Studies on hot hardness of Zr and its alloys for nuclear reactors

doi:10.1016/S0022-3115(98)00610-2

Journal of Nuclear Materials

Volume 265, Issues 1–2, February 1999, Pages 91-99

https://doi.org/10.1016/S0022-3115(98)00610-2 Get rights and content

Abstract

The hardness behaviour of pure Zr, Zircaloy 2, Zr–2.5Nb, Zr–2.5Nb–0.5Cu and Zr–1Nb–1Sn–0.1Fe alloys (composition in wt%) were evaluated from ambient temperature to 1173 K using a hot hardness tester in order to compare the hardness and to extract information on the strength behaviour from the hot hardness data using standard empirical relations. This paper summarizes the experimental hot hardness data of several Zr alloys and discusses the results with the help of their microstructure, alloy content and atomic radii of the constituents.

Introduction

Zirconium based alloys are the natural choice for both fuel element cans and in-core structural components in water-cooled nuclear reactors, since these alloys have low thermal neutron absorption cross section, adequate strength and ductility, good corrosion resistance, long-term dimensional stability in an irradiation environment, excellent compatibility with the fuel and coolant, good thermal conductivity and adequate resistance to fracture 1, 2. The most important alloy of Zr used in Pressurised Heavy Water Reactors (PHWRs) is Zircaloy (Zircaloy 2 and Zircaloy 4), which is commonly used for fuel cladding and also for pressure tubes and calendria tubes. In Zircaloy, the major alloying element is Sn. The addition of Sn lowers the stacking fault energy of Zr which is expected to have significant influence on workability 3, 4and creep strength. Zr–2.5Nb alloy is the new generation alloy for pressure tube material. Nb helps in improving the creep strength as well as oxidation and corrosion resistance. Zr–2.5Nb–0.5 Cu alloy is used for garter springs which separate the hot pressure tube from the cold calendria tube. The addition of Cu results in lowering the α–β transition temperature and partitions entirely to β-Zr. Cu is expected to modify the deformation characteristic of β alone in the α + β region [5]. The addition of Cu changes the ageing kinetics of the alloy. The quaternary Zr–1Nb–1Sn–0.1Fe alloy is an advanced cladding material which shows superior corrosion resistance and irradiation stability 6, 7, 8. It has been reported that the irradiation creep and irradiation growth of the quaternary alloy is only 80% and 60%, respectively of that of Zircaloy 4 [8]. In a typical PHWR, the fuel tube sees a temperature of ~623 K and undergoes rigorous stress conditions due to pellet clad interaction. The pressure tube operates at ~558 K and has to withstand a coolant pressure of ~10 MPa. The garter spring should support the pressure tube initially and the creep sag subsequently and therefore should have adequate strength. Such a diverse range of specifications are met only through proper control of processing parameters [5]. It is worth investigating certain properties of Zr alloys such as hot hardness, since hardness is an easily measurable property and can be converted to the more useful strength data through well-established empirical relations. It may be noted that all the materials used in this study belonged to materials of various sizes and shapes making it practically impossible to resort to the conventional tension test. Hence, hardness testing methodology was used in order to derive the tensile strength from the hardness data. The hardness test method will also be a very useful tool for extracting information on the mechanical properties of irradiated Zr alloys since sample requirement for this test is very small. The only sample requirement of this test is a metallographically polished surface. This miniaturisation of the specimen reduces the induced radioactivity so that the testing can be carried out even outside the shielded cell. Instead of yielding one set of results on a specimen of many cm² surface, statistically significant number of results can be obtained per cm² surface using hardness tests [9]. Indentation testing technique, in particular microindentation method, can be regarded as a quick, simple and non-destructive mechanical procedure [9]. In this paper, the Zr alloys were compared in terms of their hot hardness in the temperature range of 293–1173 K. The hot hardness of the matrix material, pure Zr, has also been evaluated for comparison.

Section snippets

Experimental

The details of the samples used in this study, their heat treatment conditions and chemical composition are given in Table 1Table 2. Samples were cut from the above mentioned alloys and were metallographically prepared. Hot hardness measurements were carried out using a hot hardness tester (Nikon, Model QM) with the help of a diamond Vickers pyramid indenter. Before starting the hot hardness experiment, the instrument was calibrated using a standard (Cu: SRM; National Bureau of Standards, USA)

Results

The hardness (H) vs. temperature (T) plots for Zr, Zircaloy, Zr–2.5Nb, Zr–2.5Nb–0.5Cu and Zr–1Nb–1Sn–0.1Fe alloys are shown in Fig. 2. The H–T data of the above mentioned alloys were fitted by a polynomial and the equations for the best fit are given below:
Zr $H=232.91−0.35T+1.07×10^{−4} T^{2} +2.19×10^{−8} T^{3},$ Zircaloy 2 $H=390.56−0.92T+8.39×10^{−4} T^{2} −2.88×10^{−7} T^{3},$ Zr–2.5Nb $H=400.91−0.78T+5.81×10^{−4} T^{2} −1.76×10^{−7} T^{3},$ Zr–2.5Nb–0.5Cu $H=2164.18−15.17T+0.045T^{2} −6.27×10^{−5} T^{3} +4.13×10^{−8} T^{4} −1.03×10^{−11} T^{5},$ Zr–1Nb–1Sn–0.1Fe

Hardness–temperature relation

The temperature dependence of hardness of metals and alloys has been reviewed by Westbrook [10]. The summary of his finding is that the temperature dependence of hardness is best represented by the following relation of the type $H=A exp (−BT),$ where constants A and B are called the intrinsic hardness (i.e. the value of hardness at 0 K) and softening coefficient, respectively. The constants A and B have one set of values at low temperature (A_I and B_I) and another at high temperature (A_II and B_II),

Conclusions

The hardness of Zr-based alloys used in nuclear thermal reactors was measured from ambient temperature to 1173 K and the following conclusions were drawn:

1.
The hardness of Zr and Zr–2.5Nb–0.5Cu alloy was the lowest and the highest respectively in the low temperature region below the transition. The effect of alloying elements like Cu, Nb, Sn has been discussed and found that Cu followed by Nb are the best strengthening agents in Zr.
2.
The hardness–temperature behaviour of these alloys can be

Acknowledgements

The authors wish to acknowledge Mr D.S.C. Purushotham, Head, Radiometallurgy Division, for permitting to publish this work and for his keen interest in this research programme. They wish to record their sincere thanks to Dr S. Banerjee, Head, Materials Science Division for reviewing this paper and for useful suggestions. The suggestions rendered by Messrs K.C. Sahoo, S. Anantharaman, K.S. Balakrishnan and Dr J.K. Chakravarthy are kindly acknowledged.

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Studies on hot hardness of Zr and its alloys for nuclear reactors

Abstract

Introduction

Section snippets

Experimental

Results

Hardness–temperature relation

Conclusions

Acknowledgements

Prog. Mater. Sci.

J. Nucl. Mater.

J. Nucl. Mater.

J. Nucl. Mater.

Trans. Am. Soc. Met.

J. Mater. Sci.

Proc. Royal. Soc. A

Metallurgia

J. Met.

Trans. Am. Soc. Met.