Corrosion of 316 stainless steel in high temperature molten Li2BeF4 (FLiBe) salt
Not Available
Zheng, Guiqiu↗
Engineering topics
Sridharan, Kumar
Publications and source records attributed to Sridharan, Kumar.
Observations of Ag diffusion in ion implanted SiC
The nature and magnitude of Ag diffusion in SiC has been a topic of interest in connection with the performance of tristructural isotropic (TRISO) coated particle fuel for high temperature gas-cooled nuclear reactors. Ion implantation diffusion couples have been revisited to continue developing a more complete understanding of Ag fission product diffusion in SiC. Ion implantation diffusion couples fabricated from single crystal 4H-SiC and polycrystalline 3C-SiC substrates and exposed to 1500–1625°C, were investigated in this study by transmission electron microscopy and secondary ion mass spectrometry (SIMS). The high dynamic range of SIMS allowed for multiple diffusion régimes to be investigated, including enhanced diffusion by implantation-induced defects and grain boundary (GB) diffusion in undamaged SiC. Lastly, estimated diffusion coefficients suggest GB diffusion in bulk SiC does not properly describe the release observed from TRISO fuel.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat
In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Stability Of Nanoclusters In 14YWT Oxide Dispersion Strengthened Steel Under Heavy Ion-irradiation By Atom Probe Tomography
14YWT oxide dispersion strengthened (ODS) ferritic steel was irradiated with of 5 MeV Ni2+ ions, at 300 °C, 450 °C, and 600 °C to a damage level of 100 dpa. The stability of Ti–Y–O nanoclusters was investigated by applying atom probe tomography (APT) in voltage mode, of the samples before and after irradiations. The average size and number density of the nanoclusters was determined using the maximum separation method. These techniques allowed for the imaging of nanoclusters to sizes well below the resolution limit of conventional transmission electron microscopy techniques. The most significant changes were observed for samples irradiated at 300 °C where the size (average Guinier radius) and number density of nanoclusters were observed to decrease from 1.1 nm to 0.8 nm and 12 × 1023 to 3.6 × 1023, respectively. In this study, the nanoclusters are more stable at higher temperature.
Environmental Molecular Sciences Laboratory↗
Advanced heat exchanger development for molten salts
This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700°C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850°C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.
42 ENGINEERING↗
Stability of nanoclusters in 14YWT oxide dispersion strengthened steel under heavy ion-irradiation by atom probe tomography
14YWT oxide dispersion strengthened (ODS) ferritic steel was irradiated with of 5 MeV Ni2+ ions, at 300 C, 450 C, and 600 C to a damage level of 100 dpa. The stability of Ti–Y–O nanoclusters was investigated by applying atom probe tomography (APT) in voltage mode, of the samples before and after irradiations. The average size and number density of the nanoclusters was determined using the maximum separation method. These techniques allowed for the imaging of nanoclusters to sizes well below the resolution limit of conventional transmission electron microscopy techniques. The most significant changes were observed for samples irradiated at 300 C where the size (average Guinier radius) and number density of nanoclusters were observed to decrease from 1.1 nm to 0.8 nm and 12 1023 to 3.6 1023, respectively. In this study, the nanoclusters are more stable at higher temperature.
99 GENERAL AND MISCELLANEOUS↗
Response of 9Cr-ODS Steel to Proton Irradiation at 400 °C
The stability of Y–Ti–O nanoclusters, dislocation structure, and grain boundary segregation in 9Cr-ODS steels has been investigated following proton irradiation at 400 °C with damage levels up to 3.7 dpa. A slight coarsening and a decrease in number density of nanoclusters were observed as a result of irradiation. The composition of nanoclusters was also observed to change with a slight increase of Y and Cr concentration in the nanoclusters following irradiation. Size, density, and composition of the nanoclusters were investigated as a function of nanocluster size, specifically classified to three groups. In addition to the changes in nanoclusters, dislocation loops were observed after irradiation. Finally, radiation-induced enrichment of Cr and depletion of W were observed at grain boundaries after irradiation.
99 GENERAL AND MISCELLANEOUS↗
Freeze-casting as a Novel Manufacturing Process for Fast Reactor Fuels. Final Report
Advanced burner reactors are designed to reduce the amount of long-lived radioactive isotopes that need to be disposed of as waste. The input feedstock for creating advanced fuel forms comes from either recycle of used light water reactor fuel or recycle of fuel from a fast burner reactor. Fuel for burner reactors requires novel fuel types based on new materials and designs that can achieve higher performance requirements (higher burn up, higher power, and greater margins to fuel melting) then yet achieved. One promising strategy to improved fuel performance is the manufacture of metal or ceramic scaffolds which are designed to allow for a well-defined placement of the fuel into the host, and this in a manner that permits greater control than that possible in the production of typical CERMET fuels.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Characterization of structural defects in nuclear graphite IG-110 and NBG-18
Nuclear graphite IG-110 and NBG-18 were examined using X-ray diffraction (XRD), Raman spectroscopy, scanning electron microscope (SEM) and high resolution transmission electron microscope (HR-TEM) to understand the structure and microstructure of nuclear graphite. The lattice parameter (a), degree of graphitization ( ), crystallite size parallel and perpendicular to c-direction (Lc and L ), anisotropy (B), as well as in-plane crystallite size (La) were calculated and compared based on XRD patterns and Raman spectra. Results indicate that IG-110 has a larger crystallite size and higher degree of graphitization, but lower anisotropy than NBG-18. These differences are attributed to the properties of coke source and manufacturing processes. Additionally, the shape of the pores and crystallized filler particles, the interface between binders and fillers, Mrozowski cracks and nano-cracks, and the defects of disclination were observed and characterized from SEM and HR-TEM images. The similarities and differences in microstructure between IG-110 and NBG-18 are discussed. The results in this work provide useful information to guide selection of nuclear graphite for the design of next generation nuclear plants (NGNP).
