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Vol.6 , No. 2, Publication Date: May 30, 2019, Page: 12-17
 exhibits exceptional mechanical, chemical, and electrical properties for use as component for Gas-cooled Fast Reactor (GFR) fuel. This work improved current understanding of processing and thermophysical properties of zirconium nitride, also neutronic performance of (U-Pu)N fuels. A newly developed external gelation produced zirconiumoxynitride pellets, and their structure and growth was characterized by SEM and XRD. XRD analysis of the ZrN showed the samples had formed highly crystalline solid solutions during sintering. The resulting thermal conductivity of ZrN<i>x</i> and Zr-biphasic material would still meet the melting temperature requirements of GFR fuel materials of 2200 K. The differences in the properties observed were due to the different gelation mechanisms involved and the physical form of the matrix produced. The neutronic performance of the (U-Pu)O<sub>2</sub> and (U-Pu)N fuels (80 wt.% natural uranium and 20 wt.% plutonium reactor grade) has been studied using Monte Carlo N–Particle Transport Code System (MCNPX 2.7.0 code). Due to the higher density of heavy metals in (U-Pu)N fuel than in (U-Pu)O<sub>2</sub> fuel, (U-Pu)N, this fuel achieved longer irradiation times.&picture=http://www.aascit.org/aascit/decorators/img/journal/905.jpg)
| [1] | Osama Farid, Reactor Department, Atomic Energy Authority, Cairo, Egypt. |
| [2] | Nader Mohamed, Reactor Department, Atomic Energy Authority, Cairo, Egypt. |
Materials used in fuel elements in next generation of nuclear power are expected to be more efficient than previous materials and are also expected to be subject to harsher thermal and radiation environments. Zirconium nitride (ZrN) exhibits exceptional mechanical, chemical, and electrical properties for use as component for Gas-cooled Fast Reactor (GFR) fuel. This work improved current understanding of processing and thermophysical properties of zirconium nitride, also neutronic performance of (U-Pu)N fuels. A newly developed external gelation produced zirconiumoxynitride pellets, and their structure and growth was characterized by SEM and XRD. XRD analysis of the ZrN showed the samples had formed highly crystalline solid solutions during sintering. The resulting thermal conductivity of ZrNx and Zr-biphasic material would still meet the melting temperature requirements of GFR fuel materials of 2200 K. The differences in the properties observed were due to the different gelation mechanisms involved and the physical form of the matrix produced. The neutronic performance of the (U-Pu)O2 and (U-Pu)N fuels (80 wt.% natural uranium and 20 wt.% plutonium reactor grade) has been studied using Monte Carlo N–Particle Transport Code System (MCNPX 2.7.0 code). Due to the higher density of heavy metals in (U-Pu)N fuel than in (U-Pu)O2 fuel, (U-Pu)N, this fuel achieved longer irradiation times.
Keywords
Gas-Cooled Fast Reactor, Non-oxide Ceramics, Zirconium Nitride Fabrication, Nuclear Fuels Microstructure, MCNPX 2.7.0 Code
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