Assessment of Steels for Nuclear Reactor Pressure Vessels

1964 ◽  
Vol 86 (4) ◽  
pp. 393-401 ◽  
Author(s):  
A. Cowan ◽  
R. W. Nichols
Keyword(s):  
Large Body ◽  
Stress Rupture ◽  
Crack Arrest ◽  
Pressure Vessels ◽  
Small Scale ◽  
Specimen Thickness ◽  
Stress Concentrations ◽  
Flat Plates ◽  
Reactor Pressure ◽  
Arrest Temperature

Some of the materials problems associated with the use of mild steels in large gas-cooled reactor pressure vessels are discussed. Tests to failure of 5-ft-dia 0.36 percent carbon-steel vessels with through-thickness longitudinal slots, supported by tests on 7-ft-wide centrally slotted flat plates, have indicated that rapid failure at working-stress levels can only initiate from very long cracks, feet rather than inches in length. Of the mechanisms whereby realistic defects can grow to these sizes, brittle-crack propagation is considered the most important and this can be prevented by the maintenance of a minimum pressurization temperature, based on the crack-arrest temperature. The tests used to assess the crack arrest temperature of plates up to 4 in. thick are described; compared with tests on thinner specimens the thick plate gives arrest temperatures higher by approximately 10 deg C per in. of test-specimen thickness. A comparison is made of crack-arrest temperature and data given by small-scale tests, particularly the Charpy V-notch test. Mechanical limitations of creep deformation in some current designs have been more restrictive on design stress than the values allowed by the existing BS.1500. The test data quoted for stress-rupture and fatigue indicate that these modes of crack extension are not important in current designs. Possible magnitudes and effects of stress concentrations are quoted but, other than a large body of satisfactory service operation, there is little direct evidence of the effect of operating in the creep range on these stress concentrations. The importance of work of this type in justifying higher design stresses and more economic use of material is emphasized.

French RPV PTS Assessment: An Overview of EDF Research and Development in Progress on Thermalhydraulic, Materials and Mechanical Aspects

2003 ◽  
Author(s):  
Dominique Moinereau ◽  
Jean-Michel Frund ◽  
Henriette Churier-Bossennec ◽  
Georges Bezdikian ◽  
Alain Martin
Keyword(s):  
Structural Integrity ◽  
Three Dimensional ◽  
Crack Arrest ◽  
Pressure Vessels ◽  
Short Description ◽  
Constraint Effect ◽  
Materials Properties ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

A significant extensive Research & Development work is conducted by Electricite´ de France (EDF) related to the structural integrity re-assessment of the French 900 and 1300 MWe reactor pressure vessels in order to increase their lifetime. Within the framework of this programme, numerous developments have been implemented or are in progress related to the methodology to assess flaws during a pressurized thermal shock (PTS) event. The paper contains three aspects: a short description of the specific French approach for RPV PTS assessment, a presentation of recent improvements on thermalhydraulic, materials and mechanical aspects, and finally an overview of the present R&D programme on thermalhydraulic, materials and mechanical aspects. Regarding the last aspect on present R&D programme, several projects in progress will be shortly described. This overview includes the redefinition of some significant thermalhydraulic transients based on some new three-dimensional CFD computations (focused at the present time on small break LOCA transient), the assessment of vessel materials properties, and the improvement of the RPV PTS structural integrity assessment including several themes such as warm pre-stress (WPS), crack arrest, constraint effect ....

