Fracture toughness behavior of dissimilar metal (SA508 Gr.3 Class 1 and SA312 Type 304LN) weld joint: With and without stress relieving treatment

2021 ◽  
Author(s):  
Suranjit Kumar ◽  
Rahul P. Kale ◽  
Pawan Kumar Singh ◽  
Mainak Ghosh ◽  
Jayanta Chattopadhyay
Keyword(s):  
Fracture Toughness ◽  
Weld Joint ◽  
Dissimilar Metal ◽  
Stress Relieving ◽  
Class 1

Brittle Fracture Analysis of a Dissimilar Metal Weld

2013 ◽  
Author(s):  
A. Blouin ◽  
S. Chapuliot ◽  
S. Marie ◽  
J. M. Bergheau ◽  
C. Niclaeys
Keyword(s):  
Brittle Fracture ◽  
Weld Joint ◽  
Threshold Stress ◽  
Base Alloy ◽  
Welding Process ◽  
Loading Conditions ◽  
Brittle To Ductile Transition ◽  
Dissimilar Metal ◽  
Dissimilar Metal Weld ◽  
Metal Weld

One important part of the integrity demonstration of large ferritic components is based on the demonstration that they could never undergo brittle fracture. Connections between a ferritic component and an austenitic piping (Dissimilar Metal Weld — DMW) have to respect these rules, in particular the Heat Affected Zone (HAZ) created by the welding process and which encounters a brittle-to-ductile transition. Within that frame, the case considered in this article is a Ni base alloy narrow gap weld joint between a ferritic pipe (A533 steel) and an austenitic pipe (316L stainless steel). The aim of the present study is to show that in the same loading conditions, the weld joint is less sensitive to the brittle fracture than the surrounding ferritic part of the component. That is to say that the demonstration should be focused on the ferritic base metal which is the weakest material. The bases of this study rely on a stress-based criterion developed by Chapuliot et al., using a threshold stress (σth) below which the cleavage cannot occur. This threshold stress can be used to define the brittle crack occurrence probability, which means it is possible to determine the highest loading conditions without any brittle fracture risk.

scholarly journals Experimental Characterisation and Numerical Modelling of Residual Stresses in a Nuclear Safe-End Dissimilar Metal Weld Joint

Metals ◽  
2021 ◽  
Vol 11 (8) ◽  
pp. 1298
Author(s):  
Shuyan Zhang ◽  
Zhuozhi Fan ◽  
Jun Li ◽  
Shuwen Wen ◽  
Sanjooram Paddea ◽  
...  
Keyword(s):  
Residual Stress ◽  
Neutron Diffraction ◽  
Residual Stresses ◽  
Weld Joint ◽  
Steel Tube ◽  
Contour Method ◽  
Fe Modelling ◽  
Dissimilar Metal ◽  
Dissimilar Metal Weld ◽  
Metal Weld

In this study, a mock-up of a nuclear safe-end dissimilar metal weld (DMW) joint (SA508-3/316L) was manufactured. The manufacturing process involved cladding and buttering of the ferritic steel tube (SA508-3). It was then subjected to a stress relief heat treatment before being girth welded together with the stainless steel tube (316L). The finished mock-up was subsequently machined to its final dimension. The weld residual stresses were thoroughly characterised using neutron diffraction and the contour method. A detailed finite element (FE) modelling exercise was also carried out for the prediction of the weld residual stresses resulting from the manufacturing processes of the DMW joint. Both the experimental and numerical results showed high levels of tensile residual stresses predominantly in the hoop direction of the weld joint in its final machined condition, tending towards the OD surface. The maximum hoop residual stress determined by the contour method was 500 MPa, which compared very well with the FE prediction of 467.7 Mpa. Along the neutron scan line at the OD subsurface across the weld joint, both the contour method and the FE modelling gave maximum hoop residual stress near the weld fusion line on the 316L side at 388.2 and 453.2 Mpa respectively, whereas the neutron diffraction measured a similar value of 480.6 Mpa in the buttering zone near the SA508-3 side. The results of this research thus demonstrated the reasonable consistency of the three techniques employed in revealing the level and distribution of the residual stresses in the DMW joint for nuclear applications.

