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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↗

NASA NTRS1991

Measurements and mechanisms of localized aqueous corrosion in aluminum-lithium-copper alloys

Summary information is included for electrochemical aspects of stress corrosion cracking in alloy 2090 and an introduction to the work to be initiated on the new X2095 (Weldalite) alloy system. Stress corrosion cracking (SCC) was studied in both S-T and L-T orientations in alloy 2090. A constant load TTF test was performed in several environments with a potentiostatically applied potential. In the same environments the electrochemical behavior of phases found along subgrain boundaries was assessed. It was found that rapid failure due to SCC occurred when the following criteria was met: E(sub BR,T1) is less than E(sub applied) is less than E(sub Br, matrix phase). Although the L-T orientation is usually considered more resistant to SCC, failures in this orientation occurred when the stated criteria was met. This may be due to the relatively isotropic geometry of the subgrains which measure approximately 12 to 25 microns in diameters. Initial studies of alloy X2095 includes electrochemical characterization of three compositional variations each at three temperatures. The role of T(sub 1) dissolution in SCC behavior is addressed using techniques similar to those used in the research of 2090 described. SCC susceptibility is also studied using alternate immersion facilities at Reynolds Metals Corporation. Pitting is investigated in terms of stability, role of precipitate phases and constituent particles, and as initiation sites for SCC. In all research endeavors, attempts are made to link electrochemistry to microstructure. Previous work on 2090 provides a convenient basis for comparison since both alloys contain T(sub 1) precipitates but with different distributions. In 2090 T(sub 1) forms preferentially on subgrain boundaries whereas in X2095 the microstructure appears to be more homogeneous with finer T(sub 1) particles. Another point for comparison is the delta prime strengthening phase found in 2090 but absent in X2095.

Wall, Douglas↗

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.

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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↗

NASA NTRS1987

NASCRAC - A computer code for fracture mechanics analysis of crack growth

NASCRAC - a computer code for fracture mechanics analysis of crack growth - is described in this paper. The need for such a code is increasing as requirements grow for high reliability and low weight in aerospace components. The code is comprehensive and versatile, as well as user friendly. The major purpose of the code is calculation of fatigue, corrosion fatigue, or stress corrosion crack growth, and a variety of crack growth relations can be selected by the user. Additionally, crack retardation models are included. A very wide variety of stress intensity factor solutions are contained in the code, and extensive use is made of influence functions. This allows complex stress gradients in three-dimensional crack problems to be treated easily and economically. In cases where previous stress intensity factor solutions are not adequate, new influence functions can be calculated by the code. Additional features include incorporation of J-integral solutions from the literature and a capability for estimating elastic-plastic stress redistribution from the results of a corresponding elastic analysis. An example problem is presented which shows typical outputs from the code.

Harris, D. O.↗

DOE OSTI2017

Development of Friction Stir Processing for Repair of Nuclear Dry Cask Storage System Canisters

The Nuclear Regulatory Commission has identified chloride-induced stress corrosion cracking (CISCC) of austenitic stainless steel dry cask storage systems (DCSS) as an area of great concern. Friction Stir Processing (FSP) was used to repair laboratory-generated stress corrosion cracking (SCC) in representative stainless steel 304 coupons. Results of this study show FSP is a viable method for repair and mitigation CISCC. This paper highlights lessons learned and developed techniques relative to FSP development for crack repair in sensitized thick section stainless steel 304. These include: development of process parameters, welding at low spindle speed, use of weld power and temperature control and optimization of these controls. NDE and destructive analysis are also presented to demonstrate effectiveness of the developed methods for SCC crack repair.

Friction Stir↗

DOE OSTI2019

Comparison of selective oxidation in Ni-based alloys exposed to PWR primary water and Rhines Pack environments

A mechanistic understanding of corrosion and oxidation processes is crucial to ensure long-term resistance to stress corrosion cracking of Ni-based structural alloys in PWR primary and secondary systems. Aqueous corrosion of complex Ni-based alloys under PWR water conditions as well as gaseous oxidation of low-solute Ni binary alloys at high temperatures (>800 °C) have been studied extensively. However, a significant knowledge gap exists between response of Ni-based alloys in aqueous and gaseous environmental conditions. Selective oxidation at different temperatures was investigated in a series of gaseous Rhines Pack experiments for a high-purity Ni-5Cr binary alloy and a commercial Ni-18Cr-9Fe Alloy 600. Focused ion beam milling enabled the extraction of site-specific specimens containing high-energy grain boundaries. Analytical transmission electron microscopy was employed to analyze the microstructure and chemical composition of the resulting localized oxidation. The oxidation mechanism in the gaseous Rhines Pack atmosphere at 420 °C included the formation of a protective Cr-oxide cap above grain boundaries in all cases. At higher temperatures (600 and 800 °C), penetrative intergranular and transgranular internal oxidation was observed. Direct comparisons are made to corrosion response for these same materials in simulated PWR primary water at 320-360 °C. These results reveal a temperature dependence for gaseous oxidation and indicate mechanistic differences between aqueous and gaseous degradation in Ni-Cr alloys. These results imply that selective intergranular corrosion in PWR water is governed by processes beyond mere oxidation. Consideration of complex corrosion/dissolution processes at surfaces and in open cracks is essential for the better understanding of stress corrosion cracking mechanisms in PWR primary water.

