Effects of hydrogen content and temperature on fracture toughness of Zircaloy-4

https://doi.org/10.1016/S0022-3115(03)00193-4Get rights and content

Abstract

The influence of hydrogen content and temperature on the fracture toughness of a Zircaloy-4 commercial alloy was studied in this work. Toughness was measured on CT specimens obtained from a rolled material. The analysis was performed in terms of J-integral resistance curves. The specimens were fatigue pre-cracked and hydrogen charged before testing them at different temperatures in the range of 293–473 K. A negative influence of the H content on material toughness was important even at very small concentrations, being partially restored when the test temperature increased. Except for some specimens with high H concentration tested at room temperature, the macroscopic fracture behaviour was ductile. The role of Zr-hydrides and Zr(Fe,Cr)2 precipitates in the crack growth and the dependence with hydrogen content were analysed by observation of the fracture surfaces and determination of the Zr(Fe,Cr)2 precipitates density on them.

Introduction

It has long been known that zirconium-based cladding alloys absorb hydrogen (H) as a result of service in a nuclear reactor environment. This hydrogen can be absorbed from different potential sources: H cathodically generated as a result of the corrosion of the cladding by the primary side water coolant [1], H gas dissolved in the primary side coolants in pressurized water reactors [2] and hydrogen resulting from corrosion by gases or vapours containing H in the internal plenum region of the rods [3].

Different stable phases are present in the zirconium–hydrogen (Zr–H) system, depending on the H content and temperature. For temperatures below 773 K these equilibrium phases are [4]:

  • αZr, with H in solid solution in the matrix,

  • δ hydride, ZrH1.5 to ZrH1.66, with a fcc structure which initially precipitates as platelets in αZr,

  • ε hydride, ZrH1.66 to ZrH2, fct, which comes from a martensitic transformation of δ.


Since Zircaloy-41 (Zry-4) is 98% Zr, the Zr–H phase diagram is employed when the H–Zry-4 interaction is studied, although this consideration does not account for the influence of alloying elements on the H absorption process. These alloying elements are present in solid solution in the Zr matrix, or as second phase precipitates preferentially located in the grain boundaries. When hydrogen is absorbed into Zry-4, these precipitates may play an important role at low H concentration (due to their preferential location respect to the hydrogen diffusion path) by producing deviations from the Zr–H diagram.

According to the Zr–H diagram, hydrogen has limited solubility in the Zr lattice, the terminal solid solubility (TSS), and, at higher concentration, the hydride phases precipitate [5], [6]. The dependence of TSS with temperature can be obtained from useful empirical formulations. The morphology of hydride phases depends on the formation process where the cooling rate plays an important role. The precipitation of zirconium hydrides embrittles the zirconium alloy cladding reducing its ductility as well as its fracture toughness. It has been previously documented that this detrimental effect of hydrides causes premature fracture of nuclear fuel components in service [7].

Many authors [8], [9], [10] have found a ductility reduction in tensile tests, induced by H in Zr alloys; this reduction sometimes also reported as ductile-to-brittle transition [8]. Here, fracture mechanics is the most convenient methodology to study this behaviour since it not only characterises the toughness variation, but also gives a quantitative engineering magnitude that can be used in both design and in-service defect analysis.

A study of the detrimental effects of H on the elastic–plastic fracture toughness of a Zircaloy-4 commercial alloy at different temperatures is introduced in this work. Hydrogen precipitation mechanisms and their relationship with the toughness variation were also analysed aided by fracture surface observation. The existence of a ductile-to-brittle transition is discussed.

Section snippets

Material

All specimens were obtained from a Zry-4 rolled 6 mm thick plate. Fe, Cr, Sn and O in the proportion of 0.22:0.1:1.4:0.128 wt%, respectively, were the main alloying components in the Zr matrix; Sn and O in solid solution, and Zr(Fe,Cr)2 in Laves phases precipitate form. The thermo-mechanical history of the plate included hot-rolling, annealing and final cold-rolling treatments. The final material microstructure was a α-Zry-4 fine equiaxed grains with a mean diameter of approximately 8 μm (Fig. 1

Results

The hydrogen charge employed produced hydride precipitates of δ phase type, platelet shaped, oriented in planes parallel to the rolling direction, mainly at inter-granular sites. The distribution and size of hydrides is influenced to a great extent by the cooling rate; for slow cooling rates the platelets are grouped in long bands [14] as it was well observed in this work (Fig. 4). The hydride distribution was also identified by employing a colorant treatment.

Fig. 5 shows the dependence of the

Discussion

As a consequence of the load–displacement records obtained (Fig. 5) including those that resulted in brittle fracture, elastic–plastic fracture mechanics was considered the adequate methodology to study hydrogen embrittlement over the concentration and temperature ranges analysed. Linear-elastic fracture mechanics was not employed to characterise fracture toughness in this work because its limits of applicability were violated.

Because of practical limitations, no observation of hydride

Conclusions

The main behaviour observed was ductile and only a few specimens, with very high H content, failed by brittle mode when tested at room temperature. At higher temperatures the JR curves raised, increasing both initiation and resistance to growth values. At room temperature, the JR curves dependence on H content were fitted with an exponential decay for the initiation values as well as for the slope ones. When the temperature grows, the values of JIC and ϕ1mm increased retaining the functional

Acknowledgements

To CNEA (Atomic Energy National Commission) and CONICET (National Council of Scientific and Technological Research) for having supported this project.

References (20)

  • A. Sawatzky

    J. Nucl. Mater.

    (1960)
  • J.J. Kearns

    J. Nucl. Mater.

    (1967)
  • J.H. Huang et al.

    Scr. Metall. Mater.

    (1993)
  • K.W. Lee et al.

    J. Alloys Compd.

    (2002)
  • D. Arias et al.

    J. Nucl. Mater.

    (1987)
  • B.F. Kammenzind et al.

    ASTM STP

    (1996)
  • J.C. Clayton

    ASTM STP

    (1984)
  • J.C. Clayton

    ASTM STP

    (1988)
  • E. Zuzek et al.
  • C.K. Chow et al.

    ASTM-STP

    (1986)
There are more references available in the full text version of this article.

Cited by (86)

View all citing articles on Scopus
View full text