Residual Stress Measurements in a 316L Uniaxial Samples Manufactured by Laser Powder Bed Fusion

Uniaxial samples have been manufactured for tension/compression testing from 316L stainless steel by laser powder bed fusion (LPBF). Samples manufactured by LPBF are known to contain high levels of residual stresses. These uniaxial samples were built from a solid cylindrical rod and subsequently machined to reduce the central cross section of the sample to the required gauge diameter and improve the surface finish. Finite element (FE) models have been developed to simulate the LPBF process of the rods, their removal from the build plate and subsequent machining into the tension/compression samples. High tensile residual stresses were predicted at the surface of the samples, balances by similar magnitude compressive stresses along their axis. Post machining however, these stresses were reduced by around 80% or more. Residual stress measurements were performed on the samples post machining using the neutron diffraction techniques. These measurements confirmed that negligible residual stresses remained in the samples post removal from the build plate and machining.

Creep Damage Assessment of a Low Alloy Pressure Vessel With Dissimilar Metal Welded Joint

Creep damage in low alloy steel plates welded using carbon steel filler can be of concern for petroleum and petrochemical industry due to differential thermal strain and creep behavior. An aggravated scenario can occur when a flaw is present at the weld. The current work considers a catalytic reactor vessel in steady-state operation whose design life is consumed by 80%. It is assumed that after few years of initial operation a low alloy steel patch (similar to the vessel plate) was welded to the vessel using carbon steel filler. A crack like flaw is assumed to be present currently and was sized using non-destructive testing. However, uncertainty remains regarding the time when the crack first appeared, which is often a representative case for a vessel in operation for many years. The objective of the current study is to, by following API 579-1/ASME FFS-1 code procedures, assess the creep damage at the end of the design life considering the postulated scenario as described above. Two cases are considered. For the first case, the currently observed flaw was assumed to be present as is since the beginning of time when the welded patch was made. No flaw growth is assumed for this case from initial time till present. This represents a conservative upper bound case. For the second case, it is assumed that a smaller initial flaw grew to the current size during normal operation. The initial flaw size was determined iteratively such that after growth it matches the currently determined flaw size. For the ease of calculations, the time from initiation of the flaw to the present was discretized into three time intervals during which it is assumed that the flaw size remains constant for the time interval. This removes some of the conservatism inherent in the first case. For both the cases, additional accumulated creep damage is determined considering crack growth. A Finite Element Analysis (FEA) is performed for the reactor vessel patch consisting of a single crack at the weld-patch interface to assess the accumulated creep damage from normal operation till the end of design life. Due to inherent uncertainties in the parameters, sensitivity studies on creep damage due to Adjustment Factors were also performed, based on which Adjustment Factors for Creep Strain Rate (material scatter) and Creep Ductility were chosen appropriately. Steady-state crack growth is considered for creep damage using analytical approach. The current work shows a practical approach combining FEA and analytical calculations to determine accumulated creep damage where a crack appears to have initiated sometime during the past normal operation.

Ultrasonic Nondestructive Evaluation of Stress Corrosion Crack in Welded Steel Plate

Stress corrosion cracking (SCC) has been observed in the high-level nuclear waste tanks that were constructed by welding carbon steel plates. This paper aims to establish an ultrasonic inspection system and its fundamental ability for SCC inspection and quantification on thick welded steel plates. A welded steel plate was fabricated without heat treatment by joining two carbon steel plates through gas metal arc welding (GMAW) procedure. SCC growth, which was initiated with starter cracks across the weld, were observed in a few weeks after submerging the plate in 5 molar (5M) sodium nitrate (NaNO3) solution at about 90 °C. The SCC is inspected with an ultrasonic guided wave system, which employs a piezoelectric transducer for guided wave actuation and a scanning laser Doppler vibrometer (SLDV) for wavefield sensing. The measured wavefield can immediately show wave interactions with the crack. Wavefield images are further generated for the crack length quantification. To demonstrate the crack sizing capability of using the piezoelectric transducer and SLDV, the previous results from the magnetic particle test (MT) are compared. Reasonable agreement in crack length measurement is obtained with the ultrasonic test imaging technique.

