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DOE OSTI2016

Hanford Double Shell Waste Tank Corrosion Studies - Final Report FY2015

During FY15, SRNL performed corrosion testing that supported Washington River Protection Solutions (WRPS) with their double shell tank (DST) integrity program. The testing investigated six concerns including, 1) the possibility of corrosion of the exterior of the secondary tank wall; 2) the effect of ammonia on vapor space corrosion (VSC) above waste simulants; 3) the determination of the minimum required nitrite and hydroxide concentrations that prevent pitting in concentrated nitrate solutions (i.e., waste buffering); 4) the susceptibility to liquid air interface (LAI) corrosion at proposed stress corrosion cracking (SCC) inhibitor concentrations; 5) the susceptibility of carbon steel to pitting in dilute solutions that contain significant quantities of chloride and sulfate; and 6) the effect of different heats of A537 carbon steel on the corrosion response. For task 1, 2, and 4, the effect of heat treating and/ or welding of the materials was also investigated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Comparison of the Phase Stability and Corrosion Resistance of the Ni-Based Alloys C-4 and C-276

A one-to-one correspondence is developed between the phase stability and corrosion resistance of the commercial (Ni-Cr-Mo)-based alloys C-276 and C-4. Alloy C-4 is shown to outperform alloy C-276 in acidic oxidizing media and approaches its resistance in acidic reducing media. Both alloys are found to have similar resistance to chloride-induced stress corrosion cracking in the annealed condition with short-range order. The aqueous corrosion resistance of both alloys is found to be improved by long-range order; however, they become highly susceptible to chloride-induced stress corrosion cracking. Alloy C-4 is distinguished by being stable toward precipitation of detrimental intermetallic compounds. However, alloy C-276 is shown to be prone to precipitation of mu phase which degrades its resistance to aqueous corrosion and chloride-induced stress corrosion cracking. It is concluded that the chemical composition of alloy C-4 is tailored to make it more versatile than alloy C-276 in acidic oxidizing media and the results are found to be consistent with the concept of atomic percent factor of each alloy.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Three dimensional microstructural characterization of nanoscale precipitates in AA7075-T651 by focused ion beam (FIB) tomography

The size and distribution of precipitates in Al 7075 alloys affects both the mechanical and corrosion behavior (including stress corrosion cracking and fatigue corrosion) of the alloy. Three dimensional (3D) quantitative microstructural analysis of Al 7075 in the peak aged condition (T651) allows for a better understanding of these behaviors. In this study, Focused ion beam (FIB) tomography was used to characterize the microstructure in three dimensions. Analysis of grains and precipitates was performed in terms of volume, size, and morphology. It was found that the precipitates at the grain boundaries are larger in size, higher in aspect ratios and maximum Feret diameter compared to the precipitates inside the grains, due to earlier nucleation of the precipitates at the grain boundaries. Our data on the precipitates at the interface between grains and Mg{sub 2}Si inclusion show that the surfaces of inclusion (impurity) particles can serve as a location for heterogeneous nucleation of precipitates. - Highlights: •Focused ion beam (FIB) tomography was used to characterize the microstructure of Al 7075 in three dimensions. •Analysis of grains and precipitates was performed in terms of volume, size, and morphology. •Precipitates at the grain boundaries have larger size and aspect ratio compared to the precipitates inside the grains.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage

