Author: W. H. CullenPublisher:
ISBN:
Category : Nuclear pressure vessels
Languages : en
Pages : 64
Book Description
Fatigue Crack Growth of A508 Steel in High-temperature, Pressurized Reactor Grade Water
Author: W. H. CullenPublisher:
ISBN:
Category : Nuclear pressure vessels
Languages : en
Pages : 64
Book Description
Fatigue Crack Growth of A508 Steel in High-Temperature, Pressurized Reactor-Grade Water
Author: W. H Cullen (Jr)Publisher:
ISBN:
Category :
Languages : en
Pages : 58
Book Description
Fatigue crack growth tests of A508-2 pressure vessel steel have been conducted at two test temperatures (93 degrees C and 288 degrees C) using a variety of constant amplitude waveforms. The load ratios were either 0.1 or 0.125, and the water chemistry was carefully monitored and controlled so as to simulate the nominal pressurized water reactor chemistry. The test procedures are described, and an examination of all the data indicates that the results fall into one of two rather clearly defined categories. One band of data, termed 'low', lies close to or essentially on, the ASME Section XI Code air environment default line. The other band of data, termed 'high', resides approximately midway between the ASME Sec. XI air and water environment default lines. The two bands of data are the result of certain combinations of the waveform and temperature variables listed above, and are determined by the following rules: (1) a ramp time in excess of one second is needed to obtain the high crack growth rate; and (2) the application of a hold time, together with a high temperature, serves to depress a normally high crack growth rate test (i.e. one with a long ramp time) and force it into the low category. A hydrogen embrittlement model is used as a basis for the explanation of this behavior. During the longer ramp times (approximately one second) hydrogen, formed by aqueous hydrolysis, diffuses into the plastic zone, resulting in local embrittlement.
A Review of Fatigue Crack Growth of Pressure Vessel and Piping Steels in High-temperature, Pressurized Reactor-grade Water
Author: W. H. CullenPublisher:
ISBN:
Category : Fractography
Languages : en
Pages : 134
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).
Fatigue Crack Growth Rates of A 508-2 Steel in Pressurized, High Temperature Water
Author: W. H. CullenPublisher:
ISBN:
Category : Fractography
Languages : en
Pages : 40
Book Description
Scientific and Technical Aerospace Reports
Author: Monthly Catalog of United States Government Publications
Author: Technical Abstract Bulletin
Author: Fatigue Crack Growth Rates in Pressure Vessel and Piping Steels in LWR Environments
Author: W. H. CullenPublisher:
ISBN:
Category : Light water reactors
Languages : en
Pages : 80
Book Description
Monthly Catalogue, United States Public Documents
Author: Publisher:
ISBN:
Category : Government publications
Languages : en
Pages : 1124
Book Description
Report of NRL Progress
Author: Naval Research Laboratory (U.S.)Publisher:
ISBN:
Category : Naval research
Languages : en
Pages : 628
Book Description