Microstructural Processes in Irradiated Materials: Fusion Materials
Sponsored by: TMS Structural Materials Division, TMS/ASM: Nuclear Materials Committee
Program Organizers: Thak Sang Byun, Pacific Northwest National Laboratory; Dane Morgan, University of Wisconsin-Madison; Yasuyoshi Nagai, Tohoku University; Zhijie Jiao, University of Michigan-Ann Arbor; Christine Guéneau, CEA-Saclay
Wednesday 2:00 PM
March 6, 2013
Location: Henry B. Gonzalez Convention Center
Session Chair: Roger Stoller, Oak Ridge National Laboratory; Steven Zinkle, Oak Ridge National Laboratory
2:00 PM Invited
Radiation Effects on a High Strength, High Conductivity Copper Alloy: Steven Zinkle1; 1Oak Ridge National Laboratory
High-strength, high-conductivity copper alloys are being considered for first wall heat sink and divertor structural applications in fusion energy systems. Two heats of CuNiBe alloys with room temperature yield strengths of 630-725 MPa and electrical conductivities of 65-72% IACS were irradiated with fission neutrons at 100 - 240˚C. The irradiation produced a slight increase in strength and a significant decrease in ductility. The measured tensile elongation increased with increasing irradiation temperature, with a uniform elongation of ~3.3% observed at 240˚C. The data indicate that CuNiBe alloys may be suitable for certain fusion energy structural applications at temperatures <250˚C. However, ductile intergranular fracture and low uniform elongation occurs for test temperatures above 350˚C. The change in deformation behavior from intragranular to intergranular was analyzed using an Ashby-type deformation map. The possible implications for other precipitation hardened alloys such as superalloys are discussed.
Effect of Grain Boundary Characters on Sink Efficiency: Weizhong Han1; Michael Demkowicz2; Engang Fu1; Yongqiang Wang1; Amit Misra1; 1Los Alamos National Lab; 2MIT
The dependence of the width of void-denuded zones (VDZs) on grain boundary (GB) characters was investigated in Cu irradiated with Helium ions at elevated temperature. Dislocation loops and voids formed near GBs during irradiation were characterized by transmission electron microscopy, and GB misorientations and plane normals were determined by electron backscatter diffraction. The VDZ widths at Σ3 <110> tilt GBs ranged from 0 to 24 nm and increased with the GB plane inclination angle. For non-Σ3 GBs, VDZ widths ranged from 40 to 70 nm and generally increased with misorientation angle. Nevertheless, there is considerable scatter about this general trend, indicating that the remaining crystallographic parameters also play a role in determining the sink efficiencies of these GBs. Voids were also observed within GB planes and their density and radius also appeared to depend on GB character. We conclude that GB sink efficiencies depend on the complete GB character, including both misorientation and GB plane orientation. This research is supported by US DOE, Office of Science, Office of Basic Energy Sciences.
Correlation between Irradiation Hardening and Microstructural Evolution in High Purity Reference V-4Cr-4Ti Alloy for Fusion Reactor: Takuya Nagasaka1; Takeo Muroga1; Hideo Watanabe2; Takeshi Miyazawa3; Masanori Yamazaki4; 1National Institute for Fusion Science; 2Research Institute for Applied Mechanics, Kyushu University; 3The Graduate University for Advanced Studies; 4International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University
Vanadium alloys are recognized as a candidate structural material for fusion reactor blanket. The present study seeks the correlation between the neutron irradiation hardening and the microstructural evolution for vanadium alloys to understand the mechanisms of irradiation hardening and effects of interstitial impurities on it. Base metal and weld metal of NIFS-HEAT-2, a reference high-purity V-4Cr-4Ti alloy, were irradiated in JMTR, JOYO, HFIR and BR-II reactor up to 1.3 x 10^26 n m^-2 (E > 0.1 MeV, 5.3 dpa). Irradiation temperature varied from 333 to 723 K. Observed irradiation hardening of the base metal ranged from 16 to 92 VHN compared with the hardness before irradiation, 144 VHN. Some irradiations exhibited additional hardening likely due to interstitial impurity contamination. Welding also enhanced the irradiation hardening. The irradiation hardening was recovered by post-irradiation annealing at around 873 K. Hardening factors in microstructures were identified as dislocation loops, tangle and irradiation-induced precipitates.