99 GENERAL AND MISCELLANEOUS↗
Relationship between lath boundary structure and radiation induced segregation in a neutron irradiated 9 wt.% Cr model ferritic/martensitic steel
Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.
36 MATERIALS SCIENCE↗
Study of Interfacial Interactions Using Thing Film Surface Modification: Radiation and Oxidation Effects in Materials
Interfaces play a key role in dictating the long-term stability of materials under the influence of radiation and high temperatures. For example, grain boundaries affect corrosion by way of providing kinetically favorable paths for elemental diffusion, but they can also act as sinks for defects and helium generated during irradiation. Likewise, the retention of high-temperature strength in nanostructured, oxide-dispersion strengthened steels depends strongly on the stoichiometric and physical stability of the (Y, Ti)-oxide particles/matrix interface under radiation and high temperatures. An understanding of these interfacial effects at a fundamental level is important for the development of materials for extreme environments of nuclear reactors. The goal of this project is to develop an understanding stability of interfaces by depositing thin films of materials on substrates followed by ion irradiation of the film-substrate system at elevated temperatures followed by post-irradiation oxidation treatments. Specifically, the research will be performed by depositing thin films of yttrium and titanium (~500 nm) on Fe-12%Cr binary alloy substrate. Y and Ti have been selected as thin-film materials because they form highly stable protective oxides layers. The Fe-12%Cr binary alloy has been selected because it is representative of ferritic steels that are widely used in nuclear systems. The absence of other alloying elements in this binary alloy would allow for a clearer examination of structures and compositions that evolve during high-temperature irradiations and oxidation treatments. The research is divided into four specific tasks: (1) sputter deposition of 500 nm thick films of Y and Ti on Fe-12%Cr alloy substrates, (2) ion irradiation of the film-substrate system with 2MeV protons to a dose of 2 dpa at temperatures of 300°C, 500°C, and 700°C, (3) oxidation of as-deposited and ion-irradiated samples in a controlled oxygen environment at 500°C and 700°C, (4) multi-scale computational modeling involving first- principle molecular dynamics (FPMD) and coarse-grained dissipative particle dynamics (DPD) approaches to develop theories underlying the evolution and stability of structures and phases. Samples from Tasks 1 to 3 (above) will be rigorously characterized and analyzed using scanning electron microscopy, Auger electron microscopy, x-ray diffraction, Rutherford back scatter spectroscopy, and transmission electron microscopy. Expected outcomes of the experimental work include a quantitative understanding film-substrate interface mixing, evolution of defects and other phases at the interface, interaction of interfaces with defects, and the ability of the Y and Ti films to mitigate irradiation-assisted oxidation.The aforementioned experimental work will be closely coupled with multi-scale molecular dynamics (MD) modeling to understand the reactions at the surface, the transport of oxidant through the thin film, and the stabilities of the deposited thin films under radiation and oxidation. Simulations of materials property changes under conditions of radiation and oxidation require multiple size domains and a different simulation scheme for each of these domains. This will be achieved by coupling the FPMD and coarse-grained kinetic Monte Carlo (KMC). This will enable the comparison of the results of each simulation approach with the experimental results.