Rhines Pack, Alloy 600, Ni-base alloys, Internal O↗

DOE OSTI2023

Novel effects of grain size and ion implantation on grain boundary segregation in ion irradiated austenitic steel

Nuclear reactor environments are extreme in nature due to the combination of exposure to corrosive coolant, mechanical stresses, and neutron irradiation damage. This combination of environments can lead to irradiation induced/assisted precipitation, segregation, and embrittlement coupled with enhanced creep and stress corrosion cracking. While nanostructured materials are of interest in a variety of applications due to their unique properties including high mechanical strength, for nuclear applications nanostructured materials have the added benefit of irradiation resistance due to the high volume density of sinks for irradiation-induced defects. In this study, we show unique radiation-induced segregation (RIS) behavior in nanocrystalline 304 stainless steel (SS304). Typical RIS in austenitic steels includes Ni enrichment and Cr depletion within the grain boundaries. In nanostructured SS304 this RIS is reduced. Furthermore, interestingly, some grain boundaries in the nanocrystalline SS304 in this study were found to be enriched in Cr after Fe 2+ ion irradiation at 500 °C up to 50 displacements per atom (dpa). To understand this unique observation, lattice-based atomic kinetic Monte Carlo simulations were performed to investigate the effects of grain size, thermal segregation, as well as injected interstitials (introduced as Fe self-ions). The results indicate that the Cr enrichment at grain boundaries may be caused by combined effects of grain size, self-ion injection and preferential Cr diffusion via the interstitial mechanism. In conclusion, the findings also suggest a possible difference in RIS between ion and neutron irradiations.

36 MATERIALS SCIENCE↗

DOE OSTI2023

An examination of the precipitation behavior of proton irradiated dual phase 308L weldment filler materials

Voids, G phase particles, and Ni-Si rich clusters in proton irradiated dual phase 308L groove filler of a SA508–304 L dissimilar metal weldment are analyzed using advanced characterization techniques. These weldments are often used in light water nuclear reactors and are subject to enhanced corrosion and associated stress corrosion cracking (SCC). Radiation damage is known to accelerate SCC. Ni-Si enriched clusters were observed in proton irradiated γ austenite, while G phase M 6 Ni 16 Si 7 (where M transition metal element) precipitates were observed in proton irradiated δ ferrite. Compositional analysis of the G phase precipitates and Ni-Si clusters from STEM-EDS and APT are compared. Unlike G phase particles in proton irradiated δ ferrite, Ni-Si clusters in proton irradiated γ austenite are not rich in Mn. Both STEM-EDS and APT line-scan profiles of the Ni-Si clusters show that the Fe and Cr concentration gradient between matrix γ austenite and the Ni-Si clusters is not as sharp as those between matrix δ ferrite and G phase. Further, HR-STEM imaging indicates that the lattice parameter of the Ni-Si clusters is commensurate with γ austenite and the clusters do not represent the precipitation of a second phase. Finally, our analysis demonstrates the density and volume fraction of G phase particles and the density of voids scales with proton irradiation damage and the energy to recoils.

308L groove filler↗

DOE OSTI2017

Thermal Analysis Capability of UNF-ST&DARDS

In the United States, approximately 2,500 casks are loaded with commercial spent nuclear fuel (SNF) that has transitioned from wet storage (spent fuel pools) to dry storage. The number of loaded dry storage casks is increasing by approximately 200 each year. Over time, cask designs have evolved to enhance safety and to accommodate more fuel and higher heat loads. Also, higher burnup fuel is being transitioned into dry storage. The SNF is being stored in dry casks for longer times than specified in the original certification period. Several degradation mechanisms related to fuel assemblies and canisters are affected by temperature. For the cladding, temperature-dependent phenomena include creep and annealing, hydride reorientation and embrittlement, and the ductile-to-brittle transition. Temperature can also influence phenomena that affect the long-term integrity of the storage system, including deliquescence, corrosion, and stress-corrosion cracking. Therefore, accurate determination of the temperatures of various components is needed to evaluate potential safety-related issues during transportation after extended storage and to ensure SNF retrievability. The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is being developed for the US Department of Energy Office of Nuclear Energy to streamline analyses for the waste management system [1]. Lastly, the thermal analysis capability within UNF-ST&DARDS and example results are discussed herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

