Energy Materials 2017: Materials for Nuclear Energy: Accident Tolerant Fuels & Irradiation 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 2:00 PM
March 2, 2017
Location: Marriott Marquis Hotel
Session Chair: Peter Hosemann, University of California Berkeley
Advanced ODS FeCrAl Alloys for Accident-tolerant Fuel Cladding: Sebastien Dryepondt1; Caleb Massey1; Philip Edmondson1; Kurt Terrani1; 1Oak Ridge National Laboratory
ODS FeCrAl alloys are good candidates for accident-tolerant fuel cladding because of their superior oxidation performance at very high temperature, excellent tensile strength at temperature up to ~800ºC, and limited irradiation-induced embrittlement at low temperature. A first generation of new ODS Fe-12Cr-5Al alloys exhibited, however, high hardness values and low ductility at temperature below ~400ºC, which could hinder the fabrication of very thin tubes (<500um). The impact of several fabrication parameters and minor element additions on ODS Fe-10/12Cr-6Al tensile properties was therefore studied. Decreasing the ball milling time or increasing the extrusion temperature led to an increase of the alloy ductility and decrease of the alloy strength. Addition of Zr or Zr and Ti had a limited impact on the tensile properties below 400ºC, and plastic ductility >15% at T<800ºC was achieved for many of these alloys. Research sponsored by the U.S. DOE, Office of Nuclear Energy, FCRD Program.
Minimizing Hydrogen Diffusion through FeCrAl Alloy Accident Tolerant Fuel Cladding: Raul Rebak1; Young Kim1; 1GE Global Research
The US Department of Energy is partnering with fuel vendors to study alternatives to the current UO2 – zirconium alloy system to make commercial nuclear power plants more resistant to accidents. The proposed alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric is proposing to replace zirconium based alloy cladding in current commercial power reactors with a FeCrAl cladding such as APMT. FeCrAl alloys do not react with hydrogen to form stable hydrides as zirconium alloys do. Therefore, it is possible that more tritium may be released to the coolant with the use of FeCrAl cladding. This work discusses the formation of an alumina layer on the surface of APMT cladding as an effective barrier for tritium permeation from the fuel to the coolant across the cladding wall.
The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions: Mahmut Cinbiz1; Nicholas Brown1; Kurt Terrani1; Rick Lowden1; Donald Erdman III1; 1Oak Ridge National Laboratory
This study investigates the failure mechanisms of advanced nuclear fuel cladding of FeCrAl and SiC-SiC composite at high-strain rates, similar to design basis reactivity initiated accidents (RIA). During RIA, the nuclear fuel cladding is subjected to the plane-strain to equibiaxial tension strain states. To achieve those accident conditions, the samples are to be deformed by the expansion of high strength Inconel alloy tube under pre-specified pressure pulses as occurring RIA. The mechanical response of the advanced claddings will be compared to the that of hydrided Zirconium-based nuclear fuel cladding alloys. The strain evolution during pressure pulses will be collected in situ; the failure strains of both advanced claddings and the current nuclear fuel alloys will be determined and compared to high strain states. The results will be discussed in terms of damage accumulation in the advanced cladding.
Systematic Studies on Dispersoid Stability and Swelling Resistance in ODS Alloys under Ion Irradiation Conditions: Hyosim Kim1; Jonathan Gigax1; Tianyi Chen1; Frank Garner1; Lin Shao1; 1Texas A&M University
Oxide-dispersion-strengthened (ODS) alloys are candidates for structural components of advanced reactors. Here we first compare the swelling resistance during high dose ion irradiation of a wide variety of commercially available ODS alloys, from easily-swelling MA 956 with unstable dispersoids to very swelling-resistant 12 Cr ODS variants with relatively stable dispersoids. The second material focus involves three new developmental ODS variants, seeking further optimization of swelling resistance by varying thermal-mechanical treatment and alloying elements such as Al, Ni, W, and O, to produce oxide dispersoids of different densities, sizes and interface coherency. Using 3.5 MeV Fe ion irradiation, these alloy variants were irradiated at various irradiation temperatures, displacement-per-atom (dpa) values, and dpa rates. This study provides comprehensive experimental data on dispersoid stability to compare with our previously published theoretical model in which temperature, dpa, dpa rate and interface coherency are key parameters to predict dispersoid shrinkage or growth under irradiation.
In-situ Observation on the Oxides Stability under Laser and/or Electron Beams Irradiations in 9Cr-ODS Steel: Wang Hui1; Yang Zhanbing2; Yang Subing1; Watanabe Seiichi3; Shibayama Tamaki3; 1University of Science & Technology Beijing; 2School of Metallurgical and Ecological Engineering, State Key Laboratory of Advanced Metallurgy,University of Science and Technology Beijing; 3Centre for Advanced Research of Energy and Materials, Faculty of Engineering, Hokkaido University
Oxide-dispersion-strengthened (ODS) steels have been considered as the candidate materials for Gen.Ⅳ fission reactor core components and fusion reactor first wall with their excellent radiation resistance and high temperature mechanical properties. In this paper, three kinds of irradiation, electron irradiation, laser-electron sequential irradiation and laser-electron dual-beam simultaneous irradiation, were carried out using a high-voltage electron microscope (HVEM) equipped with a laser head (Laser-HVEM) at temperatures 725 K, 775 K and 825 K to investigate the changes of oxides in ODS steel by in-situ observation. The results showed that oxides size have a decrease trend with the increase of irradiation dose, and the decrease rates of oxides vary under different temperatures and irradiation methods. Coherency between oxides and the matrix are confirmed by high resolution transmission electron microscope (HRTEM), and microscopic changes of oxides under different irradiation conditions are compared for further analysis.