This final report is summarized four years of work of awarded project number DE-NE0008442\IRP-15-9318, title as: “Innovative Approach to SCC Inspection and Evaluation of Canister in Dry” supported by Nuclear Energy University Program\Integrated Research Program, NEUP\IRP, where last year is an extension of no cost project. This research work is focused on the possible degradation of dry storage canister holding used nuclear fuel (UNF), due to exposure of marine atmospheric environment. The canister usually designed from a welded austenitic stainless-steel of types 304 or 316, encased in a concrete shell known as an overpack. This canister designed and licensed for 20 years with an extension period need to be determined, in addition 20 years or more. These casks are located throughout the country in Independent Spent Fuel Storage Installations (IFSFIs) often on the site close to nuclear power plants, that are often located in coastal or lake-side regions where a marine atmospheric salt air environment exist. As a result of several recent decisions, such as not pursuing Yucca Mountain as a long-term geologic repository storage facility for UNF and the cancellation of the construction of a reprocessing facility in the 1980s due to proliferation risks, interim dry storage facilities for long-term storage are vital option for the immediate future of nuclear energy technology. One of the primary concerns with respect to the long-term performance of the storage casks is the potential for corrosion initiation due to deliquescence of salts deposited on the canister surface as aerosols; in regions of high residual weld stresses, this may lead to localized stress corrosion cracking (SCC). Dust and aerosols in the air being drawn though ventilation openings in the overpacks of passively ventilated dry canister storage systems may be deposited on the stainless-steel canister outer surfaces. Under these conditions, localized corrosion attack can occur. Chlorides ions deliquescence on the canister surface, a process by which the salts essentially pull moisture out of the air, creating a concentrated salt brine on the canister surface. Chloride-induced stress corrosion cracking (CISCC) of welded zones is of special concern, as it a well-documented mode of attack for austenitic stainless steels marine environments, and many independent spent fuel storage installations (ISFSIs) are located in coastal areas. Recent canister inspections have shown that chloride salts are present on the surface of in-service canisters at near-marine settings. However, canister surface inspections of satisfactory resolution to detect SCC have never been carried out in details, because access to the canister surfaces through vents in the overpacks is extremely limited, and high radiation fields make removal of the canisters from the overpacks undesirable. Although no significant corrosion damage has not been detected up to date, there is concern that this may lead to pitting corrosion, which could in turn develop into stress corrosion cracking, causing canister structural failure. The initiation and propagation of pitting corrosion has been widely studied in full immersion conditions, but limited studies have been done on marine atmospheric conditions, which is a focus of this work, these includes laboratory and field experimentally and propagation rate modeling of pitting and transition to crack, and crack propagation rate. It is known, however, that the rate of atmospheric corrosion can be affected by variations in relative humidity (RH), salt density, salt type, droplet area, temperature and stainless-steel microstructure and residual stress particularly in welded zones of canisters. The integrated research program has identified potential deterioration mechanisms for the stainless-steel canisters containing the UNF in dry storage that require detailed research and investigation and will have an impact on the performance of long-term interim storage under the normal and extreme environmental conditions experienced during the duration of this storage. The prediction and monitoring of canister corrosion processes while in storage can provide important data for the assessment of interim storage performance and the safety to the public. This report comprises of mainly the published results in different journals and covering a wide range spectrum from susceptible stainless-steel types 304\316 microstructural analysis through the implant test at atmospheric environment, residual stress analysis of mockup canister and simulation, inspection methodology includes nonlinear ultrasound and vibro-thermography and stochastic modeling and simulation of pits propagation rates and transition to cracks. Description of field specimens installed at ISFSI of Main Yankee is given in the following chapters and will be analyzed in the coming future and serves as an important data for a stochastic analysis as it was drawn from the publications and reports described in this report. The purpose of this report is to bring together the project efforts that includes improving in our understanding of structural aging performance at current dry storage casks fleet at ISFSIs in the context of remedial maintenance and inspection applications technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Pitting behavior of friction stir repair-welded 304L stainless steel in 3.5% NaCl solution at room temperature: role of grain and defect structures

Stainless steel (SS) canisters used for storing spent nuclear waste are prone to chloride-induced stress corrosion cracking. Friction stir welding (FSW), being a low heat input process, can be used for repair welding of these canisters. In this study, an artificial crack was introduced on the 304L SS coupons by an electric discharge machining process. The artificial crack was repair-welded by FSW by maintaining the tool temperature a constant at two levels (725 and 825 °C). Friction stirring significantly reduced the grain size of the stirred zone (SZ) from about 47 to 2–4 μm. The fraction of low-angle grain boundaries increased in the SZ from 2 to 37–43%. Here, on the other hand, the fraction of special grain boundaries (Σ3 and Σ9) that was ~ 50% in the base metal reduced to < 10% in the SZ. During friction stirring, the oxide layer of artificial crack was broken and aligned to form a spiral defect called lazy-S structure. All these microstructural changes affected the corrosion behavior of the FSW specimens when tested in 3.5% NaCl at room temperature using cyclic polarization, chronoamperometry, and electrochemical impedance spectroscopy. The FSW specimens showed lower polarization resistance and lower transpassive potential than those of the base metal specimens. However, the pitting protection potentials of the FSW specimens were higher than that of the base metal. The pitting behavior of the FSW specimens was influenced more by the preferential attack on the lazy-S region than by passive film breakdown. The flat band potentials of the passive film formed on the FSW specimens were more positive than that of base metal. The charge carrier density of passive film formed on the FSW specimens was higher than that of the base metal. The higher fraction of low-angle grain boundaries present in the FSW specimens could supply more number of misorientation dislocations at the metal/film interface which could form anion vacancies by a climb process leading to formation of oxide at these locations without stressing the substrate. Therefore, low-angle boundaries are considered helpful for formation of stress-free passive films that will be highly stable and enhance both pitting and stress corrosion cracking resistance.

