Results 191 to 200 of about 3,105,065 (231)

The fatigue behavior of irradiated Reactor Pressure Vessel steel

Engineering Failure Analysis, 2017
Abstract Fatigue specimens of A508-3 steel were irradiated in the swimming-pool test reactor in China Institute of Atomic Energy, the fluence was 3 × 10 19  n/cm 2 at 300 °C, then low-cycle fatigue tests were carried out at ambient temperature, with the fatigue strain range is 0.32–1.8%. The results indicate that, irradiated A508-3 specimens exhibit
W. Zhong, Z. Tong, G. Ning, C. Zhang, H. Lin, W. Yang   +5 more
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Usnrc Research on Reactor Pressure Vessel Steel Embrittlement

1994
Meutron radiation embrittlement of nuclear reactor vessels continues to be a major concern of the USNRC. Research is sponsored for improving understanding of radiation embrittlement effects to predict margins of safety and to improve regulations. Regulatory issues that are affected by neutron embrittlement are examined and related to NRC research ...
A Taboada, M Mayfield
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Radiation Hardening and Embrittlement in a Reactor Pressure Vessel Steel

1969
The relationship between the increase in tensile yield stress (YS) and the increase in the ductile-brittle transition temperature (DBTT) upon irradiation was studied for ASTM A 212-65 Grade B pressure vessel steel. Smooth tension specimens and Charpy V-notch impact specimens were irradiated in the same facility of the Oak Ridge Research Reactor under ...
M. S. Wechsler, R. G. Berggreen, N. E. Hinkle, W. J. Stelzman   +3 more
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Steels for commercial nuclear power reactor pressure vessels

Nuclear Engineering and Design, 1969
Abstract The properties of carbon and low alloy steels currently used in the USA for commercial nuclear power reactor pressure vessel construction are examined in reference to higher strength-higher alloy steels having potential for advanced reactor service.
R.H. Sterne, L.E. Steele
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Analysis of Fracture Surfaces of Reactor Pressure-Vessel Steels

1982
The fracture surfaces of three A533-B pressure-vessel steels have been analyzed using Auger analysis scanning electron microscopy and X-ray microanalysis. The steels included both weld and plate materials with differing radiation sensitivity. The unirradiated specimens were fractured under different conditions to give both ductile and brittle failure ...
OC deHodgins, BL Shriver
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Validation of RTTo for German Reactor Pressure Vessel Steels

Volume 6: Materials and Fabrication, 2005
For the introduction of the new reference temperature RTTo of the ASME Code Cases N-629 and N-631 into the German KTA rules the equation RTTo = T0 + 19,4 °C (35 °F) was validated by the re-evaluation of the existing fracture toughness data base of German RPV steels, including unirradiated and irradiated base material and weld metal data.
Elisabeth Keim, Dieter Siegele, Gerhard Nagel   +2 more
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Evaluation of the reactor pressure vessel steels by positron annihilation

Journal of Nuclear Materials, 2013
Abstract This paper presents a comparison of commercially used German and Russian reactor pressure vessel steels from the positron annihilation spectroscopy (PAS) point of view, having in mind knowledge obtained also from other techniques from the last decades.
V. Slugeň, H. Hein, S. Sojak, J. Simeg Veterníková, M. Petriska, V. Sabelová, M. Pavúk, R. Hinca, M. Stacho   +8 more
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Positron Lifetime Study of Reactor Pressure Vessel Steels

physica status solidi (a), 2000
Positron lifetime (PL) study of several reactor pressure vessel (RPV) steels is presented. VVER-440 type Cr-Mo-V, VVER-1000 type Cr-Mo-Ni-V and A533-B RPV steels were included in the investigation. It was found that positrons annihilate from delocalized state or trapped at dislocations (lifetime 150 ps) in all specimens.
J. Čížek, I. Procházka, J. Kočík, E. Keilová   +3 more
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Modeling of Irradiation Embrittlement of Reactor Pressure Vessel Steels

Journal of Engineering Materials and Technology, 1999
A model to describe the change in the inelastic and fracture properties of reactor pressure vessel steels due to neutron irradiation in the ductile region (i.e., irradiation embrittlement) is developed. First, constitutive equations for unirradiated elastic-viscoplastic-damaged materials are developed within the framework of the irreversible ...
S. Murakami, A. Miyazaki, M. Mizuno
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Microstructural Analysis of Nuclear Reactor Pressure Vessel Steels

1999
One of the most fundamental tasks of nuclear-reactor safety research is assessing the integrity of the reactor’s pressure vessel (RPV). Properties of RPV steels and influence of thermal and neutron treatment on these properties are routinely investigated by macroscopic methods such as Charpy V-notch and tensile tests.
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