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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,441 records · Page 80
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

The corrosion behavior of technetium metal exposed to aqueous sulfate and chloride solutions

Here, metal waste forms are being studied as possible disposal forms for technetium and other fission products from spent nuclear fuel. As an initial step in assessing the viability of waste forms, technetium corrosion and passivity behavior was assessed across a broad pH spectrum (pH –1 to pH 13). Measurements indicate that the open circuit potential falls into the region of Tc +7 stability, more noble than the region of presumed passivity. Potentiodynamic polarization tests indicate that the Tc samples are not passive. Both electrochemical results and visual inspection suggest the presence of a nonprotective film. The corrosion rate is relatively independent of pH and low, as measured by linear polarization resistance. No evidence of passivity was observed in the Tc +4 region of the potential-pH diagram following in-situ abrasion, suggesting that Tc does not passivate, regardless of potential.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Effect of tritium on corrosion behavior of chromium in 0.01 N sulfuric acid solution

The effects of tritium on the corrosion behavior of chromium in 0.01 N sulfuric solution have been investigated in the present study. Electrochemical experiments have been carried our for pure chromium. At first, the concentration dependence of sulfuric acid solution on anodic polarization behavior of chromium was experimented, resulting in that 0.01 N one was found appropriate. The dependence of both dissolved oxygen and tritium concentration on anodic behavior of chromium were performed. It was found from that the self-passivation of chromium induced by dissolved oxygen was inhibited in tritiated solution resulting in the enhancement of the corrosion. As a consequence it is highly likely that the elution of chromium by highly oxidative radiolysis products would explain the passivation inhibitory effect of SUS304 stainless steel observed in tritiated solutions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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 OSTI2016

Investigating the effect of zirconium oxide microstructure on corrosion performance: a comparison between neutron, proton, and nonirradiated oxides - 2016-0069

Scanning precession electron diffraction in the transmission electron microscope has been used to simultaneously map the phase, orientation, and grain morphology of oxides formed on Zircaloy-2 after three and six cycles in a boiling water reactor in unprecedented detail. For comparison, a region of a preoxidized autoclave-formed oxide was also proton-irradiated at the Dalton Cumbrian Facility. The proton irradiation was observed to cause additional stabilization of the tetragonal phase that was attributed to the stabilizing effect of irradiation- induced defects in the oxide. In the reactor-formed oxides, no extra stabilization of the tetragonal grains was observed under neutron irradiation, as indicated by the similar tetragonal phase fraction and transformation twin-boundary distributions between the nonirradiated and reactor-formed oxides. It is suggested that the damage rate is too low in the newly formed oxide to cause significant stabilization of the tetragonal phase. This technique also reveals that the oxide formed under reactor conditions has a more heterogeneous microstructure, and the growth of well-oriented columnar monoclinic grains is significantly reduced compared with a nonirradiated oxide. High-angle annular dark-field scanning transmission electron microscopy also revealed the development of extensive networks of intergranular porosity and eventually grain de-cohesion in the reactor-formed oxides. These results suggest that the tetragonal-monoclinic transformation is not responsible for the accelerated corrosion exhibited under reactor conditions. It is proposed that the usual out-of-reactor oxide growth and nucleation processes are significantly modified under reactor conditions, resulting in a more heterogeneous and randomly oriented oxide microstructure with reduced columnar grain growth. It is suggested that this disordered oxide microstructure allows for the formation of extensive intergranular porosity that could lead to accelerated in-reactor corrosion. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Influence of hydride precipitation on the corrosion kinetics of Zircaloy-4: effect of the nano-structure and grain boundary properties of the zirconium oxide layer on oxygen diffusion flux - 2016-0073

Hydride precipitation at the oxide-metal interface is frequently proposed as causing the corrosion acceleration of Zircaloy-4 at high burnup in pressurized water reactors (PWRs). In order to identify the local mechanisms possibly involved, we studied the nano-structure of oxides formed on massive zirconium hydrides and reference Zircaloy-4 with an innovative grain mapping technique with the use of transmission electron microscopy (TEM). In autoclave PWR-conditions, the presence of a precipitated hydride phase, previously formed by a cathodic charging technique at the surface of Zircaloy-4, clearly increased the corrosion rate, and a higher oxygen diffusion flux along oxide grain boundaries is observed compared with the reference Zircaloy-4. The texture, grain size, and adjacent grain misorientation in the oxide layer were studied in detail along the direction of the oxide growth using an automated crystal orientation mapping technique associated with TEM for both pre-hydrided and reference Zircaloy-4 samples. The texture in the growth direction is similar on pre-hydrided and reference samples, but the grain-to-grain mis-orientations showed differences. Indeed, on the pre-hydrided sample, the misorientation of 90 deg. C with respect to the [001] monoclinic axis is less probable than in the reference oxide, and more mis-orientations of 50-70 deg. C and 120-150 deg. C, corresponding to larger mismatches between neighboring grain boundaries, are observed. A smaller average diameter of the columnar monoclinic grains is also clearly revealed for the oxide grown on the pre-hydrided sample that leads to a larger number of diffusion paths for oxidizing species. These results are discussed and used for simulating oxygen diffusion flux through the polycrystalline microstructure of the oxide layer as a function of the grain size. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2011

