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Sample records for steel-astm-a533

  1. Reactor pressure vessel steels ASTM A533B and A508 Cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Kemppainen, M.; Toerroenen, K.

    1979-11-01

    This report presents the tensile test results of steels ASTM A533B and A508 Cl.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of the structural integrity of the reactor pressure vessels. The tensile properties were studied between -196 and 300 degC varying austenitizing and tempering temperatures and having two different carbon contents for the heats of A533B. (author)

  2. Characterization by transmission electron microscopy of a JRQ steel subjected to different heat treatments

    International Nuclear Information System (INIS)

    Moreno G, N.

    2014-01-01

    In this work a study was conducted on the steel Astm A-533, Grade B, Class 1 of reference JRQ, for the purpose of carrying out a study by transmission electron microscopy on the size and distribution of precipitates in steel samples JRQ previously subjected to heat treatments. This because the reactor vessels of the nuclear power plant of Laguna Verde, are made of a steel Astm A-533 Grade B, Class 1. It is known that the neutron radiation causes damage primarily embrittlement in materials that are exposed to it. However, observable damage through mechanical tests result from microstructural defects and atomic, induced by the neutron radiation. In previous studies hardening by precipitation of a JRQ steel (provided by the IAEA) was induced by heat treatments, finding that the conditions of heat treatment that reproduce the hardness and stress mechanical properties of a steel Astm A-533, Grade B, Class 1 irradiated for 8 years to a fluence of 3.5 x 10 17 neutrons/cm 2 and to a temperature of 290 grades C are achieved with annealing treatments at 550 grades C. In the studied samples it was found that the more hardening phase both the heat treatments as the neutron radiation, is the bainite, being the ferrite practically unchanged. Which it gave the tone to believe that the ferrite is the phase that provides at level macro the mechanical properties in stress, since in the irradiated samples such properties remained unchanged with respect to the non-irradiated material, however changes were observed in material ductility, which may be attributable to the change of hardness in the bainite, which opens a possibility for modeling the micromechanical behavior of this material. (Author)

  3. Characterization by transmission electron microscopy of a JRQ steel subjected to different heat treatments; Caracterizacion por microscopia electronica de transmision de un acero JRQ sometido a diferentes tratamientos termicos

    Energy Technology Data Exchange (ETDEWEB)

    Moreno G, N.

    2014-07-01

    In this work a study was conducted on the steel Astm A-533, Grade B, Class 1 of reference JRQ, for the purpose of carrying out a study by transmission electron microscopy on the size and distribution of precipitates in steel samples JRQ previously subjected to heat treatments. This because the reactor vessels of the nuclear power plant of Laguna Verde, are made of a steel Astm A-533 Grade B, Class 1. It is known that the neutron radiation causes damage primarily embrittlement in materials that are exposed to it. However, observable damage through mechanical tests result from microstructural defects and atomic, induced by the neutron radiation. In previous studies hardening by precipitation of a JRQ steel (provided by the IAEA) was induced by heat treatments, finding that the conditions of heat treatment that reproduce the hardness and stress mechanical properties of a steel Astm A-533, Grade B, Class 1 irradiated for 8 years to a fluence of 3.5 x 10 {sup 17} neutrons/cm{sup 2} and to a temperature of 290 grades C are achieved with annealing treatments at 550 grades C. In the studied samples it was found that the more hardening phase both the heat treatments as the neutron radiation, is the bainite, being the ferrite practically unchanged. Which it gave the tone to believe that the ferrite is the phase that provides at level macro the mechanical properties in stress, since in the irradiated samples such properties remained unchanged with respect to the non-irradiated material, however changes were observed in material ductility, which may be attributable to the change of hardness in the bainite, which opens a possibility for modeling the micromechanical behavior of this material. (Author)

  4. Heat treatments in a conventional steel to reproduce the microstructure of a nuclear grade steel

    International Nuclear Information System (INIS)

    Rosalio G, M.

    2014-01-01

    The ferritic steels used in the manufacture of pressurized vessels of Boiling Water Reactors (BWR) suffer degradation in their mechanical properties due to damage caused by the neutron fluxes of high energy bigger to a Mega electron volt (E> 1 MeV) generated in the reactor core. The materials with which the pressurized vessels of nuclear reactors cooled by light water are built correspond to low alloy ferritic steels. The effect of neutron irradiation on these steels is manifested as an increase in hardness, mechanical strength, with the consequent decrease in ductility, fracture toughness and an increase in temperature of ductile-brittle transition. The life of a BWR is 40 years, its design must be considered sufficient margin of safety because pressure forces experienced during operation, maintenance and testing of postulated accident conditions. It is necessary that under these conditions the vessel to behave ductile and likely to propagate a fracture is minimized. The vessels of light water nuclear reactors have a bainite microstructure. Specifically, the reactor vessels of the nuclear power plant of Laguna Verde (Veracruz, Mexico) are made of a steel Astm A-533, Grade B Class 1. At present they are carrying out some welding tests for the construction of a model of a BWR, however, to use nuclear grade steel such as Astm A-533 to carry out some of the welding tests, is very expensive; perform these in a conventional material provides basic information. Although the microstructure present in the conventional material does not correspond exactly to the degree of nuclear material, it can take of reference. Therefore, it is proposed to conduct a pilot study to establish the thermal treatment that reproduces the microstructure of nuclear grade steel, in conventional steel. The resulting properties of the conventional steel samples will be compared to a JRQ steel, that is a steel Astm A-533, Grade B Class 1, provided by IAEA. (Author)

  5. Application of subsize specimens in nuclear plant life extension

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kumar, A.S.; Cannon, N.S.; Hamilton, M.L.

    1993-01-01

    The US Department of Energy is sponsoring a research effort through Sandia National Laboratories and the University of Missouri-Rolla to test a correlation for the upper shelf energy (USE) values obtained from the impact testing of subsize Charpy V-notch specimens to those obtained from the testing of full-size samples. The program involves the impact testing of unirradiated and irradiated full-, half-, and third-size Charpy V-notch specimens. To verify the applicability of the correlation on LWR materials, unirradiated and irradiated full-, half-, and third-size Charpy V-notch specimens of a commercial pressure vessel steel (ASTM A533 Grade B) will be tested. The correlation methodology is based on the partitioning of the USE into crack initiation and crack propagation energies. To accomplish this partition, both precracked and notched-only specimens will be used. Whereas the USE of notched-only specimens is the sum of both crack initiation and crack propagation energies, the USE of precracked specimens reflects only the crack propagation component. The difference in the USE of the two types of specimens represents a measure of the crack initiation energy. Normalizing the values of the crack initiation energy to the fracture volume of the sample produces similar values for the full-, half-, and third-size specimens. In addition, the ratios of the USE and the crack propagation energy are also in agreement for full-, half-, and third-size specimens. These two observations will be used to predict the USE of full-size specimens based on subsize USE data. This paper provides details of the program and presents results obtained from the application of the developed correlation methodology to the impact testing of the unirradiated full-, half-, and third-size A533 Grade B Charpy V-notch specimens

  6. Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

    International Nuclear Information System (INIS)

    Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

    2011-01-01

    Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter