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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

SNF Interim Storage Canister Corrosion and Surface Environment Investigations

This progress report describes work being done 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. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

New International Program to Asses the Reliability of Emerging Nondestructive Techniques (PARENT)

The Nuclear Regulatory Commission (NRC) established the Program to Assess the Reliability of Emerging Nondestructive Techniques (PARENT) to follow on from the successful Program for the Inspection of Nickel alloy Components (PINC). The goal of the PARENT is to conduct a confirmatory assessment of the reliability of nondestructive evaluation (NDE) techniques for detecting and sizing primary water stress corrosion cracks (PWSCC) and applying the lessons learned from PINC to a series of round-robin tests. These open and blind round-robin tests will comprise a new set of typical pressure boundary components including dissimilar metal welds (DMWs) and bottom-mounted instrumentation penetrations. Open round-robin tests will engage research and industry teams worldwide to investigate and demonstrate the reliability of emerging NDE techniques to detect and size flaws with a wide range of lengths, depths, orientations, and locations. Blind round-robin tests will utilize various testing organizations, whose inspectors and procedures are certified by the standards for the nuclear industry in their respective countries, to investigate the ability of established NDE techniques to detect and size flaws whose characteristics range from relatively easy to very difficult for detection and sizing. Blind and open round-robin testing started in late 2011 and early 2012, respectively. This paper will present the work scope with reports on progress, NDE methods evaluated, and project timeline for PARENT.

Bottom Mounted Instrumentation↗
DOE OSTI2019

PIE of Cycle 13 316SS Cladding Using Aberration Corrected STEM Image and Analys

Optical, SEM and TEM analysis confirmed the presence of moderate sensitization in the 20% cold worked 316SS after the iron aluminide coating was applied. The Cr depletion occurs on the grain boundaries between the closely spaced carbides, and at the interface between the carbides and matrix. Cr depletion varied from locally along the grain boundaries, in many cases showing levels down to 15 wt%, but a few isolated regions had depletion levels down to 12.5 wt%. Optical and SEM images taken from a weld heat affected zone revealed the recrystallization occurred after an end plug cap had been welded to the coated cladding. Irradiation did not affect the carbides in a noticeable manner, but radiation-induced segregation/depletion altered the Cr depletion profiles by removing the wide Cr depletion profile, and replacing it with a much deeper, more narrow profile. Cr levels were depleted to 8-10 wt%, Ni and Si were enriched to levels of 40 wt% and 12 wt% respectively. Substantial precipitation formed in the matrix, and on line dislocations in the matrix and on grain boundary dislocations. Voids were observed in the upper region of the TPBAR where the irradiation temperature exceeds 320C. The levels of Cr depletion and Ni/Si enrichment indicate a material in a state susceptible to irradiation assisted stress corrosion cracking.

316SS cladding↗
DOE OSTI2013

Technical Letter Report on the Cracking of Irradiated Stainless Steels in Low-Corrosion-Potential Environments

Irradiation-assisted stress corrosion cracking (IASCC) is a major degradation mechanism for reactor internal components of austenitic stainless steels (SSs) in light water reactors. Crack growth rate (CGR) and fracture toughness J-R curve tests were performed on four irradiated SS specimens in high-purity water with low dissolved oxygen and in a simulated pressurized water reactor (PWR) environment. The samples had been previously irradiated from ~5 to 8 dpa in the BOR-60 reactor. The materials were cold-worked (CW) 316 and solution-annealed (SA) 304L SSs, which are commonly used in PWR core internals. Cyclic and constant-load CGRs were obtained from these samples to assess their cracking behaviors in low-corrosion-potential environments. A Ti-stabilized specimen, CW 316-Ti SS, was found to be highly vulnerable to IASCC, and its cracking susceptibility increased with neutron dose between ~5 and 8 dpa. The CW 316 and SA 304L SSs showed moderate cracking susceptibilities in the low-corrosion potential environments up to ~7 dpa. Periodic partial unloading (PPU) had a significant impact on the crack growth behavior of all irradiated specimens, and the measured CGRs were more than one order of magnitude higher in the test periods with than without PPU. A “stair-step” crack growth behavior was also observed for all irradiated specimens during the test periods with PPU, suggesting a strong effect of dynamic loading on IASCC. The SA 304L SS showed the best fracture resistance, and the CW 316 SS was the most brittle among the tested specimens. The fracture toughness J values of the two CW 316-Ti SS specimens were nearly identical at ~5 and ~8 dpa.

