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Fatigue Life and Crack Growth Predictions of Irradiated Stainless Steels.

紀錄類型:	書目-語言資料,印刷品 : Monograph/item
正題名/作者:	Fatigue Life and Crack Growth Predictions of Irradiated Stainless Steels./
作者:	Fuller, Robert William.
出版者:	Ann Arbor : ProQuest Dissertations & Theses, : 2018,
面頁冊數:	182 p.
附註:	Source: Dissertation Abstracts International, Volume: 79-09(E), Section: B.
Contained By:	Dissertation Abstracts International79-09B(E).
標題:	Mechanical engineering. -
電子資源:	http://pqdd.sinica.edu.tw/twdaoapp/servlet/advanced?query=10747959
ISBN:	9780355920871

Fatigue Life and Crack Growth Predictions of Irradiated Stainless Steels.
Fuller, Robert William.

Fatigue Life and Crack Growth Predictions of Irradiated Stainless Steels. - Ann Arbor : ProQuest Dissertations & Theses, 2018 - 182 p.

Source: Dissertation Abstracts International, Volume: 79-09(E), Section: B.

Thesis (Ph.D.)--Mississippi State University, 2018.

One of prominent issues related to failures in nuclear power components is attributed to material degradation due the aggressive environment conditions, and mechanical stresses. For instance, reactor core support components, such as fuel claddings, are under prolonged exposure to an intense neutron field from the fission of fuel and operate at elevated temperature under fatigue loadings caused by start up, shut down, and unscheduled emergency shut down. Additionally, exposure to high-fluence neutron radiation can lead to microscopic defects that result in material hardening and embrittlement, which significantly affects the physical and mechanical properties of the materials, resulting in further reduction in fatigue life of reactor structural components. The effects of fatigue damage on material deterioration can be further exacerbated by the presence of thermal loading, hold-time, and high-temperature water coolant environments.

ISBN: 9780355920871Subjects--Topical Terms:

557493
Mechanical engineering.

Fatigue Life and Crack Growth Predictions of Irradiated Stainless Steels.
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300 $a 182 p.
500 $a Source: Dissertation Abstracts International, Volume: 79-09(E), Section: B.
500 $a Advisers: Haitham El Kadiri; Nima Shamsaei.
502 $a Thesis (Ph.D.)--Mississippi State University, 2018.
520 $a One of prominent issues related to failures in nuclear power components is attributed to material degradation due the aggressive environment conditions, and mechanical stresses. For instance, reactor core support components, such as fuel claddings, are under prolonged exposure to an intense neutron field from the fission of fuel and operate at elevated temperature under fatigue loadings caused by start up, shut down, and unscheduled emergency shut down. Additionally, exposure to high-fluence neutron radiation can lead to microscopic defects that result in material hardening and embrittlement, which significantly affects the physical and mechanical properties of the materials, resulting in further reduction in fatigue life of reactor structural components. The effects of fatigue damage on material deterioration can be further exacerbated by the presence of thermal loading, hold-time, and high-temperature water coolant environments.
520 $a In this study, uniaxial fatigue models were used to predict fatigue behavior based only on simple monotonic properties including ultimate tensile strength and Brinell hardness. Two existing models, the Baumel Seeger uniform material law and the Roessle Fatemi hardness method, were employed and extended to include the effects of test temperature, neutron irradiation fluence, irradiation induced helium and irradiation induced swellings on fatigue life of austenitic stainless steels. Furthermore, a methodology to estimate fatigue crack length using a strip-yield based model is presented. This methodology is also extended to address the effect of creep deformation in a presence of hold- times, and expanded to include the effects of irradiation and water environment. Reasonable fatigue life predictions and crack growth estimations are obtained for irradiated austenitic stainless steels types 304, 304L, and 316, when compared to the experimental data available in the literature. Lastly, a failure analysis methodology of a mixer unit shaft made of AISI 304 stainless steel is also presented using a conventional 14-step failure analysis approach. The primary mode of failure is identified to be intergranular stress cracking at the heat affected zones. A means of circumventing this type of failure in the future is presented.
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