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At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Waste Chemistry Envelope for the Determination of Pitting Corrosion Susceptibility - 19554

Liquid radioactive wastes at the Hanford and Savannah River sites are stored in large underground carbon steel tanks. The high-level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks, will be required to be in service for times exceeding the initial intended life (i.e., greater than 60 years). Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. Administrative programs are in place to mitigate potential corrosion mechanisms and thereby maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. Waste chemistry and temperature control are the means by which corrosion degradation of the waste tanks is minimized. Chemistry controls are based on in-situ coupon tests, waste tank experience, and laboratory testing. These requirements comprise the technical standards for corrosion control. The standards are utilized then as guidelines for operational procedures. The present waste chemistry controls for the Hanford DSTs were established in the 1980's in response to tank failures caused by stress corrosion cracking (SCC) at the DOE Savannah River Site (SRS). Portions of the SRS chemistry limits, which were developed at the Savannah River National Laboratory (SRNL) were coupled with corrosion testing results from Pacific Northwest National Laboratories (PNNL) and adopted for corrosion control of the double-shell tank (DST) waste at the Hanford Site. Since that time, the temperatures of the waste in the DSTs have decreased to less than 50 deg. C in most cases as the radioactive material has decayed. As a consequence, new chemistry controls were formulated for SCC in 2010. This testing however, did not address that potential for pitting in the supernatant and at the liquid-air interface. Indeed, initial tests indicated that the inhibitor requirements for SCC would be insufficient for the prevention of pitting. In addition to lower waste temperatures, future retrieval and process changes may impact the DST waste chemistry. Initial projections suggest that the waste chemistry of the DSTs may shift to a broader range of pH and higher aggressive anion concentrations (e.g., chloride, sulfate, etc.) than the wastes that are presently stored. Additionally, processes within the vitrification facility may result in streams that are returned to the tank farm for evaporation. These streams may also contain high concentrations of aggressive species. SRNL investigated the propensity for pitting of the carbon steel waste tanks at current and projected DST waste chemistries. A statistically designed series of tests provided a pitting factor that relates the ratio of the inhibitor species to the aggressive species and the probability of observing pitting. An acceptable pitting factor was determined and was utilized for the development of a new chemistry envelope. The new chemistry envelope was compared to the envelope that was established for SCC and was determined to be bounding. Validation of the new chemistry envelope was performed via model simulations and comparison with historical experimental results and approaches to establishing chemistry envelopes. The new chemistry envelope will eventually be part of an overall assessment of the condition of the tanks and any changes made to the overall structural integrity program for the Hanford DSTs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

On residual stress analysis and microstructural evolution for stainless steel type 304 spent nuclear fuel canisters weld joint: Numerical and experimental studies

Chloride-induced stress corrosion cracking (CISCC) in the weldments of austenitic stainless steel canisters is one of the primary safety concerns during the dry storage of used nuclear fuel at Independent Spent Fuel Storage Installations in coastal areas. In order to evaluate the CISCC behavior in the canister, a 3D sequentially coupled thermo-mechanical finite element model was built to simulate the residual stresses induced by two intersecting longitudinal and circumferential multi-pass welds in austenitic stainless steel mockup canister. Weld-induced residual stresses from simulation were in good agreement with the experimental measurement results by deep-hole drilling and contour methods. Through-wall high tensile axial and hoop stresses were observed in both the fusion zone and heat affected zone (HAZ) in the longitudinal and circumferential weld, respectively. On the other hand, the microstructure within the weldments of 304 stainless steel mockup was characterized using optical microscopy, scanning electron microscopy, electron back-scattered diffraction and energy dispersive X-ray spectroscopy methods. A few amount of MnS inclusions in both the base metal and HAZ, and a lot of (Mn,Si) x O y inclusions in the fusion zone were found. Meanwhile, there is some α'-martensite in the base metal, while the martensite was reversed in the HAZ. In the fusion zone, there is a lot of δ-ferrite. The martensite reversion phenomenon was also verified by the combination of Feritscope test and thermal simulation, which could predict a martensite reversion temperature as 400C°. The welding residual stresses and microstructure analysis results could offer the instructive information for the prediction of CISCC behavior of the spent nuclear fuel canisters.