Dry storage canister↗

DOE OSTI2018

Environmental Cracking and Irradiation Resistant Stainless Steels by Additive Manufacturing

Metal additive manufacturing (AM), or metal 3D printing is an emergent advanced manufacturing method that can create near net shape geometries directly from computer models. This technology can provide the capability to rapidly fabricate complex parts that may be required to enhance the integrity of reactor internals components. Such opportunities may be observed during a plant refueling outage and AM parts can be rapidly custom designed, manufactured and deployed within the outage interval. Additive manufacturing of stainless steel (SS) components can add business benefits on fast delivery on repair hardware, installation tooling, new design prototypes tests, etc. For the nuclear industry, the supply chain is always an issue for reactor service. AM can provide through-life supply chain (40-60 years) for high-value low-volume components. In the meantime, the capability of generating complex geometries and functional gradient materials will improve the performance, reduce the overall component cost, plant asset management cost and increase the plant reliability by the improvement in materials performance in nuclear environments. While extensive work has been conducted regarding additively manufacturing of austenitic SS parts, most efforts focused only on basic attributes such as porosity, residual stress, basic tensile properties, along with components yield and process monitoring. Little work has been done to define and evaluate the material requirements for nuclear applications. Technical gaps exist, which limit this technology adoption in the nuclear industry, which includes high manufacturing cost, unknown risks, limited nuclear related data, lack of specification and qualification methods, and no prior business experience. The main objective of this program was to generate research data to address all these technical gaps and establish a commercial practice to use AM technology in the nuclear power industry. The detailed objectives are listed as follows: (1) Evaluate nuclear related properties of AM 316L SS, including microstructure, tensile properties, impact toughness, stress corrosion cracking (SCC), corrosion fatigue (CF), irradiation effects, and irradiation assisted stress corrosion cracking (IASCC). (2) Understand the correlations among laser processing, heat treatment, microstructure and SCC/irradiation properties; (3) Optimize and improve the manufacturing process to achieve enhanced nuclear application properties; (4) Fabricate, evaluate, qualify and test a prototype reactor component to demonstrate the commercial viability and cost benefit; (5) Create regulatory approval path and commercialization plans for the production of a commercial reactor component.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Assessing the release, transport, and retention of radioactive aerosols from hypothetical breaches in spent fuel storage canisters

Interim dry storage of spent nuclear fuel involves storing the fuel in welded stainless-steel canisters. Under certain conditions, the canisters could be subjected to environments that may promote stress corrosion cracking leading to a risk of breach and release of aerosol-sized particulate from the interior of the canister to the external environment through the crack. Research is currently under way by several laboratories to better understand the formation and propagation of stress corrosion cracks, however little work has been done to quantitatively assess the potential aerosol release. The purpose of the present work is to introduce a reliable generic numerical model for prediction of aerosol transport, deposition, and plugging in leak paths similar to stress corrosion cracks, while accounting for potential plugging from particle deposition. The model is dynamic (changing leak path geometry due to plugging) and it relies on the numerical solution of the aerosol transport equation in one dimension using finite differences. The model’s capabilities were also incorporated into a Graphical User Interface (GUI) that was developed to enhance user accessibility. Model validation efforts presented in this paper compare the model’s predictions with recent experimental data from Sandia National Laboratories (SNL) and results available in literature. We expect this model to improve the accuracy of consequence assessments and reduce the uncertainty of radiological consequence estimations in the remote event of a through-wall breach in dry cask storage systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Inspection of Dry Storage Casks Using PGAA and NAA Techniques

