Raul Rebak

Worldwide, light water reactors (LWRs) have been using zirconium (Zr)-based alloys for the cladding of the uranium dioxide fuel for more than 6 decades. Zr alloys oxidize rapidly in the presence of water and steam at temperatures > 450°C; therefore, they do not respond well to scenarios of loss of coolant accidents. There is a global effort by nuclear materials technologists to find more robust or stronger cladding materials for LWRs. One option is to use an iron-chromium-aluminum (FeCrAl) alloy since they have excellent resistance to high temperature oxidation and superior mechanical properties at LWR operation temperatures. Results show that (1) FeCrAl alloys have better mechanical properties than Zr alloy and are orders of magnitude more resistant to creep at temperatures higher than LWR normal operation conditions. (2) FeCrAl alloys have better resistance to fretting wear than Zr alloys at the normal operation conditions of LWRs.

Light water reactors (LWR) have been using zirconium (Zr)-based alloys for the cladding of the fuel for more than six decades. Zirconium alloys oxidize rapidly in presence of water and steam at temperatures above 400°C and therefore they do not respond well to scenarios of loss of coolant accident conditions. In the 1950s and 1960s iron-chromium-aluminum (FeCrAl) alloys were explored to replace Zr alloys manly because the former has better oxidation resistance and better mechanical properties than the latter. The use of FeCrAl for LWR fuel cladding has been revived in the last decade following the 2011 accident at the Fukushima Daiichi site. This work will show comparative data to illustrate the advantages of FeCrAl in high temperature scenarios considering oxidation resistance, mechanical properties, and interactions with the fuel pellet.

Because of its resistance to high temperature (>1000°C) attack by steam and good mechanical properties above 400°C, iron-chromium-aluminum alloys (FeCrAl) are considered attractive candidates to replace current zirconium (Zr) based alloys as cladding for light water reactor (LWR) fuels. FeCrAl alloys were sporadically investigated for nuclear applications since the 1950s. Most of the knowledge on these alloys in the literature is for dry environments and temperatures above 1000°C and little information was available for lower temperatures or wet environments. As cladding for LWR, the FeCrAl alloys are expected to survive the nuclear fuel cycle, that it, from manufacturing to used fuel disposal. The presentation will cover the corrosion behavior aspects of FeCrAl from manufacturing using powder metallurgy and welding, to residence in the reactor core for up to ten years (including loss of coolant scenarios), to used fuel residence in cooling pools, to fuel retrievability and reprocessing followed by long term dry cask storage and eventual geologic repository disposal.

One of the main sources of clean electricity in the civilian power grid comes from light water reactors (LWRs). Some of these LWRs are being decommissioned prematurely because they may not be economically sustainable. It is proposed that the use of accident-tolerant fuel concepts may enable for LWR plant operation extension by making them safer and more economically competitive to operate. For example, the use of iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel would not only reduce the risk of catastrophic reaction between the cladding and its environment in the case of a loss of coolant accident situation but also, for example, allow for higher fuel burnup during normal plant operation.

After the Fukushima reactor accidents following Japan's March 2011 tsunami, the U.S. Department of Energy engaged with nuclear fuel vendors to develop improved fuels for the current fleet of light water power reactors. General Electric and Oak Ridge National Laboratory have proposed using iron-chrome-aluminum (FeCrAl) ferritic alloys as cladding material for the existing uranium dioxide fuel (UO2). This is a simple approach that leaves unchanged the present coolable geometry in the reactor. FeCrAl alloys have outstanding resistance, in accident conditions, to attack by superheated steam compared to the current zirconium alloys. Since ferritic FeCrAl alloys have not been used before in nuclear power reactors, extensive characterization has been performed to determine their behavior in light water reactor conditions (e.g., normal operation and accident). The present work deals with the electrochemical behavior of the newer alloys in high-temperature water (near 300°C) containing either excess hydrogen gas or excess oxygen. Results show that chromium-containing ferritic FeCrAl have similar electrochemical high-temperature behavior like other common existing reactor alloys containing chromium for passivation (such as X-750, Alloy 600, and Type 304SS). The use of FeCrAl alloy cladding would also eliminate existing common degradation mechanisms such as shadow corrosion in boiling water reactors.

Since the Fukushima incident in 2011, the nuclear industry has sought to replace Zircaloy fuel cladding with a material which will better withstand a beyond-design-basis incident. A suitable material must possess superior oxidation resistance in high temperature steam (1200-1700°) to withstand a loss of coolant accident, while maintaining good hydrothermal corrosion properties for environmental compatibility during normal operating conditions (288-320°C water). In addition to corrosion issues, a suitable material must also demonstrate sufficient mechanical strength, creep resistance, radiation tolerance, and favorable neutronics.

