Zobrazeno 1 - 10
of 14
pro vyhledávání: '"Marius Twite"'
Autor:
Matthias Bruchhausen, Gintautas Dundulis, Alec McLennan, Sergio Arrieta, Tim Austin, Román Cicero, Walter-John Chitty, Luc Doremus, Miroslava Ernestova, Albertas Grybenas, Caitlin Huotilainen, Jonathan Mann, Kevin Mottershead, Radek Novotny, Francisco Javier Perosanz, Norman Platts, Jean-Christophe le Roux, Philippe Spätig, Claudia Torre Celeizábal, Marius Twite, Marc Vankeerberghen
Publikováno v:
Metals, Vol 11, Iss 2, p 307 (2021)
A substantial amount of research effort has been applied to the field of environmentally assisted fatigue (EAF) due to the requirement to account for the EAF behaviour of metals for existing and new build nuclear power plants. We present the results
Externí odkaz:
https://doaj.org/article/40f5530529f54b26bcb39f760bd66a16
Publikováno v:
Volume 1A: Codes and Standards.
High temperature water environments typical of LWR operation are known to significantly reduce the fatigue life of reactor plant materials relative to air environments in laboratory studies. This environmental impact on fatigue life has led to the is
Publikováno v:
Volume 1A: Codes and Standards.
Environmentally assisted fatigue of nuclear plant materials in the Pressurised Water Reactor (PWR) coolant environment is a phenomenon that has been extensively studied over the past 30 years. Methods for accounting for the PWR environment in an ASME
Autor:
Marc Vankeerberghen, Matthias Bruchhausen, R. Cicero, Luc Doremus, Jean-Christophe Le-Roux, Norman Platts, Philippe Spätig, Marius Twite, Kevin Mottershead
Publikováno v:
Microsoft Academic Graph
INCEFA-PLUS stands for INcreasing safety in nuclear power plants by Covering gaps in Environmental Fatigue Assessment. It is a five year project supported by the European Commission HORIZON2020 program that commenced in mid-2015 and in which sixteen
Publikováno v:
Volume 1A: Codes and Standards.
Environmentally assisted fatigue of nuclear plant materials in the Pressurised Water Reactor (PWR) coolant environment is a phenomenon that has been extensively studied over the past 30 years. Methods for accounting for the PWR environment in an ASME
Publikováno v:
Karabaki, E, Solin, J, Twite, M, Herbst, M, Mann, J & Burke, G 2017, Fatigue With Hold Times Simulating NPP Normal Operation Results for Stainless Steel Grades 304L and 347 . in ASME 2017 Pressure Vessels and Piping Conference : Codes and Standards . vol. 1A, PVP2017-66097, American Society of Mechanical Engineers (ASME), ASME 2017 Pressure Vessels and Piping Conference, PVP 2017, Waikoloa, United States, 16/07/17 . https://doi.org/10.1115/PVP2017-66097
The cyclic behavior and endurance of austenitic stainless steels tested under NPP-relevant laboratory conditions has been studied. It had been earlier shown that long intervals between fatigue transients can affect the fatigue performance in stainles
Autor:
Keith Wright, J. C. Le Roux, Thomas Métais, Marius Twite, Andrew Morley, D. A. Steininger, G. Léopold
Publikováno v:
Volume 3B: Design and Analysis.
Research has led to an update of the various existing fatigue rules to account for aggravating effects, chief among them being the effect of PWR and BWR primary system environments on fatigue life. Among the many documents and evaluation rules publis
Publikováno v:
Volume 1A: Codes and Standards.
The Pressurized Water Reactor (PWR) primary coolant environment is known both to significantly reduce the fatigue life of austenitic stainless steels and to lead to enhanced fatigue crack propagation rates. Relationships for the impact of the PWR coo
Publikováno v:
Volume 1A: Codes and Standards.
Small specimen fatigue testing is challenging in simulated LWR coolant environments at elevated temperatures and pressures. Two approaches to isothermal uniaxial testing in such environments have been developed: use of an autoclave to contain the env
Publikováno v:
Volume 5: High-Pressure Technology; Rudy Scavuzzo Student Paper Symposium and 24th Annual Student Paper Competition; ASME Nondestructive Evaluation, Diagnosis and Prognosis Division (NDPD); Electric Power Research Institute (EPRI) Creep Fatigue Workshop.
The ASME Boiler & Pressure Vessel Code Section III method for the evaluation of fatigue in nuclear plant components uses a fatigue design curve derived from the testing of standard cylindrical specimens to describe the fatigue endurance of austenitic