A Replica Technique for Extracting Precipitates from Neutron-Irradiated or Thermal-Aged Vanadium Alloys for TEM Analysis: Ken-ichi Fukumoto1; Masahiro Iwasaki2; 1RINE/Univ. of Fukui; 2Univ. of Fukui
A carbon replica technique has been developed to extract precipitates from vanadium alloys. Using this technique, precipitation phases can be extracted from neutron-irradiated or thermal-aged V-4Cr-4Ti alloys. Precipitate identification using EDS X-ray analysis and electron diffraction was facilitated. Only NaCl type of Ti(OCN) precipitate was formed in the thermal-aged V-4Cr-4Ti alloys in the temperature regime from 800 to 1000°C and cation sub-lattice was only occupied by Ti atoms only. However the thin plate of precipitates with NaCl type of crystallographic structure could be seen in the V-4Cr-4Ti alloys irradiated in the temperature from 400 to 600°C at the JOYO fast reactor in Japan. The precipitate contained chromium and vanadium atoms on the cation sub-lattice as well as titanium atoms. It is considered that the phase of MX type (M = Ti, V, Cr and X=O, N, C) is metastable phase under neutron irradiation.
Microstructures of Heavily Neutron-Irradiated SiC/SiC Composites: Yutai Katoh1; Keith Leonard1; Peng Dou1; Lance Snead1; 1Oak Ridge National Laboratory
Continuous SiC fiber-reinforced SiC matrix composites of certain grades have been proven to be exceptionally stable in radiation environment in spite of the substantial microstructural changes occurring in the matrix, fiber, and interphase. Recently, degradation in mechanical properties for the Hi-Nicalon Type-S fiber, chemically vapor-infiltrated composite with pyrocarbon/SiC multilayer interphase was discovered after neutron irradiation in High Flux Isotope Reactor to ~70 dpa at 300, 500, and 800°C. However, exact reasons for the observed property changes still remain to be clarified. In this work, microstructures of the irradiated composite constituents were examined and discussed in relation with the potential mechanisms for the composite mechanical properties modification. This work was supported by United States Department of Energy under contract DE-C05-00OR22725 with UT-Battelle, LLC.
3:50 PM Break
4:00 PM Invited
Multiple Simultaneous Ion Beam (MSIB) Examination of Inertial Fusion Energy Candidate Materials: Michael Fluss1; Luke Hsiung1; William Choi1; Peter Hosemann2; Estelle Meslin3; Jaime Marian1; David Hoelzer4; 1LLNL; 2UC Berkeley; 3CEA-Saclay; 4ORNL
Materials for IFE can be designed for a damage-dose as low as 20 dpa rather than 100-200 dpa as in MFE. For qualification of first wall and blanket materials we must include the possible synergistic consequences of the He and H, produced by nuclear reactions, along with the displacement damage. There is no source of fusion neutrons of adequate intensity currently available. Instead, we are utilizing multiple ion beams (3) to simultaneously produce dpa (with self-ions), and to implant H and He at doses and temperatures that emulate the fusion energy environment. We will discuss some of our early results on Fe(Cr) and Fe(Cr)-ODS materials irradiated with He+Fe and He+H+Fe focusing on the evolved micro-structures (measured with HRTEM) and the mechanical properties (measured with micro-mechanical indentation and pillar compression). We will conclude with a brief discussion of special issues related to IFE and how they may be explored using MSIB.
