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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Corrosion in Heavy Liquid Metals for Energy Systems
Presentation Title Material Needs and Developments for the Westinghouse Lead Fast Reactor
Author(s) Mike Ickes, Arash Parsi, Luke Czerniak, Paolo Ferroni
On-Site Speaker (Planned) Mike Ickes
Abstract Scope Westinghouse is actively developing a Lead Fast Reactor (LFR), chosen due to the inherent safety and economic potential achievable with this Generation IV reactor type. Critical activities supporting the reactor design and development are materials testing efforts within Westinghouse and at national laboratories and universities worldwide. An overview of the Westinghouse LFR design will be given with an emphasis on aspects critical to materials performance. The Westinghouse-led materials testing efforts being performed in liquid lead will be reviewed and the available results will be summarized. These efforts include exploring the development of new materials while also investigating the use of commercially mature materials. Future activities in materials testing will also be briefly discussed.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, High-Temperature Materials,


Anubis Multiphysics: A Neutronics-Thermal Hydraulics Coupling Platform for Flow Accelerated Corrosion Modeling in Reactor Conditions
Behaviour of spallation, activation and fission products in LBE
Compatibility of Alumina-Forming Austenitic Steels in Static and Flowing Pb
Corrosion behaviour and microstructural stability of alumina-forming austenitic steels exposed to oxygen-containing molten lead
Corrosion Investigations of Materials in Antimony-Tin And Antimony-Bismuth Alloys For Liquid Metal Batteries
Corrosion of Refractory Metals and Advanced Steels in Lead-Bismuth Eutectic
Electromagnetic Flow Sensor for Heavy Liquid Metals for Energy Systems
Engineering Model of the Kinetics of the Steel Oxide Layer in A Flow of a Heavy Liquid Metal Coolant Under Various Oxygen Conditions
Exposure Tests of Different Materials in Liquid Lead for LFRs: Effect of the Dissolved Oxygen on Corrosion
Fundamental Interactions of Steels and Nickel-based Alloys with Lead-based Liquid Alloys or Liquid Tin
In-Situ Crack Growth Testing in a Liquid Metal Environments
Investigation on the evaporation rate of liquid lead and radioisotope retention capability of molten lead as coolant
Liquid metal embrittlement of Al-containing high-entropy alloys exposed to lead-bismuth eutectic
Material Needs and Developments for the Westinghouse Lead Fast Reactor
Materials Compatibility Testing with Molten Lead up to 700C
Numerical and Analytical Researches of the Formation and Accumulation of Deposits on the Circuit with HLM Coolant within the Consistent Model for Physical and Chemical Processes
Numerical modelling of coolant chemistry in lead bismuth eutectic cooled nuclear reactors
Oxide Growth Modeling in Fe-Cr-Al Alloys: A Current Review
Performance of Candidate Alloys at 500C in Flowing Lead
PILLAR: Pool-type Integral Leading facility for Lead-alloy cooled Advanced small modular Reactor, and Its Use for Natural Convection Study and Corrosion
Preliminary results on the compatibility of Fe-Cr-Al and Fe-Cr-Al-Mo steels with liquid sodium at 700 C.
Progress in LBE chemistry control and measurement techniques for MYRRHA
Review of Liquid Metal Corrosion Under Irradiation and Progress Report on the LBE-Irradiation-Corrosion Experiment (ICE)

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