Researchers at Oak Ridge National Laboratory are engaged in testing of two leading candidates for accident tolerant fuel cladding: iron-chomium-aluminum (FeCrAl) alloys, and silicon carbide ceramic matrix composites (SiC/SiC). FeCrAl alloys have shown excellent oxidation resistance due to the formation of a passive alumina film in high temperature steam, and a protective Fe-Cr spinel layer in water. SiC/SiC has shown excellent corrosive properties in high temperature steam, but a tendency to dissolve in LWR water. To improve performance of SiC normal operating conditions, several candidate coatings are being considered, including Cr, CrN, and TiN.

This work presents results of hydrothermal corrosion experiments on FeCrAl alloys and coated SiC in boiling water reactor hydrogen water chemistry (BWR-HWC) and normal water chemistry (BWR-NWC) (288°C with 150 ppb H₂ or 2 ppm O₂ respectively). FeCrAl samples generally exhibited low corrosion rates. Hydrothermal corrosion properties of current FeCrAl alloys are compared to older generations. CVD-SiC and SiC/SiC exhibited significant mass loss during exposure. SiC samples coated with Cr and CrN adhered well to their substrates, and effectively mitigated dissolution, with acceptable corrosion rates.

This work was funded by the DOE Office of Nuclear Energy Advanced Fuels Campaign

The demonstrated interest of using iron-chrome-aluminum (FeCrAl) ferritic alloys as accident tolerant fuel (ATF) cladding in light water reactors has recently resurfaced the interest in FeCrAl, which have been used for other purposes for more than eight decades. The main interest for ATF is that FeCrAl alloys are many orders of magnitude more resistant to attack in super-heated steam (>1200°C) than zirconium alloys. Recent research has shown that FeCrAl is also resistant to corrosion in high temperature (~300°C) reactor normal operation conditions hydrogenated water since it develops a protective chromia rich oxide film on its surface. Under accident conditions in superheated steam the surface oxide becomes alumina. If an alumina protected tube is flooded with fresh water after the quenching in accident, the surface oxide returns to chromia. The oxide protection versatility of FeCrAl alloys will be discussed on the several situations that may be encountered in a light water reactor both under normal operation conditions and under the unlikely situation of an accident.

Low-alloy steels (LAS) are extensively used in oil and gas (O&G) production due to their good mechanical properties and low cost. Even though nickel improves mechanical properties and hardenability with low penalty on weldability, which is critical for large subsea components, nickel content cannot exceed 1-wt% when used in sour service applications. The ISO 15156-2 standard limits the nickel content in LAS on the assumption that nickel concentrations above 1-wt% negatively impact sulfide stress cracking (SSC) resistance. This restriction excludes a significant number of high-strength and high-toughness alloys, such as Ni-Cr-Mo (e.g., UNS G43200 and G43400), Ni-Mo (e.g., UNS G46200), and Ni-Cr-Mo-V grades, from sour service applications and can be used only if successfully qualified. However, the standard is based on controversial research conducted more than 40 years ago. Since then, researchers have suggested that it is the microstructure that determines SSC resistance, regardless of Ni content. This review summarizes the advantages and disadvantages of nickel-containing LAS in terms of strength, weldability, hardenability, potential weight savings, and cost reduction. Likewise, the state of knowledge on the effect of nickel on hydrogen absorption as well as SSC initiation and propagation kinetics is critically reviewed.

The most common failure mode of commercial nuclear reactor internal components is environmentally assisted cracking (EAC). Environmentally cracking is promoted by the simultaneous effects of three affecting variables, namely (1) presence of tensile stresses and strains, (2) and aggressive environment and (3) a susceptible microstructure. If one or more of these affecting variables are removed, cracking will not occur or propagate. For example, EAC susceptibility and growth rate will be minimized if the residual tensile stresses or strains are controlled. Similarly, EAC may be minimized by controlling the chemistry of the weld metal, for example by increasing the content of chromium in the weld metal. The objective of the current research was to complete laboratory weldments of 76 mm thick plates of Alloy 600 using weld metal with three different compositions. Electron backscatter diffraction (EBSD) results of cross sections of the weld showed that approximately 10 to 15% of residual plastic strain existed at near the root of the weld and in the middle section (at approximately 38 mm); however, lower residual plastic strain existed near the top of the weld (last passes).

Austenitic stainless steels such as type 304 and 316 are susceptible to stress corrosion cracking in high temperature water environments typical of boiling water reactors (BWR) and pressurized water reactors (PWR). The accumulation over time of irradiation dose on the austenitic materials increases further their susceptibility to environmental cracking. Ferritic steels containing chromium are less susceptible to irradiation damage such as void swelling. Ferritic steels also offer desirable higher thermal conductivity and lower thermal expansion coefficient. Little is known however about the stress corrosion cracking behavior of ferritic steels in high temperature water. Crack propagation rate studies were conducted using four types of wrought and welded ferritic steels (5 to 17% Cr) in high purity water at 288°C containing dissolved oxygen or dissolved hydrogen. Results show that the ferritic steels are notably more resistant to environmental assisted cracking than the austenitic materials.

