TY - GEN
T1 - Auxiliary beam stress improved laser welding for repair of irradiated light water reactor components
AU - Chen, Jian
AU - Miller, Roger
AU - Leonard, Keith
AU - Tatman, Jonathan
AU - Tang, Wei
AU - Sutton, Benjamin
AU - Feng, Zhili
AU - Gussev, Maxim
AU - Frederick, Greg
N1 - Publisher Copyright:
Copyright © 2019 ASME
PY - 2019
Y1 - 2019
N2 - The welding task focuses on development of advanced welding technologies for repair and maintenance of nuclear reactor structural components to safely and cost-effectively extend the service life of nuclear power reactors. This paper presents an integrated research and development effort by the Department of Energy Light Water Reactor Sustainability Program through the Oak Ridge National Laboratory (ORNL) and Electric Power Research Institute (EPRI) to develop a patent-pending technology, Auxiliary Beam Stress Improved Laser Welding Technique, that proactively manages the stresses during laser repair welding of highly irradiated reactor internals without helium induced cracking (HeIC). Finite element numerical simulations and in-situ temperature and strain experimental validation have been utilized to identify candidate welding conditions to achieve significant stress compression near the weld pool during cooling. Preliminary welding experiments were performed on irradiated stainlesssteel plates (Type 304L). Post-weld characterization reveals that no macroscopic HeIC was observed.
AB - The welding task focuses on development of advanced welding technologies for repair and maintenance of nuclear reactor structural components to safely and cost-effectively extend the service life of nuclear power reactors. This paper presents an integrated research and development effort by the Department of Energy Light Water Reactor Sustainability Program through the Oak Ridge National Laboratory (ORNL) and Electric Power Research Institute (EPRI) to develop a patent-pending technology, Auxiliary Beam Stress Improved Laser Welding Technique, that proactively manages the stresses during laser repair welding of highly irradiated reactor internals without helium induced cracking (HeIC). Finite element numerical simulations and in-situ temperature and strain experimental validation have been utilized to identify candidate welding conditions to achieve significant stress compression near the weld pool during cooling. Preliminary welding experiments were performed on irradiated stainlesssteel plates (Type 304L). Post-weld characterization reveals that no macroscopic HeIC was observed.
UR - http://www.scopus.com/inward/record.url?scp=85075823662&partnerID=8YFLogxK
U2 - 10.1115/PVP2019-93667
DO - 10.1115/PVP2019-93667
M3 - Conference contribution
AN - SCOPUS:85075823662
T3 - American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP
BT - Codes and Standards
PB - American Society of Mechanical Engineers (ASME)
T2 - ASME 2019 Pressure Vessels and Piping Conference, PVP 2019
Y2 - 14 July 2019 through 19 July 2019
ER -