Constant-displacement bolt-loaded compact tension specimens of Nickel-based Alloy 52 were exposed to boiling water reactor environment for 12 years, followed by an additional 3 years in post-shutdown cold water conditions in a Swedish nuclear power plant test loop, under a stress intensity factor of 20 MPa root m. After outer surface decontamination and specimen opening, unexpected crack extensions of 3-4.5 mm were observed. The fracture surface and the cross-sectional deformation microstructure were examined by electron microscopies techniques down to the nanoscale. The oxide layer in the region exhibiting unexpected crack growth was notably thin, suggesting that it formed after exposure to elevated operating temperatures. The dominant fracture mode is transgranular, propagating along close-packed {111} planes. The grains contained heterogeneous microstructures with regions enriched in nanometer-sized Ti(N,C) and the zigzag crack paths did not traverse these regions strengthened areas. Extensive shear bands were present near the crack tips, indicating pronounced localized plasticity. Hydrogen reduces stacking fault energy, results in localized plasticity and enhances shear bands formation. Low temperature crack propagation with evident effects of hydrogen was considered as the potential cause of crack propagation in Alloy 52 in the absence of external dynamic loading under post-shutdown cold water chemistry.

Initial cases of stress corrosion cracking (SCC) in pressurized water reactors (PWRs) occurred mostly but not exclusively in stagnant areas like dead-legs, but recently more extensive IGSCC has occurred in normal free-flowing PWR primary water. Operational experience and laboratory data reveal that the main parameters in IGSCC include cold work and weld residual strain, oxygen, and residual and applied stress. Residual strain, which arises from manufacturing, surface grinding, and welding, should be limited by optimizing manufacturing procedures, minimizing alignment and fit-up stresses and using high-quality weld procedures. Preventing oxygen ingress in the make-up water should be pursued. Stresses created by thermal fluctuations (thermal mixing, low-leakage core operation, and start-ups) deserve more attention. Weld residual stress, fit-up stresses and local stresses from load follow must be maintained below the annealed yield stress. IGSCC should be considered in aging management and in-service inspection. Detection techniques capable of identifying IGSCC should be employed.

This work highlighted that a ground surface finish and the exposure to a pressurised water reactor (PWR) environment result in a decreased low-cycle fatigue lifetime, an enhanced fatigue crack initiation and an accelerated fatigue crack growth rate of 304 L austenitic stainless steel. A ground surface finish promotes fatigue crack initiation and short crack growth especially in a water environment, due to the highly deformed underlying microstructure with high-angle grain boundaries and the grinding marks on surface. Martensite was observed in the vicinity of secondary crack tips in specimens tested in a simulated PWR primary side environment. The aggregated presence of alpha'- and epsilon-martensite in the vicinity of the fatigue crack tip can enhance the material's susceptibility to hydrogen-assisted fatigue cracking. Martensite formation was rarely observed in specimens exposed to high temperature air. The phase transformation from gamma-austenite to alpha'-martensite in the PWR primary environment occurred via the intermediate epsilon-martensite phase. (C) 2021 The Author(s). Published by Elsevier B.V.

The effects of machining surface preparations on the stress corrosion cracking (SCC) susceptibility in boiling water reactor environment were evaluated for Alloy 182 weld metal. Constant extension rate tensile test, con-stant load test and electrochemical tests (passivation, current-voltage curves and electrochemical impedance spectroscopy) showed that the surface treatments strongly influence the resistance to SCC initiation. The machining-induced deformation and surface oxide layer from the surface treatments affect the re-passivation behaviour and SCC initiation susceptibility. The SCC susceptibility of the studied surface treatments was found following the sequence of shot peened > industrial surface milled > polished > advanced manufactured surface.

The applicability of an accelerated test technique using tapered tensile specimens for investigating the stress corrosion crack (SCC) initiation behaviour of structural materials in high-temperature water was assessed in the framework of a European collaborative research project (MICRIN - MItigation of CRack INitiation). The main advantage of using a tapered geometry is, that in a single test a stress gradient is obtained through the gauge length, and therefore a stress threshold for SCC initiation can be determined in a reasonable timeframe. This method was used to investigate two different materials that were known to be susceptible to SCC in light water reactor environment: a high-Si stainless steel and a Ni-base weld metal (Alloy 182). The results of the international test programme confirmed that the tapered specimen test methodology could be used to identify a SCC initiation stress threshold, albeit that significant scatter was present in the data.

Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant. As-welded and reheated regions mainly consist of acicular and polygonal ferrite, respectively. Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions (0.5–2.5 µm) at the brittle fracture primary initiation sites. High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch. The density, geometry, and chemical composition of the primary initiation inclusions were investigated. A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the de-bonded interfaces between the uncracked inclusions and weld metal matrix. Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance.

The effects of four machining surface preparations on the environmentally-assisted cracking (EAC) initiation and early crack growth were evaluated for Alloy 182 flat tapered specimens exposed to a simulated boiling water reactor (BWR) environment under constant extension rate tensile test conditions. Results showed that the apparent EAC critical threshold stress for initiation was highly dependent on the surface treatment. Detailed microstructural characterization revealed the nature of the machining-induced deformation associated with the four surface treatments. Surface treatments affected the surface oxidation during exposure in the simulated BWR environment. The machining-induced deformation structure affected the EAC initiation susceptibility.

