... Title : Reactor Pressure Vessel Steels ASTM A533B and A508 Cl.2: Crack Opening ... Personal Author(s) : Pelli,R. ; Kemppainen,M. ; Toeroenen,K. ...
DTIC Science & Technology
... 3) special A533-B steel heat for ... PRESSURE VESSELS, THERMAL STABILITY, MARAGING ... SENSITIVITY, EMBRITTLEMENT, STEEL, RESPONSE ...
This report presents the description of the research programme concerning the reactor pressure vessel steels ASTM A533B A533B and A508 Cl.2. The main aim of the programme is to gather and create adequate information needed in ...
National Technical Information Service (NTIS)
... Fatigue Crack Growth in Coarse-Grained, Martensitic A533B Pressure Vessel Steel,. ... one-step-temper-embrittlement', when tempered at 290 ...
... Title : Effect of Thermal Ageing on the ... of PWR Pressure Vessel Steels and Weldments,. ... The materials investigated include A533B plate and A508 ...
Corrosion fatigue crack growth tests have been performed on three pressure vessel steels, A533-B, Ducol W30 and a C-Mn steel, in simulated water reactor environments at ambient temperature and pressure. A533-B ...
In order to predict the stress reduction during stress relief heat treatment in welded joints of the pressure vessel steel A533B, uniaxial stress relaxation as well as creep tests have been performed. The specimens were isothermally stress relaxed between...
In order to predict the stress reduction during stress relief heat treatment in welded joints of the pressure vessel steel A533B, uniaxial stress relaxation as well as creep tests have been performed for base and weld metal. The specimens were isothermall...
Stress corrosion of the pressure vessel steel A-533-B can be induced by accelerated laboratory tests in oxygenous water at high temperature. Cracking has occurred in water with 8 ppm O sub 2 but not in water with less than 10 ppB O sub 2 . High load durin...
The susceptibility of the reactor pressure vessel steels A533B and A508 cl.2 to strain ageing has been studied using conventional tensile and impact testing of prestrained and aged specimens. The results show a modest susceptibility, seen as an increase i...
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 ...
This report describes the heat treatment programmes utilized in connection with a research programme concerning the reactor pressure vessel steels ASTM A533B and A508 Cl.2. (Atomindex citation 10:444562)
This report presents the impact and drop-weight test results of reactor pressure vessel steels ASTM A533B and A508 Cl.2 conducted at the Technical Research Centre of Finland. The ductility of the steels was studied after varying austenitizing and temperin...
Computations have been made of the relaxation of residual stresses in a thick walled tube under conditions corresponding to commercial stress relief heat treatment of the nuclear reactor pressure vessel steel A533B. The distribution of residual stresses w...
The aim with this work was to compare strain data obtained using the film-or camera method and the diffractometer method for residual stress measurement by X-ray diffraction. The measurements were carried out on a weldment of pressure vessel steel A533-B....
This report describes the crack opening displacement (COD) test results for the steels ASTM A533B and A508 C1.2 obtained in connection with a program initiated to gather and create information concerning the manufacturing variables, e.g. heat treatment, n...
Post weld heat treatments of thick-section A533B steel for nuclear pressure vessels are discussed with reference to the ASME code. The discussion is in the form of a lecture and summarized by noting that the ASME code, in particular Section III, Division ...
... Accession Number : AD0180659. Title : Radiation Resistant Weld Metal for Fabricating A533-B Nuclear Reactor Vessels. Corporate Author : ...
irradiated A533B pressure vessel steel weld has been characterized by atom probe field-ion microscopy. Many vessel steel weld metal by Burke et a1.[2s41 In this paper, an atom probe field-ion microscopy OF IRRADIATED ...
E-print Network
The report covers research for the period 1 November 1969-31 January 1970, includes: (1) demonstration of improved radiation embrittlement resistance in commercial scale melt of Type A533-B pressure vessel steel, (2) an outline of the effects of fast reac...
