227 Reber Building
University Park, PA 16802
This project is part of a collaborative research program involving universities from the United States, England, and Sweden to study hydrogen pickup during aqueous corrosion of zirconium-based alloys. This will be achieved by a combination of both experimental and theoretical/modeling studies during the thesis research of eight PhD candidates. The universities involved provide a breadth of complementary techniques and approaches that will be used during the execution of the program.
The research will focus on the relation of hydrogen pickup to corrosion kinetics and on the role of alloy microstructure and chemical composition in determining hydrogen pickup. The study will focus on identifying mechanisms for the pickup of hydrogen in autoclave sample, which allows the inclusion of a broader range of alloy chemistries and microstructures in the study. This provides the opportunity to examine samples with extremes in their hydrogen pickup behavior and enhances the likelihood of identifying differences that can be rationalized relative to their impact on hydrogen pickup. It is anticipated that the techniques developed in this program could later be applied to in-reactor specimens.
Prompt Gamma Neutron Activation Analysis (PGNAA) provides a non-destructive technique to accurately quantify the hydrogen content in autoclave specimens. The technique utilizes the facility at NIST to measure hydrogen content on selected samples that will subsequently be returned to the autoclave for continued autoclave exposure. The information will provide detailed information on hydrogen uptake as a function of specimen weight gain from single specimens rather than reliance on measurements from sister specimens. The goal of the measurements is to determine the relationship between the hydrogen pickup and the cyclic growth behavior of the oxide. Measurement of hydrogen by PGNAA will also be supplemented by use of destructive techniques such as hot vacuum extraction or inert gas fusion. The results of these measurements will better quantify the dependence of HPUF on oxide thickness and its variation as a function of time.
Synchrotron radiation provides a relatively straight forward technique to characterize both the types and size of SPPs in zirconium-based alloys. The high intensity beam can be used to obtain diffraction peaks of minor phases while the peak width can be used to estimate the size of each particle type. In addition to conventional diffraction, the availability of micro-beams at the Advanced Photon Source (APS) provides the ability to probe the oxide and oxide-metal interface at a spatial resolution of about 200 nm. Micro-diffraction provides information on the interface structure and information on oxide phases, grain size, and texture as a function of distance from the oxide-metal interface. These oxide characteristics will be interpreted relative to their potential impact on the transport of hydrogen into the metal.
The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation.
This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime.
This research project is focused on the identification of the formation mechanism and evolution for dislocation loops with Burgers vector of a<100> and determine whether the defect microstructure (predominately dislocation loop/dislocation density) saturates at high dose. Another task is to identify whether a threshold irradiation temperature or dose exists for the nucleation of growing voids that mark the beginning of irradiation-induced swelling, and begin to probe the limits of thermal stability of the tempered martensitic structure under irradiation.
Hydride reorientation in irradiated nuclear fuel cladding during drying, storage, and transportation leads to increased cladding susceptibility to failure. Thus it is important to understand under which conditions such reorientation can occur so as to predict irradiated cladding performance. Current information on hydride reorientation behavior in Zr-based alloys is based on post-test examination of samples cooled under hoop stress. The degrading effects of radial hydrides are studied by subjecting the post-cooled sample to simple mechanical properties tests. With this approach, re-orientation kinetics are difficult to quantify.
We propose to perform in-situ synchrotron radiation diffraction experiments on hydrided Zircaloy-4 as a means to elucidate the mechanisms and kinetics of hydride reorientation during cooling under stress. Synchrotron radiation diffraction can be performed on bulk samples (thickness 1-2 mm) in transmission using a high energy (80 keV) beam. This technique allows us to obtain information on the dissolution and re-precipitation of second-phase hydrides as it happens, while at temperature and under load. The technique can determine the crystal structure of the hydrides phases present, their orientation relationship with the matrix, their volume fraction and macroscopic orientation. Importantly, recent experiments indicate that analysis of the diffraction data can allow us to differentiate between radial and circumferential hydrides.
We propose to perform such studies on model Zircaloy-4 plate material, hydrided by a gaseous charging procedure, and to verify these results by performing selected experiments on irradiated material, to be furnished by Argonne National Laboratory. Such an approach will allow us to explore in detail the range of parameters of importance to hydride reorientation (maximum temperature, cooling rate, applied load) using non-irradiated hydrided samples and to confirm that these results are applicable by performing tests on irradiated samples.
Because the beamline has a mechanical load frame and a furnace, the data can be obtained in situ (under load at temperature), and thus hydride dissolution and precipitation can be followed as it happens. Since the acquisition time for the diffraction patterns is short, the hydride kinetics can be followed in detail (one pattern per second is possible). Cooling rates can be programmed, enabling the temperature and time dependence of the degree of hydride reorientation and particle connectivity to be examined as a function of temperature and load history. The specimen geometry will be designed to produce the relevant state of stress on the sample, and controlled hydriding allows the investigation of specific initial hydride distributions.
The successful completion of this research program will help delineate under what conditions hydrided and irradiated zirconium fuel cladding maintains its integrity during storage.