Dissimilar metal welds (DMWs) are a key design feature in nuclear power systems, typically involving ferritic low-alloy steels (LAS), stainless steels (SS), and nickel-base alloys. They are, however, a potential concern regarding the structural integrity of nuclear power systems. In particular, the LAS/nickel-base alloy weld metal interface is known to develop a local strength mismatch upon post-weld heat treatment (PWHT). Very limited data is available on the effect of thermal ageing on the DMW interface. The aim of this study was to investigate the effects of thermal ageing at 400 °C for up to 10,000 h on a narrow-gap DMW mock-up representative of the weld between the reactor pressure vessel nozzle and its safe-end after PWHT, with a special focus on the LAS SA 508/nickel-base Alloy 52 weld metal interface. No significant effect of thermal ageing on the appearing microstructure was observed in either LAS base material, LAS heat-affected zone or Alloy 52 weld metal. However, thermal ageing reduced the local strength mismatch at the LAS/nickel-base weld metal interface formed during PWHT. The reduction of the strength mismatch was detected using nanoindentation measurements and was concluded to be associated with a decrease in the carbon pile-up in the weld metal caused by PWHT. Based on the obtained results, thermal ageing promotes carbon diffusion from the weld metal side of the fusion boundary further away into the weld metal and thus slightly decreases the local strength mismatch.

Environmentally-assisted intergranular cracking (EAC) in austenitic stainless steel piping in nuclear power plants started with extensive cracking in boiling water reactors (BWRs) in the 1970s. These events initiated extensive research programs, which led to numerous expensive mitigation steps and lessons learned. The research, development, and mitigation actions dealt with new material(s), new welding techniques, and new water chemistry strategies, as well as developing a mechanistic understanding. The actions have been mainly, but not totally, successful, and research is still ongoing and needed. This chapter gives a review on EAC in stainless steels in light water reactors with a focus on BWRs and includes a personal opinion of where research and measures are especially needed.

Accident tolerance of LWR fuels and structures is of paramount importance, as highlighted by the accident at the Fukushima Daiichi nuclear power station. The ability of fuel cladding and core internals to resist runaway oxidation during a beyond design basis accident (BDBA), as well as to minimize corrosion during steady state operation and design basis accidents (DBAs), determines its degree of accident tolerance. A gap was identified in the accident tolerant fuel (ATF) concepts under consideration, as most consisted either of coatings which may not hold their integrity under prolonged operation, or different materials altogether which incur significant neutronic penalties. A compromise was therefore sought – by minimizing the amount of coating material while ensuring a strong microstructural bond, it was thought that a Zircaloy-steel layered composite would achieve the best of both approaches.The goal of this NEUP project was to develop a multi-metallic layered composite (MMLC) tailored to addressing accident tolerance of LWR fuel cladding and core internal structures using an innovative fabrication technology. The MMLC developed in this program is expected to enhance the accident tolerance of LWRs, thereby reducing the cost by recovering lost operating margins and/or increasing operating windows of peak cladding temperature, peak linear power, reducing steady-state corrosion, and enhancing severe accident tolerance.

Thermal ageing of nickel-base alloys can lead to the formation of brittle ordered phases, but the direct study of these phases is challenging. Nanoindentation is used in this study as an alternative technique to determine the extent of short-range ordering (SRO) in thermally aged Alloy 690 TT. Two methods are used for obtaining both qualitative (depth sensitivity) and quantitative (spatial stability) results, which are compared to data from metallography, microhardness and atomic force microscopy in order to discriminate between factors affecting the hardness. When intergranular precipitation and grain size become dominant factors with increasing indentation loads, nanoindentation at low loads enables to distinct within-grain hardness increase related to SRO.

In-service thermal aging of CF8M cast austenitic stainless steel was investigated in materials removed from the steam generator inlet and crossover elbows of the Ringhals 2 pressurized water reactor nuclear power plant unit after approximately 92kh of full operating time. The thermal aging of these materials was investigated using the double loop electrochemical potentiokinetic reactivation method, coupled with indentation hardness measurements and microstructural characterizations, to identify correlations between the electrochemical behavior and traditional methods of investigating thermal aging embrittlement effects in cast stainless steels. While this electrochemical method can be easily employed to quantify thermal aging effects in materials aged at higher temperatures (e.g. greater than 350∘C), this study highlights the difficulties encountered when electrochemically evaluating the aging of materials exposed to nuclear power plant operating conditions.

Firefighting water systems are important safety systems in all industries, including nuclear power plants (NPPs). However, they are susceptible to microbially induced corrosion, which is a degradation mode needing special attention. Leakages were observed in a fire fighting system made from stainless steel at a nuclear power plant shortly after maintenance and modernization work, which included replacement of part of the old carbon steel pipelines with stainless steel pipelines, as well as exchange of some Type 304 stainless steel pipes with Type 316 pipes due to relining parts of the system. The failure analysis revealed sub-surface corrosion cavities with pinholes at the inner surface and finally penetrating the whole pipe wall thickness. It was concluded that the reason for the leaks was due to microbially induced corrosion, (MIC). The paper will present the results from failure analyses, explain the remedial actions taken at the power plant, and discuss the implication of these findings on new similar systems, including the importance of avoiding iron deposits and optimization of water quality.

This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 degrees C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations. (C) 2016 Elsevier B.V. All rights reserved.