Energy Materials 2017: Materials for Nuclear Energy: Materials for Nuclear Applications II
Sponsored by: Chinese Society for Metals
Program Organizers: Raul Rebak, GE Global Research; Zhengdong Liu, China Iron & Steel Research Institute Group; Peter Hosemann, University of California, Berkeley; Jian Li, CanmetMATERIALS
Wednesday 2:00 PM
March 1, 2017
Location: Marriott Marquis Hotel
Session Chair: Jian Li, CanmetMATERIALS
2:00 PM Invited
Fuel Cladding Materials for Supercritical Water Cooled Reactor: Wenyue Zheng1; 1Canmet Materials
Supercritical Water-cooled Reactors (SCWR) is one of the Gen IV reactor systems being developed through an international treaty-level collaboration led by Generation IV International Forum (GIF). Following the release of the core concepts from the Japanese and the EU members, Canada published its own pressure-tube based conceptual design in 2015. With an outlet temperature of 625 C and a core pressure of 25 MPa, the Canadian SCWR concept requires cladding materials that can sustain extremely harsh in-core physical and chemical conditions. Based on initial calculations using stainless steels and nickel alloys, the maximum cladding surface temperature is predicted to be as high as 825 C; the irradiation dose can reach as much as 10 dpa and the supercritical coolant can be very oxidizing due to the production of oxygen and hydrogen peroxide by radiolysis of light water. Under these conditions, the cladding can be readily degraded by corrosion, stress-corrosion, creep, or any of the radiation-related processes such as void-swelling and embrittlement. In the course of the Canadian program (2007-2015) on R&D of in-core materials, unique experimental and computational facilities were set up specifically to probe into the behaviours of candidate alloys under these extreme conditions. Novel, and frequently surprising, results have been achieved. Key highlights of this collaborative R&D effort are presented in this paper and the challenges for the future are also discussed.
2:40 PM Cancelled
Development of the 12Cr2Mo1R Steel Plate for Metal Internal Equipment for Demonstration Project of High Temperature Gas—cooled Reactor: Hanqian Zhang1; Huibin Liu1; 1Baoshan Iron & Steel Company
The demonstration project of high temperature gas—cooled reactor (HTR) is building in Shidaowan,Rongcheng,Shandong province. This project was designed by Institute of Nuclear and New Energy Technology,Tsinghua University. The metal internal equipment is designed to manufacture by using the 12Cr2Mo1R steel plates with thickness of 10~135mm which are developed and manufactured by Baosteel. In this paper, the development of 12Cr2Mo1R steel plates are introduced and their basic mechanical property,high temperature tensile property and fatigue property listed and analyzed.
EBSD and TEM Assessment of Deformation Localization in 718 Alloy: Aida Amroussia1; Keith Leonard2; Maxim Gussev2; Jacqueline Stevens3; 1Michigan State University; 2Oak Ridge National Laboratory; 3AREVA Inc.
The performance of Ni-base alloy 718 is dependent upon many factors that include the starting microstructure and its subsequent evolution under the combination of applied stresses, irradiation, and corrosive environments. The previous characterization of in-reactor stress corrosion cracking of two different heats of 718 alloy leaf spring (LS) material showed a difference in resistance to the stress corrosion cracking. Microstructural and microhardness changes in the materials were minimal following exposure. The present work examines the effect of the starting microstructure on strain localization and deformation behavior in the two 718-alloy heats with different delta-phase amounts. The materials are characterized using SEM, EBSD, TEM, and EDS. The -phase morphology and its volume fraction is investigated. FIB liftout are prepared to identify and analyze the chemical composition of the precipitates in detail. EBSD in combination with slip trace analysis method is used to asses the delta-phase role on plastic strain localization.
Microstructure Evolution of a Reactor Pressure Vessel Steel during High-temperature Tempering: Chuanwei Li1; Jianfeng Gu1; Lizhan Han1; Qingdong Liu1; 1Shanghai Jiao Tong University
The electron microscopy techniques and thermodynamic calculation were employed to analyze the microstructure evolution during high-temperature tempering of a reactor pressure vessel steel. The results show that carbon enriched martensite/austenite (M/A) constituents decomposed into ferrite laths and accumulated carbides during initial stage of tempering. Simultaneously, the carbon atoms in the constituents diffused into the matrix continuously. With further prolonging of tempering, Mo2C carbides were found to be distributed uniformly in bainitic ferrite. In case of longer tempering, baintic ferrite would combine and broaden, and grain boundary carbides grew up sequentially and coarsened. The newly formed austenite was detected during tempering at 660 °C for 5 h, and at 650 °C for 100 h，which shown that the Ac1 is time-related. This phenomenon may be depend on the component fluctuation of M/A constituents and segregation of carbides.
3:40 PM Break
Thermal Conductivity Reduction of Tungsten Plasma Facing Material Due to Helium Plasma and Cu2+ Ion Irradiation: Shuang Cui1; Michael Simmonds1; Joseph Barton1; Yongqiang Wang2; Russ Doerner1; George Tynan1; Renkun Chen1; 1UCSD; 2LANL
Near-surface region of plasma facing material (PFM) plays an important role in thermal management of fusion reactors. We extracted thermal conductivity of tungsten (W) surface layer damaged by He plasma in PISCES at UCSD and Cu2+ ion beam at LANL with adopted 3-omega method (bulk) and Wiedemann-Franz law (thin film). Results showed that thermal conductivity in the He-plasma damaged layers (bulk and thin film) was reduced by ~80% compared to that of undamaged W. Moreover, Cu2+ ion irradiation is performed on W bulk to mimic neutron displacement damage. Results show that the thermal conductivity of irradiated surfaces (1um thick) drops from the un-irradiated value of 182±3.3 Wm-1K-1 to 37 ±2.8 Wm-1K-1 for 2dpa, 50±7.8 Wm-1K-1 for 0.6 dpa, 52.5±8 Wm-1K-1 for 0.2 dpa irradiated at 500 oC. Our result suggests that suppressed thermal conductivity in PFM needs to be taken into account in the thermal design of future plasma-facing components.