The stress corrosion cracking (SCC) of Alloy 22 in concentrated ground waters was associated to the presence of bicarbonate and chloride ions in the water. SCC occurred when an anodic peak appeared in the polarization curve of Alloy 22. The aim of this work was to investigate further which element in Alloy 22 was responsible for the anodic peak and therefore SCC. Four alloys (22, 800H, 600 and 201) were mainly used for this study. It has been found that Alloy 22, 600 and 800H show anodic peaks; which are affected by chloride and temperature. It is likely that the anodic peak is associated to the presence of either chromium, molybdenum or both.

Alloy 22 is considered as a candidate for engineered barriers of nuclear repositories. Chloride is the only species present in groundwater that is able to promote crevice corrosion, if severe conditions such as high temperatures and a tight crevice are present. Other species present in groundwater have been shown to be inhibitors or non-detrimental species. The objective of this work was to evaluate the efficiency of different species potentially found in groundwaters as possible inhibitors of crevice corrosion of Alloy 22. The crevice corrosion repassivation potential of Alloy 22 was determined in chloride plus inhibitor solutions at 90ºC. The species tested as inhibitors were nitrate, sulfate, carbonate, bicarbonate, chromate, molybdate and tungstate. Nitrate was the most efficient among tested inhibitors. The carbonate was the only species of the carbonate / bicarbonate / carbonic acid equilibrium able to inhibit the chloride-induced crevice corrosion of Alloy 22. Sulfate, chromate and molybdate were moderately good inhibitors.

All the countries that operate commercial nuclear power plants are planning to dispose of the waste in underground geologically stable repositories. The materials being studied for the fabrication of the containers include carbon steel, stainless steel, copper, titanium and nickel alloys. The aim of this work is to review results from research performed using the alloys of interest regarding their resistance to environmentally assisted cracking (EAC) under simulated repository conditions. In general, it is concluded that the environments are mild and that the studied metals may not be susceptible to cracking under the planned emplacement conditions.

Alloy 22 has been extensively studied regarding its crevice corrosion (CC) resistance both in pure chloride solutions and in solutions containing different oxyanions that may act as inhibitors of crevice corrosion. The scope of this work was to study the general and localized corrosion behavior of Alloy 22 when phosphate ions were added to a 1 M NaCl solution at 90°C. Results from the electrochemical tests indicate that the size of the passive potential range and the localized corrosion repassivation potential value increased in the presence of phosphate ions. Results from creviced specimens showed a strong inhibition effect of phosphate ions on the chloride induced crevice corrosion of Alloy 22. The critical molar concentration ratio (R_CRIT = [phosphate]/[Cl]) to inhibit crevice corrosion was 0.3.

There is a current nuclear energy renaissance around the world, mainly in Asia. However, after 60 years of using nuclear power, the issue of nuclear waste still needs to be solved. All the countries currently addressing the issue of nuclear wastes are considering disposing them in stable geologic formations. Most repositories designs in Europe are considering materials such as carbon steel and copper for the container. The US repository in Yucca Mountain is currently on hold waiting for new recommendations from the Blue Ribbon Commission. A robust corrosion and materials testing program was carried out over the last 20 years at universities and National Laboratories to support the Yucca Mountain Project. For example, the intensive corrosion and electrochemical research helped understanding of how the different variables affect the resistance of C-22 to localized corrosion. Even if the repository of Yucca Mountain is not materialized, the knowledge gained on the corrosion behavior of materials is permanent.

Sulfidic corrosion of steels in refineries is a prevalent phenomenon that occurs in oil containing sulfur species between 230°C and 425°C. There are several internal and external variables controlling the occurrence of sulfidic corrosion. The most important external factors are temperature, concentration and type of sulfur species, and presence of naphthenic acid. The most important internal or metallurgical factor to control sulfidic corrosion is the amount of chromium in the steel. The refinery industry relies today in a vast industrial experience on the variables affecting sulfidic corrosion but very little is known on the basic mechanism of attack. There is ample room for research and the basic understanding of this phenomenon.

Alloy 22 (UNS N06022) is a Ni-Cr-Mo alloy, highly resistant to localized corrosion. It has a wide range of industrial applications. Because of its versatility, Alloy 22 has been proposed as a corrosion-resistant barrier for high-level nuclear waste containers. Crevice corrosion is expected to occur in hot chloride solutions if the corrosion potential is equal to or greater than the crevice corrosion repassivation potential. For less-aggressive environments, transpassive dissolution limits the alloy performance at high potentials. The goal of the present work was to establish the effect of potential on the kinetics of chloride-induced crevice corrosion of Alloy 22 at 90°C. Electrochemical tests carried out included cyclic potentiodynamic polarization (CPP) curves, polarizations at constant applied potentials followed by repassivation, and electrochemical impedance spectroscopy (EIS) measurements at anodic applied potentials. The attack profiles were measured using a roughness checker. Crevice corrosion susceptibility of Alloy 22 was independent of bulk solution pH, but it was a strong function of potential. Likelihood of transpassive dissolution was found to increase for high pH solutions. Two different regions were distinguished in polarization tests at a constant potential. A first region was associated with crevice corrosion incubation and a second one was associated with crevice corrosion initiation and propagation. The crevice corrosion current density was estimated as a function of potential. It increased with potential up to a maximum value of approximately 20 mA/cm2. Absence of pitting corrosion was explained in terms of such a small current density, which was insufficient to satisfy a critical value of the product x·i, necessary for pitting corrosion to occur. A diagram of crevice corrosion susceptibility and transpassive dissolution for Alloy 22 in 1 M chloride solutions at 90°C was presented. If the corrosion potential exceeds the crevice corrosion repassivation potential, crevice corrosion will initiate, provided that the diffusion path associated with the crevice is long enough.

