Energy Materials 2017: Materials for Nuclear Energy: Environmental Effects
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

Thursday 8:30 AM
March 2, 2017
Room: Miramar
Location: Marriott Marquis Hotel

Session Chair: Zhengdong Liu, China Iron & Steel Research Institute Group; Yiyin Shan, Institute of Metal Research, Chinese Academy of Sciences


8:30 AM  Invited
Environmental Assisted Cracking of the Additively Manufactured Austenitic Stainless Steel in High Temperature Water: Xiaoyuan Lou1; Paul Emigh1; Michelle Othon1; 1GE Global Research
    In recent years, nuclear industry has realized the potential of deploying additive manufacturing (AM) technology to fabricate reactor internal components due to its capability to produce complex near net shape geometries with minimum tooling requirement. For current reactor service and future reactor design/fabrication, additive manufacturing technology can significantly reduce the fabrication cost and time to market while improve the component performance. In this study, austenitic stainless steel fabricated by Direct Metal Laser Melting (DMLM) is evaluated for light water reactor applications. The corrosion fatigue and stress corrosion cracking of the AM stainless steel in boiling water reactor environment are studied under different post treatment conditions. The effects of crack propagation orientation, K level, and corrosion potential on SCC crack growth rate are studied in detail. The potential risk from the new process and EAC mechanism of this grade of material are also discussed in the talk.

9:10 AM  Invited
Effect of Steam Pressure on the Oxidation Behaviour of Alloy 625: Shengli Jiang1; Xiao Huang2; Wenjing Li3; Pei Liu4; 1Institute of Metal Research, Chinese Academy of Sciences; 2Carleton University; 3Canadian Nuclear Laboratories; 4CANMET
    The preliminary design of the Canadian SCWR uses a coolant operating under a pressure of 25 MPa at 625oC, reaching a peak temperature od 800oC. This presents challenges in materials selections due to limited data on material performance at such high temperatures and pressure. Ni based alloys have been of particular interest due to their ability to maintain high strength and toughness at elevated temperatures. Nickel-based Alloy 625 was assessed at 625oC for 1000 hours in supercritical water, subcritical water, and superheated steam. The samples showed small amount of weight gains after the exposure at 29 MPa and 0.1 MPa, and a slight weight loss at 8 MPa due to pitting formation. SEM examination of the compositions of the surface oxide indicated similar oxide formation on the top surface after exposures at different pressures, likely NiO or/and Ni(Cr,Al)2O4 type spinel. The implications of these results are discussed.

9:50 AM  
First Principles Investigations of Alternative Nuclear Fuels: Barbara Szpunar1; Linu Malakkal1; Ericmoore Jossou1; J.A. Szpunar1; 1University of Saskatchewan
    The comparison of properties for U3Si2, UN and UO2 is presented. The local density approximation (LDA) and generalized gradient approximation (GGA/WC) predict that U3Si2 is non-magnetic while GGA/PBE predicts that ferromagnetic ordering (not observed experimentally) is more stable by just 0.02 eV per uranium atom. On the other hand, the ground states of Urania and UN are independent of the used functional, with non-zero magnetic moment on uranium. U3Si2 is predicted to have lower binding energy (-14.4 eV) than UN (-20.4 eV) and UO2 (-32.7 eV) and melting point (1796 K versus 2932 K and 2987 K, respectively). The calculated minimal lattice thermal conductivity (in Wm-1K-1), indicates that phonon contribution to the thermal conductivity is smaller in both UN and U3Si2 (0.64 and 0.78 versus 1.06). However calculations confirm that UN and U3Si2 are safer alternative fuels due to their electronic thermal conductivity, increasing with temperature as also observed experimentally.

10:10 AM Break

10:25 AM  
Calculation of Phase Equilibria and Properties in Multi-Component Molten Salt Systems: Shuanglin Chen1; Weisheng Cao1; Fan Zhang1; Chuan Zhang1; Jun Zhu1; 1CompuTherm LLC
    Molten salt is one of the important nuclear reactor coolants. There is a growing interest in the various thermodynamic and physical properties in the molten salt systems, especially how these properties change with temperature and chemical compositions. This presentation will use CALPAHD method to calculate the phase equilibria and the thermodynamic and physical properties, such as activities, heat of mixing, densities, surface tension and viscosity, in multi-component molten salt systems. Calculated properties will be presented as 2D graphs and contour diagrams for better visualization and understanding. Optimal properties could be found by searching a multi-dimensional property space.

