Supercritical water-cooled reactor candidate materials are typically corrosion-resistant alloys whose mass changes in supercritical water fluctuate around zero. A previous interlaboratory corrosion experiment exercise revealed large scatter in mass change results between laboratories. Here, we reduced systemic differences between laboratories by unilateral preparation of coupons and unilateral chemical cleaning. Type 310S stainless steel and Alloy 800HT test coupons were exposed for 10 0 0 h to 550 degrees C, 25 MPa, deaerated water. Average mass loss for 310S and 800HT was 38 +/- 26 and 51 +/- 31 mg/dm(2), respectively. Differences in mass transfer, galvanic and local corrosion, and average coupon temperature may explain the poor reproducibility. (c) 2022 Published by Elsevier B.V.

All in-core components in an SCWR will experience irradiation by α and β particles, neutrons and high-energy photons (γ-rays) resulting in damage at the atomic level in the form of ionization and microstructural degradation due to the development of vacancies, interstitials and voids. These microscopic defects induce changes in physical and mechanical properties such as hardening, ductility, swelling, radiation-induced segregation, and creep, and can increase the risk of cracking. In combination with thermal creep, these changes are a major factor in determining long-term component reliability. This chapter discusses the various forms of radiation damage relevant to SCWR concepts, as well as discussing thermal creep of candidate SCWR materials.

Environmentally assisted cracking (EAC) is a complex phenomenon driven by the synergistic interaction of mechanical, chemical and metallurgical factors. The complex interplay between causative factors makes experimental measurements difficult, and the state of knowledge on EAC under supercritical water-cooled reactor (SCWR) conditions is not as well advanced as that of general corrosion. This chapter discusses the effects of the three key causative factors (environment, material, and mechanical) on the occurrence of EAC in supercritical water, focussing on candidate SCWR alloys and expected SCWR in-core conditions. Possible differences in mechanisms in the near-critical and higher temperature regimes are highlighted.

The Canadian super critical water-cooled reactor (SCWR) concept requires materials to operate at higher temperatures than current generation III water-cooled reactors. Materials performance after radiation damage is an important design consideration. Materials that are both corrosion resistant and radiation damage tolerant are required. This paper summarizes the operating conditions including temperature, neutron flux, and residence time of in-core Canadian SCWR components. The focus is on the effects of irradiation on in-core components, including those exposed to a high neutron flux in the fuel assembly, the high pressure boundary between coolant and moderator, as well as the low-temperature, low-flux calandria vessel that contains the moderator. Although the extreme conditions and the broad range of SCWR in-core operating conditions present significant materials selection challenges, candidate alloys that can meet the performance requirements under most in-core conditions have been identified. However, for all candidate materials, insufficient data are available to unequivocally ensure acceptable performance and experimental irradiations of candidate core materials will be required. Research programs are to include out-of-pile tests on un-irradiated and irradiated alloys. Ideally, in-flux studies at appropriate temperatures, neutron spectrum, dose rate, duration, and coolant chemistry will be required. Characterization of the microstructure and the mechanical behavior including strength, ductility, swelling, fracture toughness, cracking, and creep on each of the in-core candidate materials will ensure their viability in the Canadian SCWR.

In any power cycle using high-temperature water as the heat transfer medium, water chemistry has a major influence on materials degradation, and it is necessary to understand those water properties that influence material degradation processes. In particular, for any supercritical water-cooled reactor concept the properties of supercritical water directly affect general and localized corrosion (such as environmentally assisted cracking) of system materials and the transport of corrosion products to and from the core. The deposition of corrosion products on surfaces can affect heat transfer to the coolant leading to higher metal temperatures, which can also affect bulk degradation mechanisms such as creep. This chapter starts by examining the question “What is supercritical water?”. The discussion then turns to feedwater chemistry, the release and transport of radioactive material out of the core, and water radiolysis. Preliminary water chemistry specifications are then presented. The chapter ends with a brief introduction to molecular dynamics simulations of the structure of SCW at surfaces, which serves as an introduction to Chapters 5 and 6Chapter 5Chapter 6.

