¹⁴C content in CANDU spent fuel cladings and its release under alkaline conditions

Abstract

The total 14C content and its partition between inorganic and organic species were measured on irradiated Zy-4 samples from a CANDU spent fuel rod transferred from Cernavoda Nuclear Power Plant (NPP). Long-term leaching tests and accelerated corrosion tests were carried out to measure the 14C release and corrosion rate, respectively, in chemical conditions relevant to cementitious environment. Experimentally measured 14C inventory was compared to the theoretically one predicted based on the irradiation history and impurity content of Zy-4 by means of ORIGEN computations. CANDU SF claddings have a 14C content of around 2 × 104 Bq/g of Zy-4, mainly as organic compounds (more than 99%). The total 14C content measured by acid dissolution/wet oxidation method is in good agreement with the value estimated by ORIGEN simulations for an average burn-up of 7 MWd/kgU. The total 14C released as dissolved species after 18 days and 18 months of Zy-4 immersing in alkaline solution are similar, indicating that a small amount of 14C was available as instant release fraction (0.05% from the initial 14C content) followed by a very low release rate that could not be measured by liquid scintillation counting. In alkaline conditions, 14C is released predominantly (∼70%) as soluble species, but also inorganic 14C was measured as gaseous species. From the soluble 14C released during leaching test, more than 60% was found to be as organic species. Generally, corrosion rates values ranging between 46 and 130 nm/yr were measured by the linear polarization resistance method. In addition, defects and cracks were observed on the oxide layer by scanning electron microscopy (SEM) investigation.

Overview OF14C release from irradiated zircaloys in geological disposal conditions

“…Most of the experiments were conducted on as-received unirradiated Zircaloy, which means that the available results concerned the pre-transition regime (cubic law) only, with an oxide thickness <<2.5 µm. This overview includes the work obtained on unirradiated preoxidized Zr-4 to explore the post-transition regime as well (Bucur et al 2017). The corrosion rates were determined by applying linear polarization measurements on both as-received and preoxidized Zr-4 (oxide thickness ~2.7 µm) in NaOH solution, pH 12.5.…”

14C Exposure from Disposal of Radioactive Waste Compared to14C Exposure from Cosmogenic Origin

“…OH radicals are produced by chemicals such as potassium persulphate (K 2 S 2 O 8 ) for the experimental determination of the organic 14 C content in samples such as leachate solutions of irradiated Zircaloy corrosion as proposed by Magnusson (e.g. Bucur 2017 and Sakuragi 2017), leachate solutions of irradiated graphite (e.g. Toulhoat 2018), acidic solutions with spent ion exchange resins (e.g.…”