Some Mechanistic Observations on the Crack Growth Characteristics of Pressure Vessel and Piping Steels in PWR Environment

Some Mechanistic Observations on the Crack Growth Characteristics of Pressure Vessel and Piping Steels in PWR Environment PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :

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The fatigue crack growth behavior of A533B and A508 pressure vessel steel and AISI Types 304 and 316 steels used in reactor coolant piping have been studied in a pressurized water reactor environment at 288°C (550°F). The influence of stress ratio (P/sub min//P/sub max/), frequency, ramp times, specimen orientation and material microstructures were included in the study. While none of the materials showed evidence of static crack growth in the environment, the ferritic steels did show an enhanced fatigue crack growth rate at test frequencies of five cycles per minute and lower. Based on fractographic examinations the enhanced growth rate is not the result of environmentally induced intergranular or cleavage modes of crack propagation. Instead, striation spacing measurements were found to agree with the macroscopic crack growth rate, demonstrating a time dependent environmental interaction which introduces a frequency dependent enhancement of the mechanically developed striations. Crack growth experiments using hold times have confirmed the absence of any superimposed contribution of static crack growth components. Fatigue crack growth tests were conducted in an environment of Hydrogen Sulfide gas to establish the contribution of hydrogen embrittlement and will also be described.

Some Mechanistic Observations on the Crack Growth Characteristics of Pressure Vessel and Piping Steels in PWR Environment

Some Mechanistic Observations on the Crack Growth Characteristics of Pressure Vessel and Piping Steels in PWR Environment PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :

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Book Description
The fatigue crack growth behavior of A533B and A508 pressure vessel steel and AISI Types 304 and 316 steels used in reactor coolant piping have been studied in a pressurized water reactor environment at 288°C (550°F). The influence of stress ratio (P/sub min//P/sub max/), frequency, ramp times, specimen orientation and material microstructures were included in the study. While none of the materials showed evidence of static crack growth in the environment, the ferritic steels did show an enhanced fatigue crack growth rate at test frequencies of five cycles per minute and lower. Based on fractographic examinations the enhanced growth rate is not the result of environmentally induced intergranular or cleavage modes of crack propagation. Instead, striation spacing measurements were found to agree with the macroscopic crack growth rate, demonstrating a time dependent environmental interaction which introduces a frequency dependent enhancement of the mechanically developed striations. Crack growth experiments using hold times have confirmed the absence of any superimposed contribution of static crack growth components. Fatigue crack growth tests were conducted in an environment of Hydrogen Sulfide gas to establish the contribution of hydrogen embrittlement and will also be described.

A Review of Fatigue Crack Growth of Pressure Vessel and Piping Steels in High-temperature, Pressurized Reactor-grade Water

A Review of Fatigue Crack Growth of Pressure Vessel and Piping Steels in High-temperature, Pressurized Reactor-grade Water PDF Author: W. H. Cullen
Publisher:
ISBN:
Category : Fractography
Languages : en
Pages : 134

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Book Description
Fatigue crack growth data sets, for pressure vessel and piping steels, in reactor-grade water environment have appeared in various reports and publications since about 1972. All of the results which have been published from 1972 through 1979 have been plotted and are presented in this report. Beginning with a discussion of the need for these data, and an explanation of the laboratory facilities which are required for this research, this report goes on to describe the overall trends which have evolved through consideration of the data sets and the conditions under which they were generated. A model for hydrogen assisted fatigue crack growth is described and applied to the pressurized water reactor type of environment. A complete listing of references is included in the report. (Author).

NUREG/CR.

NUREG/CR. PDF Author: U.S. Nuclear Regulatory Commission
Publisher:
ISBN:
Category : Nuclear energy
Languages : en
Pages : 518

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Scientific and Technical Aerospace Reports

Scientific and Technical Aerospace Reports PDF Author:
Publisher:
ISBN:
Category : Aeronautics
Languages : en
Pages : 602

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Book Description
Lists citations with abstracts for aerospace related reports obtained from world wide sources and announces documents that have recently been entered into the NASA Scientific and Technical Information Database.

Proceedings of the U.S. Nuclear Regulatory Commission ... Water Reactor Safety Research Information Meeting

Proceedings of the U.S. Nuclear Regulatory Commission ... Water Reactor Safety Research Information Meeting PDF Author:
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 776

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Energy Research Abstracts

Energy Research Abstracts PDF Author:
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 486

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Environmentally Assisted Cracking

Environmentally Assisted Cracking PDF Author: W. Barry Lisagor
Publisher: ASTM International
ISBN: 0803112769
Category : Alloys
Languages : en
Pages : 540

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Book Description
Papers included topics of phenomena, basic mechanisms, modeling, test methodologies, materials performance, engineering applications and service experience and failures and reflects the current emphasis with regard to material/environment systems.

NRL Report

NRL Report PDF Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 62

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Fatigue Crack Growth of A508 Steel in High-temperature, Pressurized Reactor Grade Water

Fatigue Crack Growth of A508 Steel in High-temperature, Pressurized Reactor Grade Water PDF Author: W. H. Cullen
Publisher:
ISBN:
Category : Nuclear pressure vessels
Languages : en
Pages : 64

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Structural Integrity of Water Reactor Pressure Boundary Components

Structural Integrity of Water Reactor Pressure Boundary Components PDF Author: Materials Engineering Associates
Publisher:
ISBN:
Category : Light water reactors
Languages : en
Pages : 112

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