In this effort, an assessment of a chloride detection method is performed for inspecting the surfaces of dry cask storage system (DCSS) canisters, with focus on the implementation of a non-destructive nuclear analysis and methodology. Some DCSS canisters (especially those located in coastal environments) will be exposed to marine environments, which can cause atmospheric stress corrosion cracking (SCC). Chloride deposits on canister surfaces provide the corrosive environment necessary for atmospheric SCC with the residual stresses and materials properties conditions. The most susceptible regions on the canister surface are the weld areas, or where cold work has been performed, and may provide the necessary driving force for crack growth. The heat-affected zones (HAZs) near welds are also potentially susceptible to atmospheric SCC due to sensitization of the material caused by chromium depletion during the welding process. Information collected from the field and from laboratory studies has not been able to rule out the possibility of atmospheric chloride induced stress corrosion cracking (CISCC), although no occurrences of atmospheric SCC in DCSS canisters have been detected to-date. A number of researchers have demonstrated that localized corrosion can take place under such situations on materials commonly used for interim storage containers, including 304SS (Cook et al. 2010; Shirai et al. 2011) and 316SS (Tani et al. 2009). In this summary, the preliminary results of a neutron activation analysis (NAA) and prompt gamma activation analysis (PGAA) are performed using MCNP6, a highly sophisticated, multi-particle Monte Carlo simulation code (with the employment of the high performance computing system at the Idaho National Laboratory), and confirm the feasibility of using PGAA and NAA as an environmental inspection methodology for detection and monitoring of small quantities of chlorine on canister surfaces. The specific signature of gamma rays due reactions of sea salt constituents are documented and summarized in this paper. The environmental constituent for passive oxide film breaching is chlorine, which can cause pitting initiation and incubation time that may lead to cracking in stainless steel. Therefore, an assessment of chloride presence at a canister surface is essential before moving to other nondestructive analysis (NDA) methods for detection of pitting and cracking. Two nonlinear acoustic techniques will also be used in our integrated research project (IRP) research study that proves to be useful in the detection of SCC: nonlinear resonant ultrasound spectroscopy (NRUS), and time revisited elastic nonlinear diagnostic (TREND). These two techniques are complementary to marine atmospheric detection techniques. In this presentation only PGAA and NAA are described in detail. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Effect of heat input on the microstructure, residual stresses and corrosion resistance of 304L austenitic stainless steel weldments

Austenitic stainless steels are widely used in high performance pressure vessels, nuclear, chemical, process and medical industry due to their very good corrosion resistance and superior mechanical properties. However, austenitic stainless steels are prone to sensitization when subjected to higher temperatures (673 K to 1173 K) during the manufacturing process (e.g. welding) and/or certain applications (e.g. pressure vessels). During sensitization, chromium in the matrix precipitates out as carbides and intermetallic compounds (sigma, chi and Laves phases) decreasing the corrosion resistance and mechanical properties. In the present investigation, 304L austenitic stainless steel was subjected to different heat inputs by shielded metal arc welding process using a standard 308L electrode. The microstructural developments were characterized by using optical microscopy and electron backscattered diffraction, while the residual stresses were measured by X-ray diffraction using the sin{sup 2}ψ method. It was observed that even at the highest heat input, shielded metal arc welding process does not result in significant precipitation of carbides or intermetallic phases. The ferrite content and grain size increased with increase in heat input. The grain size variation in the fusion zone/heat affected zone was not effectively captured by optical microscopy. This study shows that electron backscattered diffraction is necessary to bring out changes in the grain size quantitatively in the fusion zone/heat affected zone as it can consider twin boundaries as a part of grain in the calculation of grain size. The residual stresses were compressive in nature for the lowest heat input, while they were tensile at the highest heat input near the weld bead. The significant feature of the welded region and the base metal was the presence of a very strong texture. The texture in the heat affected zone was almost random. - Highlights: • Effect of heat input on microstructure, residual stresses and corrosion is studied. • HAZ and width of dendrite in the welded region increase with heat input. • Residual stresses are tensile near the welded region after the highest heat input. • Welded region has the highest pit density after highest heat input. • Dendrites and δ-ferrite were highly oriented in the welded region.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Neutron Residual Stress Mapping for Spent Nuclear Fuel Storage Canister Weldment

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Higher harmonic ultrasonic guided waves for structural integrity assessment of dry storage canisters - 15381

