Energy Materials 2017: Materials for Nuclear Energy: Materials for Nuclear Applications I
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 8:30 AM
March 1, 2017
Location: Marriott Marquis Hotel
Session Chair: Raul Rebak, GE Global Research
8:30 AM Keynote
Is There a Role for Advanced Materials in Light Water Reactors?: Kurt Terrani1; Steven Zinkle2; LL Snead3; 1Oak Ridge National Laboratory; 2University of Tennessee, Knoxville; 3Massachusetts Institute of Technology
Unlike many other industries subject to strict regulation and despite the challenging operational environments in nuclear reactors, there exists only a few instance where advanced high-performance materials have been effectively deployed in fission reactors in the past 50 years. In case of light water reactors (LWRs), in many cases the materials used today are the ones originally deployed in civilian power reactors of 1950s-1960s, themselves blueprinted after the naval reactor program in the 1940s. A few of such instances and opportunities for improvements are discussed. In particular, a detailed discussion of the advanced accident tolerant fuel systems for LWRs that may improve reactor safety while maintaining today’s operational standards is offered.
9:10 AM Keynote
Development of a Novel Structural Material (SIMP steel) for Nuclear Equipment with Balanced Resistances to High Temperature, Radiation and LBE Corrosion: Yiyin Shan1; Wei Yan1; Wei Wang1; Quanqiang Shi1; Ke Yang1; Zhiguang Wang1; 1Institute of Metal Research, Chinese Academy of Sciences
Accelerator Driven Subcritical (ADS) system has been recognized to be the most promising technology for safely treating the nuclear wastes by now. This system is composed of three parts, which are accelerator, spallation target and reactor. The biggest challenge exists in the structural material for the spallation target is to possess not only good heat-resistance and radiation resistance but also a resistance to liquid metal corrosion. A novel martensitic heat-resistant steel, SIMP steel, has been developed against this challenge. By negotiating the effects of the contents of those important elements such as C, Cr and Si in the 9-12%Cr martensitic heat-resistant steel, an optimized chemical composition was obtained for SIMP steel and a good performance balance was reached. The test results conducted on one-ton and five-tons scales SIMP steels showed that this novel steel is much potential as a candidate structural material for the spallation target in ADS system.
Enhancing the High-Cycle Fatigue Property of 316 Austenitic Stainless Steels through Introduction of Mechanical Twins by Cold-Drawing: Xingfei Xie1; 1Shanghai Jiao Tong University
The strain-controlled fatigue tests of cold-drawn 316 austenitic stainless steels used in nuclear reactors were conducted at room temperature. The interaction between mechanical twins activated by prior cold drawing and dislocation structures during fatigue of cold-drawn 316 steels was investigated by transmission electron microscope. The results reveal that high-cycle fatigue life of 30% cold-drawn steels is higher than that of 20% cold-drawn steels. The complex dislocation structures, like walls, channels and cells, were generated during fatigue. The mechanical twins can segment the austenitic grains and hinder the dislocations motion between two mechanical twins. The dislocations pile up and slip along the mechanical twin boundaries. The extrusion on austenitic grain boundaries resulting from dislocations motion can be reduced by mechanical twin boundaries, leading to effectively improve deformation homogeneity and delay fatigue micro-crack initiations. Thus, the high-cycle fatigue property of 316 steels is enhanced through the mechanical twins activated by cold drawing.
10:10 AM Break
10:25 AM Invited
Research and Development of Pressure Vessel Steels for Advanced Pressurized Water Reactors in China: Xikou He1; Zhengdong Liu1; Wenhui Zhang2; Deli Zhao3; Ying Luo4; Xiaobin Wang5; 1China Iron & Steel Research Institute Group; 2China First Heavy Industries ; 3China First Heavy Industries; 4Nuclear Power Institute of China ; 5Nuclear Power Institute of China
Application of reactor pressure vessel (RPV) steels with higher strength and toughness might be an effective way to insure the integrity and increase the efficiency of Advanced Pressurized Water Reactors (APWR). This paper introduces the research and development of commercial RPV steel (SA508Gr.3 steel) currently used for PWR pressure vessel construction and higher strength steel (SA508Gr.4N steel) having strong potential for future APWR service in china. Particular emphasis is placed on the comparison of the SA508Gr.3 low alloy steel and SA508Gr.4N low alloy steels, including chemical composition, microstructural features and correlated with mechanical properties and environmental influences on the properties of these steels.
