Fission Gas Release from Irradiated UO2+x During Transient Annealing. (3rd July 2019)

• Record Type: Journal Article
• Authors: Massih, A. R.
• Is Part Of: Nuclear technology. Volume 205:Number 7(2019)
• Page Start: 992
• Page End: 1001
• Publication Date: 2019-07-03
• Subjects: Fission gas release -- hyperstoichiometric uranium dioxide, modeling; Nuclear engineering -- Periodicals; Nuclear engineering; Nuclear Physics; Periodicals; Periodicals; 621.48
• DOI: 10.1080/00295450.2019.1568102
• Languages: English
• ISSNs: 1943-7471

Abstract

Abstract: Oxidation of UO2 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 UO2+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 UO2+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 UO2+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 UO2+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.