Traditionally, the susceptibility of Alloy 22 (N06022) to suffer crevice corrosion has been measured using the Cyclic Potentiodynamic Polarization (CPP) technique (ASTM G 61). When the alloy is not very susceptible to crevice corrosion, the values of repassivation potential obtained using the CPP technique are not highly reproducible. To circumvent the large uncertainty in the values of the repassivation potential by the CPP method, the repassivation potential of Alloy 22 may be measured using a slower method that combines sequentially potentiodynamic, galvanostatic, and potentiostatic treatments (this method is called the Tsujikawa-Hisamatsu Electrochemical or THE method). In the THE method the anodic charge is applied to the specimen in a more controlled manner, which avoids driving the alloy to transpassivity and therefore results in more reproducible repassivation potential values. Results using THE method under various testing conditions are presented. A new standard has been prepared for ASTM balloting for the THE method. The round robin matrix results are also discussed.

Artificially creviced Alloy 22 (N06022) may be susceptible to crevice corrosion in the presence of high-chloride aqueous solutions, especially at higher temperatures and at anodic potentials. The presence of oxyanions in the electrolyte, particularly nitrate, inhibits the nucleation and growth of crevice corrosion. The current results show that crevice corrosion will develop in Alloy 22 when a constant potential above the crevice repassivation potential is applied to a creviced specimen. The analyses of the current output showed the presence of three characteristic domains: (1) passivation or induction time, (2) nucleation and growth, and (3) stifling and arrest. That is, under the tested conditions, crevice corrosion did initiate but after it reached a critical stage of growth, further damage stalled and the output anodic current returned to the passive values before the nucleation of the attack.

It is planned to use the highly corrosion resistant titanium grade 7 (Ti Gr 7) and a high strength titanium alloy (Ti Gr 29) to fabricate the drip shield for the Yucca Mountain repository. Ti Gr 7 contains 0.15% Palladium (Pd) to increase its corrosion performance, mainly under reducing conditions. It was important to determine the corrosion behavior of Ti Gr 7 in concentrated brines at temperatures higher than 100°C, which may represent the behavior of dust deliquescence solutions. Tests were performed in concentrated NaCl + KCl solutions containing also nitrates and fluorides. Results show that Ti Gr 7 was highly resistant to general and localized corrosion. Some specimens were polarized to potentials higher than 4 volts. None of the tightly creviced specimens suffered crevice corrosion. The presence of fluoride promoted localized corrosion around the edges of the crevice former.

The nitrate ion (NO3−) is an inhibitor for crevice corrosion of Alloy 22 (N06022) in chloride (Cl−) aqueous solutions. Naturally formed electrolytes may contain both chloride and nitrate ions. The higher the ratio R = [NO3−]/[Cl−] in the solution the stronger the inhibition of crevice corrosion. Atmospheric desert dust contains both chloride and nitrate salts, generally based on sodium (Na+) and potassium (K+). Some of these salts may deliquescence at relatively low humidity at temperatures on the order of 150°C and higher. The resulting deliquescent brines are highly concentrated and especially rich in nitrate. Electrochemical tests have been performed to explore the anodic behavior of Alloy 22 in high chloride high nitrate electrolytes at temperatures as high as 150°C at ambient atmospheres. Naturally formed brines at temperatures higher than 120°C do not induce crevice corrosion in Alloy 22 because they contain high levels of nitrate. The inhibitive effect of nitrate on crevice corrosion is still active for temperatures higher than 100°C.

Iron-based amorphous alloys possess enhanced hardness and are highly resistant to corrosion, which make them desirable for wear applications in corrosive environments. It was of interest to examine the behavior of amorphous alloys during anodic polarization in concentrated salt solutions and in the salt-fog testing. Results from the testing of one amorphous material (SAM2X5) both in ribbon form and as an applied coating are reported here. Cyclic polarization tests were performed on SAM2X5 ribbon as well as on other nuclear engineering materials. SAM2X5 showed the highest resistance to localized corrosion in 5 M CaCl2 solution at 105°C. Salt fog tests of 316L SS and Alloy 22 coupons coated with amorphous SAM2X5 powder showed resistance to rusting. Partial devitrification may be responsible for isolated pinpoint rust spots in some coatings.