... have been performed on three pressure vessel steels, A533-B, Ducol W30 and a C-Mn steel, in simulated water reactor environments at ambient ...
... Abstract : The effects of long term thermal aging treatments on notched ... commercially produced PWR pressure vessel steels, A533B Class 1 ...
This report describes the crack opening displacement (COD) test results for the steels ASTM A533B and A508 C1.2 obtained in connection with a program initiated to gather and create information concerning the manufacturing variables, e.g. heat treatment, needed in the assessment of the structural integrity of reactor ...
Energy Citations Database
Tensile testing has been performed at three different strain rates (2,4times10 exp -2 , 2,4times10 exp -3 , 2,4times10 exp -4 s exp -1 ) in the temperature range 20-375degreeC on a quenched-and-tempered nuclear-reactor pressure-vessel steel, A533 B. The a...
Three pressure vessel steels of type A508 class 3/A533B class 1, containing varying copper and phosphorus contents, have been examined by a Positron Annihilation technique before and after irradiation to doses of (approximately) 1 x 10 sup 23 n m sup -2 (...
After exposure to low and intermediate fluence neutron irradiation, a variety of reactor pressure vessel (RPV) steels, including A533B-type surveillance specimens, Gundremmigen KRB-A ex-service plate, and test reactor irradiated materials, have been analy...
Stress corrosion cracking (SCC) studies were conducted in two commonly used pressure vessel steels: A533 Grade B Class 1 (A533-B-1) plate and an equivalent A508 Class 2 (A508-2) forging. The purpose of these studies was to determine the response of the ma...
The effects of long term thermal ageing treatments on notched impact fracture properties has been studied in two commercially produced PWR pressure vessel steels, A533B Class 1 and A508 Class 3. Heat treatments of up to 10,000h duration at temperatures be...
This ENEA Data-Base regards mechanical properties, chemical composition and heat treatments of nuclear pressure vessel materials: type A533-B, A302-B, A508 steel plates and forgings, submerged arc welds and HAZ before and after nuclear irradiation. Irradi...
Four unirradiated A533B pressure vessel steels in the form of small three-point-bend (TPB) specimens were precracked for fracture toughness testing. Crack growth rates, obtained from eight specimens of each type of steel, were measured and Paris law const...
Exploratory assessments were made of the Charpy-V notch ductility characteristics of heavy section A533-B and A533-C steel plate and submerged arc weldments following neutron irradiation at 550F. The experimental evaluations were performed largely with co...
Mechanical properties of nine heats of manual arc welded A533B-1 steel are reported. The specific tests are: tensile, Charpy V-notch, and static compact and dynamic one-inch and four-inch fracture toughness results. (GHT)
The effects of thermal ageing of fatigue crack growth behavior have been evaluated in three experimental melts of ASTM A533B MnMoNi steel containing differing amounts of copper and phosphorus. Tests have been conducted under vacuum and in a high purity, l...
An important aspect of the Light Water Reactor-Pressure Vessel-Surveillance Dosimetry Improvement Program is the Hanford Engineering Development Laboratory effort to develop and test trend curve exposure parameter data. Progress in these trend curve-data correlation analysis activities at HEDL is described. The exposure parameters of primary interest are ...
The principal objectives are: (a) to characterize the dynamic fracture toughness of three heats of nuclear pressure vessel steel, namely, 6-in. A302-B plate, 8-in. A533-B Class 1 plate, and 9-in. A508 Class 2 forging; and (b) to develop techniques for the interpretation and analysis of notched, ...
The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced ...
Experience with ASTM A533-B steel and the HSST program are described; ASTM A543 is seen as the material for pressure vessel fabrication. Nuclear applications and problems of stainless steels are also discussed. Various other material selection topics are discussed: RTGR materials, core ...
... improved steel production (current practice). ... VESSELS, NUCLEAR REACTORS, RADIATION ... METALLOGRAPHY FISSION REACTOR MATERIALS. ...
