Fission Gas Release from Irradiated UO2+x During Transient Annealing

• Published in: Nuclear technology 2019-07, Vol.205 (7), p.992-1001
• Main Author: Massih, A. R.
• Format: Article
• Language: English
• Subjects: Fission gas release; hyperstoichiometric uranium dioxide, modeling
• ISSN: 0029-5450; 1943-7471
• DOI: 10.1080/00295450.2019.1568102

Oxidation of UO 2 fuel under off-normal and normal reactor conditions occurs when fuel cladding fails, thereby allowing steam/water to enter the fuel rod. The steam/water will react with the fuel to produce UO 2+x thus releasing hydrogen, with x standing for the amount of interstitial oxygen ions above the stoichiometric value. In this technical note the impact of fuel oxidation on fission gas release (FGR) is modeled and discussed. The classical diffusion equation is used to describe migration and release of fission product gas (Xe) in UO 2+x under time-varying postirradiation annealing conditions. We assume that the main quantity affected by fuel oxidation is the effective diffusivity of fission gas. Fuel oxidation enhances the diffusivity as a function of x in UO 2+x in a parabolic fashion for 0.005 ≤ x ≤ 0.12 in the temperature range of 1000 ≤ T ≤ 1600 K. We first benchmark our model against an annealing test in which for x = 0.004 the Xe release fraction was measured as a function of time (temperature) during annealing. Furthermore, we apply the model to simulate a series of postirradiation annealing tests on UO 2+x fuel, in which FGR fractions were measured for a given thermal ramp history in the range 0.00 ≤ x ≤ 0.66. The results of our computations in the range 0.004 ≤ x ≤ 0.20 show good agreement with measurements.