William K. Soppet

Uniaxial creep behavior of V-4Cr-4Ti alloy

We are undertaking a systematic study at Argonne National Laboratory to evaluate the uniaxial creep behavior of V-Cr-Ti alloys in a vacuum environment as a function of temperature in the range of 650-800 C and at applied stress levels of 75-380 MPa. Creep strain in the specimens is measured by a linear-variable-differential transducer, which is attached between the fixed and movable pull rods of the creep assembly. Strain is measured at sufficiently frequent intervals during testing to define the creep strain/time curve. A linear least-squares analysis function is used to ensure consistent extraction of minimum creep rate, onset of tertiary creep, and creep strain at the onset of tertiary creep. Creep test data, obtained at 650, 700, 725, and 800 C, showed power-law creep behavior. Extensive analysis of the tested specimens is conducted to establish hardness profiles, oxygen content, and microstructural characteristics. The data are also quantified by the Larson-Miller approach, and correlations are developed to relate time to rupture, onset of tertiary creep, times for 1 and 2% strain, exposure temperature, and applied stress.

Effects of oxygen and oxidation on tensile behavior of V–4Cr–4Ti alloy

Abstract Vanadium-base alloys are potential candidates for applications such as the first wall and other structural components of fusion reactors, but a good understanding of the oxidation behavior of the alloys intended for elevated-temperature use is essential. We conducted a systematic study to determine the effects of time and temperature of air exposure on the oxidation behavior and microstructure of V–4Cr–4Ti alloy. Uniaxial tensile tests were conducted at room temperature and at 500°C on preoxidized specimens of the alloy to examine the effects of oxidation time and oxygen migration on maximum engineering stress and uniform and total elongation. The effect of preexposure of the specimens to environments with varying oxygen partial pressures on the tensile properties of the alloy was investigated. Extensive microstructural analyses of the oxygen-exposed/tensile-tested specimens were conducted to evaluate the cracking propensity for the alloy. In addition, tensile-property data for the alloy were correlated with oxygen pressure in the exposure environment, test temperature, and exposure time.

Performance of V–Cr–Ti alloys in a hydrogen environment

A systematic study is underway at Argonne National Laboratory to evaluate the mechanical properties of several V-Cr-Ti alloys after exposure to environments containing hydrogen at various partial pressures. The goal is to correlate the chemistry of the exposure environment with hydrogen uptake by the samples and with the resulting influence on microstructures and tensile properties of the alloys. Other variables examined are specimen cooling rate and synergistic effects, if any, of oxygen and hydrogen on tensile behavior of the alloys. Experiments were conducted to evaluate the effect of pH{sub 2} in the range of 3 x 10{sup {minus}6} and 1 torr on tensile properties of two V-Cr-Ti alloys. Up to pH{sub 2} of 0.05 torr, negligible effect of H was observed on either maximum engineering stress or uniform and total elongation. However, uniform and total elongation decreased substantially when the alloys were exposed at 500 C to 1.0 torr of H{sub 2} pressure. Preliminary data from sequential exposures of the materials to low-pO{sub 2} and several low-pH{sub 2} environments did not reveal adverse effects on the maximum engineering stress or on uniform and total elongation when the alloy contained {approx} 2,000 wppm O and 16 wppm H. Furthermore, tests in H{sub 2}-exposed specimens, initially annealed at various temperatures, showed that grain-size variation by a factor of {approx} 2 had little or no effect on tensile properties. Also, specimen cooling rate had a small effect, if any, on the tensile properties of the alloy.

Effect of oxygen and oxidation on tensile properties of V-5Cr-5Ti alloy

Oxidation studies were conducted on V-5Cr-5Ti alloy specimens in an air environment to evaluate the oxygen uptake behavior of the alloy as a function of temperature and exposure time. The oxidation rates calculated from parabolic kinetic measurements of thermogravimetric testing and confirmed by microscopic analyses of cross sections of exposed specimens were 5, 17, and 27 {mu}m per year after exposure at 300, 400, and 500{degrees}C, respectively. Uniaxial-tensile tests were conducted at room temperature and at 500C on preoxidized specimens of the alloy to examine the effects of oxidation and oxygen migration on tensile strength and ductility. Microstructural characteristics of several of the tested specimens were determined by electron optics techniques. Correlations were developed between tensile strength and ductility of the oxidized alloy and microstructural characteristics such as oxide thickness, depth of hardened layer, depth of intergranular fracture zone, and transverse crack length.

