• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.1
• /
• pp.53-64
• /
• 2007

The dynamic release behavior of Cs from high burn-up spent PWR fuel was experimentally performed under the conditions of a thermal treatment process such as voloxidation and sintering conditions. In voloxidation process, influence of the oxidation and reduction atmosphere on the Cs release characteristic using fragment type of spent fuel heated up to $1,500^{\circ}C$ was compared. In sintering process, temperature history effect on Cs release behavior was evaluated using green pellet under 4% $H_2/Ar$ environment. Temperature range for complete Cs release from spent fuel fragment under voloxidation condition was about $800^{\circ}C{\sim}1,200^{\circ}C$, but that of green pellet under the reduction atmosphere was $1,100^{\circ}C{\sim}1,400^{\circ}C$. Key parameters on Cs release behavior from spent fuel was powder formation as well as the diffusion rate of Cs compound to grain boundary and fuel surface.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.385-386
• /
• 2005

• Journal of Energy Engineering
• /
• v.10 no.3
• /
• pp.278-285
• /
• 2001

In this study, the effects of the variables on sintering of simulated fuel to simulate the spent fuel are described. Mainly, the effects of compaction pressure, sintering temperature and time on the density of pellet are described. The experimental is performed with compaction pressure of 1 ton/$\textrm{cm}^2$~4 ton/$\textrm{cm}^2$, sintering temperature of 167$0^{\circ}C$, 173$0^{\circ}C$ and 178$0^{\circ}C$ and sintering time of 4 hr, 8 hr and 24 hr. The green density of simulated fuel is proportional to the one third power of compaction pressure and the sintered density is 90.5~99.6% of theoretical density. The grain growth exponent and activation energy of simulated fuel is 2.5 and 287.97 kJ/mol, respectively.

• Resources Recycling
• /
• v.11 no.4
• /
• pp.3-10
• /
• 2002

The characteristics of dry and wet milled powder prepared by 1 cycle OREOX (oxidation and reduction of oxide fuels) treatment were investigated using the simulated spent fuel pellet. Sintered pellets simulating spent nuclear fuel burned in reactor were fabricated from $UO_2$ powder using as a starting material in fabrication of nuclear fuel. The 1 cycle OREOX-treated powder was prepared by only one path of oxidation md reduction of the simulated pellet. Powder having average particle size of less than 1 $\mu\textrm{m}$ could be easily obtained by dry milling, but not be achieved by wet milling. And, specific surface area of dry milled pow-der was higher than that of wet milled powder. Dry milled powder formed loose agglomerate, while wet milled powder showed the shape of irregular and angular particles. Dry milled powder provided higher green density, resulting in higher sintered density of higher than 95% TD and average grain size of larger than 8 $\mu\textrm{m}$ satisfying the standard specification of sintered pellets.