Irradiation Hardening and Microstructure Characterization of Zr -1% Nb During Low Dose Neutron Irradiation

Due to low neutron absorption cross section, high mechanical strength, high thermal conductivity and good corrosion resistance in water and steam, Zirconium alloys are widely used as fuel cladding material in nuclear reactors. During life-time of a reactor the microstructure of this alloy is affected due to, among other factors, radiation damage and hydrogen damage. In this work mechanical properties changes on neutron irradiated Zr-1wt.% Nb at low temperatures (< 100 °C) and low dose (3.5 ´ 1023 n m-2 (E > 1 MeV)) were correlated with hydrides and crystal defects evolution during irradiation. To achieve this propose, tensile tests of: 1) Non-hydrided and non-irradiated material, 2) Hydrided and non-irradiated material and 3) Hydrided and irradiated material were performed at 25 ºC and 300 ºC. Different phases, hydrides and second phase precipitates were characterized by transmission electron microscopy (TEM) techniques. For the hydrided and irradiated material, the ductility decreased sharply with respect to the hydrided and non-irradiated material, among other factors, due to the change in the microstructure produced mainly by neutron irradiation. Even if the presence of the hydride ζ (zeta) was observed, both in the irradiated and non-irradiated material, tensile tests showed that ζ-hydrides did not affect ductility, since hydrided samples are more ductile than non-hydrided samples.


Introduction
Zirconium alloys are commonly used as structural material of components in nuclear industry, due to their excellent properties such as low thermal neutron absorption cross section, corrosion, and creep resistances in pressure environments at reactor operating temperatures [1][2][3].
One of the effects of irradiation which mostly affects mechanical behavior is irradiation hardening. Since fast neutron irradiation leads to growth of dislocation loops, hardening, loss of ductility, growth, and creep. [4][5][6][7][8], it is very important to determine the degree of embrittlement of irradiated components. In addition, it is necessary to establish how radiation affects the phases present in the precipitated material during and after irradiation [9].
Another effect to consider is that produced by hydrogen (H). When Zr-based alloys are used in PWR (pressurized water reactors) or PHWR (pressurized heavy water reactors) H is released as a by-product of the hydrolytic reaction between cooling water and Zr. When hydrogen is absorbed into the Zr matrix and exceeds its solid solubility at reactor temperature, zirconium hydrides precipitate in the matrix, affecting physical and mechanical properties of the alloy. These effects have been extensively studied in the last years [10,11] including the embrittlement due to hydride-assisted cracking [12,13].
Under irradiation, all these phases undergo microstructural and microchemical modifications that could affect their in-service performance [16][17][18]. Thus, there is interest to analyse the behavior of Zr-1wt% Nb alloy in a radioactive environment that combines ζ hydride's nucleation, hydrogen in solution and defects created by irradiation [19].
In this work, the interaction between hydrides, second phase particles and radiation at 25 ºC y 300 ºC in Zr-1wt.% Nb has been analyzed. A group of samples of this material was hydrided and irradiated in the CNEA-RA3 nuclear reactor. They were placed in a capsule located in one of the reactor irradiations channels. The neutron fluence was 3.5 × 10 23 n m -2 after an irradiation of 30 days. Microstructural analysis was performed by transmission electron microscopy techniques.