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 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°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 (ZrC) by using a combination of the state-of-the-art experimental methods and atomistic modeling. This project will combine (i) in-situ ion irradiation at a specialized facility at a National Laboratory, (ii) controlled temperature proton irradiation on bulk samples, and (iii) 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 ºC and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperature and dose allows 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 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.
This project is a collaborative effort between Penn State University, Queen's University in Canada and CNEA (National Atomic Energy Commission in Argentina) to perform in-situ synchrotron radiation diffraction experiments and thermodynamic and mechanical modeling to elucidate the kinetics of hydrogen migration and hydride precipitation in zirconium alloys. It is proposed to examine the kinetics of hydride precipitation under high temperatures and stresses, such as can occur near a crack tip. This hydride precipitation and the associated crack advancement are at the root of the cracking process known as delayed hydride cracking (DHC).
In the experiments proposed, compact tension (CT) specimens of hydrided Zircaloy with a fatigue pre-crack will be subjected to temperature and stress while being examined using microbeam synchrotron radiation at the Advanced Photon Source in Argonne National Laboratory. The precipitation of the hydrides at the crack tip will be directly monitored in-situ by monitoring the diffraction signal of hydride precipitates forming at the crack tip. The crack growth behavior of the specimen will be monitored simultaneously using potential drop and acoustic emission. The role of texture will be investigated using three different CT samples, with the crack advance direction parallel or perpendicular to the basal pole in textured samples and also using more randomly oriented sample textures. Benchmarking experiments will be done first. Thermodynamic and mechanical modeling will be performed to assist in the interpretation the results.
This research will provide unique data on the kinetics of hydride precipitation at a crack tip, and as such, will help identify and quantify the mechanisms and operational limits for delayed hydride cracking. The unique aspect of this research project is that the measurements will be performed in situ, and with great spatial resolution, near the crack tip so that these localized processes can be directly monitored. The researchers are well well-qualified for the work, and have complementary research capabilities such that a collaboration should prove fruitful. There is presently much uncertainty on the DHC mechanisms, at least partly because the examinations have been conducted post-facto (after cooldown and stress relaxation), which can confound the interpretation of results. As a result, parameters such as the actual concentrations of hydrides at the crack tip necessary for advancement, the kinetics of precipitation under different states and levels of stress as well as temperatures are not well known.
The successful completion of this research project will result in increased knowledge about the fundamental mechanisms governing the precipitation of hydrides at crack tips under a stress field in hydride-forming metals such as Zr- and Ti-base alloys. The development of such in-situ techniques for directly monitoring of processes occurring at crack tips may be applicable to many other phenomena, which lack detailed local data for deriving mechanistic understanding. A greater understanding of the phenomena underlying delayed hydride cracking achieved by this data will have an impact on the design and operation of components for nuclear power plants, and it will also lead to enhancement of the research infrastructure by creating new partnerships. Finally the involvement of doctoral students and young post doc researchers will result in the education of young scientists in this important field.
Participants: University of Michigan, Alabama A&M University, University of California, Berkeley, University of California, Santa Barbara, Pennsylvania State University, University of Wisconsin
The goal of this consortium is to address key materials issues in the most promising advanced reactor concepts that are yet be unresolved, or that are beyond the existing experience (dose/burnup) base, in order to 1) provide for a sound fundamental and engineering basis for operation in the intended application, 2) bring together key university, national laboratory and industry capability and support in order to provide the most comprehensive approach possible, and 3) create a long term, evolutionary program that seeks to address these and future nuclear materials issues in a progressive manner. This consortium to serve as a nucleation site, about which materials research activities will be catalyzed and grown among the leading individuals and institutions from academia, the national laboratories and industry. It represents an unprecedented opportunity to combine expertise and facilities in an effort to attack the challenge of materials behavior under irradiation on a scale that is not feasible for a single individual or institution.
The objectives of the initial three-year phase of the consortium are to:
The objectives will be accomplished in a research program consisting of three major thrusts: 1) high dose radiation stability of advanced fast reactor fuel cladding alloys, 2) irradiation creep at high temperature and 3) innovative cladding concepts embodying functionally-graded barrier materials. While the initial 3-yr program will emphasize ion irradiation and irradiated microstructures, we expect that, if successful, the second 3-yr program will increasingly emphasize reactor irradiations and will include mechanical property determination through national user facilities.
Industry partners (EPRI and GE) will utilize the core program as leverage to guide or support additional activities that are of special interest to them, and that fall within the scope of the core program. National laboratory partners (ANL, INL, LANL, ORNL and PNNL) will provide additional capability and direction to various aspects of the core program that are of interest to them. Our technical society partner, ASME, will introduce the data generated by the consortium into the ASME Codes & Standards (C&S) process. Beyond scientific achievements, this consortium will provide substantial additional outcomes that are expected to provide long term benefits to the advanced rector program, including the education of ~8 graduate students and several post-docs, inclusion of minority students into the radiation effects and reactor materials fields through the participation of Alabama A&M University, creation of new working relationships between universities, laboratories and industry in an unprecedented manner and to an unprecedented degree, and establishment of a pathway to begin to incorporate data generated by the research thrusts into the ASME codes and standards that will be crucial for success of the advanced reactor programs.