Corrosion behavior of alloy 316L stainless steel capsule was studied by exposure to the supercritical water (SCW) at 500°C and 25MPa for 20,000h. The microstructural observations have been conducted on the cross section of the exposed surfaces to the SCW to perceive the internal oxidation of the grains and/or grain boundaries. Transmission electron microscopy (TEM) observations as well as elemental analyses such as energy dispersive spectroscopy (EDS) and Electron energy loss spectroscopy (EELS) were used to study the internal oxidation and micro-crack initiation on the surface. Elemental analyses indicated that long-term exposure to the SCW resulted in formation of scales identified as Fe3O4 (outer layer), Fe-Cr spinel/(Fe,Ni)Cr2O4/(Mn,Cr)2O3/SiO2 (inner layer) on the substrate, and Ni-enrichment (chrome depleted region) in the alloy 316L. Micro-crack initiation was observed ahead of the oxidized grain boundaries in which elemental enrichments happened ahead of the crack tip. The relevance of the observed oxidation phenomena on the crack susceptibility of Alloy 316L was discussed. Finally, prolonging the exposure time up to 20,000h has shown that the alloy 316L might be susceptible to micro-crack initiation in the supercritical water.

The various supercritical water-cooled reactor concepts being developed under the Generation IV International Forum are the natural evolution of the water-cooled reactor technology that has successfully supplied the majority of nuclear-based electricity since the dawn of commercial nuclear power generation. The materials challenges that must be addressed in the development of a supercritical water-cooled reactor are in most respects the same as those experienced by the current generations of water-cooled reactors. This chapter summarizes current knowledge of corrosion and environmentally assisted cracking phenomena under the conditions expected in the core of a supercritical water-cooled reactor, with an emphasis on recent advances in experiment and modeling.

Oxidation behavior of alloy 316L stainless steel static capsules were investigated by long-term exposure to the supercritical water at 500 degrees C and 25 MPa for various exposure times. It was found that the thickness and weight gain vs. exposure time followed parabolic law. Elemental and phase analyses indicated that long-term exposure to the SCW resulted in formation of scales identified as Fe3O4 (outer layer), Mn-Fe-Ni-Cr spinel (inner layer) on the substrate, and Ni-enrichment (chrome depleted region) in the alloy 316L. Oxidation phenomena and scales grown on the Alloy 316L as the result of long term exposure to the supercritical water was discussed.

Investigation of the role of irradiation on the corrosion resistance of structural alloys is of vital importance for selection of supercritical water-cooled reactor (SCWR) materials. Gamma heating under SCWR conditions, which induces enhancement of radiolysis and corrosion kinetics at interfaces, can be efficiently simulated by electron beam irradiation over a wide range of deposited dose and temperature. The NSC KIPT-sited Canada–Ukraine Electron Irradiation Test Facility (CU-EITF) still remains the only operating facility capable of in situ irradiation of specimens in a supercritical water (SCW) natural circulation loop. This paper reports the results of postirradiation studies of Zr–1%Nb and Ni–Cr Inconel 690/52MSS alloys after a ~500-h-long exposure in the CU-EITF in the near-critical (23.5 MPa/360–385°C) regime. Results of scanning electron microscopy (SEM) studies of the sample microstructure are presented along with those of the electron-irradiated loop piping, SS X18H10T. The results of corrosion tests under electron-irradiation are correlated to the calculated three-dimensional (3D) fields of absorbed dose and temperature and to the reference data obtained in-pile for topical materials. The paper also discusses the prospects for the use of the CU-EITF facility within a cooperative SCWR program and presents an outlook of the facility development.

Monte Carlo track chemistry simulations of the low linear energy transfer radiolysis of pure, deaerated liquid water have been used in combination with a spherical spur model to examine the effect of temperature on the in situ formation of H3O+ ions and the corresponding abrupt transient "acid-spike" response that is observed after irradiation. The magnitude and duration of this acid-spike effect were found to be a very sensitive function of temperature. At 25 degrees C, it is most intense at times less than similar to 1 ns, with a pH of 3.3 remaining nearly constant. In contrast, at higher temperatures, there is an increasingly acidic but much shorter pH response. At 350 degrees C, the pH is around 1.7 on a time scale of similar to 10 ps. At longer times, the pH gradually increases for all temperatures, ultimately reaching a constant value corresponding to the non-radiolytic, pre-irradiation concentration of H3O+ arising through water's autoprotolysis at similar to 1-10 mu s following irradiation. It does not appear that this transient acidspike effect has been explored in water subject to ionizing radiation, either at ambient or at elevated temperatures. As many in-core processes in a water-cooled nuclear reactor critically depend on pH, the present work raises the question whether such abrupt highly acidic pH variations contribute to material corrosion and damage.

