Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Structural Materials I
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Ramprashad Prabhakaran, Pacific Northwest National Laboratory; Dennis Keiser, Idaho National Laboratory; Raul Rebak, GE Global Research; Clarissa Yablinsky, Los Alamos National Laboratory
Tuesday 2:00 PM
February 28, 2017
Location: Marriott Marquis Hotel
Session Chair: Brian Cockeram, Bechtel-Bettis; Stuart Maloy, Los Alamos National Laboratory
The Increase in Fatigue Crack Growth Rates Observed for Zircaloy-4 in a PWR Environment: Brian Cockeram1; B.F. Kammenzind1; 1Bechtel-Bettis
The cyclic stresses produced by reactor operation result in the extension of cracks by processes of fatigue. Although the fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR environment. In this work FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using an Electric Potential Drop method. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. The non-hydrided material exhibited much lower rates of FCGR, suggesting the presence of hydrides is required for the increase in FCGR rates in water. An increase in FCGR rates are observed at the slower cycle rates. Comparisons between the FCGR rates in water and air are used to quantify the enhancement from the PWR environment.
Wear Results for Zirconium Alloys and Their Oxides: William Howland1; Paolo Zafred1; Gene Lucadamo1; Natalia Tymiak-Carlson1; Richard Smith1; 1Bechtel Marine Propulsion Company
The in-pile wear behavior of nuclear reactor core materials, particularly zirconium- based alloys, is influenced by a number of processes which take place in parallel such as corrosion, irradiation hardening, and hydrogen uptake. To better understand these interactions, wear testing in an autoclave environment has been conducted on various zirconium alloys processed in different ways, some intended to simulate the effects of in-pile exposure. Results are presented on the determination of wear coefficients (K) measured for a range of stroke lengths and material conditions. Over 300 samples were characterized by accurate weight gain measurements, laser microscope topography, TEM and focused ion beam (FIB) analysis. The results indicated that thin oxides films grown on some samples exhibit superior wear resistance. A unique high temperature, high pressure autoclave based wear system was designed for this purpose and it was successfully operated for over 500 million cycles.
Characterization and Simulation of Wear-tested Zirconium Alloy Surfaces: Gene Lucadamo1; Natalia Tymiak-Carlson1; William Howland1; Richard Smith1; Clinique Brundidge1; 1Bettis Laboratory, Bechtel Marine Propulsion Corporation
Refinements to the fundamental understanding of wear in zirconium alloys requires a detailed treatment of both the topology and mechanical properties of the interacting surfaces, as well as contact models capable of simulating localized deformation at appropriate length scales. The in-pile behavior of nuclear reactor core materials is of particular interest since wear may occur in parallel with corrosion, irradiation hardening, and hydrogen uptake. Wear-testing was performed under autoclave conditions on zirconium-based alloys processed to simulate the effects of in-pile exposure. The topography of the test surfaces was characterized using microscopy methods and the extraction of relevant roughness and anisotropy parameters via fractal analysis is presented. Computer-simulated surface profiles for mechanical modeling and preliminary results of FEM contact analyses are described. Finally, the effects of surface roughness on the evolution of the actual contact area and contact pressure are evaluated and the potential implications for wear are discussed.
Determination of Material Properties of Ion-irradiated and Corroded Zircaloy-4 by Using Nanomechanical Raman Spectroscopy: Debapriya Mohanty1; Yang Zhang1; Vikas Tomar1; 1Purdue University
Zircaloy-4 is a zirconium based alloy which is used extensively for fuel cladding and core structural material in nuclear reactor, and it is subjected to irradiation and corrosion during the operation. Importance and the effect of grain orientation and grain size on the mechanical properties of Zircaloy-4 at microscale is studied in this work using a combination of nano-indentation and nanomechanical Raman spectroscopy (NMRS). The effect of ion-irradiation and corrosion on the material properties is investigated. The stress distribution as a function of microstructure is analyzed using NMRS. The stress distribution obtained from NMRS is used to validate finite element method (FEM) simulations. The microstructural based material properties for FEM simulations are obtained from the nano indentation experiments. Once validated, FEM is used to predict and approximate the stress distribution at various loading conditions. The effect of corrosion and irradiation on material failure is predicted.
Evolution of Stress and Fracture During Oxidation of Zirconium Alloys: Natalia Tymiak Carlson1; Jason Gruber1; John Seidensticker1; Ram Bajaj1; Douglas Rishel1; William Howland1; Richard Smith1; 1Bettis Atomic Power Laboratory
Stress evolution during aqueous corrosion of zirconium alloys is recognized as a significant factor influencing loss of the protective capability of zirconium oxide. The evolution of oxide stress is linked to the evolving curvature of the metal/oxide interface, oxide stoichiometry, phase content, and grain structure. The presented computational analysis incorporates the evolution of the interface topography with the inclusion of the experimentally observed “transition zone” between zirconium alloy and the stoichiometric ZrO2. The effects of oxidation expansion strain and grain structure are also examined. The evolution of stress and strain energy indicate that equiaxed grains are favored at film thicknesses <100nm. Columnar grains are preferable for thicker films, consistent with the experimental studies. Oxide creep is linked to stress localization near the metal/oxide interface promoting oxide cracks parallel to the interface. The Extended Finite Element Method calculations evaluate the dependence of fracture initiation on properties of the evolving oxide layer.