36 MATERIALS SCIENCE↗
DOE OSTI2013

The Effect of Weld Residual Stress on Life of Used Nuclear Fuel Dry Storage Canisters

With the elimination of Yucca Mountain as the long-term storage facility for spent nuclear fuel in the United States, a number of other storage options are being explored. Currently, used fuel is stored in dry-storage cask systems constructed of steel and concrete. It is likely that used fuel will continue to be stored at existing open-air storage sites for up to 100 years. This raises the possibility that the storage casks will be exposed to a salt-containing environment for the duration of their time in interim storage. Austenitic stainless steels, which are used to construct the canisters, are susceptible to stress corrosion cracking (SCC) in chloride-containing environments if a continuous aqueous film can be maintained on the surface and the material is under stress. Because steel sensitization in the canister welds is typically avoided by avoiding post-weld heat treatments, high residual stresses are present in the welds. While the environment history will play a key role in establishing the chemical conditions for cracking, weld residual stresses will have a strong influence on both crack initiation and propagation. It is often assumed for modeling purposes that weld residual stresses are tensile, high and constant through the weld. However, due to the strong dependence of crack growth rate on stress, this assumption may be overly conservative. In particular, the residual stresses become negative (compressive) at certain points in the weld. The ultimate goal of this research project is to develop a probabilistic model with quantified uncertainties for SCC failure in the dry storage casks. In this paper, the results of a study of the residual stresses, and their postulated effects on SCC behavior, in actual canister welds are presented. Progress on the development of the model is reported.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Out-of-reactor test of corrosion and hydrogen pickup in fuel cladding materials in contact with nickel metal - 2016-0038

Increased reactor coolant hydrogen concentration in pressurized water reactors (PWRs) is under consideration for potential benefits in reducing primary water stress corrosion cracking (PWSCC) crack growth rate as well as increasing PWSCC crack initiation time in nickel-based alloys. Potential concerns related to fuel performance include possible increases in hydrogen pickup (HPU) by zirconium based alloys and the enhancement of the pickup through the so-called nickel-window effect. Although the risk of this occurrence may be low, the consequences of increased HPU could be substantial. We examined the impact of dissolved hydrogen (DH) concentrations on Zircaloy-4 corrosion and HPU when the samples are in contact with nickel sources. The work was completed in conjunction with a large project that evaluated the impact of dissolved hydrogen on various fuel cladding and spacer weld specimens of M5{sup R}, ZIRLO{sup R}, Optimized ZIRLO{sup TM}, Low-Tin ZIRLO{sup TM}, Optimized Low-Tin Zircaloy-4, and Zircaloy-2 in either as-received or pre-oxidized conditions for exposures up to 730 days in high-pressure, out-of-reactor autoclave systems operating with a simulated PWR coolant at three different DH concentrations: 30, 100, and 500 cm{sup 3} (cc) H{sub 2}/kg H{sub 2}O. A significant increase in HPU was observed for some Zircaloy-4 specimens in contact with synthetic crud at elevated DH levels. The HPU increased with increasing DH and a metallic nickel fraction in the synthetic crud. The formation of metallic nickel in the synthetic crud samples at elevated DH levels was confirmed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2014

Corrosion in low dielectric constant Si-O based thin films: Buffer concentration effects

Organosilicate glass (OSG) is often used as an interlayer dielectric (ILD) in high performance integrated circuits. OSG is a brittle material and prone to stress-corrosion cracking reminiscent of that observed in bulk glasses. Of particular concern are chemical-mechanical planarization techniques and wet cleans involving solvents commonly encountered in microelectronics fabrication where the organosilicate film is exposed to aqueous environments. Previous work has focused on the effect of pH, surfactant, and peroxide concentration on the subcritical crack growth of these films. However, little or no attention has focused on the effect of the conjugate acid/base concentration in a buffer. Accordingly, this work examines the “strength” of the buffer solution in both acidic and basic environments. The concentration of the buffer components is varied keeping the ratio of acid/base and therefore pH constant. In addition, the pH was varied by altering the acid/base ratio to ascertain any additional effect of pH. Corrosion tests were conducted with double-cantilever beam fracture mechanics specimens and fracture paths were verified with ATR-FTIR. Shifts in the threshold fracture energy, the lowest energy required for bond rupture in the given environment, G{sub TH}, were found to shift to lower values as the concentration of the base in the buffer increased. This effect was found to be much larger than the effect of the hydroxide ion concentration in unbuffered solutions. The results are rationalized in terms of the salient chemical bond breaking process occurring at the crack tip and modeled in terms of the chemical potential of the reactive species.