Alloy 22 is a nickel base alloy highly resistant to all forms of corrosion. In very aggressive conditions (e.g. hot concentrated chloride containing brines) Alloy 22 could suffer localized attack, namely pitting and crevice corrosion. The occurrence of localized corrosion in a given environment is governed by the values of the critical potential (Ecrit) for crevice corrosion and the corrosion potential (Ecorr) that the alloy may establish in the studied environment. If Ecorr is equal or higher than Ecrit, localized corrosion may be expected. This paper discusses the evolution of Ecorr of Alloy 22 specimens in 5 m CaCl2 + 5 m Ca(NO3)2 brines at 100°C and 120°C. Two types of specimens were used, polished as-welded (ASW) creviced and non-creviced specimens and as-welded plus solution heat-treated (ASW+SHT) creviced specimens. The latter contained the black annealing oxide film on the surface. Results show that, for all types of Alloy 22 specimens the Ecorr was higher at 120°C than at 100°C, probably because a more protective film formed at the higher temperature. Specimens with the black oxide film on the surface showed more oscillations in the potential. None of the tested specimens suffered crevice corrosion probably because of the relatively high concentration of nitrate in the electrolyte, R = [NO3]/[Cl] = 1.

Double U-bend specimens of Alloy 22 (N06022) and Titanium Grade 7 (R52400) were exposed to a naturally aerated concentrated Basic Saturated Water (BSW) electrolyte at 105°C for over six years. Different type of discoloration of the Ti Gr 7 and Alloy 22 specimens was observed. General Corrosion was minimal and not distinguishable under a scanning electron microscope. None of the tested specimens suffered environmentally assisted cracking (EAC) or localized corrosion under the tested conditions. The specimens retained their residual stress after the long environmental exposure.

The long-term corrosion test facility (LTCTF) at the Lawrence Livermore National Laboratory (LLNL) consisted of 22 vessels that housed more than 7,000 corrosion test specimens from carbon steels to highly corrosion resistant materials such Alloy 22 and Ti Grade 7. The specimens from LTCTF range from standard weight-loss coupons to U-bend specimens for testing susceptibility to environmentally assisted cracking. Each vessel contained approximately 1000 liters of concentrated brines at 60°C or 90°C. The LTCTF started its operations in late 1996. Thousands of specimens were removed from the LTCTF in August-September 2006. The specimens are being catalogued and stored for future characterization. Previously removed specimens (e.g. 1 and 5 years) are also archived for further studies.

Zirconium (Zr) alloys were mostly developed for nuclear power applications. The most common commercial alloys are Zircaloy, which are used as cladding for fuel pellets in water-cooled reactors. These alloys have adequate corrosion resistance in service under irradiation in presence of high-purity water at approximately 300–350°C but in a few cases they may suffer environmental degradation. The most common types of degradation are hydriding, shadow corrosion and nodular corrosion. After the fuel rod bundles are removed from the reactors they are temporarily stored in water pools until their permanent disposition nuclear waste repositories. Simulated laboratory testing and modeling show that the long-term storage in the sealed containers would not cause further damage to the cladding material until the waste containers are breached by corrosion many thousands of years later. And even after water incursion, since the temperature will be low, it is predicted that the cladding material would survive for many thousands of years more, thus delaying the release of the radionuclides to the atmosphere.

Alloy 22 (N06022) has been extensively tested for general and localized corrosion behavior both in the wrought annealed condition and in the as-welded condition. In general, the specimens for laboratory testing are mostly prepared from flat plates of material. It is important to determine if the process of fabricating a container will affect the corrosion performance of this alloy. Thus, specimens for corrosion testing were prepared directly from a fabricated full-diameter Alloy 22 container. Results show that both the anodic corrosion behavior and the localized corrosion resistance of specimens prepared from a welded container were the same as those from flat welded plates.

An iron-based amorphous metal, Fe_49.7Cr_17.7Mn_1.9Mo_7.4W_1.6B_15.2C_3.8Si_2.4 (SAM2X5), with very good corrosion resistance has been developed. This material was prepared as a melt-spun ribbon, as well as gas atomized powder and a thermal-spray coating. During electrochemical testing in several environments, including seawater at 90 °C, the passive film stability was found to be comparable to that of high-performance nickel-based alloys and superior to that of stainless steels, based on electrochemical measurements of the passive film breakdown potential and general corrosion rates. This material also performed very well in standard salt fog tests. Chromium (Cr), molybdenum (Mo), and tungsten (W) provided corrosion resistance, and boron (B) enabled glass formation. The high boron content of this particular amorphous metal made it an effective neutron absorber and suitable for criticality control applications. This material and its parent alloy maintained corrosion resistance up to the glass transition temperature and remained in the amorphous state during exposure to relatively high neutron doses.