Master Curve and J-R Fracture Toughness of SA508/SA533-B-1 Weld and HAZ

2013 ◽  
Author(s):  
Randy K. Nanstad ◽  
Xiang Chen ◽  
Mikhail A. Sokolov ◽  
Barry H. Rabin ◽  
Ying Yang
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Master Curve ◽  
Potential Drop ◽  
Material Behavior ◽  
Normalization Procedure ◽  
Threshold Temperatures ◽  
Class 1 ◽  
Large Heat ◽  
Grade 3

A large heat of low-alloy steel that met both specifications for SA508 Grade 3 Class1 forging steel and SA533 Type B Class 1 plate steel (A508/A533) was procured and used to fabricate a submerged-arc weldment for potential application in high temperature gas-cooled reactors. Compact specimens, 1TC(T), were machined from the weld metal and from the heat-affected-zone (HAZ) of the weldment. Tests of both materials were performed to obtain the fracture toughness reference temperature, To, using the Master Curve procedure of ASTM E-1921, and J-R curves to evaluate material behavior at various threshold temperatures in Code Case N-499-2 (2001) of the ASME Boiler and Pressure Vessel Code. Tests were performed at various temperatures up to 593°C. Unloading compliance was the primary technique used, although dc-potential drop was also monitored during the tests, and the normalization procedure of E1820 was used to compare the results from each procedure. Moreover, many tests at the highest temperatures were performed with no unloading and the normalization procedure provided in E1820 was used to analyze the load-displacement measurements. The fracture toughness for the HAZ is superior to that of the weld metal both in terms of transition temperature and ductile fracture toughness.

Application of Gurson Model to Ni-Based Alloy Weld Joint Pipe With an Axial or a Circumferential Surface Flaw: Phase II

2013 ◽  
Author(s):  
Kiminobu Hojo ◽  
Daigo Watanabe
Keyword(s):  
Phase Ii ◽  
Weld Joint ◽  
Surface Flaw ◽  
Test Results ◽  
Fracture Test ◽  
Gurson Model ◽  
Dissimilar Metal ◽  
Fracture Behaviors ◽  
Parameter Values ◽  
Alloy Weld

The previous paper ASME PVP2012[1] reported application of Gurson model to the fracture test results using pipe models with part-through wall cracks on the dissimilar metal (DM) welds. The predicted maximum loads and the crack behaviors after penetration did not agree well. These results may originate from improper parameter values of Gurson model. This paper revised these parameters and improved the estimated fracture behaviors of the pipe models. A suitable fitting process of Gurson parameters was also proposed.

Thermal Aging Assessment and Microstructural Investigations of Alloy 52 Dissimilar Metal Welds for Nuclear Components

2019 ◽  
Author(s):  
Miguel Yescas ◽  
Pierre Joly ◽  
François Roch
Keyword(s):  
Fracture Toughness ◽  
Alloy Steel ◽  
Thermal Aging ◽  
Base Alloy ◽  
Narrow Gap ◽  
Low Alloy Steel ◽  
Nickel Base Alloy ◽  
Pressurized Water ◽  
Dissimilar Metal ◽  
Nickel Base

Abstract Dissimilar Metal Welds (DMW) are commonly found between the ferritic low alloy steel heavy section components and the austenitic stainless steel piping sections in nuclear power plants. In the EPR™ design which is the latest FRAMATOME Pressurized water reactor (PWR) these DMW involve a narrow gap technology with no buttering, and only one bead per layer of a nickel base alloy weld filler metal (Alloy 52). In order to assess the thermal aging performance of this relatively new narrow gap DMW design, a significant internal R&D program was launched some years ago. Several representative mock-ups were thoroughly characterized in the initial condition as well as in the thermal aged condition, up to 50,000 hours aging at 350°C. The characterisations were focused on the fusion line between the ferritic low alloy steel (LAS) and the nickel base alloy since a particular microstructure is present in this area, especially in the carbon depleted area of the Heat Affected Zone (HAZ) which is often regarded as the weak zone of the weld joint. Metallography, hardness, nanohardness, chemical analyses, and Atom Probe Tomography, as well as fracture toughness tests were carried out on different specimens in different thermal aging conditions. The results show that the fracture toughness behaviour in the ductile-brittle domain of the low alloy steel carbon depleted HAZ at the interface with the alloy 52 weld metal of the DMWs is excellent, even for a thermal ageing equivalent to 60 years at service temperature. This was found in spite of the carbon depleted zone of the HAZ, the variations of hardness, chemical composition, particularly the carbon gradients, and the thermal aging effect induced by phosphorous segregation at grain boundaries.