Thermodynamic Development of Corrosion Rate Modeling in Iron Phosphate Glasses

A two-year research program investigated links between the thermodynamic properties of phosphate glasses and their corrosion rates in different solutions. Glasses in the Na 2 O-CaO-P 2 O 5 and Na 2 O-Fe 2 O 3 -PO 5 systems were prepared and characterized. These glasses were then exposed in bulk and powder form to acid (0.1M HCl), basic (0.1M KOH) and neutral (deionized water) solutions at varying exposure times and temperatures. Analysis of the solution and the glass after exposure determined the rate and type of corrosion that occurred. Simultaneously, efforts were made to determine the thermodynamic properties of solid iron phosphate compounds. This included measurement of low temperature (5-300 K) heat capacities, measured at Brigham Young University; the attempted use of a Parr calorimeter to measure ambient temperature enthalpies of formation; and attempted measurement of temperature heat capacities. Only the first of the three tasks was successfully accomplished. In lieu of experimental measurement of enthalpies of formation, first-principles calculation of enthalpies of formation was performed at Missouri S&T; these results will be used in subsequent modeling efforts.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Effect of Chloride and Sulfate Concentration on Probability Based Corrosion Control for Liquid Waste Tanks- Part IV

A series of cyclic potentiodynamic polarization tests was performed on samples of A537 carbon steel in support of a probability-based approach to evaluate the effect of chloride and sulfate on corrosion susceptibility. Testing solutions were chosen to build off previous experimental results from FY07, FY08, FY09 and FY10 to systemically evaluate the influence of the secondary aggressive species, chloride, and sulfate. The FY11 results suggest that evaluating the combined effect of all aggressive species, nitrate, chloride, and sulfate, provides a consistent response for determining corrosion susceptibility. The results of this work emphasize the importance for not only nitrate concentration limits, but also chloride and sulfate concentration limits as well.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Erosion-Corrosion of Iron and Nickel Alloys at Elevated Temperature in a Combustion Gas Environment

This paper reports on the results of a study that compares the erosion-corrosion behavior of a variety of alloys (Fe- 2¼Cr 1Mo, 304 SS, 310 SS, Incoloy 800, Haynes 230 and a Fe 3 Al) in a combustion environment. Advanced coal combustion environments, with higher temperatures, are driving re-examination of traditional and examination of new alloys in these hostile environments. In order to simulate conditions in advanced coal combustion boilers, a special erosion apparatus was used to allow for impingement of particles under a low abrasive flux in a gaseous environment comprised of 20 % CO 2 , 0.05 % HCl, 77 % N 2 , 3 % O 2 , and 0.1 % SO 2 . Tests were conducted at room temperature and 700 °C with ~ 270 μm silica, using an impact velocity of 20 m/s in both air and the simulated combustion gas environment. The erosion-corrosion behavior was characterized by gravimetric measurements and by examination of the degraded surfaces optically and by scanning electron microscopy (SEM). At room temperature most of the alloys had similar loss rates. Not surprisingly, at 700 °C the lower chrome-iron alloy had a very high loss rate. The nickel alloys tended to have higher loss rates than the high chrome austenitic alloys.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2015

Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re-evaluate thermophysical properties of flibe and flinak. Pacific Northwest National Laboratories has focused on evaluating the fluorinating gas nitrogen trifluoride as a potential salt purification agent. Work there was performed on removing hydroxides and oxides from flinak salt under controlled conditions. Lastly, the University of California Berkeley has spent considerable time designing and simulating reactor components with fluoride salts at high temperatures. Despite the hurdles presented by the innate chemical hazards, considerable progress has been made. The stage has been set to perform new research on salt chemical control which could advance the fluoride salt cooled reactor concept towards commercialization. What were previously thought of as chemical undesirable, but nuclear certified, alloys have been shown to be theoretically compatible with fluoride salts at high temperatures. This preliminary report has been prepared to communicate the construction of the basic infrastructure required for flibe, as well as suggest original research to performed at the University of Wisconsin. Simultaneously, the contents of this report can serve as a detailed, but introductory guide to allow anyone to learn the fundamentals of chemistry, engineering, and safety required to work with flibe salt.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Report on Status of Shipment of High Fluence Austenitic Steel Samples for Characterization and Stress Corrosion Crack Testing