The ASTM standard B 265 provides the requirements for the chemical composition of titanium (Ti) alloys. It is planned to use corrosion resistant and high strength titanium alloys to fabricate the drip shield at the proposed Yucca Mountain Repository. Titanium grade (Gr) 7 (R52400) and other Ti alloys are currently being characterized for this application. Ti Gr 7 contains 0.15% Palladium (Pd) to increase its corrosion performance. In this article we report results on the comparative short term corrosion behavior of Ti Gr 7 and a Ruthenium (Ru) containing alloy (Ti Gr 33). Ti Gr 33 also contains a small amount of Pd. Limited electrochemical testing such as polarization resistance and cyclic potentiodynamic curves showed that both alloys have a similar corrosion behavior in the tested environments.

New amorphous-metal thermal-spray coatings have been developed recently that may provide a viable coating option for spent nuclear fuel & high-level waste repositories [Pang et al. 2002; Shinimiya et al. 2005; Ponnambalam et al. 2004; Branagan et al. 2000–2004]. Some Fe-based amorphous-metal formulations have been found to have corrosion resistance comparable to that of high-performance alloys such as Ni-based Alloy C-22 [Farmer et al. 2004–2006]. These materials rely on Cr, Mo and W for enhanced corrosion resistance, while B is added to promote glass formation and Y is added to lower the critical cooling rate (CCR). Materials discussed in this paper include yttrium-containing SAM1651 with CCR ∼ 80 K/s and yttrium-free Formula 2C with CCR ∼ 600 K/s. While nickel-based Alloy C-22 and Type 316L stainless steel lose their resistance to corrosion during thermal spraying, Fe-based SAM1651 and Formula 2C amorphous-metal coatings can be applied with thermal spray processes without any significant loss of corrosion resistance. In the future, such corrosion-resistant thermal-spray coatings may enable the development of less expensive containers for spent nuclear fuel (SNF) and high-level waste (HLW), including enhanced multipurpose containers (MPCs), protected closure welds, and shields to protect containers from drips and falling rocks. These materials are extremely hard and provide enhanced resistance to abrasion and gouges from backfill operations. For example, Type 316L stainless steel has a hardness of approximately 150 VHN, Alloy C-22 has a hardness of approximately 250 VHN, while the Fe-based amorphous metals typically have hardness values of 1100–1300 VHN. Both Formula 2C and SAM1651 have high boron content which allow them to absorb neutrons, and therefore be used for enhanced criticality control. Cost savings can also be realized through the substitution of Fe-based alloy for Ni-based materials. Applications are also envisioned in oil & gas industry.

The ASTM standard B 575 provides the requirements for the chemical composition of Nickel-Chromium-Molybdenum (Ni-Cr-Mo) alloys such as Alloy 22 (N06022) and Alloy 686 (N06686). The compositions of each element are given in a range. For example, the content of Mo is specified from 12.5 to 14.5 weight percent for Alloy 22 and from 15.0 to 17.0 weight percent for Alloy 686. It was important to determine how the corrosion rate of welded plates of Alloy 22 using Alloy 686 weld filler metal would change if heats of these alloys were prepared using several variations in the composition of the elements even though still in the range specified in B 575. All the material used in this report were especially prepared at Allegheny Ludlum Co. Seven heats of plate were welded with seven heats of wire. Immersion corrosion tests were conducted in a boiling solution of sulfuric acid plus ferric sulfate (ASTM G 28 A) using both as-welded (ASW) coupons and solution heat-treated (SHT) coupons. Results show that the corrosion rate was not affected by the chemistry of the materials in the range of the standards.

Alloy 22 (N06022) is the material selected for the fabrication of the outer shell of the nuclear waste containers for the Yucca Mountain high-level nuclear waste repository site. A key technical issue in the waste package program has been the integrity of the container weld joints. The currently selected welding process for fabricating and sealing the containers is the traditional gas tungsten arc welding (GTAW) or TIG method. An appealing faster alternative technique is reduced pressure electron beam (RPEB) welding. It was of interest to compare the corrosion properties of specimens prepared using both types of welding techniques. Standard electrochemical tests were carried on GTAW and RPEB welds as well as on base metal (non-welded) to determine their relative corrosion behavior in simulated concentrated water (SCW) at 90°C (alkaline), 1 M HCl at 60°C (acidic) and 1 M NaCl at 90°C (neutral) solutions. Results show that for all practical purposes, the three tested materials had the same electrochemical behavior in the three tested electrolytes.

The ASTM standard B 575 provides the requirements for the chemical composition of Nickel-Chromium-Molybdenum (Ni-Cr-Mo) alloys such as Alloy 22 (N06022) and Alloy 686 (N06686). The compositions of each element are given in a range. For example, the content of Mo is specified from 12.5 to 14.5 weight percent for Alloy 22 and from 15.0 to 17.0 weight percent for Alloy 686. It is important to determine how the corrosion rate of welded plates of Alloy 22 using Alloy 686 weld filler metal would change if heats of these alloys were prepared using several variations in the composition of the elements even though still in the range specified in B 575. Seven heats of plate were welded with seven heats of wire. Immersion corrosion tests were conducted in a boiling solution of sulfuric acid plus ferric sulfate (ASTM G28 A) using both as-welded (ASW) coupons and solution heat-treated (SHT) coupons. Results show that, for practical purposes, the corrosion rate was not affected by the chemistry of the materials in the range specified in the standard B 575.

