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Localized Deformation as a Primary Cause of Irradiation Assisted Stress Corrosion Cracking

Description: The objective of this project is to determine whether deformation mode is a primary factor in the mechanism of irradiation assisted intergranular stress corrosion cracking of austenitic alloys in light watert reactor core components. Deformation mode will be controlled by both the stacking fault energy of the alloy and the degree of irradiation. In order to establish that localized deformation is a major factor in IASCC, the stacking fault energies of the alloys selected for study must be measured. Second, it is completely unknown how dose and SFE trade-off in terms of promoting localized deformation. Finally, it must be established that it is the localized deformation, and not some other factor that drives IASCC.
Date: March 31, 2009
Creator: Was, Gary S.
Partner: UNT Libraries Government Documents Department

Corrosion and Creep of Candidate Alloys in High Temperature Helium and Steam Environments for the NGNP

Description: This project aims to understand the processes by which candidate materials degrade in He and supercritical water/steam environments characteristic of the current NGNP design. We will focus on understanding the roles of temperature, and carbon and oxygen potential in the 750-850 degree C range on both uniform oxidation and selective internal oxidation along grain boundaries in alloys 617 and 800H in supercritical water in the temperature range 500-600 degree C; and examining the application of static and cyclic stresses in combination with impure He environments in the temperature rang 750-850 degree C; and examining the application of static and cyclic stresses in combination with impure He environments in the temperature range 750-850 degree C over a range of oxygen and carbon potentials in helium. Combined, these studies wil elucidate the potential high damage rate processes in environments and alloys relevant to the NGNP.
Date: June 21, 2013
Creator: Was, Gary & Jones, J. W.
Partner: UNT Libraries Government Documents Department

Accelerator-Based Irradiation Creep of Pyrolytic Carbon Used in TRISO Fuel Particles for the (VHTR) Very Hight Temperature Reactors

Description: Pyrolytic carbon (PyC) is one of the important structural materials in the TRISO fuel particles which will be used in the next generation of gas-cooled very-high-temperature reactors (VHTR). When the TRISO particles are under irradiation at high temperatures, creep of the PyC layers may cause radial cracking leading to catastrophic particle failure. Therefore, a fundamental understanding of the creep behavior of PyC during irradiation is required to predict the overall fuel performance.
Date: July 30, 2010
Creator: Wang, Lumin & Was, Gary
Partner: UNT Libraries Government Documents Department


Description: Creep and IG cracking of nickel-base alloys depend principally on two factors--the deformation behavior and the effect of the environment. We have shown that both contribute to the observed degradation in primary water. The understanding of cracking does not lie wholly within the environmental effects arena, nor can it be explained only by intrinsic mechanical behavior. Rather, both processes contribute to the observed behavior in primary water. In this project, we had three objectives: (1) to verify that grain boundaries control deformation in Ni-16Cr-9Fe at 360 C, (2) to identify the environmental effect on IGSCC, and (3) to combine CSLBs and GBCs to maximize IGSCC resistance in Ni-Cr-Fe in 360 C primary water. Experiments performed in hydrogen gas at 360 C confirm an increase in the primary creep rate in Ni-16Cr-9Fe at 360 C due to hydrogen. The creep strain transients caused by hydrogen are proposed to be due to the collapse of dislocation pile-ups, as confirmed by observations in HVEM. The observations only partially support the hydrogen-enhanced plasticity model, but also suggest a potential role of vacancies in the accelerate creep behavior in primary water. In high temperature oxidation experiments designed to examine the potential for selective internal oxidation in the IGSCC process, cracking is greatest in the more oxidizing environments compared to the low oxygen potential environments where nickel metal is stable. In Ni-Cr-Fe alloys, chromium oxides form preferentially along the grain boundaries, even at low oxygen potential, supporting a potential role in grain boundary embrittlement due to preferential oxidation. Experiments designed to determine the role of grain boundary deformation on intergranular cracking have established, for the first time, a cause-and-effect relationship between grain boundary deformation and IGSCC. That is, grain boundary deformation in Ni-16Cr-9Fe in 360 C primary water leads to IGSCC of the deformed boundaries. As ...
Date: February 13, 2004
Creator: Was, Gary S.
Partner: UNT Libraries Government Documents Department