Probabilistic Fracture Mechanics Evaluations That Consider Nozzles in the Extended Beltline Region of Reactor Pressure Vessels

2015 ◽  
Author(s):  
Gary L. Stevens ◽  
Mark T. Kirk ◽  
Terry Dickson
Keyword(s):  
Fracture Mechanics ◽  
National Laboratory ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Stress Concentrations ◽  
Oak Ridge ◽  
Probabilistic Fracture Mechanics ◽  
Regulatory Issue ◽  
Reactor Pressure

For many years, ASME Section XI committees have discussed the assessment of nozzle penetrations in various flaw evaluations for reactor pressure vessels (RPVs). As summarized in Reference [1], linear elastic fracture mechanics (LEFM) solutions for nozzle penetrations have been in use since the 1970s. In 2013, one of these solutions was adopted into ASME Code, Section XI, Nonmandatory Appendix G (ASME App. G) [2] for use in developing RPV pressure-temperature (P-T) operating limits. That change to ASME App. G was based on compilation of past work [3] and additional evaluations of fracture driving force [4][5]. To establish the P-T limits on RPV operation, consideration should be given to both the RPV shell material with the highest reference temperature as well as regions of the RPV (e.g., nozzles, flange) that contain structural discontinuities. In October 2014, the U.S. Nuclear Regulatory Commission (NRC) highlighted these requirements in Regulatory Issue Summary (RIS) 2014-11 [6]. Probabilistic fracture mechanics (PFM) analyses performed to support pressurized thermal shock (PTS) evaluations using the Fracture Analysis Vessels Oak Ridge (FAVOR) computer code [7] currently evaluate only the RPV beltline shell regions. These evaluations are based on the assumption that the PFM results are controlled by the higher embrittlement characteristic of the shell region rather than the stress concentration characteristic of the nozzle, which does not experience nearly the embrittlement of the shell due to its greater distance from the core. To evaluate this assumption, the NRC and the Oak Ridge National Laboratory (ORNL) performed PFM analyses to quantify the effect of these stress concentrations on the results of the RPV PFM analyses. This paper summarizes the methods and evaluates the results of these analyses.

Predicting crack arrest in reactor pressure vessels

1992 ◽  
Vol 137 (3) ◽  
pp. 305-314 ◽  
Author(s):  
E. Smith ◽  
T.J. Griesbach
Keyword(s):  
Crack Arrest ◽  
Pressure Vessels ◽  
Reactor Pressure

Fast and brittle fracture studies related to steel pressure vessels

1965 ◽  
Vol 285 (1400) ◽  
pp. 104-119 ◽  
Keyword(s):  
Brittle Fracture ◽  
Small Variation ◽  
Crack Arrest ◽  
Pressure Vessels ◽  
Specimen Thickness ◽  
Transition Range ◽  
Fracture Crack ◽  
Shear Lip ◽  
Slit Length ◽  
Fracture Tests

Work aimed at the assessment of failure mechanisms in steel pressure vessels has included bursting tests on 5 ft. diameter cylinders containing longitudinal slits of varying lengths. Catastrophic failure resulted from partially pneumatic pressurization to hoop stress levels below those for general yield with slits of 12 in. overall length or greater. For a given slit length there was only a small variation of failure stress with variation of temperature through the transition range. Tests on different steels indicated that the failure strength is dependent on the yield strength and Charpy energy values at the test temperature. Some implications of this work are discussed. Mechanisms which could lead to the dangerous long defects include fatigue and brittle fracture. Crack-arrest tests on specimens up to 5 in. thick and 7 ft. long have shown that the isothermal temperature can be considerably higher than the gradient ‘shear lip' temperature, and that the isothermal arrest temperature increases with specimen thickness even with material of constant metallurgical quality. Comparisons have been made between these results and those obtained in more conventional brittle fracture tests.

Comparative analysis and verification of engineering methods of shallow cracks effect for fracture toughness prediction for reactor pressure vessels

2020 ◽  
pp. 140-165
Author(s):  
V. I. Kostylev ◽  
B. Z. Margolin
Keyword(s):  
Experimental Data ◽  
Fracture Toughness ◽  
Comparative Analysis ◽  
Nuclear Reactor ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Local Methods ◽  
European Method ◽  
Shallow Crack ◽  
Reactor Pressure