The general corrosion rate may be measured using immersion tests or electrochemical tests. The electrochemical tests are fast and can be used for a rapid screening of environmental effects such as temperature and electrolyte composition. The electrochemical tests are described in ASTM standards G 59 and G 102. The basis of these tests is to calculate the resistance to polarization (Rp) in a voltage vs. current plot and to convert these values to corrosion rates using the Faraday law. Commercial software can calculate the corrosion rate based on inputs from the operator. This paper discusses three ways of calculating the corrosion rate (Methods 1, 2 and 3) based on a fixed set of acquired data of voltage vs. current. The conclusions are that the way the corrosion rate is calculated does not impact greatly on the absolute value of the corrosion rate. Variations in the acquired data (current, potential) from one experiment to another seem more important that the manner the data is fitted with the Rp slope.

Alloy 22 (N06022) has been extensively tested for general and localized corrosion behavior both in the wrought and annealed condition and in the as-welded condition. The specimens for testing were mostly prepared from flat plates of material. It was important to determine if the process of fabricating a full diameter Alloy 22 container will affect the corrosion performance of the alloy. Specimens were prepared directly from a fabricated container and tested for corrosion resistance. Results show that both the anodic corrosion behavior and the localized corrosion resistance of specimens prepared from a welded fabricated container was the same as from flat welded plates.

When metallic plates are welded, residual tensile stresses may develop in the vicinity of the weld seam. Processes such as Low Plasticity Burnishing (LPB) and Laser Shock Peening (LSP) could be applied locally to eliminate the residual stresses produced by welding. In this study, Alloy 22 (N06022) plates were welded and then the above-mentioned surface treatments were applied to eliminate the residual tensile stresses. The aim of the current study was to compare the corrosion behavior of as-welded (ASW) plates with the corrosion behavior of plates with stress mitigated surfaces. Immersion and electrochemical tests were performed. Results show that the corrosion resistance of the mitigated plates was not affected by the surface treatments applied.

The ASTM standard B 575 provides the requirements for the chemical composition of Nickel-Chromium-Molybdenum (Ni-Cr-Mo) alloys such as Alloy 22 (N06022) and Alloy 686 (N06686). The compositions of each element are given in a range. For example, the content of Mo is specified from 12.5 to 14.5 weight percent for Alloy 22 and from 15.0 to 17.0 weight percent for Alloy 686. It was important to determine how the corrosion rate of welded plates of Alloy 22 using Alloy 686 weld filler metal would change if heats of these alloys were prepared using several variations in the composition of the elements even though still in the range specified in B 575. All the material used in this report were especially prepared at Allegheny Ludlum Co. Seven heats of plate were welded with seven heats of wire. Immersion corrosion tests were conducted in a boiling solution of sulfuric acid plus ferric sulfate (ASTM G 28 A) using both as-welded (ASW) coupons and solution heat-treated (SHT) coupons. Results show that the corrosion rate was not affected by the chemistry of the materials within the range of the standards.

Alloy 22 (N06022) is a nickel-based alloy highly resistant to corrosion. In some aggressive conditions of high chloride concentration, temperature and applied potential, Alloy 22 may suffer crevice corrosion, a form of localized corrosion. There are several electrochemical methods that can be used to determine localized corrosion in metallic alloys. One of the most popular for rapid screening is the cyclic potentiodynamic polarization (CPP). This work compares the results obtained by measuring the localized corrosion resistance of Alloy 22 using both the CPP and the more cumbersome Tsujikawa-Hisamatsu Electrochemical (THE) method. The electrolytes used were 1 M NaCl and 5 M CaCl2, both at 90°C. Results show that similar repassivation potentials were obtained for Alloy 22 using both methods. That is, in cases where localized corrosion is observed using the faster CPP method, there is no need to use the THE method since it takes ten times longer to obtain comparable results in spite of the mode of corrosion attack is sometimes different in the tested specimens.

Boron containing stainless steels are used in the nuclear industry for applications such as spent fuel storage, control rods and shielding. It was of interest to compare the corrosion resistance of three borated stainless steels with standard austenitic alloy materials such as type 304 and 316 stainless steels. Tests were conducted in three simulated concentrated ground waters at 90°C. Results show that the borated stainless were less resistant to corrosion than the witness austenitic materials. An acidic concentrated ground water was more aggressive than an alkaline concentrated ground water.