Fracture Toughness in the Ductile-Brittle Transition and Thermal Ageing Behavior of Decarburized Heat Affected Zone of Alloy 52 Dissimilar Metal Welds of Nuclear Components

2014 ◽  
Author(s):  
Pierre Joly ◽  
Miguel Yescas ◽  
Elisabeth Keim
Keyword(s):  
Fracture Toughness ◽  
Stainless Steel ◽  
Alloy Steel ◽  
Heat Affected Zone ◽  
Power Plants ◽  
Base Alloy ◽  
Thermal Ageing ◽  
Low Alloy Steel ◽  
Dissimilar Metal ◽  
Brittle Transition

Dissimilar metal welds (DMW) are used in nuclear power plants between the nozzles of main components in low alloy steel and stainless steel pipes, or safe-ends connected to the main coolant line pipes. AREVA proposes for EPR™ an improved design of DMW involving narrow gap welding without buttering between the low alloy steel nozzles and the stainless steel safe-ends, and the use of a corrosion resistant weld filler metal (Alloy 52). AREVA performed a thorough characterization of this type of welds, which shows a particular microstructure close to the fusion line between the low alloy steel and the nickel base alloy, where the heat affected zone of the low alloy steel is decarburized. This paper presents results of fracture toughness tests performed with the crack tip located in this area, in the ductile to brittle transition in the as post-welded heat treated condition and after thermal ageing. The results show an excellent fracture toughness behavior of this particular area, compared to that of low alloy steel parent metal.

Fatigue-Strength-Reduction Factors for Welds in Pressure Vessels and Piping

2000 ◽  
Vol 122 (3) ◽  
pp. 297-304 ◽  
Author(s):  
Carl E. Jaske
Keyword(s):  
Fatigue Strength ◽  
Weld Joint ◽  
Pressure Vessels ◽  
Strength Reduction ◽  
Current Status ◽  
Published Data ◽  
Weld Joints ◽  
Class 1 ◽  
Asme Code ◽  
Reduction Factors

Fatigue-strength-reduction factors (FSRFs) are used in the design of pressure vessels and piping subjected to cyclic loading. This paper reviews the background and basis of FSRFs that are used in the ASME Boiler and Pressure Vessel Code, focusing on weld joints in Class 1 nuclear pressure vessels and piping. The ASME Code definition of FSRF is presented. Use of the stress concentration factor (SCF) and stress indices are discussed. The types of welds used in ASME Code construction are reviewed. The effects of joint configuration, welding process, cyclic plasticity, dissimilar metal joints, residual stress, post-weld heat treatment, the nondestructive inspection performed, and metallurgical factors are discussed. The current status of weld FSRFs, including their development and application, are presented. Typical fatigue data for weldments are presented and compared with the ASME Code fatigue curves and used to illustrate the development of FSRF values from experimental information. Finally, a generic procedure for determining FSRFs is proposed and future work is recommended. The five objectives of this study were as follows: 1) to clarify the current procedures for determining values of fatigue-strength-reduction factors (FSRFs); 2) to collect relevant published data on weld-joint FSRFs; 3) to interpret existing data on weld-joint FSRFs; 4) to facilitate the development of a future database of FSRFs for weld joints; and 5) to facilitate the development of a standard procedure for determining the values of FSRFs for weld joints. The main focus is on weld joints in Class 1 nuclear pressure vessels and piping. [S0094-9930(00)02703-7]

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