Chen, Y.↗
DOE OSTI2021

Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

3013 DE Inner Container Closure Weld Corrosion Evaluation

Destructive evaluation (DE) of 3013 containers is one part of the U. S. Department of Energy Integrated Surveillance Program. During standard DE of 3013 containers, visual examinations for pitting and stress corrosion cracking (SCC) are performed on the accessible surfaces of the outer, inner, and convenience containers, which make up the 3013 container. As a result of 3013 DE additional analysis, the area near the inner container closure weld has been identified as being a region of increased corrosion susceptibility, which may provide a pathway for corrosive gases to the outer container. This area has a higher residual stress, an altered microstructure, and less corrosion resistant weld oxides as a result of the welding process as well as a lower temperature than other areas of the container, which may increase the absorption of moisture on the surface. The deposition of moisture in this stressed region could lead to pitting and stress corrosion cracking. During FY2013, the inner container closure weld area was more closely evaluated on several archived samples from DE containers. These containers included FY09 DE2, FY12 DE4, FY12 DE6 and FY12 DE7 and the Hanford High Moisture Container. The additional examinations included visual observations with a stereomicroscope, scanning electron microscopy along with energy dispersive spectroscopy for chemical analysis, and serial metallography of the sidewall and lid that are part of the inner container closure weld region. Pitting was observed in all the samples taken from the closure weld regions of the examined inner containers. This pitting was generally less 20 µm with most less than 5µm. These pits were similar in depth to those observed in the vapor exposed surfaces of teardrops in the shelf life corrosion testing. Cracking was not observed on either the vapor-exposed surfaces of the teardrop coupons or the inner container closure weld region. Further testing is necessary to determine if the conditions in the welded inner container could support SCC during the 50 year life time for the 3013 container.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Aging Hanford Nuclear Waste Using Carbon Dioxide to Evaluate Corrosion at Lower pH Limits - 18467

Samples of Hanford Nuclear Reservation (HNR) tank waste and simulated Hanford tank waste were exposed to carbon dioxide gas (CO{sub 2}) in the laboratory in order to decrease the pH of the solution. Sodium hydroxide is required to be added to some tanks according to waste handling specifications to keep tanks protected from pitting corrosion, but the maximum-allowed volume in some tanks is currently approaching capacity. Decreasing the pH of tank solutions using CO{sub 2} gas mimics the environmental exposure chemistry, does not add sodium, and allows for testing on solutions currently categorized as out-of-specification according to technical waste handling documents. Legacy alloy A537 Class 1 carbon steel was used to construct the AN tanks on the HNR and was utilized for this testing. The planned testing of three solutions included a simulant for feasibility and a waste solution from Tank 241-AN-102 (AN-102), taken in 2013, which was run in duplicate. The solutions were evaluated for corrosion potential using cyclic potentiodynamic polarization (CPP) scans to determine the propensity for stress corrosion cracking and pitting corrosion. Results in all solutions showed that pitting potential was low when the pH was above 11 and temperatures were below 40 deg. C. As solution pHs dipped into the 10- 10.5 range, there were mixed results. CPP scans were initially run at the higher temperature of 50 deg. C, accelerating corrosion rates. Solution pHs that did not indicate corrosion were further exposed to CO{sub 2}. Solutions with mixed or positive pitting results were rerun at 40 deg. C. The first failed run on the tank solution was seen when testing the duplicate sample at 50 deg. C and a pH of 13.01. When this solution was retested at 40 deg. C, the CPP scan did not show corrosion. A mixed corrosion result was obtained from a scan run at the final pH of 10 when tested at 50 deg. C. Upon completion of testing a visual inspection of the coupon did not show corrosion products on its surface. These results provide assurance that tanks containing slightly reduced pHs are still protected from pitting corrosion. Tank integrity assurance at a lower pH limit in Tank AN-102 could slow sodium addition and the rising volume of hazardous waste. Other benefits of reducing the pH specification would be reduced costs associated with retrieving and storing smaller volumes of hazardous waste, a reduced number of radiation exposure events (required for pH modification of tank waste), and a potential for a decreased volume required to be immobilized for long-term storage in a vitrified (glass) phase. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

