Role of grain boundary engineering in the SCC behavior of ferritic–martensitic alloy HT-9

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Abstract

This paper focuses on the role of grain boundary engineering (GBE) in stress corrosion cracking (SCC) of ferritic–martensitic (F–M) alloy HT-9 in supercritical water (SCW) at 400 °C and 500 °C. Constant extension rate tensile (CERT) tests were conducted on HT-9 in as-received (AR) and coincident site lattice enhanced (CSLE) condition. Both unirradiated and irradiated specimens (irradiated with 2 MeV protons at 400 °C and 500 °C to a dose of 7 dpa) were tested. Ferritic–martensitic steel HT-9 exhibited intergranular stress corrosion cracking when subjected to CERT tests in an environment of supercritical water at 400 °C and 500 °C and also in an inert environment of argon at 500 °C. CSL-enhancement reduces grain boundary carbide coarsening and cracking susceptibility in both the unirradiated and irradiated condition. Irradiation enhanced coarsening of grain boundary carbides and cracking susceptibility of HT-9 for both the AR and CSLE conditions. Intergranular (IG) cracking of HT-9 results likely from fracture of IG carbides and seems consistent with the mechanism that coarser carbides worsen cracking susceptibility. Oxidation in combination with wedging stresses is the likely cause of the observed environmental enhancement of high temperature IG cracking in HT-9.

Introduction

One of the most promising advanced reactor concepts for Generation IV nuclear reactors is the supercritical water reactor (SCWR). Operating above the thermodynamic critical point of water (374 °C, 22.1 MPa), the SCWR offers many advantages compared to current LWRs including the use of a single phase coolant with high enthalpy, the elimination of components such as steam generators, steam separators, dryers, and a low coolant mass inventory resulting in smaller components, and a much higher efficiency. Since SCW has never been used in nuclear power applications, there are numerous potential problems, particularly with materials. Depending upon the species present and the oxygen content in the solution, SCW can become a very aggressive oxidizing environment. This is a cause of concern in regards to both general corrosion and stress corrosion cracking (SCC) of the structural materials and fuel elements of the reactors.

Ferritic–martensitic (F–M) alloy HT-9 has been identified as a candidate core structural alloy in SCWR. Preliminary studies have shown that HT-9 experiences high corrosion rates in the SCWR [1]. HT-9 also exhibits IGSCC in SCW at 400 °C and 500 °C [1], [2]. Initial studies have shown that proton irradiation at 400 °C and 500 °C enhances cracking in SCW [2]. Irradiation enhances diffusion and/or precipitate redistribution, which can enhance recovery and coarsening.

Grain boundary engineering (GBE) is being explored as a means of reducing the susceptibility of HT-9 to IGSCC in SCW. GBE involves a series of thermo-mechanical treatments designed to change the grain boundary structure by increasing the low angle boundary or coincident site lattice boundary (CSLB) fraction. Due to an increased structural order and reduced free volume, these boundaries exhibit relatively low energy and less segregation, thus providing resistance to intergranular corrosion. Another potential benefit of GBE is that these special boundaries induce slip in neighboring grains by either transmitting or absorbing and re-emitting lattice dislocations, thereby reducing grain boundary stresses and propensity for crack formation [3].

According to work done by Hertzberg et al. [4] fracture in steels is generally initiated at the inclusions or precipitate particles. The critical stress to propagate a crack is inversely proportional to the length the crack [4]. If it is assumed that the crack length at initiation is equal to the diameter of the carbide particle then the fracture stress will decrease with increasing precipitate size. Irradiation leads to coarsening of the precipitates (including carbides) in addition to the coarsening which occurs under load and temperature. Lechtenberg et al. [5] observed that embrittlement of HT-9 is associated with increased grain boundary carbide precipitation and coarsening. Gelles et al. [6] and Kai et al. [7] attributed a decrease in fracture stress and an increase in DBTT to particle coarsening during irradiation. It is believed that coarsening of these particles would be less in grain boundary engineered specimens as these low misorientation boundaries have a lower diffusivity that should result in slower coarsening rate. Thus if cracking in HT-9 is initiated at carbide particles, GBE should reduce susceptibility to SCC.

