## 1 Introduction  
Owing to their low neutron absorption, good mechanical strength, and corrosion resistance, Zr alloys are widely used in nuclear power reactors. However, neutron damage of zirconium-based core reactor components leads to in-reactor deformation such as irradiation growth, hardening, loss of ductility, and localised plastic instability [1]. The fundamental mechanism of these degradations is neutron induced atomic displacement in alloys. One of most important damage microstructures is called a \"dislocation loop\", which evolves from atomic displacement cascades [2]. The study of the dislocation loop evolution of Zr alloys under irradiation is of active interest. In Zr alloys, prismatic (<a>-type) dislocation loops are the basic and most well-known component of radiation damage structures and are observed in a wide range of irradiation temperatures (80-500 \({}^{\circ}\)C) [2]. These loops are visible at lower to higher neutron doses, with densities saturating at relatively low doses, and have been extensively studied by neutron, charged particle irradiation, and MD simulations [2; 3; 4; 5; 6; 7; 8]. It has been reported that the point-defect clusters formed in Zircaloy-2 at temperatures between 250 and 400 \({}^{\circ}\)C and for irradiation doses lower than 5 \(\times\) 10\({}^{25}\) nm\({}^{-2}\) (E > 1 MeV), which can be observed by using transmission electron microscopy (TEM) (>2 nm), and consist of perfect dislocation loops, either of vacancy or interstitial nature, with Burgers vector of \(\frac{1}{3}\langle 11\overline{2}0\rangle\), lying in the prismatic planes (\(\overline{1}010\)) and (\(11\overline{2}0\)) planes with typical diameters from 5 to 20 nm, depending on the fluence and irradiation temperature [9; 10]. Neutron irradiated microstructures reveal that the loop dimensions are smaller in Zr alloys compared to pure Zr irradiated under similar conditions (1 \(\times\) 10\({}^{24}\) n/m\({}^{2}\), E \(>\) 1 MeV, 400 \({}^{\circ}\)C) [11]. A number of studies on pure Zr and various Zr alloys of different compositions have been conducted in the past, either by using charged ion irradiation or fast neutron irradiation. The early study conducted by [PERSON] and [PERSON] [12] on pure Zr, Zircaloy, and Zr-Nb alloys focused on the temperature and solute concentration dependencies of prismatic loops using high voltage electron microscope (HVEM). It was concluded that the presence of solutes increases the density and decreases the growth rate of loops generated during electron irradiation by an order of magnitude, although in this case, the damage mechanism is significantly different from neutron irradiation [11]. Similarly, [PERSON] et al. [13] conducted in-situ 1 MeV electron irradiation experiments on pure Zr, Zr-1% Nb, and Zircaloy-4, at higher irradiation temperature (\(>\)400 \({}^{\circ}\)C). Loop size increased with increasing irradiation temperature, whereas higher prismatic loop densities in the presence of alloying elements (O, Nb, Sn) were attributed to the preferential nucleation of loops at the impurity atoms. Due to the nature of displacement cascades produced in Zr, irradiation by heavy ions is a reasonable simulation of neutron irradiation in reactors and has been the tool for experimental studies in several different materials. A Ni ion bombardment study of Zircaloy-2 with a wide range of variables (O content, 300-600 \({}^{\circ}\)C, 0.65-87 dpa) was conducted by [PERSON] et al. [14], where ex-situ ion irradiated microstructures were investigated and compared to neutron-irradiated microstructure. The dominant defects were difficult to individually identify due to the defect strain field overlap and presence of small hydrides induced during electro-polishing. Higher irradiation temperature seemed to cause an alignment of defects with the trace of the basal plane, which was attributed to the interaction of defect clusters with the solutes and to self-climb and/or prismatic glide.