36 MATERIALS SCIENCE↗
NASA NTRS1994

Modeling time-dependent corrosion fatigue crack propagation in 7000 series aluminum alloys

Stress corrosion cracking and corrosion fatigue experiments were conducted with the susceptible S-L orientation of AA7075-T651, immersed in acidified and inhibited NaCl solution, to provide a basis for incorporating environmental effects into fatigue crack propagation life prediction codes such as NASA FLAGRO. This environment enhances da/dN by five to ten-fold compared to fatigue in moist air. Time-based crack growth rates from quasi-static load experiments are an order of magnitude too small for accurate linear superposition prediction of da/dN for loading frequencies above 0.001 Hz. Alternate methods of establishing da/dt, based on rising-load or ripple-load-enhanced crack tip strain rate, do not increase da/dt and do not improve linear superposition. Corrosion fatigue is characterized by two regimes of frequency dependence; da/dN is proportional to f(exp -1) below 0.001 Hz and to F(exp 0) to F(exp -0.1) for higher frequencies. Da/dN increases mildly both with increasing hold-time at K(sub max) and with increasing rise-time for a range of loading waveforms. The mild time-dependence is due to cycle-time-dependent corrosion fatigue growth. This behavior is identical for S-L nd L-T crack orientations. The frequency response of environmental fatigue in several 7000 series alloys is variable and depends on undefined compositional or microstructural variables. Speculative explanations are based on the effect of Mg on occluded crack chemistry and embritting hydrogen uptake, or on variable hydrogen diffusion in the crack tip process zone. Cracking in the 7075/NaCl system is adequately described for life prediction by linear superposition for prolonged load-cycle periods, and by a time-dependent upper bound relationship between da/dN and delta K for moderate loading times.

Mason, Mark E.↗
DOE OSTI2018

Effect of Retrogression Heat Treatment Time on Microstructure and Mechanical Properties of AA7010

The effect of retrogression time during retrogression and re-aging (RRA) treatment of AA7010 is evaluated by performing tensile tests and characterizing the microchemistry of the grain boundary precipitates (GBPs) using transmission electron microscope coupled with the energy-dispersive spectroscopy. Retrogression time is evaluated so that the ultimate tensile strength of the RRA-treated sample is equal to that of the T6-treated sample and the grain boundary microstructure similar to that of the over-aged (T7451) condition. The investigation reveals that the sample retrogressed at 200 °C for 20 min has UTS of 586 MPa which is equivalent to that of the T6 sample and 11.5% higher than that of the T7451 condition. The fracture toughness of the RRA-treated sample was 41 MPa√m. Microstructure of the RRA-treated sample is similar to T7451, along the grain boundaries and in the grain interior similar to that of the T6-treated sample. Energy-dispersive spectroscopy confirmed the increment of Cu content on the GBP’s with increase in the retrogression time, which is expected to improve the stress corrosion cracking resistance of the alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2015

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.

Crack Growth↗
DOE OSTI2013

EFFECTS OF LASER SHOCK PEENING ON SCC BEHAVIOR OF ALLOY 600

In this study, the effects of laser shock peening (LSP) on stress corrosion cracking (SCC) behavior of Alloy 600 in tetrathionate solution were investigated. The degree of sensitization was quantified using double loop electrochemical potentiokinetic reactivation (DLEPR) tests. The sensitized Alloy 600 was demonstrated to be susceptible to intergranular SCC in tetrathionate solution. Following LSP, residual stresses and the amount of plastic strain introduced in Alloy 600 were characterized. The effects of LSP on SCC susceptibility of Alloy 600 in tetrathionate solution were evaluated by slow strain rate tests and constant load tests. Results indicate a significant increase in resistance to crack initiation and decreased susceptibility to SCC after LSP.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Reduced yield stress for zirconium exposed to iodine: Reactive force field simulation

