Concept

Plutonium 240

Publications associées (16)

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Cermet sphere-pac concept for inert matrix fuel

Manuel Alexandre Pouchon

In the inert matrix fuel concept, plutonium reprocessed from spent fuel is burned in an inert matrix, e.g. yttria-stabilized zirconia. Coming from wet reprocessing, the internal gelation can perform an easy micro-spheres production. Utilization of these pa ...

Elsevier2003

Inert matrix fuel for the utilisation of plutonium

Manuel Alexandre Pouchon

Utilising fuel resources responsibly, reducing waste volume and emissions as well as conflict potentials within the international community (non-proliferation, energy demand) are among the principles for the judgment of sustainable development. Utilising a ...

Elsevier2001

Preparation of simulated inert matrix fuel with different powders by dry milling method

Manuel Alexandre Pouchon

Simulated zirconium oxide-based inert matrix fuel (IMF) has been prepared using cerium oxide replacing plutonium oxide, following the conventional pelletising route, in which an improved two-stage attrition mill was employed. Different methods have been ap ...

Elsevier1999

Comparison of pressurized water reactor core characteristics for 100% plutonium-containing loadings

Rakesh Chawla

The large quantities of reactor-grade (RG) and weapons-grade (WG) Pu accumulated worldwide could be reduced by employing 100% mixed-oxide (MOX) cores in light water reactors. The buildup of new Pu from the U present in the MOX, however remains disadvantage ...

American Nucl. Soc1998

Integral testing of 237Np cross-sections in GCFR spectra

Rakesh Chawla

Integral measurements of 237Np capture and fission rates have been carried out relative to the 239Pu fission rate in two GCFR lattices with different neutron spectra. The results have been compared with calculated values based, respectively, on ENDF/B-IV a ...

Elsevier1979

Determination of 238U capture rates in mixed-oxide fast reactor lattices

Rakesh Chawla

Measurements of 238U capture rates (C8), relative to 239Pu fission (F9), have been carried out in 25% and 16% PuO2/UO2 fuels in the fast reactor ZEBRA using Ge(Li) detectors. A number of experimental techniques have been developed and applied, with individ ...

1972