The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructural and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The preliminary work for sample shipment between Halden and Oak Ridge includes fabrication of an inner cask sample container, decontamination and preparation of a Type A container, preparation of new activity calculations, all necessary paperwork, and handling. ORNL will continue to work to track progress of sample preparation and shipment status, and to work toward an agreement that covers material shipping costs between the Halden Reactor and the Oak Ridge National Laboratory.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Impact of glycolate anion on aqueous corrosion in DWPF and downstream facilities

Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to other high level waste (HLW) facilities. The impact of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system has not been previously evaluated. A literature review had revealed that corrosion data were not available for the MoCs in glycolic-bearing solutions applicable to SRS systems. Data on the material compatibility with only glycolic acid or its derivative products were identified; however, data were limited for solutions containing glycolic acid or the glycolate anion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Effect of Swelling on Irradiation-Assisted Stress Corrosion Cracking

A highly irradiated stainless steel 304 material irradiated to 27 dpa in the Experimental Breeder Reactor-II was used to investigate the relation between void swelling, grain boundary cohesion and intergranular crack growth rate under solicitation by irradiation assisted stress corrosion cracking. The material selected offered two different swelling amounts (about 2% and 3.7%) in the same material at high dose. The material had been irradiated at the same time, same condition, only the temperature gradient led to different swelling. This material, however, proved to be very resistant to irradiation assisted stress corrosion cracking under pressurized water reactor conditions. The crack growth rates measured were very low with the lowest being obtained with the material exhibiting the most amount of swelling. Micromechanical testing performed with the material exhibiting the higher crack propagation rate demonstrated that the irradiation damages (including void swelling) did not decrease grain boundary cohesion enough to make the grain boundaries the weak points under load solely. Crack path and crack tip analysis showed evidence of material modification ahead of the crack but it is not clear, at this point, if the changes observed were caused by the crack advance or if their presence in front of the crack was causing the low crack growth rate measured.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Advanced surface plasma nitriding for development of corrosion resistant and accident tolerant fuel cladding: Final Report

Although various surface coating techniques have been proposed to increase oxidation and corrosion resistance of fuel cladding materials, the de-bonding of the coating layer with the original cladding matrix under exposure to coolants makes such approaches unsuitable for reactor applications. Furthermore, the feasibility of techniques for large scale processing on cladding tubes remains another technological bottleneck. This project aims to develop a hollow cathode plasma nitriding technique to solve the above issues. Starting with alloys of DOE interest, the team will apply an advanced surface plasma nitriding technique to convert alloy surface layers into nitride layers for better structural integrity and compatibility with both coolants and nuclear fuels. The technique can form uniform nitride layers on both inner and outer surfaces of cylindrical hollow tubes. Both treated and untreated samples will be irradiated by using Fe self ions or dual beams of Fe ions and He ions. Various microstructural characterizations and mechanical property measurements will be performed. For outside surfaces, samples will be tested in high temperature water and liquid molten salt to evaluate their corrosion resistance and compatibility with coolants for LWR and fast reactor applications. For inside surfaces, solid-to-solid diffusion couple experiments will be used to study interface reactions of nitrided samples with uranium, to test whether a nitrided layer can slow down fuel-cladding interactions. The project will impact both the development of advanced methods for manufacturing and the development of advanced reactor in-core structural materials. The proposed technique can form nitride layers on alloys of irregular shapes without an edging effect - an issue encountered in traditional plasma nitriding processes, and is suitable for surface processing of other in-core components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Stress Corrosion cracking Testing in Supercritical Carbon Dioxide

This document presents a summary of the ldrd project. The project consists in developing testing capability to address stress corrosion cracking issues in supercritical carbon dioxide. In addition, this project is generating corrosion data. This work aims to help select structural materials for supercritical carbon dioxide energy systems.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Impact of Permanganate Oxidation of Glycolate on Corrosion of the Defense Waste Processing Facility (DWPF) Recycle Collection Tank (RCT), Transfer-line and Waste Tank Materials of Construction (MoC)