3:40 PM Break
Damage Rate Dependence of Oxide Evolution on Zircaloy-4 under Simultaneous Irradiation-corrosion Experiment: Peng Wang1; Gary Was1; 1University of Michigan
In-situ high temperature irradiation-corrosion experiments conducted on Zircaloy-4 at 320˚C in 3 wt. ppm hydrogenated water have demonstrated that the corrosion rate was influenced by the proton irradiation. At high proton damage rate, i.e. ~10E-7 dpa/s, corrosion kinetics are similar to the breakaway corrosion rate after 3-4 cycles in the reactor which is several times faster than conventional autoclave exposure. While autoclave corrosion exposure results in a crystallographically textured oxide with elongated columnar grains, the in-situ experiment generated an equiaxed nanocrystallite oxide structure with a limited preferred growth direction. In-reactor grown oxides experience a lower damage rate, ~10E-8 dpa/s, and display an oxide structure that sits somewhere between these two extremes. To understand how irradiation can transform the conventional columnar-grained oxide to an equiaxed grained oxide, a set of proton irradiations with various damage rates were conducted to study the oxide microstructure evolution as a function of damage rate.
Modeling Activation and Radionuclide Decay in Proton Irradiated Zirconium Alloys: Jesse Carter1; Diane Moran1; Richard Smith1; 1Bettis Laboratory, BMPC
Proton bombardment is an important tool for studying radiation effects in materials due to increased damage rates and greatly reduced activation compared to in-pile neutron irradiation. While activation by protons is generally low, and activation products short-lived, radionuclides certainly are produced, and the residual activity increases with dose and proton energy. Due to the nature of absorption cross sections, activation can be quite sensitive to the alloy composition of the samples, and detailed modeling is required to track the activation and decay of the various species. To this end a computer simulation was developed to follow proton induced transmutation and the subsequent decay of zirconium alloys subjected to proton bombardment in the range up to several MeV. The model is benchmarked to gamma spectroscopy results for Zircaloy-4 samples bombarded at 2 MeV, and predictions are made for select Zr-Nb alloys containing typical levels of minor alloying elements and impurities.
Study on Texture Evolution of As-hydrided Zircaloy-4 Cladding under Low Temperature Biaxial Creep Test: Kuan-Che Lan1; Xiang Liu1; Huan Yan1; Hoon Lee1; Hsiao-Ming Tung2; Chih-Pin Chuang3; Kun Mo3; Yinbin Miao3; James Stubbins1; 1University of Illinois at Urbana-Champaign; 2Institute of Nuclear Energy Research; 3Argonne National Laboratory
To study the relationship between texture of Zircaloy-4 cladding and of zirconium hydride under nuclear spent fuel (UNF) interim dry storage facility, a synchrotron wide–angle x-ray scattering (WAXS) technique was used to evaluate texture evolution before and after biaxial creep test for 2000 hours. As-hydrided Zircaloy-4 tubular specimens with hydrogen content close to levels of normal burnup (300 wt ppm) and high burnup conditions (750 wt ppm) were adopted. Creep temperatures and effective stresses at mid-wall of cladding ranged from 573 to 673 K and from 40 to 65 MPa, respectively. Low temperature creep rupture of high burnup UNF (>45 GWD/MTU) cladding is the most possible failure mechanism during long-term dry storage. Since the crystallographic orientations of Zircaloy-4 cladding significantly correlated with the properties, a better understanding of texture evolution of this material can support the design basis of UNF interim and long-term dry storage facilities.
The Recovery of Irradiation Damage for Zircaloy-2 and Zircaloy-4 Following Irradiation at Higher Temperatures of 377-410C: Brian Cockeram1; T.S. Byun2; K.J. Leonard3; J.L. Hollenbeck1; B.F. Kammenzind1; 1Bechtel-Bettis; 2PNNL; 3Oak Ridge National Laboratory
Irradiation temperature is shown to have a significant effect on irradiation hardening and recovery following post-irradiated annealing. Zircaloy-2 and Zircaloy-4 were irradiated at 377°C to 421°C in the Advanced Test Reactor to fluences between 0.31-3.1x1025 n/m2 (E>1 MeV). Irradiation at these temperatures results in a 25%-60% lower irradiation hardening than reported in literature for irradiations at 260-326°C. A strong effect of neutron flux on irradiation hardening and recovery is observed that is not present at lower irradiation temperatures. Post-irradiation annealing at 343-427°C produces an increase in irradiation hardening in the first 1-10 hours of annealing. This Radiation Anneal Hardening (RAH) was followed by slower recovery of irradiation damage than observed for irradiations at 260-326°C. Faster recovery with no RAH was observed for post-irradiation annealing at temperatures of 454-510°C. Examinations of microstructure were used to understand the differences in the irradiation hardening and recovery of irradiation damage.