Dry storage of used fuel from nuclear power plants has gone from an interim storage solution to a longer-term solution, which necessitates monitoring the integrity of the stainless steel canister that confines the used fuel. Since the storage facilities are located around the country in a variety of different environments there is a wide range of degradation mechanisms that could occur over time in the cylindrical canister. One potential degradation mode that has created significant concern is stress corrosion cracking. We propose using higher harmonic ultrasonic guided waves as a tool to nondestructively inspect used fuel canisters. Higher harmonics are generated when a wave propagates through a nonlinear elastic media. These higher harmonics have been shown to be quite sensitive to microstructural features such as dislocation substructures, persistent slip bands, precipitates, etc. The use of ultrasonic guided waves to inspect the canister is attractive due to the size of the canister and the ability of guided waves to inspect a reasonably large domain relative to traditional bulk wave ultrasonics. We will highlight our recent progress on theory, numerical simulations, and laboratory experiments to develop higher harmonic guided waves as a tool to detect microstructural changes that precede macro-scale damage. Analyzed structures include plates and hollow cylinders, and we focus on localized damage initiation, as is the case for stress corrosion cracking. Recent experiments focus on using magnetostrictive transducers in activating the fundamental shear horizontal mode (SH0) and the third harmonic that is generated due to the material nonlinearity. The magnetostrictive transducer comprises a thin iron-cobalt foil, a meandering electric coil, and a permanent magnet. These transducers are inexpensive, easy to use, and are very efficient at generating shear waves. In applications where the wavelength is larger than the plate thickness, the waves are guided by the traction free boundary conditions and propagate in the plane of the plate. For this canister application, the shear horizontal waves energize the entire thickness of the canister and propagate along the shell in the direction dictated by the meander coil. Experiments indicate that magnetostrictive transducers can generate waves that propagate in plate thicknesses of 1 mm to at least 16 mm. The SH0 mode is nondispersive and generates a third harmonic of the same mode that is shown to be sensitive to localized plastic strain and unseen fatigue damage. It has a strong potential to detect the very early stages of stress corrosion cracking. Due to the geometric configuration of dry storage casks, the transducers will need to be delivered to the inspection points by a robotic system that navigates through the cask ventilation system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Development of Surface Sampling Techniques for the Canister Deposition Field Demonstration (FY22 Update)

This report describes the proposed surface sampling techniques and plan for the multi-year Canister Deposition Field Demonstration (CDFD). The CDFD is primarily a dust deposition test that will use three commercial 32PTH2 NUHOMS welded stainless steel storage canisters in Advanced Horizontal Storage Modules, with planned exposure testing for up to 10 years at an operating ISFSI site. One canister will be left at ambient condition, unheated; the other two will have heaters to achieve canister surface temperatures that match, to the degree possible, spent nuclear fuel (SNF) loaded canisters with heat loads of 10 kW and 40 kW. Surface sampling campaigns for dust analysis will take place on a yearly or bi-yearly basis. The goal of the planned dust sampling and analysis is to determine important environmental parameters that impact the potential occurrence of stress corrosion cracking on SNF dry storage canisters. Specifically, measured dust deposition rates and deposited particle sizes will improve parameterization of dust deposition models employed to predict the potential occurrence and timing of stress corrosion cracks on the stainless steel SNF canisters. The size, morphology, and composition of the deposited dust and salt particles will be quantified, as well as the soluble salt load per unit area and the rate of deposition, as a function of canister surface temperature, location, time, and orientation. Previously, a preliminary sampling plan was developed, identifying possible sampling locations on the canister surfaces and sampling intervals; possible sampling methods were also described. Further development of the sampling plan has commenced through three different tasks. First, canister surface roughness, a potentially important parameter for air flow and dust deposition, was characterized at several locations on one of the test canisters. Second, corrosion testing to evaluate the potential lifetime and aging of thermocouple wires, spot welds, and attachments was initiated. Third, hand sampling protocols were developed, and initial testing was carried out. The results of those efforts are presented in this report. The information obtained from the CDFD will be critical for ongoing efforts to develop a detailed understanding of the potential for stress corrosion cracking of SNF dry storage canisters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Fracture Toughness Evaluation for Sandia Mock-up Stainless Steel Canister Weldment Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials that have no current outlet than storage. These materials are often held in long-term storage as an interim stage within their lifecycle. Lifecycles can include reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. Stress corrosion cracking of interim storage containers has been identified as a high priority data gap by the Department of Energy. However, little has been done with regards to the canister weld material properties, including the fracture toughness, and their impact on stress corrosion, until recently from Sandia National Laboratory canister mockup program. Furthermore, because no post-weld heat treatment was required, the associated high tensile residual stress within these canister welds can drive the initiation and growth of stress corrosion cracking (SCC) cracks. Thus, the fracture mechanical property, such as fracture toughness, is an essential information to understand the canister weldment fracture resistance behavior and to support mitigation protocol development to counter SCC. In this work, we carried out fracture toughness evaluation on the received mock-up canister weldments from Sandia, using spiral notch torsion fracture test technology. The details of the weldment fracture toughness evaluation program development at ORNL are presented in this progress report. The objective of this research is to collect fracture toughness data on the mock-up canister weldment from Sandia National Laboratory. Data will be collected under quasi-static fracture loading using the Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support the ongoing SNF dry storage canister modeling activities and to address regulatory issues associated with canister container integrity and the follow-on SNF system transport reliability. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre-cracks are needed. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient crack length to increase the sample fracture tip localized constraint. Fatigue pre-crack growth techniques are employed to introduce sharp crack front into the sample. Previously, only rough calculations were applied to estimate the compliance evolution in the SNTT crack growth process; while accurate quantitative descriptions have never been attempted. This generates an urgent need to understand the crack evolution during the SNTT fracture testing process of ductile materials. The newly developed governing equations for SNTT crack growth estimate are used to control the fatigue crack growth to effectively reaching the target crack length. The detailed SNTT approach and its estimated fracture toughness for the as-received Sandia canister weldment and the associated baseline 304 stainless steel are presented in this report. In the past SNTT technology has also been successfully applied to investigate the fracture behavior of X52 and X80 steels and the X52 welded materials used for hydrogen infrastructures. The current study in supporting Sandia SNF Dry Storage Mock-up Canister Program is the first attempt of applying SNTT to the highly ductile stainless steel and the associated weld materials. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to SS304 steel and SS304/308 weld material. The estimate JQ upon fracture for the baseline SS304 steel is at 1618 lb/in. The estimate JQ upon fracture for the SS304/308 weld from HAZ and weld specimens are 720 lb/in and 850 lb/in, respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