10:45 AM  
IASCC Behavior of Nickel-based Alloys in Light Water Reactors (LWRs): Mi Wang1; Miao Song1; Gary Was1; 1University of Michigan
    Nickel-based alloys 625Plus, 725 and 625DA (Direct Age) have exhibited excellent mechanical properties and good resistance to corrosion, however, no data exists on their susceptibility to irradiation assisted stress corrosion cracking (IASCC) in light water reactor environments. IASCC behavior was studied using constant extension rate tensile (CERT) tests on tensile bars that were irradiated with 2 MeV protons to doses of ~5 dpa at 360C at a strain rate of 1 x 10-7 s-1 in both PWR (320C, 1000 ppm [B], 2 ppm [Li], 35 cc/kg H2) and BWR (288C BWR NWC) environments to a fixed strain. Alloy 725 had the best resistance to IASCC in both environments while 625DA had the worst. The cracking behavior was interpreted in the context of the irradiated microstructure (dislocation loops, radiation induced precipitates, and RIS, etc…) and testing environments.

11:05 AM  
Oxidation of Alloy 690 in Simulated Pressurized Water Reactor Primary Environment: Wenjun Kuang1; Miao Song1; Peng Wang1; Gary Was1; 1University of Michigan
    The microstructures of oxides formed on alloy 690 in a simulated pressurized water reactor (PWR) primary environment were characterized to further understand the mechanism of internal oxidation. Following exposure at 360C for 1270 h, the surface was found to consist of NiFe2O4 particles which are epitaxial with the matrix. The internal oxide layer is composed of penetrative Cr rich spinel and Cr2O3. It forms by the solid state reactions of substrate with the inward diffusing oxygen. The spinel oxide has cube-on-cube orientation relationship with the matrix. The Cr2O3 penetrates deeper into the substrate which forms along the {111} planes of the metal. The redundant Ni in the metal diffuses out and enters into the solution. The compactness of Cr2O3 layer formed at low temperature is dependent on the defect density of the substrate. The internal oxide structure on the surface is similar to that along an IGSCC crack wall.

11:25 AM  
Compatibility Research of Fission Product Tellurium and Alloy N in Molten Salt Reactor: Z.J. Li1; 1Shanghai Insitute of Applied Physics CAS
    In this study, the diffusion behavior of Te in alloy GH3535, a Ni-Cr-Mo alloy designed specifically for the Chinese Thorium Molten Salts Reactor, was investigated by thermally exposing alloy samples in Te atmosphere at 800oC. Subsequently, room temperature tensile tests were conducted on the Te diffused alloy samples to study the degradation of mechanical properties and cracking behaviors of the alloy. Results show that Te preferred to diffuse along the random high angle grain boundaries in the alloy, and no Te was present in Σ3 grain boundaries. Cr-Te intermetallic compounds were observed at the grain boundary and the matrix/intergranular carbides interface, whereas more complicated mixture tellurides were identified at the grain boundary triple junctions. Under tensile stresses, the grain boundaries contained Te exhibited intergranular brittle cracking, resulting a reduction in the ultimate tensile stress of the alloy.

11:45 AM  
Friction Stir Processing of Degraded Austenitic Stainless Steel Nuclear Fuel Dry Cask Storage System Canisters: Ben Sutton1; Kenneth Ross2; Glenn Grant2; Gary Cannell3; Greg Frederick1; Robert Couch1; 1Electric Power Research Institute; 2Pacific Northwest National Laboratory; 3Fluor Enterprises, Inc.
    Chloride-induced stress corrosion cracking (CISCC) of austenitic stainless steel dry cask storage system (DCSS) canisters has been identified as an industry concern. Typical DCSS canisters are constructed from Types 304 or 316 stainless steel or their variants via conventional fusion welding processes. The presence of residual tensile stress and Cr-carbide precipitation within the weld heat affected zone (HAZ) places canisters near salt-bearing environments at an elevated risk for CISCC. The current study evaluates the suitability of friction stir processing (FSP) to repair SCC and remediate sensitized fusion weld HAZs. FSP was applied to furnace sensitized Type 304 specimens containing laboratory-generated SCC and evaluated using liquid penetrant inspection, phased array ultrasonic inspection, and optical microscopy. In addition, fusion welded Type 304L specimens were fabricated, subjected to FSP, and destructively analyzed via ASTM A262 and optical microscopy. Results demonstrate that FSP is a viable option for SCC repair and sensitization remediation.