LWRSP FY09 testing and analysis of reactor metal degradation

Description: Current regulations require RPV steels to maintain conservative margins of fracture toughness so that postulated flaws do not threaten the integrity of the RPV during either normal operation and maintenance cycles or under accident transients, like pressurized thermal shock. Neutron irradiation degrades fracture toughness, in some cases severely. Thermal aging, while not generally considered a significant issue for a 40-y operating life, must be an additional consideration for operation to 60 or 80 years. Regulations, codified in the ASME Boiler and Pressure Vessel Code, Regulatory Guide 1.99 Rev 2, etc., recognize that embrittlement has a potential for reducing toughness below acceptable levels. The last few decades have seen remarkable progress in developing a mechanistic understanding of irradiation embrittlement. This understanding has been exploited in formulating robust, physically-based and statistically-calibrated models of CVN-indexed transition-temperature shifts (TTS). These semi-empirical models account for key embrittlement variables and variable interactions, including the effects of copper (Cu), nickel (Ni), phosphorous (P), fluence ({phi}t), flux ({phi}), and irradiation temperature (T{sub i}). However, these models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues. Over the past three decades, developments in fracture mechanics have led to a number of consensus standards and codes for determining the fracture toughness parameters needed for development of databases that are useful for statistical analysis and establishment of uncertainties. The CVN toughness, however, is a qualitative measure, which must be correlated with the fracture toughness and crack-arrest toughness properties, K{sub Ic} and K{sub Ia}, necessary for structural integrity evaluations. Where practicable, direct measurements of the fracture toughness properties are desirable to reduce the uncertainties associated with correlations. Moreover, fracture-toughness data have been obtained in sufficient quantity to permit probabilistic application. The progress notwithstanding, however, there are still significant technical issues ...
Date: September 1, 2009
Creator: Busby, Jeremy T; Nanstad, Randy K; Odette, G. & Was, Gary
Partner: UNT Libraries Government Documents Department

Radiation-Induced Segregation and Phase Stability in Candidate Alloys for the Advanced Burner Reactor

Description: Major accomplishments of this project were the following: 1) Radiation induced depletion of Cr occurs in alloy D9, in agreement with that observed in austenitic alloys. 2) In F-M alloys, Cr enriches at PAG grain boundaries at low dose (<7 dpa) and at intermediate temperature (400°C) and the magnitude of the enrichment decreases with temperature. 3) Cr enrichment decreases with dose, remaining enriched in alloy T91 up to 10 dpa, but changing to depletion above 3 dpa in HT9 and HCM12A. 4) Cr has a higher diffusivity than Fe by a vacancy mechanism and the corresponding atomic flux of Cr is larger than Fe in the opposite direction to the vacancy flux. 5) Cr concentration at grain boundaries decreases as a result of vacancy transport during electron or proton irradiation, consistent with Inverse Kirkendall models. 6) Inclusion of other point defect sinks into the KLMC simulation of vacancy-mediated diffusion only influences the results in the low temperature, recombination dominated regime, but does not change the conclusion that Cr depletes as a result of vacancy transport to the sink. 7) Cr segregation behavior is independent of Frenkel pair versus cascade production, as simulated for electron versus proton irradiation conditions, for the temperatures investigated. 8) The amount of Cr depletion at a simulated planar boundary with vacancy-mediated diffusion reaches an apparent saturation value by about 1 dpa, with the precise saturation concentration dependent on the ratio of Cr to Fe diffusivity. 9) Cr diffuses faster than Fe by an interstitial transport mechanism, and the corresponding atomic flux of Cr is much larger than Fe in the same direction as the interstitial flux. 10) Observed experimental and computational results show that the radiation induced segregation behavior of Cr is consistent with an Inverse Kirkendall mechanism.
Date: May 29, 2011
Creator: Was, Gary S. & Wirth, Brian D.
Partner: UNT Libraries Government Documents Department

Improving Corrosion Behavior in SCWR, LFR and VHTR Reactor Materials by Formation of a Stable Oxide

Description: The objective of this study is to understand the influence of the alloy microstructure and composition on the formation of a stable, protective oxide in the environments relevant to the SCWR and LFR reactor concepts, as well as to the VHTR. It is proposed to use state-of-the art techniques to study the fine structure of these oxides to identify the structural differences between stable and unstable oxide layers. The techniques to be used are microbeam synchrotron radiation diffraction and fluorescence, and cross-sectional transmission electron microcopy on samples prepared using focused ion beam.
Date: December 21, 2009
Creator: Motta, Arthur T.; Comstock, Robert; Li, Ning; Allen, Todd & Was, Gary
Partner: UNT Libraries Government Documents Department

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term Irradiation at Elevated Temperature: Critical Experiments

Description: 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 microchemistry changes due to radiation-induced segregation, dislocation loop formation and growth, radiation induced precipitation, destabilization of the existing precipitate structure, as well as the possibility for 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 to 200 dpa and beyond). 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. This project builds upon joint work at the proposing institutions, under a NERI-C program that is scheduled to end in September, to understand the effects of radiation on these important materials. The objective of this project is to conduct critical experiments to understand the evolution of microstructural and microchemical features ...
Date: December 20, 2013
Creator: Was, Gary; Jiao, Zhijie; Allen, Todd & Yang, Yong
Partner: UNT Libraries Government Documents Department

Alloys for 1000 degree C service in the Next Generation Nuclear Plant NERI 05-0191