FY23 Update: Surface Sampling Activities for the Canister Deposition Field Demonstration

This report describes the results of a field demonstration of the proposed surface sampling techniques and plan for the multi-year Canister Deposition Field Demonstration (CDFD). The CDFD will evaluate salt deposition rates on three commercial 32PTH2 NUHOMS welded stainless steel storage canisters in Advanced Horizontal Storage Modules. Exposure testing is planned for up to 10 years and will incorporate periodic surface sampling campaigns. The goal of the planned dust sampling and analysis is to determine important environmental parameters that impact the potential occurrence of stress corrosion cracking on spent nuclear fuel (SNF) dry storage canisters. Specifically, measured dust deposition rates and deposited particle sizes will improve parameterization of dust deposition models employed to predict the potential occurrence and timing of stress corrosion cracks on the stainless steel SNF canisters. Previously, a preliminary sampling plan was developed, identifying possible sampling locations on the canister surfaces and sampling intervals; possible sampling methods were also described. Building from previous work, this report documents hand sampling from a spent nuclear fuel canister on a transfer skid mockup designed by Sandia National Laboratories. The sampling took place from a boom lift and salts were collected from mounted sample plates. The results of these efforts are presented in this report and compared to previous laboratory-controlled tests. The information obtained from the CDFD will be critical for ongoing efforts to develop a detailed understanding of the potential for stress corrosion cracking of SNF dry storage canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Experimental Methods for the Performance Test of Environment-Assisted Cracking (EAC) in Nuclear Industry and Examples through NSUF

Environmentally assisted cracking (EAC) is a critical failure mode in nuclear reactor environments, where materials are exposed to high temperatures, radiation, and corrosive media. Specific EAC studies in nuclear environments include stress corrosion cracking in primary water chemistry, irradiation-assisted stress corrosion cracking (IASCC) under neutron exposure, and hydrogen induced cracking (HIC) in reactor pressure vessels. The experimental methodologies of EAC include slow strain rate testing (SSRT), constant load, and cyclic loading techniques, conducted in autoclaves designed to simulate the reactor’s operational environments. This paper highlights several experimental case studies, especially on IASCC, conducted within the Nuclear Science User Facilities (NSUF).

36 - MATERIALS SCIENCE↗

DOE OSTI2014

Materials for Consideration in Standardized Canister Design Activities.

