One of the most important aspects of the growth of nuclear power has been the development and use of quantitative methods of analysis for insuring the reliability of plant components and for protecting the public health and safety. Nuclear pressure vessels have always received close attention with regard to their structural strength and reliability. This paper summarizes some of the fracture mechanics concepts now being considered for application to the safety analysis of nuclear pressure vessels. The effects of thickness, temperature, strain rate, and irradiation on fracture toughness are discussed in terms of their relevance to a safety analysis. The amount of neutron irradiation likely to be received by a light water reactor pressure vessel is examined in terms of reactor type and the thermal power level. The ranges of temperature and fluence that are of most importance with regard to irradiation effects data on pressure vessel steels are identified. An example safety analysis problem is included to illustrate the application of the concepts discussed.
Skip Nav Destination
Article navigation
Research Papers
Fracture Safety Analysis Concepts for Nuclear Pressure Vessels, Considering the Effects of Irradiation
J. G. Merkle
J. G. Merkle
Oak Ridge National Laboratory, Oak Ridge, Tenn.
Search for other works by this author on:
J. G. Merkle
Oak Ridge National Laboratory, Oak Ridge, Tenn.
J. Basic Eng. Jun 1971, 93(2): 265-273 (9 pages)
Published Online: June 1, 1971
Article history
Received:
July 30, 1970
Online:
October 27, 2010
Citation
Merkle, J. G. (June 1, 1971). "Fracture Safety Analysis Concepts for Nuclear Pressure Vessels, Considering the Effects of Irradiation." ASME. J. Basic Eng. June 1971; 93(2): 265–273. https://doi.org/10.1115/1.3425223
Download citation file:
13
Views
Get Email Alerts
Cited By
Related Articles
Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel
J. Pressure Vessel Technol (June,2016)
Analysis of Radiation-Induced Embrittlement Gradients on Fracture Characteristics of Thick-Walled Pressure Vessel Steels
J. Eng. Ind (November,1971)
Probability of Fracture and Life Extension Estimate of the High-Flux Isotope Reactor Vessel
J. Pressure Vessel Technol (August,1998)
Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel
J. Pressure Vessel Technol (August,2021)
Related Proceedings Papers
Related Chapters
Long.Term Reactivity Change and Control: On.Power Refuelling
Fundamentals of CANDU Reactor Physics
Lessons Learned: NRC Experience
Continuing and Changing Priorities of the ASME Boiler & Pressure Vessel Codes and Standards
New Generation Reactors
Energy and Power Generation Handbook: Established and Emerging Technologies