In-service corrosion and grain boundary oxidation in neutron-irradiated 316 stainless steel baffle-former bolts

Reactor core internal components such as baffle-former bolts (BFBs) are subjected to significant mechanical stress, corrosive environment, and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and of the bolts. In this work, characterization was performed on the oxidized surface of stainless steel BFBs harvested from a commercial pressurized water reactor (PWR) after 40 years of operation. The analysis shows that a complex multilayered surface oxide with six identified layers formed that is different from 2-layer structure commonly observed in model experiments. The oxide varies by composition – predominantly Fe, Cr, and Ni, grain size, and phase, and has features resembling both unirradiated and radiation/ corrosion experiments likely due to the low radiation flux compared to ion-irradiation or the test reactor radiation. In addition, grain boundary oxidative attack featured a pathway for Fe and other elements to move from the metal matrix to the outermost oxide. In conclusion, the results help assess PWR lifetime extension, put into context previous experimental studies, and provide input for designing experiments combining radiation and corrosion effects.

Baffle-former bolt↗
DOE OSTI2022

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray did not find evidence of LRO in any of the three heats aged to an equivalent of 60 years of service. Testing in a primary water environment of a heat of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. However, the SCC CGR response of the aged sample appears to show a deterioration in performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

The Effect of Thermal Aging on the Performance of High-Cr Ni-based Alloys in Primary Water Environment

Nickel-based Alloy 690, and the associated weld Alloys 52 and 152, are used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels. They are chosen for their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate beyond their first or second license renewal (e.g., 40 to 60 years or beyond 60 years). Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants. Moreover, Ni-Cr alloys are being considered for high-level radioactive waste disposal containers that will be required to contain the waste for at least 10,000 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Environmentally Assisted Cracking Research for Current and Advanced Nuclear Structural Materials

Understanding environmentally assisted cracking is an important aspect of nuclear materials research, as it provides information about failure mechanisms that lead to safety concerns or shutdowns. Idaho National Laboratory has ongoing research to understand how materials behave in the harsh nuclear environments for both the current fleet of light water reactors, as well as advanced reactors, where materials are exposed to more extreme environments. Corrosion and cracking of current materials, advanced materials, and materials formed through advanced manufacturing techniques are being examined, with particular interest in irradiation effects on corrosion and cracking, as well as the effects of environmental impurities. Testing, such as stress corrosion cracking, corrosion fatigue, and in-situ crack growth measurements, as well as a suite of characterization tools, including electron microscopy techniques and x-ray computed tomography, are being utilized to better understand the material response, limitations, and cracking mechanisms to ensure safe and reliable nuclear plants.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Summary of available data for estimating chloride-induced SCC crack growth rates for 304/316 stainless steel.

The majority of existing dry storage systems used for spent nuclear fuel (SNF) consist of a welded 304 stainless steel container placed within a passively-ventilated concrete or steel overpack. More recently fielded systems are constructed with dual certified 304/304L and in some cases, 316 or 316L. In service, atmospheric salts, a portion of which will be chloride bearing, will be deposited on the surface of these containers. Initially, the stainless steel canister surface temperatures will be high (exceeding the boiling point of water in many cases) due to decay heat from the SNF. As the SNF cools over time, the container surface will also cool, and deposited salts will deliquesce to form potentially corrosive chloride-rich brines. Because austenitic stainless steels are prone to chloride-induced stress corrosion cracking (CISCC), the concern has been raised that SCC may significantly impact long-term canister performance. While the susceptibility of austenitic stainless steels to CISCC in the general sense is well known, the behavior of SCC cracks (i.e., initiation and propagation behavior) under the aforementioned atmospheric conditions is poorly understood.

36 MATERIALS SCIENCE↗
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↗