The role of grain boundary microchemistry in irradiation-assisted stress corrosion cracking of a Fe-13Cr-15Ni alloy

Abstract

A novel explanation based on the variation of grain boundary composition is proposed to elucidate the localized nature of irradiation-assisted stress corrosion cracking, i.e., intergranular cracking is observed only at specific sites of random high-angle grain boundaries, in a Fe-13Cr-15Ni austenitic model alloy subjected to proton irradiation and straining in a high-temperature water environment. Specifically, this work presents electron microscopy characterization of multiple cracked and un-cracked grain boundary sites and the neighboring oxides. The depletion of Cr and enrichment of Ni, as expected due to radiation-induced segregation, is observed at grain boundaries far from the crack tip (for cracked sites) or the metal surface (for un-cracked sites), and such modification of grain boundary composition is enhanced in the vicinity of the corrosion reaction front. Unexpectedly, grain boundary sites beyond the crack tip always present a lower Ni content and higher Cr content than the un-cracked sites on the same grain boundary. Overall, it is proposed that the site-specific susceptibility to stress corrosion cracking is governed by the grain boundary microchemistry, which determines not only the quantity of Cr, but also the efficiency of Cr transport to the reaction front and thus the protectiveness of the inner oxide. These effects of local composition may be further coupled with the local structure of and the local stresses interacting with the random high-angle grain boundaries.

Introduction

Irradiation-assisted stress corrosion cracking (IASCC) describes the accelerated initiation and/or propagation of intergranular cracks (IGCs) in alloys concurrently subjected to irradiations, mechanical loads, and aggressive (e.g., high-temperature water) environments [1], [2]. This phenomenon has been recognized as a primary cause of failure for the austenitic stainless steels (SS) and Ni-based alloys widely used in light-water reactors [3], [4], the reactor type of more than 80% of operating nuclear power plants worldwide [5]. Understanding the governing mechanism of IASCC is thus essential for the access to reliable and sustainable nuclear energy, but has been impeded by the complex and synergetic nature of radiation-induced changes in materials [1], [4]. These changes include the production and evolution of radiation-induced defects and the attendant hardening, localized deformation in the form of dislocation channels (DCs), and radiation-induced segregation (RIS) at grain boundaries (GBs).

To date it is accepted that oxidation and deformation are the most important factors that govern IASCC susceptibility [4]. The process of oxidation is controlled by the water chemistry and the alloy composition, especially the RIS at GBs [6], [7]. A weak correlation has been found between the GB Cr content in proton-irradiated 300-series SS and their IASCC susceptibility, as measured by the fraction of IGCs generated during constant extension rate tensile (CERT) tests, in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments [8]. For example, a GB Cr content lower than ∼12 wt. % always led to cracking, and a GB Cr content higher than ∼18 wt. % generally provided a resistance against IASCC. However, alloys with a GB Cr content between these limits exhibited very scattered results and no direct correlation was obtained [1], [8]. Post-irradiation annealing experiments further revealed that IASCC susceptibility can be rapidly mitigated well before the magnitude of RIS was changed [9]. Therefore, RIS alone may be a necessary but not sufficient factor to govern IASCC susceptibility [1], [9].

On the other hand, there is a growing body of evidence identifying localized deformation as a primary contributor to IASCC [10], [11], [12], [13], [14], [15], [16]. A significant and universal correlation between the weighted surface step height of DCs and IASCC susceptibility was observed in various austenitic SS and model alloys [12]. Specifically, IGCs were preferentially initiated at discontinuous intersections between DCs and random high-angle boundaries (RHABs) [11], [13], where no slip transfer occurred and a high local stress was developed [14], [16]. In consequence, the protective oxide layer over the GBs was ruptured and the metal was exposed to the water environment [10]. However, it was further noticed that not all discontinuous DC-GB intersections led to cracking. Indeed, recent analysis of all DC-GB intersections in a Fe-13Cr-15Ni model alloy subjected to proton-irradiation and CERT tests in a simulated BWR environment showed that only 5–20% (dependent on the total plastic strain) of discontinuous intersections resulted in crack initiation, even though this fraction was 4–6 times larger than that of continuous intersections, where strain can be transferred across the GBs [15]. These observations indicate that IASCC susceptibility of GBs and the site selectivity are ultimately governed by the local properties (such as composition, structure, and/or stresses) at the GBs, which can be spatially inhomogeneous but have been poorly characterized to date. It is also noted that these local factors do not necessarily refer to the initial state of GBs nor the entire GBs, but reflect the changes induced by irradiation and deformation.

In this paper, the results of a detailed and comparative characterization of multiple cracked and un-cracked sites of discontinuous DC-GB intersections and their neighboring oxides are reported. The objective is to discover the role of local composition of GBs in the IASCC susceptibility of a specific site of discontinuous DC-GB intersection.