Kim et al. [8], Palumbo et al. [9], Lehockey et al. [10] and Alexandreanu et al. [11] have demonstrated the beneficial influence of GBE in mitigating SCC in austenitic alloys and Ni-base alloys. Little work has been reported on F–M alloys. Gupta et al. [12], [13] have demonstrated the beneficial influence of GBE on creep properties of F–M alloy T91. It is envisaged that CSL-enhancement would improve the SCC behavior of F–M alloy HT-9 in SCW both with and without the effect of irradiation due to the reasons discussed above.

The purpose of the current study is to investigate the effect of GBE on SCC of F–M alloy HT-9 for use as structural material in the SCWR. CERT tests were conducted in deaerated water at 400 °C and 500 °C on unirradiated and irradiated F–M alloy HT-9 in the as-received (AR) and CSL-enhanced (CSLE) condition to observe the effect of GBE on the susceptibility of HT-9 in SCW. Further, microstructural characterization was performed on the specimens in the AR and CSLE condition both before and after the CERT test to understand the role of GBE on the SCC behavior of HT-9.

Because the dissolved oxygen concentration has been shown to affect the growth of the oxide layer, experiments were also conducted on samples implanted with oxygen to modify the oxide growth during exposure to SCW [14]. These studies showed a smaller weight gain for the surface modified (SM) HT-9 during exposure experiments at 500 °C in SCW as compared to unmodified HT-9. It was envisaged that surface modification might also improve the SCC resistance of HT-9 in SCW. Therefore, SCC experiments were conducted on SM HT-9 in SCW at 500 °C in both the irradiated and unirradiated condition.

Section snippets

Material and sample fabrication

The chemical composition of HT-9 heat used in this study is given in Table 1. Samples in the form of tensile bars with a gage length of 21 mm and a cross-section of 2 mm × 2 mm were fabricated via electric discharge machining, Fig. 1. The standard heat treatment consisted of a solution anneal at 1040 °C for 30 min to completely austenitize the microstructure and dissolve the carbides, followed by air-cooling and then tempering at 760 °C for 60 min to relieve the stresses and enhance toughness (Note

Irradiated microstructure

Carbide coarsening during irradiation at 500 °C was measured at PAGBs. The carbide size at PAGBs in the AR condition increased from 365  ± 15 nm to 405 ± 17 nm after irradiation (Fig. 7). Error bars depict the standard deviation from a set of 30 readings. Minimal or no coarsening was observed for carbides on the lath boundaries.

Surface modified corrosion samples

The SM samples used in exposure tests were characterized for weight gain and the structure and morphology of the oxide layer itself were characterized by scanning electron

Intergranular fracture in HT-9

In this study, IG cracking in HT-9 was observed in an inert environment of argon, suggesting that the microstructure of HT-9 plays a role in the inherent susceptibility to IG cracking. Alamo et al. [19] observed intergranular fracture in addition to some ductile regions with dimples in HT-9 irradiated to a dose of 3.4 dpa at 325 °C in the Osiris reactor and tested at 20 °C in air. It is well established that ductile failure is initiated by the nucleation of voids at second phase particles. There

Conclusions

  • Ferritic–martensitic steel HT-9 exhibited intergranular stress corrosion cracking when subjected to constant extension rate tensile tests in an environment of supercritical water at 400 °C and 500 °C.

  • HT-9 exhibited cracking in an inert environment of argon, emphasizing the role of microstructure in the inherent susceptibility to cracking.

  • CSL-enhancement reduces cracking susceptibility of HT-9 in both the irradiated and unirradiated condition in SCW.

  • Carbide coarsening was reduced in the CSLE

Acknowledgements

Support for this work was provided by the United States Department of Energy under the NERI (Grant # 3 F-00202) and the I-NERI project (Grant # 3 F-01041). The authors gratefully acknowledge the facilities provided by the Electron Microbeam Analysis Laboratory and Michigan Ion Beam Laboratory at the University of Michigan. The authors would also like to thank Ovidiu Toader and Victor Rotberg of the Michigan Ion Beam Laboratory for their invaluable support in conducting the irradiations.

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