Impact of Grooves in Hydraulically Expanded Tube-to-Tubesheet Joints

The stress corrosion cracking of tube-to-tubesheet joints is one of the major faults causing heat exchanger failure. After the expansion process, the stresses are developed in a plastically deformed tube around the tube-to-tubesheet joint. These residual stressed joints, exposed to tube and shell side fluids, are the main crack initiation sites. Adequate contact pressure at the tube-to-tubesheet interface is required to produce a quality joint. Insufficient tube-to-tubesheet contact pressure leads to insufficient joint strength. Therefore, a study on the residual stress and contact pressure that have a great significance on the quality of the tube-to-tubesheet joint is highly demanded. In this research, a 2D axisymmetric numerical analysis is performed to study the effect of the presence of grooves in the tubesheet and the expansion pressure length on the distribution of contact pressure and stress during loading and unloading of 400 MPa expansion pressure. The results show that the maximum contact pressure is independent of the expansion pressure length. However, the presence of grooves significantly increased the maximum contact pressure. It is proven that the presence of grooves in the tubesheet is distinguishable from the maximum contact pressure and residual von mises equivalent stress. The tube pull-out strength increases with the expansion pressure and the number of grooves. In conclusion, the presence of the grooves affects the tube-to-tubesheet joints.

Investigation Into Residual Stresses in a Small Bore Pipe Weld With Stacked Stop/Start Locations

Small bore austenitic stainless steel pipework is used in a number of nuclear plant systems. Many of these locations are subjected to large thermal shocks and therefore have high fatigue usage factors. Their justification therefore often includes a fatigue crack growth and fracture assessment, for which a key input is the residual stress associated with the welding process, in UK assessments these are typically taken from the R6 compendium. A common process used for these welds is manual tungsten inert gas welding, due to access difficulties each pass is usually completed in two halves. The stop-start locations for each weld run are sometimes stacked, especially in horizontal pipe runs where each weld operation starts at the bottom of the pipe and progresses upwards. The stack up of stop-start locations is likely to lead to considerable circumferential variation in weld residual stress, potentially resulting in stresses that locally exceed the R6 profiles. This paper presents results from a series of FE models for a single small bore pipe weld. The simulated weld is a 3-pass manual TIG weld with an EB insert in a 2 inch (50 mm) nominal diameter pipe. Both 2D and 3D models were run. The results of the modelling are then compared with measurements of weld mock-ups of the same weld (both with and without the stop-start stack-up). The results show that, local to the assumed stop location the predicted stresses do exceed even the R6 level 1 profile (a membrane stress equal to the 1% proof stress of the material). However, the locally enhanced stresses drop off quickly away from the peak location, so for defects of a size that may be a concern for a defect tolerance assessment, the R6 Level 1 and 2 profiles remains appropriate or bounding.

Probabilistic Creep Modeling of 304 Stainless Steel Using a Modified Wilshire Creep-Damage Model

Pressure vessel components subject to high temperature and pressure are susceptible to life-limiting creep and/or creep-induced failure. Traditional continuum damage mechanics (CDM) based creep-damage model are used extensively for the prediction and design against creep in these components. Conventional creep experiments show considerable uncertainty in the creep response of materials where scatter can span decades of creep life. The objective of this paper is to introduce the probabilistic methods into a deterministic creep-damage model in order to predict experimental uncertainty. In this study, a modified Wilshire model capable of creep deformation, damage, and rupture prediction is selected. Creep deformation data for 304 stainless steel is collected from the literature consisting of quintuplicate (five) tests at 600°C with varying stress levels. It is hypothesized that the scatter in creep data is due to: test condition (temperature fluctuations and eccentric loading), initial damage (pre-existing surface and sub-surface defects), and metallurgical (local variation in microstructure) uncertainties. Probability distribution functions (pdfs) and Monte Carlo simulations are applied to introduce the uncertainties into the modified Wilshire equations. The domain of each source of uncertainty must be defined. A systematic calibration approach is followed where the material constant for each creep curve (in the quintuple) are obtained and statistical analysis is performed on the material properties to assess the random distribution associated with each uncertain material parameter. The probabilistic calibration begins with the introduction of test condition randomness (±2°C and ±3.2% MPa of nominal temperature/stress) in accordance with the ASTM standards. Cross calibration of temperature-stress variability proceeds the approximation of initial damage uncertainty which captures the remaining scatter in the data. Deterministic calibration unveils the range of variabilities associated with the material properties. The best-fitted pdfs are assigned to each uncertain parameter and subsequently, the deterministic model is converted into a probabilistic model where reliability is a tunable factor. A large number of Monte Carlo simulation are conducted to generate probabilistic creep deformation, minimum-creep-strain-rate (MCSR), and stress-rupture (SR) predictions. It is demonstrated that the probabilistic model produces quantitatively and qualitatively good fits when compared with experimental data. Future work will be directed towards the inclusion of service condition related uncertainty (power plant, turbine blade, Gen IV nuclear reactor application) into the probabilistic framework where the uncertainties are more robust.