This study investigates the oxidation behavior of several stainless steels and nickel-based superalloys exposed to supercritical water at 800°C for 12h. Characterization of the resulting oxide layers were conducted using weight change measurements, X-ray diffraction, scanning/transmission electron microscopy, and energy dispersive spectroscopy. Although the exposure time is only 12h, the thickness of the oxide layers formed was as high as 1μm, comprising different spinel structures. The influence of alloying elements such as Al, Nb, Mo, Mn and Ti on the corrosion behavior is investigated and possible corrosion mechanisms for each candidate alloy are discussed.

The effect of high-temperature microstructure degradation (thermal ageing) on the corrosion resistance of austenitic stainless steels in supercritical water (SCW) was evaluated in this study. Mill-annealed (MA) and thermally treated (TT) samples of Type 316L and Type 310S stainless steel were exposed in 25 MPa SCW at 550°C with 8 ppm dissolved oxygen in a flowing autoclave testing loop. The thermal treatments applied to Type 316L (815°C for 1000 hr + water quench) and Type 310S (800°C for 1000 hr + air cool) were successful in precipitating the expected intermetallic phases in each alloy, both within the grains and on the grain boundaries. It was found that a prolonged time at relatively high temperature was sufficient to suppress significant compositional variation across the various intermetallic phase boundaries. This paper presents the results of the gravimetric analysis and oxide scale characterization using scanning electron microscopy (SEM) coupled with X-ray energy-dispersive spectroscopy (EDS). The role played by the fine precipitate structure on formation of the oxide scale, and thus corrosion resistance, is discussed. The combined role of dissolved oxygen and flow (revealed by examining the differences between Type 316L samples exposed in a static autoclave and in the flowing autoclave loop) is also addressed. It was concluded that formation of intermetallic phase precipitates during high-temperature exposure is not likely to have a major effect on the apparent corrosion resistance because of the discontinuous nature of the precipitation.

This study concentrated on the investigation of oxide scale grown on alloys 304 and 304-oxide dispersion strengthened steel in supercritical water (650 degrees C/25 MPa) environment. The corrosion rate was evaluated by measuring the weight change of the samples and by cross-section examinations. Results showed that weight gains vs. supercritical exposure time follows a parabolic law for alloy 304-ODS contrary to alloy 304 which possibly follows a cubic rate law. The general weight gain after 550 h exposure to the SCW was 131.8 and 621.6 mg/dm(2) for alloys 304-ODS and 304, respectively. Electron microscopy observations and elemental analyses as well as X-ray diffraction and time-of-flight secondary ion mass spectrometry results revealed that the oxide scales formed on 304-ODS alloy composed of three distinct layers including Fe3O4 (outer layer), FeCr2O4/(Fe,Cr)(2)O-3 spinel structures (inner layer) and Cr2O3 (transition layer). However, alloy 304 had two layers including Fe3O4 (outer layer) and Fe-Cr spine] structures (inner layer). It is believed that the formation of Cr2O3 layer at the diffusion layer/metal interface becomes the rate limiting step for oxide advancement, since this change in oxide structure also corresponds to a decrease in corrosion rate in the alloy 304-ODS compared to alloy 304. Based on the results and observations, the oxidation mechanisms are discussed. (C) 2016 Elsevier B.V. All rights reserved.

This paper addresses some of the overarching aspects of microstructure instability expected from both high temperature and radiation exposure that could affect the corrosion and stress corrosion cracking (SCC) resistance of the candidate austenitic Fe-Cr-Ni alloys being considered for the fuel cladding of the Canadian supercritical water-cooled reactor (SCWR) concept. An overview of the microstructure instability expected by both exposures is presented prior to turning the focus onto the implications of such instability on the corrosion and SCC resistance. Results from testing conducted using pre-treated (thermally-aged) Type 310S stainless steel to shed some light on this important issue are included to help identify the outstanding corrosion resistance assessment needs.