36 MATERIALS SCIENCE↗
Corrosion in Supercritical Carbon Dioxide: Materials, Environmental Purity, Surface Treatments, and Flow Issues (Project No. 10-872)
The supercritical CO 2 Brayton cycle is gaining importance for power conversion in the Generation IV fast reactor system because of its high conversion efficiencies. When used in conjunction with a sodium fast reactor, the supercritical CO 2 cycle offers additional safety advantages by eliminating potential sodium-water interactions that may occur in a steam cycle. In power conversion systems for Generation IV fast reactors, supercritical CO 2 temperatures could be in the range of 30°C to 650°C, depending on the specific component in the system. Materials corrosion primarily at high temperatures will be an important issue. Therefore, the corrosion performance limits for materials at various temperatures must be established. The proposed research will have four objectives centered on addressing corrosion issues in a high-temperature supercritical CO 2 environment: Task 1: Evaluation of corrosion performance of candidate alloys in high-purity supercritical CO 2 : The following alloys will be tested: Ferritic-martensitic Steels NF616 and HCM12A, austenitic alloys Incoloy 800H and 347 stainless steel, and two advanced concept alloys, AFA (alumina forming austenitic) steel and MA754. Supercritical CO 2 testing will be performed at 450°C, 550°C, and 650°C at a pressure of 20 MPa, in a test facility that is already in place at the proposing university. High purity CO 2 (99.9998%) will be used for these tests. Task 2: Investigation of the effects of CO, H 2 O, and O 2 impurities in supercritical CO 2 on corrosion: Impurities that will inevitably present in the CO 2 will play a critical role in dictating the extent of corrosion and corrosion mechanisms. These effects must be understood to identify the level of CO 2 chemistry control needed to maintain sufficient levels of purity to manage corrosion. The individual effects of important impurities CO, H 2 O, and O 2 will be investigated by adding them separately to high purity CO 2 . Task 3: Evaluation of surface treatments on the corrosion performance of alloys in supercritical CO 2 : Surface treatments can be very beneficial in improving corrosion resistance. Shot peening and yttrium and aluminum surface treatments will be investigated. Shot peening refines the surface grain sizes and promotes protective Cr-oxide layer formation. Both yttrium and aluminum form highly stable oxide layers (Y 2 O 3 and Al 2 O 3 ), which can get incorporated in the growing Fe-oxide layer to form an impervious complex oxide to enhance corrosion resistance. Task 4: Study of flow-assisted corrosion of select alloys in supercritical CO 2 under a selected set of test conditions: To study the effects of flow-assisted corrosion, tests will be conducted in a supercritical CO 2 flow loop. An existing facility used for supercritical water flow studies at the proposing university will be modified for use in this task. The system is capable of flow velocities up to 10 m/s and can operate at temperatures and pressures of up to 650°C and 20 MPa, respectively. All above tasks will be performed in conjunction with detailed materials characterization and analysis using scanning electron microscopy/energy dispersive spectroscopy (SEM-EDS), x-ray diffraction (XRD), Auger electron spectroscopy (AES) techniques, and weight change measurements. Inlet and outlet gas compositions will be monitored using gas chromatography-mass spectrometry (GCMS).
36 MATERIALS SCIENCE↗
Understanding the Irradiation Behavior of Zirconium Carbide
Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450{degree}C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC-based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory, (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800{degree}C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of Diffusion Barrier Coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)
The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Improved LWR Cladding Performance by EPD Surface Modification Technique
This project will utilize the electro-phoretic deposition technique (EPD) in conjunction with nanofluids to deposit oxide coatings on prototypic zirconium alloy cladding surfaces. After demonstrating that this surface modification is reproducible and robust, the team will subject the modified surface to boiling and corrosion tests to characterize the improved nucleate boiling behavior and superior corrosion performance. The scope of work consists of the following three tasks: The first task will employ the EPD surface modification technique to coat the surface of a prototypic set of zirconium alloy cladding tube materials (e.g. Zircaloy and advanced alloys such as M5) with a micron-thick layer of zirconium oxide nanoparticles. The team will characterize the modified surface for uniformity using optical microscopy and scanning-electron microscopy, and for robustness using standard hardness measurements. After zirconium alloy cladding samples have been prepared and characterized using the EPD technique, the team will begin a set of boiling experiments to measure the heat transfer coefficient and critical heat flux (CHF) limit for each prepared sample and its control sample. This work will provide a relative comparison of the heat transfer performance for each alloy and the surface modification technique employed. As the boiling heat transfer experiments begin, the team will also begin corrosion tests for these zirconium alloy samples using a water corrosion test loop that can mimic light water reactor (LWR) operational environments. They will perform extended corrosion tests on the surface-modified zirconium alloy samples and control samples to examine the robustness of the modified surface, as well as the effect on surface oxidation
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation
This project addresses both practical and fundamental scientific issues of direct relevance to operational challenges of the molten LiCl-KCl salt pyrochemical process, while providing avenues for improvements in the process. In order to understand the effects of the continually changing composition of the molten salt bath during the process, the project team will systematically vary the concentrations of rare earth surrogate elements, lanthanum, cerium, praseodymium, and neodymium, which will be added to the molten LiCl-KCl salt. They will also perform a limited number of focused experiments by the dissolution of depleted uranium. All experiments will be performed at 500 deg C. The project consists of the following tasks. Researchers will measure density of the molten salts using an instrument specifically designed for this purpose, and will determine the melting points with a differential scanning calorimeter. Knowledge of these properties is essential for salt mass accounting and taking the necessary steps to prevent melt freezing. The team will use cyclic voltammetry studies to determine redox potentials of the rare earth cations, as well as their diffusion coefficients and activities in the molten LiCl-KCl salt. In addition, the team will perform anodic stripping voltammetry to determine the concentration of the rare earth elements and their solubilities, and to develop the scientific basis for an on-line diagnostic system for in situ monitoring of the cation species concentration (rare earths in this case). Solubility and activity of the cation species are critically important for the prediction of the salt's useful lifetime and disposal.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Development of Diffusion Barrier Coatings for Miti
Fuel Cladding Chemical Interactions (FCCI) in a nu
99 GENERAL AND MISCELLANEOUS↗