Review of Life Assessment and Repair Strategies for Hydrogen Reformer Furnace Outlet Header Castings

Steam methane reforming is the most common method of hydrogen production relevant for plants in the petroleum upgrading, downstream refining, methanol, and ammonia industries. Owner-operators of steam methane reformer furnaces continue to make repair and replacement decisions that involve the cast outlet manifold fittings. One key part of these plans is assessment of the weldability and remaining life of the cast components. The 20Cr-32Ni-1Nb alloy casting materials typically used in the outlet manifolds are usually operated in the low end of their creep temperature range but are subject to metallurgical aging mechanisms which reduce their ductility, weldability, homogeneity, and fracture toughness. This paper covers the practices employed by several owner-users to optimize the lifecycle costs of the outlet manifold castings. These practices include but are not limited to controlled materials specifications, in-situ weldability tests, non-destructive testing in-situ and destructive testing post service, and repair practices such as annealing heat treatments. This paper also includes a limited survey of several owner-users and their fleets of reformer heaters. The details in the survey include the population of affected cast manifold components, alloy grades for the castings and welds, operating temperature ranges, number of startup and shutdown cycles, ranges of time in service, generic design details, and repair case studies. Also discussed are recent improvements in the state of the art for high temperature materials property data-gathering, as well as the structural modeling via Finite Element Methods. These new technologies are opportunities for future work to develop better strategies in the areas of condition assessment, repair planning, and remaining life prediction, taking into account the relevant parameters of installed manifold components, including: specific aging behavior of the casting chemistry, component mechanical design details, as well as the welding and heat treatment parameters during initial fabrication and subsequent maintenance activities.

Investigation of Circumferential Internal Surface-Cracked Pipes and Elbows Tested at Conditions Similar to PWR and an “Apparent Net-Section-Collapse” Prediction Methodology

