Temelin Irradiated Cladding Project – TIRCLAD. Issue 1 (August 2021)

• Record Type: Journal Article
• Authors: Linhart, Stanislav; Bĕláč, Josef; Ševeček, Martin; Řeháček, Radomír; Starý, Vladislav
• Is Part Of: IOP conference series. Volume 1178:Issue 1(2021)
• Publication Date: 2021-08
• Subjects: fuel cladding -- irradiation -- fuel performance -- post-irradiation examination -- FEM models; Materials science -- Periodicals; 620.1105
• DOI: 10.1088/1757-899X/1178/1/012041
• Languages: English
• ISSNs: 1757-8981

Abstract

Abstract: It is widely known, that zirconium-based alloys have been used for nuclear fuel cladding fabrication for decades [1], [2]. Therefore, it is crucial for fuel vendors, utilities as well as safety authorities to understand their behaviour during reactor operation with strong fast neutron and gamma irradiation. To complement the information about material properties for current as well as prospective advanced zirconium-based alloys, a long term research project implemented by ALVEL has been carried out, in cooperation with its partners (ČEZ, Škoda Nuclear Machinery, UJV Rez, Research Centre Rez) [3], [4]. The project is implemented on the background of a large material research program sponsored by JSC TVEL and ČEZ (this program started in 2012 and it is planned to be finished in 2024). It is fully independently implemented in the Czech Republic. Irradiation of material samples started at Temelin NPP in 2014 and finished in spring 2020. The first three batches were already transported to UJV Rez and Research Centre Rez for post-irradiation evaluation. After finishing testing of reference non-irradiated samples testing in 2018 (determination of initial state, mechanical properties and as-received microstructure), six batches of irradiated material have been analysed sequentially with their increasing neutron fluence. The final irradiation damage of the last sixth batch corresponds approximately to fuel burnup of 80 MWd/kgU. This paper presents the developed methodologies and the expected experimental and other important outcome. Several material properties are being studied on different scales including high resolution microstructural and chemical analysis, creep and mechanical behaviour, nanoindentation, oxide layer morphology. The main objectives of the project include evaluation of materials' microstructural and bulk properties and the derivation of dose- and temperature-dependent correlations that can be implemented into FEM models and fuel performance codes such as FRAPCON/FRAPTRAN [5], [6] or Transuranus [7]. These models can be then used to support licensing of new designs influenced by fuel cladding behaviour during long term dry storage, justification of dry storage for an extended period of time or justification of new safety criteria that can lead to enhanced operation of VVER reactors and spent nuclear fuel storage [8].