Iodine-induced stress-corrosion cracking (ISCC), a known failure mode for nuclear fuel cladding, occurs when iodine generated during the irradiation of a nuclear fuel pellet escapes the pellet through diffusion or thermal cracking and chemically interacts with the inner surface of the clad material, inducing a subsequent effect on the cladding’s resistance to mechanical stress. To complement experimental investigations of ISCC, a reactive force field (ReaxFF) compatible with the Zr-I chemical and materials systems has been developed and applied to simulate the impact of iodine exposure on the mechanical strength of the material. The study shows that the material’s resistance to stress (as captured by the yield stress of a high-energy grain boundary) is related to the surface coverage of iodine, with the implication that ISCC is the result of adsorption-enhanced decohesion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Integrated Simulation of Weld Residual Stress Evolution and Crack Propagation Using XFEM

Nuclear power plant components operate in environments that promote multiple degradation mecha- nisms, several of which involve crack initiation and growth. An ongoing effort in the U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is developing a general capability within the Multiphysics Object Oriented Simulation Environment (MOOSE) framework for simulating three-dimensional crack growth under a range of driving conditions, including fatigue, stress corrosion cracking (SCC), brittle fracture, and stress-relaxation cracking. This report demonstrates an end-to-end workflow that uses this capability to model weld-residual-stress-driven SCC in the J-groove weld of a pressurized-water reactor control rod drive mechanism penetration in the vessel head. The workflow consists of a thermomechanical welding simulation with temperature-dependent plasticity, followed by cooldown to ambient conditions, and a restart of the simulation using the MOOSE extended finite element method (XFEM) module to propagate a three-dimensional crack through the residual stress field. New welding capabilities were developed to properly initialize newly activated elements in the weld region, and robustness improvements were made to the mesh-based algorithm for defining cutting planes in the 3D XFEM algorithm, allowing it to handle complex crack fronts and stress fields. Together these advances allowed the simulated SCC crack to grow from an initial elliptical flaw in the weld, across the weld, through the tube wall, and almost to the triple point (where the weld, tube, and reactor pressure vessel head intersect) over roughly 36 years of simulated service. These results demonstrate a workflow that can be extended to fully three-dimensional welding simulations and more complex crack interaction problems.

42 - ENGINEERING↗
DOE OSTI2014

Light Water Reactor Sustainability Program: BWR High-Fluence Material Project: Assessment of the Role of High-Fluence on the Efficiency of HWC Mitigation on SCC Crack Growth Rates

As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a program plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A New General Paradigm for Understanding and Preventing Li Metal Penetration through Solid Electrolytes

The use of lithium (Li) or sodium (Na) metal anodes together with highly ion-conductive solid electrolytes (SEs) could provide batteries with a step improvement in volumetric and gravimetric energy densities. Unfortunately, these SEs face significant technical challenges, in large part because Li and Na dendrites can penetrate through SEs, leading to short circuits. The ability of such a soft material (Li or Na metal) to penetrate through ceramic is surprising from the point of view of models widely used in the Li-battery field. We introduce a concept, new to the battery field, for preventing penetration of lithium dendrites through SEs by putting the SE surfaces into a state of residual compressive stress. For a sufficiently high compressive stress, cracks have difficulty forming, and cracks that do form are forced to close, inhibiting dendrite penetration. This approach is widely used to solve commercially important stress corrosion cracking problems in metals and static fatigue problems in ceramics and glasses (e.g., Gorilla Glass). However, the technique will not be useful for SEs if the Li-ion transport rate through a SE is substantially reduced when the SE is under compression. Our molecular dynamics calculations for Li-ion transport through a common SE demonstrate that the introduction of even very high residual compressive stresses (~10 GPa) has only a modest effect on Li-ion transport kinetics, suggesting that this approach is viable and capable of providing a new paradigm for developing high-performance and mechanically stable SEs.

25 ENERGY STORAGE↗
DOE OSTI2016

Final Report: Characterization of Canister Mockup Weld Residual Stresses

Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