“Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & Gen IV”
William J. O’Donnell and Donald S. Griffen, ASME Standards Technology, LLC, May 7, 2007
Keywords: ACRS; CRBR; high-temperature reactors; structural integrity; cyclic design; design analysis; creep fatigue; creep rupture
The U.S. Nuclear Regulatory Agency (NRC) and Advisory Committee on Reactor Safeguards (ACRS) raised issues in conjunction with the licensing of the high-temperature Clinch River Breeder Reactor (CRBR). A construction permit for CRBR was supported by the ACRS with the stipulation that numerous ACRS/NRC technical issues be resolved prior to requesting an operating license. The research and development program that was agreed upon to resolve elevated temperature structural integrity license issues was never implemented because Congress halted the construction of CRBR. The technical issues included materials, design analysis, weldment integrity, creep ratcheting, creep cracking, and creep fatigue-creep rupture damage evaluations.
Since the 1980’s, the ASME Code has made numerous improvements in elevated temperature structural integrity technology. These advances have been incorporated into Subsection NH of Section III of the code “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400 F (760 C) to about 1742 F (950 C) where creep effects limited structural integrity, safe allowable operating conditions, and design life. Materials that are more more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses.
Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III and NRC design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking. This report identifies the safety issues relevant to the ASME Boiler and Pressure Vessel Code including Sections III and VIII, Section III Subsection NH (Class 1 Components in Elevated Temperature Service) and Code Cases that must be resolved in order to support licensing of Generation IV Nuclear Reactors particularly Very-High-Temperature Gas-Cooled Reactors (VHTRs).
Company President, Bill, Sr. is active on the ASME Subcommittee on Design, and serves as a Contributing Member of the ASME (BPV III) Working Group on Fatigue Strength. He has also co-authored numerous papers on various engineering topics.