Experimental procedure
Materials were provided by the company Teledyne Wah Chang Albany, as 1 mm thickness straps. The straps were rolled in order to obtain a final thickness of 0.5 mm. Tensile test specimens were then prepared by cutting with a numerical control machine down to a size small enough in order to minimize the dose of activity after irradiation.
The alloy composition can be observed in Table 1. All specimens were annealed at 450 ºC for 24 h and distributed in three groups: 1) non-hydrided and non-irradiated material 2) hydrided non-irradiated material 3) hydrided and irradiated material The samples were hydrided with a gaseous charge of 200 ppm of hydrogen, using a Sievert device at 350 ºC. They were subsequently annealed for three days in argon atmosphere at 380 ºC.
One of the batches of hydrided samples was irradiated at room temperature in the CNEA-RA3 nuclear reactor. The neutron fluence was 3.5 × 10 23 n m -2 (E > 1 MeV) after an irradiation of 30 days.
In all groups, tensile tests were carried out at 25 ºC and 300 ºC. Thin sheets were obtained from the heads of each specimen for their microstructural characterization.
Tensile tests were performed in an Instron machine with a maximum load cell of 50 kN, using an inverted traction system. Ad-hoc clamps were designed for the assemblage of irradiated specimens to minimize the manipulation of irradiated material.
All the tensile tests were performed at the same nominal crosshead speed of 0.2 mm min -1 , being the strain rate 1.5 × 10 -4 s -1 .
Wedge shape specimens for electron microscopy were mechanically polished down to a thickness of 0.2 mm. Discs 3 mm in diameter were punched out and mounted in a polishing device. The disks were then electropolished in a Struers Tenupol 5 twin jet electropolisher using an electrolyte of 90% ethanol and 10% perchloric acid at -30 ºC and 18 V.
Wedge shape thin films thus obtained allowed to study the hydrides in different planes of α-Zr matrix. TEM images were obtained in a Philips CM200, operated at 160 kV, and high-resolution images in a FEI TECNAI G2 operated at 200 kV.
The simulation image was obtained using a JEMS software [20].

Results
TEM micrograph of the starting material (non-rolled strip) is shown in Figure 1 [21]. Figure 2 shows the micrograph of the 50% laminated material that was cut with a numerical control machine and subjected to a relieving stress heat treatment at 450 °C for 24 hours.

Mechanical properties
Uniaxial tensile tests were performed on Zr-1%Nb at room temperature and at the reactor operation temperature (~300 °C), see Figure 3 a) and b).
The change in ultimate tensile strength (UTS), yield strength (YS), total elongation (εtot) and uniform elongation (εunif) values were extracted from the stress-strain curves, see Table 2.

Non-irradiated material
When comparing the curves corresponding to hydrided and non-hydrided specimens, it was found that at room temperature there is no noticeable effect on hardening (YS and UTS vary approximately 2%), Fig. 3a). Total elongation due to the presence of this type of hydrides was 10 % (see Table 2). On the other hand, when performing the same comparison with the tensile tests carried out at 300 ºC, it can be seen an increase of 6% in yield stress and ultimate strength, due to the presence of those hydrides Fig. 3b). Furthermore, the total elongation increased because the hydrides improve the total elongation by 10%. This phenomenon has been observed more clearly at 300 °C.

Irradiated material
After low-temperature (<100 ºC) neutron irradiation at 3.5 × 10 23 n m -2 (E > 1 MeV), the samples showed an increase of 67% (at 25 °C) and 75% (at 300 °C) in UTS and YS as compared to the case of hydrided and nonirradiated material. Simultaneously, a drastic decrease in ductility is observed at both temperatures (Figure 3a), b)), of around 35%, in εtot showing the relevance of the radiation effects on mechanical properties.

Non-irradiated material
The microstructural analysis of the first group, i.e., non-hydrided and non-irradiated, shows the typical α-Zr phase with equiaxial grains (see Figure 4 (a)). Spherical precipitates of the phase β-Zr were observed randomly distributed throughout the matrix. This phase is body centered cubic (BCC) with lattice parameter a = 0.3568 nm [17]. The average size of these precipitates was 199 nm with a dispersion of +/-10 nm (Figure 4 (b)).
ζ (zeta) hydrides were observed for the hydrided samples. They were characterized as a needle-shaped HCP structure, with lattice parameters a = 0.33 nm and c = 1.029 nm, corresponding to a trigonal crystal with spatial group P3m1. In general, they were in planes [0 0 1] of the α-Zr matrix [22]. The average length of these hydrides was 186 nm with a dispersion of +/-10 nm (see Figure 5 (a), (b), (c)).