Effect of Pressurized Water Reactor Environment on Material Parameters of 316 Stainless Steel: A Cyclic Plasticity Based Evolutionary Material Modeling Approach

At present, the fatigue life of nuclear reactor components is estimated based on empirical approaches, such as stress/strain versus life (S∼N) curves and Coffin-Manson type empirical relations. In most cases, the S∼N curves are generated from uniaxial fatigue test data, which may not truly represent the multi-axial stress state at the component level. Also, the S∼N curves are based on the final life of the specimen, which may not accurately represent the mechanistic time-dependent evolution of material behavior. These discrepancies lead to large uncertainties in fatigue life estimations. We propose a modeling approach based on evolutionary cyclic plasticity that can be used for developing finite element models of nuclear reactor components subjected to multi-axial stress states. These models can be used for more accurately predicting the stress-strain evolution over time in reactor components and, in turn, fatigue life. The model parameters were estimated for 316 stainless steel material, which are widely used in U.S. nuclear reactors. The model parameters were estimated for different test conditions to understand their evolution over time and their sensitivity to particular test conditions, such as the pressurized water reactor coolant condition.Copyright

Status report on assessment of environmentally assisted fatigue for LWR extended service conditions

.......................................................................................................................................i List of Figures ..............................................................................................................................v List of Tables ...............................................................................................................................x Abbreviations ...............................................................................................................................xi Acknowledgements ......................................................................................................................xii

Thermal-Mechanical Stress Analysis of PWR Pressure Vessel and Nozzles under Grid Load-Following Mode: Interim Report on the Effect of Cyclic Hardening Material Properties and Pre-existing Cracks on Stress Analysis Results

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable under the work package for environmentally assisted fatigue as part of DOE’s Light Water Reactor Sustainability Program. In a previous report (September 2015), we presented tensile and fatigue test data and related hardening material properties for 508 low-alloys steel base metal and other reactor metals. In this report, we present thermal-mechanical stress analysis of the reactor pressure vessel and its hot-leg and cold-leg nozzles based on estimated material properties. We also present results from thermal and thermal-mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting cracks in the reactor nozzles (axial or circumferential crack). In addition, results from validation stress analysis based on tensile and fatigue experiments are reported.

Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

i Table of

Pressurized Water Reactor Environment Effect on 316 Stainless Steel Stress Hardening/Softening: An Experimental Study

In USA there are approximately 100 operating light water reactors (LWR) consisting fleet of both pressurized water reactors (PWR) and boiling water reactors (BWR). Most of these reactors were built before 1970 and the design lives of most of these reactors are 40 years. It is expected that by 2030, even those reactors that have received 20 year life extension license from the US nuclear regulatory commission (NRC) will begin to reach the end of their licensed periods of operation. For economical reason it is be beneficial to extend the license beyond 60 to perhaps 80 years that would enable existing plants to continue providing safe, clean and economic electricity without significant green house gas emissions. However, environmental fatigue is one of the major aging related issues for these reactors, and may create hurdles in long term sustainability of these reactors. To address some of the environmental fatigue related issues, Argonne National Laboratory (ANL) with the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program trying to develop mechanistic approach for more accurate life estimation of LWR components. In this context ANL conducted many fatigue experiments under different test and environment conditions on 316 stainless steel (316SS) material that is or similar grade steels are widely used in US reactors. Contrary to the conventional S∼N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to understand material ageing more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help to develop computer based advanced modeling tools to better extrapolate stress-strain evolution of reactor component under multi-axial stress states and hence to help predicting their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In another paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.Copyright

Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components

This report provides an update of an earlier assessment of environmentally assisted fatigue for components in light water reactors. This report is a deliverable in September 2016 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2016 report, we presented a detailed thermal-mechanical stress analysis model for simulating the stress-strain state of a reactor pressure vessel and its nozzles under grid-load-following conditions. In this report, we provide stress-controlled fatigue test data for 508 LAS base metal alloy under different loading amplitudes (constant, variable, and random grid-load-following) and environmental conditions (in air or pressurized water reactor coolant water at 300°C). Also presented is a cyclic plasticity-based analytical model that can simultaneously capture the amplitude and time dependency of the component behavior under fatigue loading. Results related to both amplitude-dependent and amplitude-independent parameters are presented. The validation results for the analytical/mechanistic model are discussed. This report provides guidance for estimating time-dependent, amplitude-independent parameters related to material behavior under different service conditions. The developed mechanistic models and the reported material parameters can be used to conduct more accurate fatigue and ratcheting evaluation of reactor components.