4:30 PM Student
The Microstructure Development of Dispersion-Strengthened Tungsten due to Neutron Irradiation: Makoto Fukuda1; Akira Hasegawa1; Shuhei Nogami1; Kiyohiro Yabuuchi1; 1Tohoku University
The objective of this work is to investigate the microstructure development of dispersion-strengthened tungsten (W) due to neutron irradiation. The examined dispersion-strengthened W are potassium(K) doped W and lanthanum(La) doped W. Pure W was also examined as a reference material. The neutron irradiation experiment was conducted using JOYO at Japan Atomic Energy Agency (JAEA). Irradiation conditions were 0.44 dpa at 531 °C, 0.47 dpa at 583 °C and 0.42 dpa at 756 °C. After the irradiation, void and dislocation loop were observed in these specimens. The vickers hardness (Hv) was increased about 100 to 300 due to neutron irradiation and varied by the irradiation temperature. The effect of neutron irradiation on the microstructure development and relationship between the microstructure development and hardness change will be discussed.
Theoretical and Experimental Study of Spatial Effects in 3He Implantation in W: Andrée De Backer1; Christophe Ortiz2; Christophe Domain3; Marie France Barthe4; Charlotte Becquart1; 1UMET, UMR 8207, EM2VM; 2Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT; 3EDF, EM2VM; 4CNRS, UPR3079 CEMHTI
Tungsten is candidate for the divertor in ITER. As such it is currently under investigations using both experimental and theoretical approaches. We study, with an Object Kinetic Monte Carlo approach (OKMC), the effect of fluence on the vacancy cluster distribution in the track region created at room temperature for 800 keV 3He irradiations in tungsten further annealed. Our results show that the width and position of the maximum size of the distribution strongly depend on the implantation fluence. For the high implantation fluence, the vacancy clusters are the largest in the first 100 nm close to the surface, whereas, for the low fluence, the distribution is more uniform in size between 100 and 300 nm from the surface. Our simulations predict that vacancy clusters are always smaller very close to the surface; this whatever the fluence. The trends observed are in good agreement with experimental Positron Annihilation Spectroscopy (PAS) results.
5:10 PM Student
The Change in Mechanical Properties of Tungsten after Self-Ion Irradiation: James Gibson1; David Armstrong1; Steve Roberts1; 1Oxford University
Tungsten is a key material for the plasma-facing components of future fusion reactors. The lifetime of these components is determined by the rate of embrittlement under neutron radiation. Here, ion irradiation has been used to model the cascade damage produced by fusion neutrons. Self-ion and simultaneous self-ion and helium ion irradiations have been performed at 800oC on commercially pure tungsten.The change in mechanical properties has been determined using FIB-machined micro-cantilevers. A nanoindenter is used to displace the free end of a 30µm long cantilever in order to extract mechanical data from the small volume of ion-implanted material.An increase in modulus is seen after implantation, with values increasing by ∼10% compared to unimplanted material. An yield stress of ∼2.0 GPa is measured for unimplanted tungsten, rising to ∼3.1 GPa for both implantations, suggesting the effect of helium on the plasticity of tungsten is minimal.
Deuterium Retention in Ion Damaged Tungsten with and without the Presence of Helium: Yongqiang Wang1; Chunping Xu1; Joseph Barton2; Nate Mara1; Russ Doerner2; George Tynan2; 1Los Alamos National Laboratory; 2University of California
Candidate plasma facing material, Tungsten (W), was submitted to a fusion relevant irradiation environment where neutron displacement damage was emulated with heavy ion collisions and helium (He) build-up by (n,α) reactions was concurrently accomplished through He ion implantation. The damaged W samples were then exposed to deuterium plasma in the PISCES linear plasma device under typical divertor conditions to a fluence of 1E26 ions/m2 while maintaining the sample temperature below 373 K. Bulk retention of deuterium was measured by thermal desorption spectroscopy (TDS), and the D(3He,p)4He reaction was used to obtain concentration profiles of deuterium near the surface of the W. Deuterium retention and diffusion behavior in the irradiated W with and without the presence of helium are studied and discussed.