Description: The objective of the proposed research is to define strategies for the improvement of alloys for structural components, such as the intermediate heat exchanger and primary-to-secondary piping, for service at 1000 degree C in the He environment of the NGNP. Specifically, we will investigate the oxidation/carburization behavior and microstructure stability and how these processes affect creep. While generating this data, the project will also develop a fundamental understanding of how impurities in the He environment affect these degradation processes and how this understanding can be used to develop more useful life prediction methodologies.
Date: January 15, 2009
Creator: Was, Gary S.; Jones, J.W. & Pollock, T.
Partner: UNT Libraries Government Documents Department

Novel Concepts for Damage-Resistant Alloys in Next Generation Nuclear Power Systems

Description: The discovery of a damage-resistant alloy based on Hf solute additions to a low-carbon 316SS is the highlight of the Phase II research. This damage resistance is supported by characterization of radiation-induced microstructures and microchemistries along with measurements of environmental cracking. The addition of Hf to a low-carbon 316SS reduced the detrimental impact of radiation by changing the distribution of Hf. Pt additions reduced the impact of radiation on grain boundary segregation but did not alter its effect on microstructural damage development or cracking. Because cracking susceptibility is associated with several material characteristics, separate effect experiments exploring strength effects using non-irradiated stainless steels were conducted. These crack growth tests suggest that irradiation strength by itself can promote environmental cracking. The second concept for developing damage resistant alloys is the use of metastable precipitates to stabilize the microstructure during irradiation. Three alloys have been tailored for evaluation of precipitate stability influences on damage evolution. The first alloy is a Ni-base alloy (alloy 718) that has been characterized at low neutron irradiation doses but has not been characterized at high irradiation doses. The other two alloys are Fe-base alloys (PH 17-7 and PH 17-4) that have similar precipitate structures as alloy 718 but is more practical in nuclear structures because of the lower Ni content and hence lesser transmutation to He.
Date: December 27, 2002
Creator: Bruemmer, Stephen M.; Andersen, Peter L. & Was, Gary
Partner: UNT Libraries Government Documents Department

Mechanism of Irradiation Assisted Cracking of Core Components in Light Water Reactors

Description: The overall goal of the project is to determine the mechanism of irradiation assisted stress corrosion cracking (IASCC). IASCC has been linked to hardening, microstructural and microchemical changes during irradiation. Unfortunately, all of these changes occur simultaneously and at similar rates during irradiation, making attribution of IASCC to any one of these features nearly impossible to determine. The strategy set forth in this project is to develop means to separate microstructural from microchemical changes to evaluate each separately for their effect on IASCC. In the first part, post irradiation annealing (PIA) treatments are used to anneal the irradiated microstructure, leaving only radiation induced segregation (RIS) for evaluation for its contribution to IASCC. The second part of the strategy is to use low temperature irradiation to produce a radiation damage dislocation loop microstructure without radiation induced segregation in order to evaluate the effect of the dislocation microstructure alone. A radiation annealing model was developed based on the elimination of dislocation loops by vacancy absorption. Results showed that there were indeed, time-temperature annealing combinations that leave the radiation induced segregation profile largely unaltered while the dislocation microstructure is significantly reduced. Proton irradiation of 304 stainless steel irradiated with 3.2 MeV protons to 1.0 or 2.5 dpa resulted in grain boundary depletion of chromium and enrichment of nickel and a radiation damaged microstructure. Post irradiation annealing at temperatures of 500 � 600°C for times of up to 45 min. removed the dislocation microstructure to a greater degree with increasing temperatures, or times at temperature, while leaving the radiation induced segregation profile relatively unaltered. Constant extension rate tensile (CERT) experiments in 288°C water containing 2 ppm O2 and with a conductivity of 0.2 mS/cm and at a strain rate of 3 x 10-7 s-1 showed that the IASCC susceptibility, as measured by the crack ...
Date: April 28, 2003
Creator: Was, Gary S.; Atzmon, Michael & Wang, Lumin
Partner: UNT Libraries Government Documents Department

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

Description: The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.
Date: February 13, 2005
Creator: MacDonald, Philip; Buongiorno, Jacopo; Sterbentz, James; Davis, Cliff; Witt, Robert; Was, Gary et al.
Partner: UNT Libraries Government Documents Department

Feasibility Study of Supercritical Light Water Cooled Reactors for Electrical Power Production, 5th Quarterly Report, October - December 2002

Description: The overall objective of this project is to evaluate the feasibility of supercritical light water cooled reactors for electric power production. The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies for the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR that can also burn actinides. The project is organized into three tasks:
Date: January 1, 2003
Creator: MacDonald, Philip; Buongiorno, Jacopo; Davis, Cliff; Herring, J. Stephen; Weaver, Kevan; Latanision, Ron et al.
Partner: UNT Libraries Government Documents Department