High efficiency thermal storage system for solar plants (HELSOLAR). Final report

The project objective was to develop a high temperature Thermal Storage System (TES) based on graphite and able to provide both economical and technical advantages with respect to existing solutions contributing to increase the share of Concentrated Solar Plants (CSP). One of the main disadvantages of most of the renewable energy systems is their dependence to instantaneous irradiation and, thus, lack of predictability. CSP plants with thermal storage have proved to offer a good solution to this problem although still at an elevated price. The identification of alternative concepts able to work more efficiently would help to speed up the convergence of CSP towards grid parity. One way to reduce costs is to work in a range of temperatures higher than those allowed by the actual molten salt systems, currently the benchmark for TES in CSP. This requires the use of alternative energy storage materials such as graphite, as well as the utilization of Heat Transfer Fluids (HTF) other than molten salts or organic oils. The main technical challenges identified are derived from the high temperatures and significant high pressures, which pose risks such as potential graphite and insulation oxidation, creep, fatigue, corrosion and stress-corrosion in the pipes, leakages in the joints, high blower drivers’ electrical power consumption, thermal compatibility or relative deformations of the different materials. At the end, the main challenge of the project, is to identify a technical solution able to overcome all these problems but still at a competitive cost when compared to already existing thermal storage solutions. Special attention is given to all these issues during this project.

14 SOLAR ENERGY↗

DOE OSTI2013

Crack Growth Rate and Fracture Toughness Tests on Irradiated Cast Stainless Steels

Cast austenitic stainless steels are used in the cooling system of light water reactors for components with complex shapes, such as pump casings, valve bodies, and coolant piping. In the present study, crack growth rate and fracture toughness JR curve tests were performed on irradiated cast stainless steels in low-corrosion-potential environments (low-dissolved-oxygen high-purity water or simulated pressurized water reactor environment) at 320°C. Both as received and thermally aged materials were included to investigate the combined effect of thermal aging and irradiation embrittlement on the fracture behavior of cast stainless steels. The samples were irradiated to approximately 0.08 dpa at the Halden reactor. Good resistance to corrosion fatigue and stress corrosion cracking was observed for all samples. Thermal aging had little effect on the crack growth behavior at this dose level. Cleavage-like fracture was the dominant cracking morphology during the crack growth rate tests, and the ferrite phase was deformed to a lesser extent compared with the surrounding austenite phase. The fracture toughness results showed a dominant effect of neutron irradiation, and the fracture resistances were decreased considerably for all cast specimens regardless of their thermal aging conditions. The reduction in fracture toughness was more significant in the unaged than thermally aged materials. Nonetheless, the fracture toughness values of thermally aged specimens were about 30% lower than their unaged counterparts, suggesting a combined effect of thermal aging and neutron irradiation in cast stainless steel.

Chen, Y.↗

NASA NTRS1995

Replacement of corrosion protection chromate primers and paints used in cryogenic applications on the Space Shuttle with wire arc sprayed aluminum coatings

With the advent of new environmental laws restricting volatile organic compounds and hexavalent chrome emissions, 'environmentally safe' thermal spray coatings are being developed to replace the traditional corrosion protection chromate primers. A wire arc sprayed aluminum coating is being developed for corrosion protection of low pressure liquid hydrogen carrying ducts on the Space Shuttle Main Engine. Currently, this hardware utilizes a chromate primer to provide protection against corrosion pitting and stress corrosion cracking induced by the cryogenic operating environment. The wire are sprayed aluminum coating has been found to have good potential to provide corrosion protection for flight hardware in cryogenic applications. The coating development, adhesion test, corrosion test and cryogenic flexibility test results will be presented.

Daniel, R. L.↗

NASA NTRS1978

Method of producing complex aluminum alloy parts of high temper, and products thereof

Fully annealed aluminum sheet is first stretch formed to the complex, doubly compound shape of a previously prepared forming die, e.g., an ejection seat blowout panel of a shuttlecraft. The part is then marked with a series of grid lines for monitoring later elongation. Thereafter it is solution heat treated and refrigerated to retard hardening. While still soft, it is stretched a second time on the same die to induce a modicum of work hardening, after which it is aged to the desired stress corrosion resistant temper, preferably the T8 level, to provide the desired hardness and stress corrosion resistance.