This paper is concerned with the recent study of determining fatigue damage on medium strength A533B nuclear reactor pressure vessel steel by magnetic nondestructive testing methods. Because of the nature of fatigue, which always begins with surface damage, it was found that the Barkhausen ...
A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, ASTM type A 533 Grade B (A533B) having a low USE (USE 1 MeV) by 78 {degree}, 83{degree}, and 70{degree}C for full, half, and ...
DOE Information Bridge
The fission-track etching technique was used to determine the B microdistribution in an ASTM A533B steel weldment. B segregates on gamma grain boundaries in the heat-affected zone; the B content is 12 ppM in the deposit and l ppM in the base metal. The effect of the B segregation on the radiation embrittlement was ...
Post weld heat treatments of thick-section A533B steel for nuclear pressure vessels are discussed with reference to the ASME code. The discussion is in the form of a lecture and summarized by noting that the ASME code, in particular Section III, Division 1, imposes a post weld heat treatment ...
The report describes progress in the following areas: (a) fatigue crack propagation in reactor pressure vessel steels in an air environment, (b) dynamic fracture toughness of 1-in. (25-mm) and precracked Charpy-V bend specimens under impact loading, (c) postirradiation notch ductility and properties recovery in reactor vessel steels, ...
This paper addresses the mechanisms of cleavage fracture in the pressure-vessel steel A533B. Microstructures of single bainite microstructures exhibit a higher propensity for brittle cleavage fracture than do those of auto-tempered martensites. The K{sub 1c} values of mixed microstructures are determined by the ...
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 thermodynamics ...
This report describes progress in a continuing program to characterize material properties performance with respect to structural integrity of light water reactor pressure boundary components. Progress under fracture mechanics highlights J-R curve trends from low upper shelf A533-B weld deposits irradiated under ...
Irradiation embrittlement in nuclear reactor pressure vessel steels results from the formation of a high number density of nanometer sized copper rich precipitates and sub-nanometer defect-solute clusters. We present positron annihilation spectroscopy (PAS) results to characterize the compositions and magnetic character of these defects in model ...
This report describes research progress for Fiscal Year 1979 in a continuing program to characterize material properties performance with respect to structural integrity of light water reactor pressure boundary components. Progress under fracture mechanics investigations includes the first J-R curves from irradiated ...
The operation of some early-generation light water reactors may be limited by the irradiation-induced embrittlement of their reactor vessels. Additional nondestructive methods of measuring the actual embrittlement are desirable to support limits placed on the operation of these vessels. Previous studies have indicated that the increase in microhardness ...
Specimens from 33 heats of three types of nuclear pressure vessel steel were heat treated to produce a simulated weld HAZ microstructure and tested to determine their resistance to stress relief cracking (SRC). The method employed a special jig in which notched specimens were loaded in 3-point bending to a fixed displacement and then heated to 610/sup 0/C ...
Results of a program seeking (i) dynamic analyses of crack arrest in thermally stressed nuclear pressure vessels, (ii) standardization of a laboratory test method for measuring the crack arrest toughness, and (iii) a crack arrest toughness data base for unirradiated and irradiated nuclear steels and weldments are described. Dynamic finite difference ...
After exposure to low and intermediate fluence neutron irradiation, a variety of reactor pressure vessel (RPV) steels, including A533B-type surveillance specimens, Gundremmigen KRB-A ex-service plate, and test reactor irradiated materials, have been analyzed in the atom probe field-ion ...
Stress corrosion cracking (SCC) studies were conducted in two commonly used pressure vessel steels: A533 Grade B Class 1 (A533-B-1) plate and an equivalent A508 Class 2 (A508-2) forging. The purpose of these studies was to determine the response of the materials in simulated ...
The work described in this paper is part of a continuing program to produce well-characterized data for use in the development of models for the prediction of irradiation shift in pressurized water reactor steels. Results are presented for around 50 batches of Charpy specimens, irradiated either in the DIDO or PLUTO heavy water, or in the HERALD light water, research reactors, ...
