Elsevier

Journal of Nuclear Materials

Volumes 386–388, 30 April 2009, Pages 503-506
Journal of Nuclear Materials

Irradiation hardening and embrittlement in high-Cr oxide dispersion strengthened steels

https://doi.org/10.1016/j.jnucmat.2008.12.144Get rights and content

Abstract

The effects of neutron irradiation on the mechanical properties of high-Cr oxide dispersion strengthened (ODS) ferritic steels were observed. The materials used were produced by varying the Cr content from 14 to 22 wt% while maintaining the yittria contents within the range of 0.36–0.38 wt%. The ODS steels were irradiated in JMTR at 300, 420 and 550 °C to 5.56 × 1020 n/cm2 (>1 MeV). Charpy impact tests were performed between −150 °C and room temperature. The upper shelf energy (USE) of all ODS steel samples were reduced after irradiation. However, the reduction became smaller as the irradiation temperature increased. When the ODS steels were irradiated at 300 and 420 °C, the ductile-to-brittle transition temperature increased significantly. Tensile tests of the ODS steels irradiated at 300 °C were performed at room temperature with a strain rate of 6.7 × 10−4 s−1. The yield stress and tensile strength of the irradiated ODS alloys increased significantly, and the level of irradiation hardening increased with the Cr content. Anisotropy was not observed in the yield stress and tensile strength but was observed in the elongation.

Introduction

Oxide dispersion-strengthened (ODS) ferritic/martensitic steels have been developed as a fuel cladding material for sodium-cooled fast breeder reactors (SFRs) [1], [2]. Due to the dispersion of oxide particles, the ODS steels show high-strength at high temperatures [3]. In terms of the irradiation effects on the mechanical properties, recent irradiation experiments clearly showed that the ODS steels are highly resistant to irradiation embrittlement at temperatures between 300 and 500 °C up to 15 dpa [4]. The ODS steels developed as SFR fuel cladding material contain at most 12% chromium. It is well known that the level of corrosion resistance in high-temperature water diminishes significantly as the chromium concentration decreases to less than 13% [5]. High-Cr ODS steels are noted for their improved corrosion resistance properties in super critical pressurized water (SCPW) [5], [6]. However, irradiation embrittlement caused by Fe–Cr phase decomposition under neutron irradiation can be a critical issue for high-Cr steels. The objective of this study is to investigate the effect of neutron irradiation on the mechanical properties of the high-Cr ferritic ODS steels.

Section snippets

Experimental procedure

The materials used were five types of ODS steels (K1–K5) that were produced by varying the Cr content from 14 to 22 wt% while maintaining the yittria content in the range of 0.36–0.38 wt%. The main chemical compositions of the K1, K2, K3, K4 and K5 steels are 19Cr, 14Cr–4Al, 16Cr–4Al, 19Cr–4Al and 22Cr–4Al, respectively. The chemical composition of each type of steel is summarized in detail in Table 1. All specimens were irradiated with fast neutron fluence of 5.56 × 1020 n/cm2 (>1 MeV) in the Japan

Ductile-to-brittle transition behavior

Charpy impact energy curves of 14Cr–4Al (K2) ODS steel as a function of the test temperature after irradiation at 300, 420, 550 °C are shown in Fig. 2. The upper shelf energies (USE) of the irradiated samples were lower than these values of unirradiated samples, and the ductile–brittle transition temperatures (DBTT) were higher. For the samples irradiated at 300 °C, USE was reduced and DBTT increased significantly compared with unirradiated samples; however, the USE reduction and the increase in

Conclusion

The effects of neutron irradiation on the Charpy impact energy and tensile properties of five high-Cr ODS steels were investigated after JMTR irradiation at 300, 420 and 550 °C. The main results are as follows:

  • (1)

    The reduction in the USE and the increase in the DBTT decreased as the irradiation temperature increased. Below an irradiation temperature of 420 °C, the degree of irradiation-induced embrittlement increased. On the other hand, the degree of irradiation embrittlement was reduced

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