Strain rate effects in nuclear steels at room and higher temperatures (original) (raw)
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Strain Rate Effects of Nuclear Steels in Room and Higher Temperatures
IASMiRT, 2001
An investigation of strain rate, temperature and size effects of three nuclear steels has been undertaken. The materials are: ferritic steel 20MnMoNi55 (vessel head), austenitic steel X6CrNiNb1810 (Upper Internal Structure), and ferritic steel 26NiCr Mo146 (bolting). Smooth cylindrical tensile specimens of three sizes have been tested at strain rates from 0.001/s up to 300/s, at room and elevated temperatures (400 o C-600 o C). Full stress-strain diagrams have been obtained, and additional parameters have been calculated based on them. The results demonstrate a clear influence of temperature, which amounts into reducing substantially mechanical strengths with respect to R.T. conditions. The effect of strain rate is also shown. It is observed that at R.T. the strain rate effect causes up-shifting of the flow stress curves, whereas at the higher temperatures a mild down-shifting of the flow curves is manifested. Size effect tendencies have also been observed. Some implications when assessing the pressure vessel structural integrity under severe accident conditions are considered.
Size Effects in Deformation and Fracture of a Ferritic Reactor Pressure Vessel Steel
2001
SHEET, Paper Number 1224 Size Effects in Deformation and Fracture of a Ferritic Reactor Pressure Vessel Steel T. Malmberg, K. Krompholz, D. Kalkhof, G. Solomos, E.C. Aifantis 5) (1) Forschungszentrum Karlsruhe GmbH, Institut für Reaktorsicherheit, Postfach 3640, D-76021 Karlsruhe, Germany (2) Paul Scherrer Institut, Labor für Werkstoffverhalten, CH-5232 Villigen PSI, Switzerland (3) European Commission, Joint Research Centre, Institute for Systems, Informatics and Safety, I-21020 Ispra/VA, Italy (4) Aristotle University of Thessaloniki, Laboratory of Mechanics and Materials, GR-54006 Thessaloniki, Greece (5) Michigan Technological University, Center for Mechanics of Materials and Instabilities, Houghton, MI 49931, USA ABSTRACT In support of small scale tests of reactor structures and to extend basic knowledge, screening material tests of different, geometrically similar specimens have been performed to examine the influence of size on the mechanical response. Among other nuclear ste...
Analysis of warm prestressing effect on fracture toughness of reactor pressure vessel steels
Strength of Materials, 2010
Ôîíä ïðèêëàäíûõ èññëåäîâàíèé èì. Çîëòàíà Áàÿ, Èíñòèòóò ëîãèñòèêè è ïðîìûøëåííûõ ñèñòåì, Ìèøêîëüö-Òàïîëüöà, Âåíãðèÿ Âûïîëíåí êîíå÷íîýëåìåíòíûé ðàñ÷åò íàïðÿaeåííî-äåôîðìèðîâàííîãî ñîñòîÿíèÿ îáðàçöîâ òèïà Øàðïè ñ òðåùèíîé ïðè èñïûòàíèÿõ íà òðåõòî÷å÷íûé èçãèá ñ ó÷åòîì âëèÿíèÿ ïðåäâàðèòåëüíîé ãîðÿ÷åé îïðåññîâêè íà âÿçêîñòü ðàçðóøåíèÿ ðåàêòîðíûõ ñòàëåé. Ïðè ýòîì èñïîëüçîâàëèñü äâà ðàçëè÷íûõ çàêîíà óïðî÷íåíèÿ ìàòåðèàëà. Äëÿ êàaeäîãî ñëó÷àÿ ïîëó÷åíû ðàñ÷åòíûå çíà÷åíèÿ J-èíòåãðàëà è ïðîâåäåí èõ ñðàâíèòåëüíûé àíàëèç. Êëþ÷åâûå ñëîâà: ñîñóäû äàâëåíèÿ ðåàêòîðà, ïðåäâàðèòåëüíàÿ ãîðÿ÷àÿ îïðåññîâêà, êîíå÷íûé ýëåìåíò, çàêîí óïðî÷íåíèÿ, J-èíòåãðàë. Introduction. The presented research work is connected to a running project NESC-VII, which is a European cooperative action in support of warm prestressing (WPS) usage in the reactor pressure vessel (RPV) integrity assessment. WPS is a phenomenon which justifies that crack does not propagate during unloading cycle in case of PTS. Bay Zoltan Foundation for Applied Research (BZF) takes part in the modeling work where the WPS effects on J-integral should be determined. In the first stage of the project, a preliminary calculation has been performed. The discussed problem is very actual, since many nuclear power plants are getting closer to the end of their lifetime. The keypoint of lifetime extension projects is to determine which method should be applied, in order to ensure the expected safety service period. The purpose of this project is to assess the WPS effect on RPV steel properties. Scarce data are available on WPS, and it is known that this phenomenon improves fracture toughness of materials. However, the mechanism of this effect is not explained yet, although much research has been dedicated to this issue. Within framework of this project, two different material hardening laws are compared, in order to estimate the one which describes better the mechanical behavior under WPS conditions.