3:40 PM Break
A Preliminary Investigation on the Phase Transformation Kinetics Behavior of an U-10wt%Mo Cast and Homogenized Alloy: Saumyadeep Jana1; Arun Devaraj1; Vineet Joshi1; Curt Lavender1; 1PNNL
Phase transformation kinetics has been investigated in a cast and homogenized U-10wt%Mo alloy through isothermal heat-treatment studies carried out at 500° C and 400° C. The cast and homogenized alloy fully consisted of metastable bcc γ-UMo phase. Isothermal heat-treatment led to formation of a lamellar microstructure along the grain boundaries of initial γ-UMo phase. At 500° C, 1.4 % transformation was observed after one hour of thermal exposure, and almost complete transformation of the initial γ-UMo phase was noted after 100 hours of thermal treatment. At 400° C, only 5.1 % of the initial γ-UMo phase transformation has been noted after 100 hours of exposure. The lamellar microstructure formed at 400° C is significantly finer compared to 500° C condition. However, after 10 days and 21 days of exposure at 400° C, a very different looking microstructure forms. Results of these heat-treatment studies will be reported here.
First Principles Study of Electronic Structure and Thermo-mechanical Properties of the Components of Accident Tolerant Nuclear Fuel: UO2 and UB2: Ericmoore Jossou1; Linu Malakkal1; Dotun Oladimeji1; Barbara Szpunar1; Jerzy Szpunar1; 1University of Saskatchewan
The nuclear accident in Fukushima clearly illustrates the risks associated with the present design of reactors based on pure uranium oxide fuel and justify the research towards Accident Tolerant Fuel (ATF). ATFs are fuels with enhanced thermal properties, which can withstand the loss of coolant for a long time by allowing faster dissipation of heat, thus lowering the fuel centerline temperature and preventing fuel meltdown. Also, there is renew interest in using UO2-UB2 composite due to the high melting point, and thermal stability compared to pristine urania. In this work, theoretical calculations are performed using density functional theory codes. Elastic properties analysis shows UB2 to be intrinsically ductile due to B/G > 1.75 also the Poisson ratio of 0.35 implies that ionic contribution is dominant. The lattice thermal conductivity at room temperature is calculated to be an order of 5.7 larger compared to urania at room temperature.
Irradiation Defects in UO2, CeO2 and (U, Ce)O2 Leached in Oxidizing Water: An In-situ Raman Study: Ritesh Mohun1; Lionel Desgranges1; Christophe Jégou1; Sandrine Miro1; Patrick Simon2; Aurélien Canizarès2; Nicole Raimboux2; 1CEA (French Alternative Energies and Atomic Energy Commission), France; 2CNRS(French National Centre for Scientific Research), France
Our study is focused on an incidental scenario of a defective fuel rod stored in water pools, which will allow water to interact with the irradiated fuel matrix. We simulated this scenario under laboratory conditions using He irradiation of a UO2/H2O system. An in-situ Raman installation was used to measure the kinetics of irradiation induced defects in UO2. The acquired data were then compared with a reference UO2/(Ar+H2) system. The results revealed that the presence of either aerated deionized water or reducing environment modifies the formation kinetics of irradiation defects under alpha irradiation. In addition, the precipitation of the studtite phase is also observed during the leaching experiment and detailed solution analysis showed that the dissolution of UO2 proceeds without the formation of an oxidized UO2+x layer. Similar studies carried out with CeO2 and (U, Ce)O2, which is a surrogate material for MOX fuel, will also be presented.
Comparative Study of Thermal Conductivity of SiC and BeO from Ab Initio Calculations: Linu Malakkal1; Barbara Szpunar1; Jerzy Szpunar1; 1University of Saskatchewan
SiC and BeO are materials proposed to use in accident tolerant fuel. Therefore, we did a systematic study of the thermal conductivity of SiC and BeO and its dependence on temperatures and structure by solving the Boltzmann transport equation using the shengBTE a solver for phonon thermal conductivity (kL) with ab initio techniques. We also predict the structural, elastic, and thermodynamic properties of alpha-SiC, wurtzite (w)-SiC and w-BeO by first principles calculation using Quantum ESPRESSO within quasi-harmonic approximation. kL is also predicted using the Slack model. The thermo-mechanical properties of these materials show significant improvement over Urania with one order of magnitude higher thermal conductivity. Wurtzite structure shows the directional dependence of kL. Hence, we provide the directional thermal conductivity of w-BeO, w-SiC and compare with the thermal conductivity of cubic SiC. The simulated results are compared with the available experimental data and showed excellent agreement.
Morphology of Y-Ti Nano-oxides in ODS Alloys Irradiated with High Energy Heavy Ions: Vladimir Skuratov1; Alexander Sohatsky1; Jacques O'Connell2; Kateryna K. Kornieieva11; Jan Neethling2; Alexey Volkov3; Maxim Zdorovets4; 1FLNR JINR; 2CHRTEM, Nelson Mandela Metropolitan University; 3Nazarbaev University; 4Institute of Nuclear Physics, Astana, Kazakhstan
Irradiation with heavy ions of fission fragment energy leads to amorphization of Y-Ti nano-oxides particles in ODS alloys. At present there is no any experimental data demonstrating that amorphized nanoparticles in ODS materials will assure the same properties and the same excellent radiation resistance like observed for steels containing crystalline nanoparticles. This assumes that mechanical properties of a few microns thick layer at the surface of ODS alloy based cladding material contacting with nuclear fuel may be significantly affected by fission fragments exposure. Aim of this report is to study the morphology of swift (167 and 220 MeV) Xe ion induced latent tracks in Y2Ti2O7 nanoparticles during post-irradiation heat treatment and after irradiation at different temperatures