Irradiation Effects on ZR-2.5NB in Power Reactors

Abstract

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, “High Power Channel-type Reactor”) reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr-2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge.

Phase stability and precipitation modeling in neutron irradiated Zr–2% Nb alloy

“…Values of estimated precipitate density and the mean radius of radiation induced precipitates at quasi-stationary regime are: N * p ≈ 10 14 cm −3 , R * p ≈ 12 nm at T = 550 K, K = 10 −6 dpa s −1 ; N * p ≈ 4.1 × 10 13 cm −3 , R * p ≈ 13 nm at T = 550 K, K = 5 × 10 −7 dpa s −1 ; N * p ≈ 7 × 10 13 cm −3 , R * p ≈ 14 nm at T = 570 K, K = 10 −6 dpa s −1 . Obtained values relate well to experimental observations discussed in references [1,8,10]. Moreover, obtained results relate well to experimental data from OSIRIS reactor (CEA, Saclay, France) where an increase in the minimal precipitate size from 6 nm up to 18 nm at doses 18 dpa with densities ≈10 14 cm −3 was observed [9].…”

“…Therefore, the total dynamics of the system is described by a set of equations ( 5)- (10). The following normalizations for numerical calculations are used: r = r/ , t = tD v / 2 , where is the interface width.…”

“…By studying Zr-based alloys (used as cladding materials) it was found that the during radiation, the flux of point defects accelerates diffusion of doping (Nb with (1-3)at%, and Sn, Fe, Cr, Cu with less than 1 at%) resulting in formation of myriads of new precipitates [3][4][5][6][7]. Detailed experimental and numerical studies have shown that the size of β-precipitates (of body centered cubic structure with >80% of Nb content) emerged in α-Zr matrix (of hexagonal close packed structure) after incubation dose around 1 dpa (displacement per atom) [8] and their density depends on irradiation conditions [1,9,10], whilst large spherical precipitates are not strongly affected by irradiation [11]. Nonequilibrium defects produced at irradiation migrate by forces governed by both chemical potential gradients and elastic interactions resulting in radiation induces segregation, precipitation and dislocation loops formation affecting the material performance in the reactors [1,2,10].…”

Solid-State Phase Transformation Explains the Mixed Crystallographic Character of Zr(Nb,Fe)2 Nanoprecipitates in Zr-2.5Nb

“…Zirconium alloys are widely used in most commercial nuclear reactors worldwide. − The presence of alloying elements and impurities like Nb, Fe, and Cr in these alloys typically results in the formation of intermetallic nanoprecipitates (NPs) under processing heat treatments. ,,, The presence of NPs in Zr alloys affects the mechanical properties of the material through their interaction with dislocations . Moreover, these impurities, alloying elements, and the intermetallic NPs they form impact the corrosion resistance of some zirconium alloys; this effect appears to be dependent on the NP size and distribution. − …”

Fundamental Mechanisms for Irradiation-Hardening and Embrittlement: A Review

“…The study of irradiation-hardening and embrittlement is non-trivial as it intrinsically denotes a typical multi-scale problem [16][17][18][19]. Under exposure to high-energy particles, lattice atoms in the pristine materials are displaced from their original sites, which further interact with the surrounding matrix that result in the formation of the so-called collision cascade, and after its cooling, a number of point defects such as interstitials and vacancies are recombined.…”