Irradiated material
Regarding hydrided and irradiated specimens, small precipitates of Zr (Nb, Fe)2 were identified. These precipitates are known to be hexagonal compact structure (HCP), with spacial group symmetry P63/mmc and lattice parameters a = 0.5335 nm and c = 0.866 nm [2,23]. Due to their small size this phase was identified by electron diffraction patterns. The superposition of the matrix accounts for the uncertainty of the data that give compositional values higher than those corresponding to a Laves phase [14]. However, when observing high resolution images and its respective simulation, (Figure 6a) and b)), it is confirmed that the contrast obtained is consistent with the presence of a crystalline structure of the precipitated corresponds to a Laves phase known as C14-structure type MgZn2 [2].
The presence of these precipitates was observed surrounding the ζ-hydrides at the above-mentioned irradiation dose (see Figure 7(a)).
The metastable hexagonal ω-Zr phase was identified, distributed randomly in the α-Zr matrix. This phase has spatial group symmetry P6/mmm and lattice parameters a = 0.5034 nm and c = 0.3124 nm. The mean size of these precipitate was measured as 7 nm with a dispersion of +/-1 nm. Due to their small size (see Figure 8) this phase was identified as such by electron diffraction.

Non-irradiated material
When analyzing the stress-strain curves behavior of the hydrided and non-hydrided specimens (see Figure 3), it was observed that at room temperature the presence of hydrides did not generated an increase in hardening nor ductility. Instead at 300 °C a slight increase can be seen in hardening, and ductility seems too slightly improve.
This anomalous behavior could be attributed, among other factors, to the type of hydride and its possible location. In this work, a type of hydride known as ζ-hydride was found, see Figure 5. This hydride was characterized by Zhao et.al (2008) [22] in Zircaloy-4. It is needle-shaped with HCP structure, corresponding to a trigonal crystal with space group P3m1 [22,24]. Also, by studies carried out by Thuinet et.al (2012), it was oberved that ζ-hydrides nucleated more easily than γ-hydrides because they have a lower interfacial energy [25]. According to the cooling rate, the ζ-hydrides transform to give rise to metastable γ-hydrides.
Another aspect which should be considered is the hydride orientation in the matrix. In this work, the texture of laminated sample favors hydrides to be located in certain planes. Zhao proposed that a new phase may play an important role in the stress-reorientation of hydrides. [1] since the hydride orientation is determined by the nucleation process and is maintained constant during growth. In presence of external stresses, elastic interaction between the precipitation-induced strain and the applied stress may have an influence on the orientation of the hydrides [26][27][28][29]. In addition, Zhu and coworkers completed a study on the ductility of each zirconium hydrides. Their research indicated that ζ-hydrides and γ-hydrides are more ductile than the α-Zr matrix [1,30]. The observed stress-strain curves behavior, in this study, could be explained as the sum of these factors.
It is also necessary to consider the microstructure of this alloy, especially the presence of β-Nb precipitates [31], as a result of the heat treatments carried out. Dislocation pile-ups at these precipitates are believed to play an important role in work hardening [32]. However, the amount of precipitated phase does not seem to be significantly.

