oxide Nuclear Fuels: First Principles Modeling
2010
UO2 and (U,Pu)O2 solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein. We present density functional theory (DFT) calculations for UO2, PuO2 and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model …
New measurement of the 242Pu(n,γ) cross section at n-TOF-EAR1 for MOX fuels: Preliminary results in the RRR
2016
The spent fuel of current nuclear reactors contains fissile plutonium isotopes that can be combined with 238U to make mixed oxide (MOX) fuel. In this way the Pu from spent fuel is used in a new reactor cycle, contributing to the long-term sustainability of nuclear energy. The use of MOX fuels in thermal and fast reactors requires accurate capture and fission cross sections. For the particular case of 242Pu, the previous neutron capture cross section measurements were made in the 70’s, providing an uncertainty of about 35% in the keV region. In this context, the Nuclear Energy Agency recommends in its “High Priority Request List” and its report WPEC-26 that the capture cross section of 242Pu…
Radiative neutron capture on Pu242 in the resonance region at the CERN n_TOF-EAR1 facility
2018
The spent fuel of current nuclear reactors contains fissile plutonium isotopes that can be combined with uranium to make mixed oxide (MOX) fuel. In this way the Pu from spent fuel is used in a new reactor cycle, contributing to the long-term sustainability of nuclear energy. However, an extensive use of MOX fuels, in particular in fast reactors, requires more accurate capture and fission cross sections for some Pu isotopes. In the case of Pu242 there are sizable discrepancies among the existing capture cross-section measurements included in the evaluations (all from the 1970s) resulting in an uncertainty as high as 35% in the fast energy region. Moreover, postirradiation experiments evaluat…
Validation of a method for neutron dosimetry and spectrometry using neutron activation of metal discs
2009
A technique for neutron dosimetry and spectrometry based on neutron activation of different metal discs has been studied. After exposure to a neutron field, the radionuclides produced in the discs are detected using low-level gamma-ray spectrometry and the neutron spectrum is obtained using a spectrum unfolding technique. In order to validate the method, irradiation was performed in a well-characterised (252)Cf neutron reference field. Furthermore, the detector was used to determine the neutron fluence rate and spectrum at a storage place for MOX nuclear fuel. The results of the two measurements are reported and discussed.