Savannah River Remediation (SRR) is evaluating the use of glycolic acid to replace formic acid as an alternate reductant for the Defense Waste Processing Facility (DWPF) at Savannah River Site (SRS). The replacement is aimed at reducing facility hazards, improving pH stability and rheological control, and lowering off gas production. Glycolate, which is the product that may remain, can also be carried downstream to High-Level Waste and Low-Level Waste facilities. Complete oxidation of glycolate and other organic species from the Sludge Receipt and Adjustment Tank (SRAT) may be necessary to minimize thermolytic hydrogen generation that can cause flammability issues. The glycolate oxidation may be performed in the Recycle Collection Tank (RCT). The RCT heel would be adjusted with sodium hydroxide to maintain the solution caustic and sodium nitrite as corrosion inhibitor prior to the addition of condensate that can be from the contents of the Slurry Mix Evaporator Condensate Tank (SMECT) or Offgas Condensate Tank (OGCT). The RCT tank collects the waste to be transferred via the Low Point Pump Pit – Recycle Pump Tank (LPPP-RPT) through a jacketed transfer-line to the Tank Farm facility. Scoping studies were performed to evaluate an oxidizing agent and sodium permanganate was selected as the best option for glycolate destruction. Testing was recommended to determine the impact on the Materials of Construction (MoC) due to the addition of sodium permanganate. The targeted concentration of sodium permanganate used equated to a 7:1 molar permanganate to glycolate ratio (nominally 0.002 M permanganate). However, to conservatively maximize the permanganate corrosivity, in some cases no glycolate was added.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Corrosion of Steel During Long-Term Exposure to Evolving Cementitious Environments

SRS is proceeding with closure of the H-Area Tank Farm (HTF) and F-Area Tank Farm (FTF). Closure consists of removing the bulk waste, heel removal, and filling the tank with tailored grout formulations. The long-term performance of the concrete and steel materials of construction and the closure grout is an important consideration for Performance Assessment (PA) of closed tanks in FTF and HTF. Over PA timeframes of hundreds, thousands, to tens of thousands of years, the chemical and physical properties of the concrete and steel materials will slowly degrade due to environmental exposure and material aging. In the interim these materials will provide a barrier to the leaching of radionuclides into the soil. Savannah River Remediation (SRR) is updating relevant portions of these analysis inputs to support an imminent revision to the HTF PA and a future update to the FTF PA. This analysis provides an update to the previous PA inputs for steel corrosion. The Central Scenario for the PA analysis includes three postulated modeling cases: 1) Realistic Case, 2) Compliance Case, and 3) Pessimistic Case. This analysis also included a Fast Flow Path Case. This latter case considered the circumstance where the initial condition of the concrete was a completely degraded state and that grout shrinkage exposed the steel to the soil environment immediately. In effect, the steel was unprotected by the concrete and grout. The cases have various degrees of conservatism considered. Chemical, physical and tank configuration parameters were investigated to understand their effects on the predicted time to release of the contaminants. The assessment reviewed the initial tank and steel configuration, service life degradation of the steel, and potential corrosion mechanisms associated with degradation of the concrete materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Combined modeling and experiments to predict corrosion and embrittlement in dual-phase stainless steels within the MARMOT framework

Extending service of Light Water Reactors (LWRs) to beyond 60 years will demand high integrity of materials and components in the reactors. The accurate evaluation and prediction of materials performance under anticipated operating conditions are of particular importance for ensuring the safety of nuclear power plants over their extended lifetime. Stainless steels are extensively used as structural alloys in light water reactor (LWR) systems because of their excellent combination of mechanical properties and corrosion resistance. Austenitic stainless steel welds and cast austenitic stainless steels (CASS) contain significant amounts of ferrite ~5-30% and are ubiquitous in LWR piping (elbows, pump casings, valves) and internal structural components. During their lifetime, these components are subjected to thermal aging at the temperature range of 288-327°C for pressurized water reactors (PWR). A phase transformation within the ferrite phase of these materials leads to embrittlement and degradation in corrosion resistance, which undermines the materials selection criteria.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Development of In-Situ Corrosion Kinetics and Salt Property Measurements (Final Technical Report)

The goal of this research is to fill the knowledge gaps of salt properties and gain a fundamental understanding of corrosion mechanisms, thereby to guiding material selections of salts and containment materials. The proposed research is focused on building unique cross-cutting research capabilities that can perform research and analysis relevant to the following three research topics important for Generation 3 Concentrating Solar Power Systems: (1) Material characterization including investigations of fluid thermophysical properties and stability, (2) Durability testing of containment materials, and (3) Corrosion behavior characterization relative to levels of known contaminants (e.g., water and oxygen) in Heat Transfer Fluid (HTF).

14 SOLAR ENERGY↗