Doctoral Dissertations

Keywords and Phrases

Bending fatigue test; FEA code ABAQUS; Krouse-type; S-N curve

Abstract

"New materials with superior radiation and corrosion resistance are needed to improve the economy and performance of current nuclear reactors as well as future nuclear reactors. Measurement of mechanical properties of the material of equipment is required to estimate its safe operating life. Studying fatigue of irradiated specimens is challenging due to space limitation in research reactors (e.g. ATR). The mini-specimen bending fatigue (Krouse-type) of nuclear materials was used to study fatigue properties and compare the obtained results with that of reference data of full size specimen. These Krouse-type were made of austenitic SS304, SS316, HT9 ferritic martensitic steel, and Incoloy alloy MA956 oxide dispersion strengthens (ODS). A comparison between the mini-specimens and the standard specimen ASTM B593 had similar design but a smaller size. The bending fatigue machine VSS-40H was designed and LEF-150 was modified with a specially designed adapter to test the fatigue of the mini-specimen. Consequently, the S-N curve for both HT9 and Incoloy alloy MA956 were defined. The endurance limits were measured to be 94 MPa for HT9 and 132 MPa for Incoloy alloy MA956. The finite element model code ABAQUS was used to estimate the level of stress in the bent mini specimen. The plots indicated that the 3- parameter Weibull distribution fits the data well. The correlation coefficient value for 3-parameter Weibull was improved and increased by 33.66 % for the HT9, 30.88% for the as-received and 26.51% for the thermal aging for Incoloy alloy MA956. The thermal aging process had little impact on the mini-specimens fatigue life. The analysis of error propagation of specimen's stress has been observed that the most of the error was in deflection measurements. Thus this Krouse-type was seen as the solution to studying fatigue of irradiated specimens as it can solve the challenge of limitation of space in research reactors."--Abstract, page iv.

Advisor(s)

Castano, Carlos H.
Newkirk, Joseph William

Committee Member(s)

Castano, Carlos H.
Newkirk, Joseph William
Lee, Hyoung-Koo
Al-Dahhan, Muthanna H.
Liu, Xin

Department(s)

Mining and Nuclear Engineering

Degree Name

Ph. D. in Nuclear Engineering

Sponsor(s)

King Abdul-Aziz City of Science & Technology (KACST)

Publisher

Missouri University of Science and Technology

Publication Date

Spring 2015

Journal article titles appearing in thesis/dissertation

  • An experimental study on bending fatigue test with a Krouse - type fatigue specimen
  • Characterization a bending fatigue mini-specimen technique (Krouse-type) of nuclear materials
  • Weibull statistical analysis of Krouse -type bending fatigue in nuclear materials of HT9
  • Influence of thermal aging on the mini specimen (Krouse - type) bending fatigue of nuclear materials
  • A comparison of bending fatigue specimens (Krouse - type) for nuclear materials

Pagination

xiii, 119 pages

Note about bibliography

Includes bibliographic references.

Rights

© 2015 Ahmed Suliman R. Haidyrah, All rights reserved.

Document Type

Dissertation - Open Access

File Type

text

Language

English

Library of Congress Subject Headings

Nuclear reactors -- Materials -- Fatigue -- Testing
Bending stresses
Abaqus (Electronic resource)
Weibull distribution

Thesis Number

T 10713

Electronic OCLC #

913392450

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