Bonding Characteristics and Site Occupancies of Si Atoms in M6C Carbides from First Principles and Experimental Study: Li Jiang1; 1Shanghai Institute of Applied Physics, Chinese Academy of Sciences
Hastelloy N is a Ni-Mo-Cr based superalloy developed in Oak Ridge National Laboratory for Molten-Salt Reactor Experiment. There is no more than 1% silicon in the air-melted heats of this alloy for deoxidants. Noticeably, Si additions were proved to promote the precipitation of M6C carbides in both as-cast and wrought Ni-Mo-Cr based superalloy in recent studies. In view of the obvious effects of Si, it is very necessary to understand the physical meaning of such effects, which can be abstracted as: (1) which sites or atoms are replaced by Si atoms in M6C carbide, (2) how they interact with neighboring atoms and affect the chemical properties of M6C carbide. The purpose of the present study is to use first-principles calculations together with NEXAFS spectroscopy technique to elucidate the site preference of Si, and its effect on the electronic structure and bonding characteristics.
Ductile Phase Toughening of 90-97W-NiFe Heavy Alloys: Md Ershadul Alam1; G. R. Odette1; 1University of California, Santa Barbara
Tungsten (W) is the leading candidate for the plasma-facing components for future fusion reactors. However, the inherent brittleness and high ductile-to-brittle transition temperature limits use of W as a structural material. One promising approach to improve W is ductile phase toughening. Here, the fracture toughness of commercial liquid phase sintered alloys with 90, 92.5, 95 and 97 wt.% W and a balance of 7:3 Ni:Fe solid solution ductile phase has been investigated. Pre-cracked 4-point bend resistance curve fracture tests were conducted at room temperature. Maximum load initiation toughness is scattered with the W fraction up to 95 wt.%, but drops somewhat 97%W alloy. However, the toughness of the 97 wt.% alloy is still at least ≈ 10 times higher than the monolithic W. Fracture occurs by stable crack growth and the steep resistance curve slopes average ≈200 MPa√m/mm. Toughening is due to extensive crack bridging and process zone deformation.
Investigation of Oxidation/Carburisation Mechanisms of 9Cr Ferritic Steel Heat Exchanger Tubes: Sabrina Yan1; Scott Doak1; Aya Shin2; Jonathan Pearson2; Rebecca Higginson1; 1Loughborough University; 2EDF Energy Generation
This paper presents a detailed experimental study of the oxidation/carburisation mechanisms of 9Cr ferritic steel heat exchanger tubes in advanced gas cooled reactor (AGR) primary CO2 coolant. During service, these materials typically form a duplex oxide scale comprised of an outer magnetite layer and an inner spinel enriched with Cr, which is defined as protective oxidation in this study. After a prolonged exposure time, a transition to a rapid ‘breakaway’ oxidation is known to occur, however detailed mechanism of breakaway oxidation is not fully understood. The purpose of the research is, by studying samples that are exposed to a simulated AGR environment for both short and long term periods, to develop an in-depth understanding of the metallurgical factors that influence the transition from protective oxidation to breakaway oxidation. Microstructural evolution and carbon-content measurements as a function of time and temperature have been investigated, in terms of C-profiling and carbides distribution.
12:05 PM Invited
Comparison of Corrosion Properties of Alloy 800 and Alloy 690 by In-situ Scratching Repassivation Behavior in High-temperature Pressurized Water: En-Hou Han1; Jiazhen Wang1; Jianqiu Wang1; 1Institute of Metal Research, Chinese Academy of Sciences
The stainless steel, nickel-base alloys and iron-nickel-base alloy were used for steam generator’s material. However, there exist the debate which materials are best for the steam generator to against corrosion and stress corrosion cracking (SCC) for long term operation. The rapid scratching method was used to evaluate the corrosion and SCC. However, it is difficult to achieve rapid scratch and measure the electrochemical signals during repassivation in high temperature high pressure (HTHP) water because the rapid scratching method requires sophisticated instrumentation to record the high current density within milli- or micro-seconds and the system manufacture and control difficulties. The scratch electrode system in HTHP water has been built. The design temperature and design pressure of this system are 350oC and 20 MPa respectively, and the maximum scratch speed is 3.3 m/s. By using this scratch electrode system, the in-situ scratching repassivation kinetics of steam generator tube materials Alloy 800 and 690 are studied in high temperature pressurized water with different dissolved hydrogen (DH) and dissolved oxygen (DO) at 300oC. The research results demonstrate that under DO condition the cBV value increases with the rise of DO concentration; and under DH condition, the cBV value passes through a local maximum at DH = 1.0 ppm. In addition, the repassivation results in the present work confirm the applicability of the slip-dissolution model for explaining stress corrosion cracking by means of electrochemical measurement. The scratch electrode system is valid and the scratch technique in HTHP water is a new effective and reliable method to study the SCC susceptibility of nuclear materials. Compared with other methods of evaluating stress corrosion cracking susceptibility of engineering materials, studying the repassivation kinetics by rapid scratch technique is more economical and time-efficient evaluation method for nuclear materials in high temperature pressurized water.