Effects of Fe Concentration on Ion-irradiation Induced Defect Evolution and Hardening in Ni-Fe Binary Alloys: Ke Jin1; Wei Guo2; Mohammad Ullah1; Yanwen Zhang1; William Weber3; Jonathan Poplawsky2; Hongbin Bei1; 1Materials Science & Technology Division, Oak Ridge National Laboratory; 2Center for Nanophase Materials Sciences, Oak Ridge National Laboratory; 3University of Tennessee
Effects of composition on the irradiation response of structural alloys have gained much interest recently due to the development of single-phase concentrated solid-solution alloys, in which radiation tolerance can be enhanced with controlled compositional complexity. To explore the effects of Fe concentration on ion-irradiation induced damage evolution in face-centered-cubic Ni-Fe binary alloys, single-crystalline Ni-xFe (x=0-70 at%) have been grown and irradiated with 1.5 MeV Ni-ions at room temperature, covering fluences from 3x10^13 to 3x10^16 cm-2. The defect evolution has been characterized using an ion channeling technique, which shows reduced damage accumulation with increasing Fe concentration. Irradiation induced hardening has been measured by nanoindentation tests. Ni and the Ni-Fe alloys have largely different initial hardness, but all follow a similar trend for the increase of hardness as a function of irradiation fluence. This work was supported by Energy Dissipation to Defect Evolution (EDDE), an Energy Research Frontier Center supported by the U.S. Department of Energy, Basic Energy Sciences. APT was conducted at ORNL's Center for Nanophase Materials Sciences (CNMS), which is a U.S. DOE Office of Science User Facility.
Impact of Neutron Irradiation on Helium Desorption Behavior in Iron: Xunxiang Hu1; Kevin Field1; David Woodley2; Yutai Katoh1; 1ORNL; 2University of Michigan
The synergistic effect of neutron irradiation and helium production is an important concern for the use of steels in fusion reactors. We investigated the impact of neutron irradiation on helium desorption behavior of pure iron. Two pure iron samples were neutron irradiated in HFIR at ORNL and BOR-60 in Russia, respectively. Following neutron irradiation, 10keV He ion implantation was performed on both samples to a fluence of 7×1014 He/cm2. Thermal desorption spectrometry (TDS) was conducted to obtain the helium desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during the irradiation, which appeared to be voids through TEM examination. The underlying mechanisms controlling the helium desorption behaviors could be revealed by the detailed microstructure characterization before and after TDS runs.
Size Effects in Ion-irradiated 800H Steel at High Temperatures Utilizing Nanoindentation and Microcompression Testing: Anya Prasitthipayong1; Shraddha Vachhani2; Scott Tumey3; Andrew Minor4; Peter Hosemann5; 1Department of Materials Science and Engineering, University of California, Berkeley; 2Hysitron, Inc.; 3Center of Accelerator Mass Spectrometry, Lawrence Livermore National Laboratory; 4Department of Materials Science and Engineering, University of California, Berkeley; National Center for Electron Microscopy, The Molecular Foundry, Lawrence Berkeley National Laboratory; 5Department of Nuclear Engineering, University of California, Berkeley
Small damage layers originating from ion-irradiation necessitate the development of small-scale mechanical testing to evaluate the properties of irradiated materials. However, understanding the influence of mechanical size effects remains a major obstacle for relating nanoscale mechanical measurements from small irradiated volumes to macroscopic mechanical properties. 800H was irradiated with 70 MeV Fe9+ at 4500C to 20.68 dpa and using small-scale testing we investigate the influence of temperature on both the indentation and the sample size effects. At high temperatures, indentation size effects are expected to be less significant due to the larger plastic zone size, while sample size effects are expected to be more pronounced owing to a greater contribution of the source length. This implies that testing methods determine how significant size effects are as temperatures increase. Our study will directly compare these effects using nanoindentation and microcompression at high temperatures, combined with in-situ observations of the deformation mechanisms.
Understanding Transuranic Binding Mechanisms and Speciation on Stainless Steel: Tim Kerry1; Clint Sharrad1; Andreas Geist2; Dieter Schild2; 1University of Manchester; 2Institute for Nuclear Waste Disposal
At the end of the life of a nuclear power facility handling radioactive species there is the need to safely decommission. In most cases this will leave a vast quantity of steel that will be reused, recycled or disposed of. All scenarios may require the use of some decontamination process in order to reduce associated radioactivity. Through determining the mechanism of binding and speciation of contaminants the decontamination process can be developed as effectively as possible. The contamination of stainless steels by transuranic material in conditions replicating those found in a reprocessing plant will be presented. An understanding of contamination by radionuclides on both corroded and unaffected steel surfaces has been established. All samples in contact with contaminant containing solutions have shown uptake with dependence on acid molarity and surface morphology. Both solution and surface analysis has been used to show the uptake mechanism and kinetics of the process.