Nickel (Ni) can dissolve a large amount of alloying elements while still maintaining its desirable austenitic microstructure. The resulting alloys are generally divided in families depending on the type of alloying elements they contain. Each one of these families is aimed to specific applications. Corrosive environments in industrial applications are generally divided for example in reducing acids, oxidizing acids, contaminated acids, caustic environments, oxidizing salts, etc. Depending on the application and the environment (electrolyte composition and temperature) several or single alloys may be recommended to fabricate components. The Ni-chromium-molybdenum (Ni-Cr-Mo) series contains a balanced selection of beneficial alloying elements so it can handle a variety of aggressive environments. By design, Alloy 22 or N06022 is one of the most versatile corrosion resistant nickel alloys since it has outstanding corrosion resistance both in reducing and oxidizing conditions.

Alloy 22 (UNS N60622) is a nickel-based alloy, which is extensively used in aggressive industrial applications, especially due to its resistance to localized corrosion and stress corrosion cracking in high chloride environments. The purpose of this work was to characterize the anodic behavior of Alloy 22 in concentrated calcium chloride (CaCl2) brines and to evaluate the inhibitive effect of nitrate, especially to localized corrosion. Standard electrochemical tests such as polarization resistance and cyclic polarization were used. Results show that the corrosion potential of Alloy 22 was approximately −360 mV in the silver-silver chloride (SSC) scale and independent of the tested temperature. Cyclic polarization tests showed that Alloy 22 was mainly susceptible to localized attack in 5 M CaCl2 at 75°C and higher temperatures. The addition of nitrate in a molar ratio of chloride to nitrate equal to 10 increased the onset of localized corrosion to approximately 105°C. The addition of nitrate to the solution also decreased the uniform corrosion rate and the passive current of the alloy.

Yucca Mountain (Nevada) is designated as a high-level nuclear waste repository. The nuclear waste will be isolated by a series of engineered barriers. The metallic engineered barriers will consist of a double-wall container with a detached drip shield. The material for the external wall of the container is Alloy 22, a corrosion-resistant Ni-Cr-Mo alloy. Titanium grade 7 has been proposed for the drip shield. Ti alloys are highly resistant to all forms of corrosion due to the formation of a stable, protective and strongly adherent oxide film. The aim of this research was to characterize the general and localized corrosion behavior of Ti Gr 7, 16 and 12 in simulated concentrated ground waters. Welded and non-welded coupons were exposed for up to 5 years to the vapor and liquid phases of acidic and alkaline multi-ionic solutions at 60°C and 90°C. This paper describes the results obtained after approximately 2-1/2- to 5-1/2-year exposure to the testing electrolyte solutions. In general, the highest corrosion rate was obtained for Ti Gr 12; however, in all of the tested conditions, the corrosion rate was generally lower than 100 nm/yr. For all alloys, the highest corrosion rate was obtained in the concentrated alkaline solution.

In its current design, the high-level nuclear waste container includes an external layer of Alloy 22 (Ni-22Cr-13Mo-3W-3Fe). Since the containers may be exposed to multi-ionic aqueous environments over their lifetime, a potential degradation mode of the outer layer could be environmentally assisted cracking (EAC). The objective of the current research is to characterize the effect of applied potential and temperature on the susceptibility of Alloy 22 to EAC in simulated concentrated water (SCW) using the slow strain rate test (SSRT). Results show that Alloy 22 may suffer EAC at applied potentials approximately 400 mV more anodic than the corrosion potential (Ecorr).

Alloy 22 (UNS N06022) is a candidate material for the external wall of the high-level nuclear waste containers for the potential repository site at Yucca Mountain. In the mill-annealed (MA) condition, Alloy 22 is a single face centered cubic phase. When exposed to temperatures on the order of 600°C and above for times higher than 1 h, this alloy may develop secondary phases that reduce its mechanical toughness and corrosion resistance. The objective of this work was to age Alloy 22 at temperatures between 482°C and 760°C for times between 0.25 h and 6,000 h and to study the mechanical and corrosion performance of the resulting material. Aging was carried out using wrought specimens as well as gas tungsten arc welded (GTAW) specimens. Mechanical and corrosion testing was carried out using ASTM standards. Results show that the higher the aging temperature and the longer the aging time, the lower the impact toughness of the aged material and the lower its corrosion resistance. However, extrapolating both mechanical and corrosion laboratory data predicts that Alloy 22 will remain corrosion resistant and mechanically robust tbr the projected lifetime of the waste container.

As part of proposed geological repository at Yucca Mountain, Nevada, Alloy 22 (Ni-22Cr-13Mo-3W-3Fe) has been chosen as the candidate material for a 2-cm outer layer on the high-level nuclear waste containers. During the repository period, the container materials will be subject to corrosion due to their exposure to multi-ionic aqueous environments. Although Alloy 22 has demonstrated excellent corrosion resistance, accumulation of a small, yearly corrosion rate for 10,000 or more years can be significant. When subjected to the conventional weight loss technique for corrosion studies, Alloy 22 requires many years to demonstrate a detectible weight loss. The goal of this research is to seek alternative techniques to determine a reasonably confident corrosion rate. This paper will discuss the latest experimental results using the potentiostatic technique to determine passive dissolution rates.