The Original Net-Section-Collapse (NSC) analysis was developed in the 1970s for prediction of the maximum (failure) moment for a circumferential flaw in a pipe, and is used widely in pipe flaw assessments. A large number of past pipe tests show that deep surface cracks can break through the thickness and result in leaks; hence, the maximum moment of that surface-cracked pipe was below the maximum moment for the circumferential through-wall crack with the same length. In these cases, the applied moment has to be increased for the resulting leak to grow as a through-wall crack. Hence, load-controlled leak-before-break (LBB) fracture behavior has been experimentally observed although it is not predictable by the Original NSC analysis. Recently, Original NSC analysis for circumferential surface-cracked pipes under combined bending and axial tension were enhanced through the development of the “Apparent Net-Section Collapse” methodology to explain inconsistencies with the Original NSC. “Apparent NSC” methodology was developed considering surface-cracked pipe test data developed from external (OD) surface-cracked pipe tests conducted at room temperature (RT) with a vast majority conducted under pure bending and unpressurized conditions. Since it is undesirable to have leakage in many applications, the deficiency in the Original NSC analysis was shown experimentally, and the recently developed “Apparent NSC” methodology applied to a carefully planned matrix of pipe and elbow tests conducted on TP304 stainless steel and Alloy600 materials with different flaw dimensions (composed of short and shallow to long and deep surface cracks), in the range of normalized crack depth, a/t = 0.4 to 0.8 and crack length, 2θψ = 90° to 180°. The tests were conducted under conditions similar to a pressurized water reactor (PWR), and consistent with the International Piping Integrity Research Group (IPIRG-2) [1] test conditions, namely a temperature of 550°F (288°C) and an internal pressure of 2,250 psi. The loads corresponding to the surface-crack initiation, maximum load, and leakage events were recorded from each of the surface-cracked pipe and elbow tests. The data were used to understand the predictable nature of the “Apparent NSC” methodology and to develop an understanding of the fracture behavior of surface-cracked pipes leading to correlation of these results to LBB behavior. Further, the results were correlated between the material composition and the variation of the experimental and predicted bending stress from NSC loads to observations from the previous IPIRG-2 program, where the experimental burst loads were characterized with respect to the flow stress assumptions. The material composition such as variation in sulfur content, and the crack-initiation and crack growth based on elastic-plastic fracture mechanics were used to explain the variability of the flow stress assumption when used in a NSC/limit-load type of analysis. The investigation also showed comparison of predictions based on various flow stress (σf) definitions assumed using yield and ultimate stresses obtained from the tensile tests conducted on the pipe and elbow materials at 550°F (288°C) and applied to the Original NSC and “Apparent NSC” methodologies. The moment predictions using ASME elbow stress indices (B2, C2 used in design) or the IPIRG-2 parameter (Ψec) for the circumferentially surface-cracked elbows were also compared to the experimental maximum moments for the tested elbows.

Study of the Influence of Microstructure and Intergranular Carbides on the Oxidation Behavior of a Nickel Base Alloy 690 TT in Supercritical Water Nuclear Reactor Conditions

Alloy 690, which was designed as a replacement for the Alloy 600, is widely used in the nuclear industry due to its optimum behavior to stress corrosion cracking (SCC) under nuclear reactor operating conditions. Because of this superior resistance, alloy 690 has been proposed as a candidate structural material for the Supercritical Water Reactor (SCWR), which is one of the designs of the next generation of nuclear power plants (Gen IV). In spite of this, striking results were found [1] when alloy 690 was tested without intergranular carbides. These results showed that, contrary to expectations, the crack growth rate is lower in samples without intergranular carbides than in samples with intergranular carbides. Therefore, the role of the carbides in the corrosion behavior of Alloy 690 is not yet well understood. Considering these observations, the aim of this work is to study the effect of intergranular carbides in the oxidation behavior (as a preliminary stage of degenerative processes SCC) of Alloy 690 in supercritical water (SCW) at two temperatures: 400 °C and 500 °C and 25 MPa. Oxide layers of selected specimens were studied by different techniques like Scanning Electron Microscope (SEM) and Auger Electron Spectroscopy (AES).

Fracture Mechanics Assessment of Reactor Pressure Vessel Irradiated Structural Steel for Short Column Type Supports and Neutron Shield Tank

As plants apply for 80 year licensure (subsequent license renewal), the United States Nuclear Regulatory Commission (U.S. NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) structural steel supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of Generic Safety Issue No. 15 (GSI-15) in NUREG-0933 Revision 3 [1], NUREG-1509 [2] (published in May 1996), and NUREG/CR-5320 [3] (published in January 1989) for design life (40 years) and for first license renewal (20 additional years). The conclusions in NUREG-0933, Revision 3 stated that there were no structural integrity concerns for the RPV support structural steels; even if all the supports were totally removed (i.e. broken), the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for 80 year life licensure, the U.S. NRC has requested an evaluation to show structural integrity of the RPV supports by accounting for radiation embrittlement (radiation damage) for continued operation into the second license renewal period (i.e. 80 years). The RPV support designs in light water reactors are grouped into one of five categories or types of supports: (1) skirt; (2) long-column; (3) shield-tank; (4) short column; and (5) suspension. In this paper, two of these RPV support configurations (short column supports and neutron shield tank) will be investigated using fracture mechanics to evaluate the effect of radiation embrittlement of the structural steel supports for long term operations (i.e. 80 years). The technical evaluation of other support configurations will be provided in a separate technical publication at a future date.