This document identifies materials and material mitigation processes that might be used in new designs for standardized canisters for storage, transportation, and disposal of spent nuclear fuel. It also addresses potential corrosion issues with existing dual-purpose canisters (DPCs) that could be addressed in new canister designs. The major potential corrosion risk during storage is stress corrosion cracking of the weld regions on the 304 SS/316 SS canister shell due to deliquescence of chloride salts on the surface. Two approaches are proposed to alleviate this potential risk. First, the existing canister materials (304 and 316 SS) could be used, but the welds mitigated to relieve residual stresses and/or sensitization. Alternatively, more corrosion-resistant steels such as super-austenitic or duplex stainless steels, could be used. Experimental testing is needed to verify that these alternatives would successfully reduce the risk of stress corrosion cracking during fuel storage. For disposal in a geologic repository, the canister will be enclosed in a corrosion-resistant or corrosion-allowance overpack that will provide barrier capability and mechanical strength. The canister shell will no longer have a barrier function and its containment integrity can be ignored. The basket and neutron absorbers within the canister have the important role of limiting the possibility of post-closure criticality. The time period for corrosion is much longer in the post-closure period, and one major unanswered question is whether the basket materials will corrode slowly enough to maintain structural integrity for at least 10,000 years. Whereas there is extensive literature on stainless steels, this evaluation recommends testing of 304 and 316 SS, and more corrosion-resistant steels such as super-austenitic, duplex, and super-duplex stainless steels, at repository-relevant physical and chemical conditions. Both general and localized corrosion testing methods would be used to establish corrosion rates and component lifetimes. Finally, it is unlikely that the aluminum-based neutron absorber materials that are commonly used in existing DPCs would survive for 10,000 years in disposal environments, because the aluminum will act as a sacrificial anode for the steel. We recommend additional testing of borated and Gd-bearing stainless steels, to establish general and localized corrosion resistance in repository-relevant environmental conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

SNF Interim Storage Canister Corrosion and Surface Environment Investigations (FY2020 Status Report)

This progress report describes work performed during FY20 at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. Work in FY20 further defined our understanding of the potential chemical and physical environment present on canister surfaces, evaluated the relationship between the environment and the resultant corrosion that occurs, and initiated crack growth rate testing under relevant environmental conditions. In FY20, work to define dry storage canister surface environments included several tasks. First, collection of dust deposition specimens from independent spent fuel storage installation (ISFSI) site locations helped to establish a more complete understanding of the potential chemical environment formed on the canister. Second, the predicted evolution of canister surface relative humidity RH) values was estimated using ISFSI site weather data and the horizontal canister thermal model used by the SNL probabilistic SCC model. These calculations determined that for typical ISFSI weather conditions, seasalt deliquescence to produce MgCl2-rich brines could occur in less than 20 years at the coolest locations on the canister surface, and, even after nearly 300 years, conditions for NaCl deliquescence (75% RH) are not reached. This work illustrates the importance of understanding the stability of MgCl2-rich brines on the heated canister surface, and the potential impact of brine composition on corrosion processes, including pitting and stress corrosion cracking. In an additional study, the description of the canister surface environment was refined in order to define more realistic corrosion testing environments including diurnal cycles, soluble salt chemistries, and inert mineral particles. The potential impacts of these phenomena on canister corrosion are being evaluated experimentally. Finally, work over the past few years to evaluate the stability of magnesium chloride brines continued in FY20. MgCl2 degassing experiments were carried out, confirming that MgCl2 brines slowly degas HCl on heated surfaces, converting to less deliquescent magnesium hydroxychloride phases and potentially leading to brine dryout.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

Corrosion Assessment of Candidate Materials for the SHINE Subcritical Assembly Vessel and Components FY15 Report

In the previous report of this series, a literature review was performed to assess the potential for substantial corrosion issues associated with the proposed SHINE process conditions to produce 99 Mo. Following the initial review, substantial laboratory corrosion testing was performed emphasizing immersion and vapor-phase exposure of candidate alloys in a wide variety of solution chemistries and temperatures representative of potential exposure conditions. Stress corrosion cracking was not identified in any of the exposures up to 10 days at 80°C and 10 additional days at 93°C. Mechanical properties and specimen fracture face features resulting from slow-strain rate tests further supported a lack of sensitivity of these alloys to stress corrosion cracking. Fluid velocity was found not to be an important variable (0 to ~3 m/s) in the corrosion of candidate alloys at room temperature and 50°C. Uranium in solution was not found to adversely influence potential erosion-corrosion. Potentially intense radiolysis conditions slightly accelerated the general corrosion of candidate alloys, but no materials were observed to exhibit an annualized rate above 10 μm/y.

36 MATERIALS SCIENCE↗