Wilson, I. J.↗

DOE OSTI2015

Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example. Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the size constraints of the MITR water loop are described. The safety case for operation of the high pressure gas-driven bellows mechanism is also presented. Key issues are the design and response of systems to limit gas flow in the event of a high pressure gas leak in the in-core autoclave. Integrity of the autoclave must be maintained and reactivity effects due to voiding of the loop coolant must be shown to be within the reactor technical specifications. The technical development of the crack growth monitor for application in the INL Advanced Test Reactor or the MITR can act as a template for adaptation of this technology in other reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2014

Simulation of the ultrasonic array response from real branched cracks using an efficient finite element method

A hybrid model to simulate the ultrasonic array response from stress corrosion cracks is presented. These cracks are branched and difficult to detect so the model is required to enable optimization of an array design. An efficient frequency-domain finite element method is described and selected to simulate the ultrasonic scattering. Experimental validation results are presented, followed by an example of the simulated ultrasonic array response from a real stress corrosion crack whose geometry is obtained from an X-ray Computed Tomography image. A simulation-assisted array design methodology, which includes the model and use of real crack geometries, is proposed.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2024

FY24 Progress Report: Disposition of Cracks & Features Observed in the Inner Can Closure Weld Region (ICCWR) of the 3013 Package

The long-term integrity of the 3013 containers is of interest for the safe storage of Pu materials. Although the 3013 standard embodies multiple barrier concept, the integrity of the inner container is considered to be crucial since it protects the Safety Class outer container from its contents. The container is designed to withstand high pressures that could result from the complete radiolysis of the maximum water content permissible by the 3013 standard. Shelf-life studies and destructive examination have not found high gas pressures but have found that corrosive gases are generated. Pitting and stress corrosion cracking (SCC) are considered to be critical corrosion modes for the performance of the inner container and have been observed during destructive examinations. Considerable research has been conducted on the possibility of aqueous electrolytes condensing on the inner container that can promote SCC. These studies indicate the uncertainties surrounding the formation of corrosive environments and the corrosion behavior of container materials. In this initial report, a Bayesian network (BN) model is described that can consider the uncertainties and the causal connections between various factors influencing the corrosion modes of the inner container. The BN model is preliminary and provides an initial framework to identify the necessary information. The report also provides initial experimental results on the electrochemical behavior of stainless steels in anticipated condensed environments from gas phase migration of acidic gases. The experimental results are consistent with the corrosion model. Recommendation for further work on the BN model include assembling an expert group to provide input to the BN structure and quantification of the conditional probability matrix, experimental studies to characterize the microstructure of the container, electrochemical studies to identify critical potentials for localized corrosion and SCC, and crack growth rate studies

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DOE OSTI2015

Evaluation of DUSTRAN Software System for Modeling Chloride Deposition on Steel Canisters

The degradation of steel by stress corrosion cracking (SCC) when exposed to atmospheric conditions for decades is a significant challenge in the fossil fuel and nuclear industries. SCC can occur when corrosive contaminants such as chlorides are deposited on a susceptible material in a tensile stress state. The Nuclear Regulatory Commission has identified chloride-induced SCC as a potential cause for concern in stainless steel used nuclear fuel (UNF) canisters in dry storage. The modeling of contaminant deposition is the first step in predictive multiscale modeling of SCC that is essential to develop mitigation strategies, prioritize inspection, and ensure the integrity and performance of canisters, pipelines, and structural materials. A multiscale simulation approach can be developed to determine the likelihood that a canister would undergo SCC in a certain period of time. This study investigates the potential of DUSTRAN, a dust dispersion modeling system developed by Pacific Northwest National Laboratory, to model the deposition of chloride contaminants from sea salt aerosols on a steel canister. Results from DUSTRAN simulations run with historical meteorological data were compared against measured chloride data at a coastal site in Maine. DUSTRAN’s CALPUFF model tended to simulate concentrations higher than those measured; however, the closest estimations were within the same order of magnitude as the measured values. The decrease in discrepancies between measured and simulated values as the level of abstraction in wind speed decreased suggest that the model is very sensitive to wind speed. However, the influence of other parameters such as the distinction between open-ocean and surf-zone sources needs to be explored further. Deposition values predicted by the DUSTRAN system were not in agreement with concentration values and suggest that the deposition calculations may not fully represent physical processes. Overall, results indicate that with parameter refinement, DUSTRAN has the potential to simulate atmospheric chloride dispersion on steel canisters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