Sixteen wide-plate crack-arrest tests have been completed, ten utilizing specimens fabricated from A533B class 1 material and six fabricated from a low-upper-shelf base material. Each test utilized a single-edge notched specimen that was subjected to a linear thermal gradient along the plane of crack propagation. Test results exhibit ...
Atom probe field-ion microscopy has been used to characterize the microstructure of a neutron-irradiated A533B pressure vessel steel weld. The atomic spatial resolution of this technique permits a complete structural and chemical description of the ultra-fine features that control the ...
Metals that can sustain plastic deformation homogeneously throughout their bulk tend to be tough and malleable. Often, however, if a metal has been hardened it will no longer deform uniformly. Instead, the deformation occurs in narrow bands on a microscopic scale wherein stresses and strains become concentrated in localized zones. This strain localization degrades the mechanical properties of the ...
The dose dependence of true stress parameters has been investigated for nuclear structural materials: A533B pressure vessel steels, modified 9Cr 1Mo and 9Cr 2WVTa ferritic martensitic steels, 316 and 316LN stainless steels, and Zircaloy-4. After irradiation to significant doses, these alloys ...
NASA Astrophysics Data System (ADS)
Pre and postirradiation fatigue crack growth rates in A302-B, A533-B, and A543 steel plate and in two A543 welds (submerged arc and electroslag) were determined at 550$sup 0$F in air. The fracture mechanics approach was used to analyze the experimental data. Neutron irradiation at 585$sup 0$F to 2.5 x 10$sup 19$ n/cm$sup 2$, ...
Previous work by the authors described a micromechanics fracture model to correct measured J sub c-values for the mechanistic effects of large-scale yielding. This new work extends the model to also include the influence of ductile crack extension prior to cleavage. Ductile crack extensions of 10-15 times the initial crack tip opening displacement at initiation are considered in plane-strain, ...
A comparison between the BCL one-dimensional and two-dimensional dynamic fracture models was made using experimental crack propagation results from a transparant plastic as a basis for the comparison. Both models gave essentially the same results in good agreement with experiment with regard to the important features of the fracture events, e.g. length of crack jump, and the variation in the ...
This investigation employed laboratory melts of pressure vessel steels (A 533-B or A 302-B base) to probe suspect interactions between copper impurities and manganese, molybdenum, chromium and nickel alloying as influencing elevated temperature, radiation sensitivity development. Radiation ...
A533B-1 are used in the construction of coal conversion vessels and were furnished .... heat treatment than to carbon content although more data is needed before any ... 1010 STEEL. 400 TORR. IA-N) b= 0.365 Cm. a = 0.239 Cm. ...
NASA Website
The Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. Recently, it has been shown that, in notched beam testing, shallow cracks tend to exhibit an elevated toughness as a result of a loss of constraint at the ...
... Accession Number : ADD117706. Title : Kinetics of Void Development in Fracturing A533B Tensile Bars,. Corporate Author : ...
... Title : Evaluation of Commercial Production A533-B Steel Plates and Weld Deposits with Extra-Low Copper Content for ... Report Date : OCT 1977. ...
Neutron irradiation is known to cause embrittlement of iron-based materials; in the nuclear industry, this effect can be detrimental for reactor pressure vessel steels. In this paper, we investigate the variations of the magnetic hysteretic behavior due to neutron irradiation, for four materials, i.e. nominally pure Fe, Fe-0.1 wt% Cu and Fe-0.3 wt%Cu model ...
The objective of this work was to investigate the relationship between the micromechanisms of ductile crack growth, the microstructural constituent phases present in nuclear pressure vessel steel, and the observed fracture behavior as determined by impact and fracture mechanics tests. Results from a microstructural and mechanical property comparison of an ...
