Simulating Distribution of Carbon in Spent Fuel Cladding of Zirconium Alloys

Spent fuel cladding made of zirconium (Zr) alloys would be mechanically compressed as cladding hulls in Japan. In view of waste disposal of the hulls, distribution of radioactive carbon-14 (C-14) produced in a Zr matrix is one of important factors to estimate release behavior of C-14 accompanied with oxidation of the Zr matrix [1]. Although C-14 should be distributed in the Zr cladding matrix by diffusion and precipitation under gradients of concentrations and temperatures during operation of nuclear reactors, it would be potentially relocated by isothermal annealing during a mid-term storage becoming longer before processing of the hulls.<br/><br/>In the present study, diffusion, solution and thermo-transport data for carbon (C) as well as hydrogen (H), nitrogen (N) and oxygen (O) in Zr have been revisited and diffusion data of C in oxygen dissolved Zr (Zr(O)) have been experimentally obtained. Using the transport parameters for C in Zr and Zr(O), distributions of C in the Zr matrix have been simulated under various temperature conditions and periods using numerical calculation of the one-dimensional diffusion equation.<br/><br/>To date, no data of heat of transport, Q*, for C both in αZr and βZr have been available whereas those for N and O in βZr were reported as Q*=+0.925 and +1.090, respectively [2]. Assuming Q* for C in Zr is positive, initially uniform concentration of C in a wall thickness (0.5 mm) along a radial direction of the Zr cladding tube and diffusion of C for 3 years under the temperature gradient from 573 K to 773 K, C tends to segregate at the cooler side and deplete at the hotter side by thermo-transport but diffuse to the opposite direction with a driving force to minimize a concentration gradient, resulting in formation of a maximum concentration of C in a middle of the wall thickness at around 0.3 mm from the cooler side. Such the concentration non-uniformity disappears during isothermal annealing at 773 K for more than 10 years. In the presentation, effects of dissolved N and O on distribution of C in the Zr matrix will be discussed.<br/><br/>References<br/>[1] T. Sakuragi, Y. Yamashita, M. Akagi, R. Takahashi, Carbon 14 Distribution in Irradiated BWR Fuel Cladding and Released Carbon 14 after Aqueous Immersion of 6.5 Years, Procedia Chemistry, 21 (2016) 341-348.<br/>[2] Vogel, D. L., Thermotransport of oxygen and nitrogen in beta-zirconium, beta-titanium, niobium and tantalum, Technische Hogeschool Eindhoven (1969).<br/><br/>This study was carried out as a part of R&D supporting program titled “Advanced technology development for geological disposal of TRU waste (2021 FY)” under the contract with Ministry of Economy, Trade and Industry (METI) (Grant Number: JPJ007597).