Irradiated material
In the stress-strain curves of hydrided and irradiated material it was observed a large increase of ultimate tensile strengths (UTS) and a considerable decrease in the total elongation with respect to the non-irradiated samples. This indicates that the behavior of the mechanical properties of the irradiated and hydrided material is opposite to those corresponding to non-irradiated and hydrided samples.
The irradiated samples were analyzed by transmission electron microscopy (TEM). The presence of the same hydrides as in the previous case was observed, as well as microstructural changes resulting from radiation damage [33].
One of the microstructural changes that have been noted is the presence of second-phase precipitated particles (SPP), around the ζ-hydrides. The incipient precipitation has been crystallographically characterized by two different techniques, HRTEM and SAD, as Laves-C14 phases. Due to of the small size of the precipitates observed in the present work, was not possible to determine their composition.
At present, determining the composition of these precipitates is of great interest. The Laves-C14 phases have strict stoichiometries AB2. However, the compositional designation of Zr (Nb, Fe)2 has a very wide range [34]. Studying the interaction between these precipitates and hydrides is of vital importance to understand how it would affect the mechanical properties of this alloy. Idress et al. [35] demonstrated that irradiation induces the Fe dissolution, and this dissolved Fe was found in the form of spherical precipitates in the α-Zr matrix [35]. As an H atom is smaller than a Fe atom, it moves more easily occupying the interstitial sites of the cells delaying the Fe atoms movement in the α-Zr matrix [36].
Another possibility was proposed by Burr and coworkers [36], who studied the effects of SPP on the H absorption of Zr alloys [36][37][38]. Their results have showed that these particles could potentially be used as H sinks by decreasing the H availability for hydriding. However, under irradiation, Fe atoms spread out of the SPP faster than other elements [38][39][40]. Then, in the case of Nb-containing Laves phases, it would increase the affinity of residual SPP for H, suggesting that H will likely segregate to them, thus depleting the H content in the Zirconium metal, this phenomenon is limited to low doses (3.5 × 10 23 n m -2 ) [36].
The presence of the ω phase has also been observed due to decomposition of the β-Zr phase induced by radiation (see Figure 8) [41].
In the present work, it was not possible to observe the presence of dislocation loops. The first loops to nucleate at low temperatures and low doses (5 × 10 25 n m -2 ) are type-〈 〉 loops, varying from 5 to 20 nm in size [42]. Also, Cockeram et al. [43,44] have reported that〈a〉 and 〈c〉-type dislocation loops as well as groups of point defects can only be observed if the neutron irradiation fluence is greater than 0.058 × 10 24 n m -2 . If dislocation loops were present, then they were below the size detection limit for the TEM (nominally ≤ 0.4 nm). It could also be expected to find these loops lying on the prism planes {1 0 1 � 0} or {1 1 2 � 0} [45], although recent studies suggested that most 〈 〉-type dislocation loops do not perfectly lie on these prism planes but tilt a little towards the basal planes [46]. These were not the planes that were analyzed in this work (see Figure 3a)).
In Figure 9 the ductility of hydrided and irradiated and non-irradiated alloy is compared at 25 °C and 300 °C. It is observed that the total elongation of the irradiated and non-irradiated material varies by approximately 60% at both temperatures, however the difference in non-uniform elongation between the irradiated and non-irradiated material is greater at 300 °C.
Mechanical property degradations of the irradiated specimens may be attributed to the ζ-hydrides orientation produced during the nucleation process and by the radiation damage in matrix. The presence of radiation generated loops, Zr (Nb, Fe)2 precipitates and ω-phase produce microstructural changes at the nano scale, which could explain the decrease in mechanical properties such as a sharp increase in ultimate tensile strength. The irradiation produced defects represent barriers to the movement of dislocations so that the radiation hardening is based upon the interaction of mobile dislocations with the irradiation produced defects.
It can be concluded that, at the present dose (3.5 × 10 23 n m -2 (E > 1 MeV)), the microstructural changes generated by radiation are the most responsible for the damage, as measured through the mechanical stress changes, rather than that produced by ζ-hydrides.

Conclusions
Mechanical tests and microstructure studies of Zr-1wt.% Nb hydrided with 200 ppm of hydrogen and nonirradiated and irradiated were carried out at 25 ºC and 300 ºC.
The presence of the metastable ζ-hydride was observed in both cases. However, tensile behavior is different. In the case of hydrided and non-irradiated, no significant change in hardening and ductility were noted at room temperature. Instead at 300 ºC a slight increase can be seen in ductility of the hydrided alloy. This anomalous behavior could be attributed to the type of hydride and its possible orientation in the matrix.
In the mechanical tests of hydrided and irradiated Zr-1wt.% Nb, a great decrease in the ductility and great increase in hardening were observed at both temperatures. The mechanical properties behavior suggested a reordering of ζ-hydride and the presence of the microstructural changes due to the radiation damage acting according to the scattered barrier model.