A study has been made of the effects of prior hydrogen attack damage on fatigue crack propagation behavior in commercial pressure vessel steels. Quenched and tempered Mn-Mo-Ni steel (ASTM A533B Class 2) and normalized and tempered 2.25Cr-1Mo steel (ASTM A387 Class 2 Grade 22) were exposed to ...
Experimental studies demonstrate a significant effect of specimen size, a/W ratio and prior ductile tearing on cleavage fracture toughness values (J{sub c}) measured in the ductile-to-brittle transition region of ferritic materials. In the lower-transition region, cleavage fracture often occurs under conditions of large-scale yielding but without prior ductile crack extension. The increased ...
Steel plates and weld deposits exhibiting relatively low Charpy-V C sub v upper-shelf energy levels in the preirradiation condition are present in certain older reactor pressure vessels. Because the postirradiation notch ductility behavior of such materials is not well understood, a series of investigations has been undertaken for representative ...
Some recent studies of material response have identified an issue that crosses over and blurs the boundary between ASME Boiler and Pressure Vessel Code Section III Subsection NB and Subsection NH. For very long design lives, the effects of creep show up at lower and lower temperature as the design life increases. Although true for the temperature at which ...
... FOR GENERATING AE HAS BEEN USED TO PRODUCE REPEATABLE SIGNALS IN TEMPER AND HYDROGEN EMBRITTLED A533B STEEL. ...
Irradiation embrittlement in nuclear reactor pressure vessel steels results from the hardening by a high number density of nanometer scale features. In steels with more than ?0.10% Cu, the dominant features are often Cu-rich precipitates typically alloyed with Mn, Ni and Si. At low-Cu and low-to-intermediate Ni levels, so-called matrix hardening features ...
The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that ...
The precracked Charpy single-edge notched bend, SE(B), specimen (PCC) is the most likely specimen type to be used for determination of the reference temperature, T0, with reactor pressure vessel (RPV) surveillance specimens. Unfortunately, for many RPV steels, significant differences have been observed between the T0 temperature for the PCC specimen and ...
The 550 deg F irradiation response characteristics of several NiCrMo steel plates, forgings, and weldments were evaluated. Primary objectives were to explore variable radiation-embrittlement tendencies and to assess notch ductility performance relative to that of the MnMoNi steel (ASTM Type A533-B) currently used in nuclear ...
... Descriptors : *UNDERWATER VEHICLES, *PRESSURE VESSELS, PRESSURE VESSELS, HYDROSTATIC PRESSURE, CYLINDRICAL BODIES ...
... Title : Indentation Loading Studies of Acoustic Emission from Temper and Hydrogen Embrittled A533B Steel,. Corporate Author : ...
... temperature. Strain aging is found at 200 C for A508 but at 288 C for A533B, connected with a Portevin-Le Chatelier effect. ...
... Descriptors : (*PRESSURE VESSELS, TEST FACILITIES), (*GLASS TEXTILES, PRESSURE VESSELS), (*SONAR EQUIPMENT, CALIBRATION ...
Eleven wide-plate crack-arrest tests have been completed to date, seven utilizing specimens fabricated from A533B class 1 material (WP-1 series), and four fabricated from a low upper-shelf base material (WP-2 series). With the exception of one test in the WP-1 series and two tests in the WP-2 series which utilized 152-mm-thick ...
... the program is the pressurization until rupture of large pressure vessels having dimensions resembling those of a nuclear reactor pressure vessel. ...
A dual-wall pressure balanced vessel for processing high viscosity slurries at high temperatures and pressures having an outer pressure vessel and an inner vessel with an annular space between the vessels pressurized at a ...
DOEpatents
... Accession Number : AD0097481. Title : THE DESIGN OF PRESSURE VESSELS FOR EXTREME PRESSURES. Corporate ...
... in the paper include: Boiler and Pressure Vessel Code and ... During Fabrication; Visual Inspection of Pressure Vessels; Radiographic Inspection of ...