Metallurgical Transactions A, 1980
Critical fracture stress and stress modified fracture strain models are utilized to describe the variation of lower and upper shelf fracture toughness with temperature and strain rate for two alloy steels used in the manufacture of nuclear pressure vessels, namely SA533B-1 (HSST Plate 02) and SA302B (Surveillance correlation heat). Both steels have been well characterized with regard to static and dynamic fracture toughness over a wide range of temperatures (-190 to 200~ although valid JIc measurements at upper shelf temperatures are still somewhat scarce. The present work utilizes simple models for the relevant fracture micromechanisms and local failure criteria to predict these variations in toughness from uniaxial tensile properties. Procedures are discussed for modelling the influence of neutron fluence on toughness in irradiated steel, and predictions are derived for the effect of increasing fluence on the variation of lower shelf fracture toughness with temperature in SA533B-1.
Small specimen predictions of fracture toughness for nuclear pressure vessel steels
Nuclear Engineering and Design, 1980
The use of fracture mechanics in the fracture-safe design and continued safe operation of nuclear reactor pressure vessels has provided an incentive for the development of small specimens for obtaining pertinent fracture toughness data. Small specimens are required for economic reasons when a large number of heats are involved and for space limitation reasons such as in surveillance programs. Several approaches to obtaining fracture toughness from small specimens by either direct measurements or indirect correlations and calculations are reviewed, and their merits and limitations are discussed. Emphasis is placed on techniques which have been developed to determine static and dynamic fracture toughness from surveillance-type specimens. Recently developed techniques for obtaining J-initiation values from a single test specimen and methods for estimating lower and upper shelf fracture toughness from tensile properties are also presented.
Characterization of Reactor Pressure Vessel Steel by ABI Testing
Procedia Materials Science, 2016
Microstructure of the base metal, the multilayer welding seam and the two-layer cladding was characterized through the thickness of the WWER 440 reactor pressure vessel wall. Mechanical properties were determined by performing a series of instrumented indentations across the weld at room temperature. The results were treated by so-called automated ball indentation technique. Mechanical properties obtained by instrumented indentation from the local stress-strain behavior were compared with minimum values required by the standard.
A review of work related to reheat cracking in nuclear reactor pressure vessel steels
International Journal of Pressure Vessels and Piping, 1978
This review was completed in 1976 and describes the then-current knowledge on the problem of reheat cracking in weldments of nuclear pressure vessel steels. The incidence of underclad cracks and structural weld cracks in heat affected zones (HA Z) is described and current theories on the cracking mechanism and controlling factors discussed. Problems concerned with detecting the cracks by destructive and nondestructive methods are outlined. The paper then deals in detail with testing techniques for reheat crack susceptibility and with methods of controlling the problem of underelad and structural weld cracking. Finally, the engineering significance of the cracks in relation to hydrotesting and service is discussed. An appendix gives recommendations for additional work on the subject. ~Study completed in 1976 under Contract 248-76-V-EGB for the Commission of European Communities by the European Research Institute for Welding.