This paper describes stress analysis and fracture mechanics work performed to assess boiling water reactor (BWR) and pressurized water reactor (PWR) nozzles located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Various RPV nozzle geometries were investigated: 1. BWR recirculation outlet nozzle; 2. BWR core spray nozzle; 3. PWR inlet nozzle; 4. PWR outlet nozzle; and 5. BWR partial penetration instrument nozzle. The above nozzle designs were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-license (EOL) to require evaluation as part of establishing the allowed limits on heatup, cooldown, and hydrotest (leak test) conditions. These nozzles analyzed represent one each of the nozzle types potentially requiring evaluation. The purpose of the analyses performed on these nozzle designs was as follows: • To model and understand differences in pressure and thermal stress results using a two-dimensional (2-D) axi-symmetric finite element model (FEM) versus a three-dimensional (3-D) FEM for all nozzle types. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated; • To verify the accuracy of a selected linear elastic fracture mechanics (LEFM) hand solution for stress intensity factor for a postulated nozzle corner crack for both thermal and pressure loading for all nozzle types; • To assess the significance of attached piping loads on the stresses in the nozzle corner region; and • To assess the significance of applying pressure on the crack face with respect to the stress intensity factor for a postulated nozzle corner crack.
Skip Nav Destination
ASME 2011 Pressure Vessels and Piping Conference
July 17–21, 2011
Baltimore, Maryland, USA
Conference Sponsors:
- Pressure Vessels and Piping Division
ISBN:
978-0-7918-4451-9
PROCEEDINGS PAPER
Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles
Shengjun Sean Yin,
Shengjun Sean Yin
Oak Ridge National Laboratory, Oak Ridge, TN
Search for other works by this author on:
Gary L. Stevens,
Gary L. Stevens
U.S. Nuclear Regulatory Commission, Rockville, MD
Search for other works by this author on:
B. Richard Bass,
B. Richard Bass
Oak Ridge National Laboratory, Oak Ridge, TN
Search for other works by this author on:
Mark T. Kirk
Mark T. Kirk
U.S. Nuclear Regulatory Commission, Rockville, MD
Search for other works by this author on:
Shengjun Sean Yin
Oak Ridge National Laboratory, Oak Ridge, TN
Gary L. Stevens
U.S. Nuclear Regulatory Commission, Rockville, MD
B. Richard Bass
Oak Ridge National Laboratory, Oak Ridge, TN
Mark T. Kirk
U.S. Nuclear Regulatory Commission, Rockville, MD
Paper No:
PVP2011-57014, pp. 963-972; 10 pages
Published Online:
May 21, 2012
Citation
Yin, SS, Stevens, GL, Bass, BR, & Kirk, MT. "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles." Proceedings of the ASME 2011 Pressure Vessels and Piping Conference. Volume 1: Codes and Standards. Baltimore, Maryland, USA. July 17–21, 2011. pp. 963-972. ASME. https://doi.org/10.1115/PVP2011-57014
Download citation file:
34
Views
Related Proceedings Papers
Related Articles
Safety Assessment of Reactor Pressure Vessel Integrity for Loss of Coolant Accident Conditions
J. Pressure Vessel Technol (February,2012)
CFD Tool for Assessment of the Reactor Pressure Vessel Integrity in Pressure Thermal Shock Conditions: Influence of Turbulence Model and Mesh Refinement on the Vessel Thermal Loading During PTS Transient
J. Pressure Vessel Technol (June,2011)
Brittle Failure Assessment of a PWR-RPV for Operating Conditions and Loss of Coolant Accident
J. Pressure Vessel Technol (August,2008)
Related Chapters
Development of Nuclear Boiler and Pressure Vessels in Taiwan
Companion Guide to the ASME Boiler and Pressure Vessel Code, Volume 3, Third Edition
Lessons Learned: NRC Experience
Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards
Subsection NG—Core Support Structures
Companion Guide to the ASME Boiler & Pressure Vessel Codes, Volume 1 Sixth Edition