The main features of shallow cracks fracture are considered, and a brief analysis of methods allowing to predict the temperature dependence of the fracture toughness KJC (T) for specimens with shallow cracks is given. These methods include DA-method, (JQ)-method, (J-T)-method, “local methods” with its multiparameter probabilistic approach, GP method uses power approach, and also two engineering methods – RMSC (Russian Method for Shallow Crack) and EMSC (European Method for Shallow Crack). On the basis of 13 sets of experimental data for national and foreign steels, a detailed verification and comparative analysis of these two engineering methods were carried out on the materials of the VVER and PWR nuclear reactor vessels considering the effect of shallow cracks.

scholarly journals Influence of the Heat Input on the Dendritic Solidification Structure and the Mechanical Properties of 2.25Cr-1Mo-0.25V Submerged-Arc Weld Metal

2021 ◽  
Author(s):  
Hannah Schönmaier ◽  
Ronny Krein ◽  
Martin Schmitz-Niederau ◽  
Ronald Schnitzer
Keyword(s):  
Mechanical Properties ◽  
Weld Metal ◽  
Heat Input ◽  
Stress Rupture ◽  
Pressure Vessels ◽  
Rupture Time ◽  
Solidification Structure ◽  
Welding Parameters ◽  
Austenite Grain Size ◽  
Submerged Arc

AbstractThe alloy 2.25Cr-1Mo-0.25V is commonly used for heavy wall pressure vessels in the petrochemical industry, such as hydrogen reactors. As these reactors are operated at elevated temperatures and high pressures, the 2.25Cr-1Mo-0.25V welding consumables require a beneficial combination of strength and toughness as well as enhanced creep properties. The mechanical properties are known to be influenced by several welding parameters. This study deals with the influence of the heat input during submerged-arc welding (SAW) on the solidification structure and mechanical properties of 2.25Cr-1Mo-0.25V multilayer metal. The heat input was found to increase the primary and secondary dendrite spacing as well as the bainitic and prior austenite grain size of the weld metal. Furthermore, it was determined that a higher heat input during SAW causes an increase in the stress rupture time and a decrease in Charpy impact energy. This is assumed to be linked to a lower number of weld layers, and therefore, a decreased amount of fine grained reheated zone if the multilayer weld metal is fabricated with higher heat input. In contrast to the stress rupture time and the toughness, the weld metal’s strength, ductility and macro-hardness remain nearly unaffected by changes of the heat input.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

An Analysis of Filament Overwound Toroidal Pressure Vessels and Optimum Design of Such Structures

2002 ◽  
Vol 124 (2) ◽  
pp. 215-222 ◽  
Author(s):  
Shuguang Li ◽  
John Cook
Keyword(s):  
Optimum Design ◽  
Equivalent Stress ◽  
Mechanical Damage ◽  
Stress Rupture ◽  
Optimization Procedure ◽  
Pressure Vessels ◽  
Thickness Variation ◽  
Von Mises Equivalent Stress ◽  
Metal Liner ◽  
Von Mises

This paper is concerned with the membrane shell analysis of filament overwound toroidal pressure vessels and optimum design of such pressure vessels using the results of the analysis by means of mathematical nonlinear programming. The nature of the coupling between overwind and linear has been considered based on two extreme idealizations. In the first, the overwind is rigidly coupled with the liner, so that the two deform together in the meridional direction as the vessel dilates. In the second, the overwind is free to slide relative to the linear, but the overall elongations of the two around a meridian are identical. Optimized designs with the two idealizations show only minor differences, and it is concluded that either approximation is satisfactory for the purposes of vessel design. Aspects taken into account are the intrinsic overwind thickness variation arising from the winding process and the effects of fiber pre-tension. Pre-tension can be used not only to defer the onset of yielding, but also to achieve a favorable in-plane stress ratio which minimizes the von Mises equivalent stress in the metal liner. Aramid fibers are the most appropriate fibers to be used for the overwind in this type of application. The quantity of fiber required is determined by both its short-term strength and its long-term stress rupture characteristics. An optimization procedure for the design of such vessels, taking all these factors into account, has been established. The stress distributions in the vessels designed in this way have been examined and discussed through the examples. A design which gives due consideration of possible mechanical damage to the surface of the overwind has also been addressed.

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