ASME Conference Presenter Attendance Policy and Archival Proceedings

2016;():V06BT00A001. doi:10.1115/PVP2016-NS6B.

This online compilation of papers from the ASME 2016 Pressure Vessels and Piping Conference (PVP2016) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Materials and Fabrication: Manufacturing and Mitigation Process Simulation

2016;():V06BT06A001. doi:10.1115/PVP2016-63240.

The modeling of residual stresses induced by drilling remains an issue. Indeed, even if some models are under development, their validation is limited by the absence of universal characterization method of residual stresses inside the hole. This paper aims at presenting a procedure to characterize the residual stress profile induced by drilling. The principle is based on the preparation of a reference sample that has been pre-heat treated in order to remove bulk residual stresses without modifying the microstructure. Then the sample is instrumented with strain gauges before being cut in two parts. This enables on one hand to estimate the elastic recovery after the cut and on the other hand to provide an easy access to the investigated surface. Finally a standard X-Ray diffractometer and an electropolishing technique are combined to estimate residual stresses profiles in two directions (circumferential and axial directions). Some numerical analyses are also performed to estimate stress relaxation during thermal treatment and during sample cutting to give a better interpretation of measurements.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A002. doi:10.1115/PVP2016-63285.

This paper presents a modelling study and analysis performed on a stud welding including thermal, microstructure and stress calculation. The main concern of this work is toward controlling undesirable residual stresses and the evolution of material properties, as well as the chance of estimating cracks especially with regard to future services of structures. Historically, prediction of welding features is being pursued by welding engineers to enable them for optimal design and mitigation of adverse effects. Stud welding is among the welding processes that are not often addressed by means of modelling and associated activities to develop a comprehensive valid prediction. The aim of this research is to present a modelling practice for a stud weld joint to capture the transient thermal profile, consequent evolution of microstructural phase fractions, and stress calculation using a thermomechanical model based on FE methods (SYSWELD package). The material properties are fed into the model as temperature dependent. The microstructure model is based on t8/5 cooling trajectory on CCT diagram that captures transformation from Austenite phase, and the residual stress calculation is compared to experimental measurement for the sake of validation.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A003. doi:10.1115/PVP2016-63364.

The residual stress and distortion found in multi-pass welds are often influenced by the prescribed inter-pass temperature. In ferritic steel components, the severity of this influence is affected by solid-state phase transformation (SSPT) kinetics, which will depend on the overall heat input and cooling conditions. The development of an optimised welding procedure to mitigate weld residual stress (WRS) and distortion in these components can therefore necessitate an extensive test matrix, varying both preheat and inter-pass temperatures as well as the transient weld heat input.

Computational parametric studies provide an opportunity to dramatically reduce the cost and time associated with the development of welding procedure specifications. Welding procedures can be simulated using validated modelling approaches to examine parametric sensitivity and gain insights into optimal conditions for a given welding task. In the present study, a three-pass tungsten inert gas (TIG) groove weld in SA508 Gr.3 Cl.1 ferritic steel is numerically investigated using the ABAQUS finite element code with a user defined subroutine to incorporate the effect of SSPT kinetics. Parametric sensitivity is assessed whereby a representative heat input is applied to simulate weld deposition for each pass, and the inter-pass temperature is varied to examine its effect on WRS and distortion in the weldment. The implications of the overall heat input on cross-weld microstructure are also presented using this approach.

Topics: Temperature , Steel , Stress
Commentary by Dr. Valentin Fuster
2016;():V06BT06A004. doi:10.1115/PVP2016-63517.

In nuclear industry, most of heavy components and reactor coolant lines have a large thickness and their manufacturing processes require multi-pass welding. When low-alloy steel components are concerned, the assembly process is often performed in several stages, such as a cladding, a buttering and a Post Weld Heat Treatment (PWHT) before joining two materials without phase transformations. The distortions induced by the welding operation might be an issue and residual stresses could be significant and play a role on the weldability. For these reasons, AREVA has placed a lot of effort to improve the reliability of numerical simulation of its welding processes, in order to have a better understanding of the involved phenomena and also to predict the residual stress state through the structure [1], [2] and [3] because this numerical simulation can be used to select the manufacturing process in the early phase of welded component design.

The aim of the simulations presented in this paper is the investigation of the final residual state of a nozzle placed at the central position of the vessel head. The computations are performed according a robust methodology packaged in AREVA OSS tool [4] which is based on SYSWELD™ Finite Element solver [5]. Two welding configurations are investigated. The first one is a mock-up with an “open” narrow gap, the groove filling being performed by a manual welding process using electrode coatings. The second one is a mock-up with a “closed” narrow gap, the groove filling being performed by an automatic TIG process. After the comparison of these two configurations, a special investigation is performed on the “open” narrow gap mock-up. The influence of the vessel head buttering before the groove filling is investigated, as well as the efficiency of the PWHT performed after the buttering operation.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A005. doi:10.1115/PVP2016-63524.

The large thickness of most of the heavy components in Pressurized Water Reactors often lead to use multi-pass welding processes. Distortion tolerance and maximal residual stress requirements if any are difficult to fulfill by simply adjusting welding processes with a trial-and-error procedure. This is the main reason why AREVA has developed a robust methodology to perform welding numerical simulations leading to have a better understanding of the phenomena involved during the process.

The present work details a 3D method successfully used to simulate a peripheral adapter J-Groove attachment weld in a vessel head [1] and compares the results to those obtained with a 3D simplified method.

The first method is the state of the art method use for solving Computational Welding Mechanics problems. It is a transient “step by step” 3D simulation using an equivalent moving heat source as input. The second method is a simplified one. First, an equivalent thermal cycle is obtained from a 3D stationary thermal simulation. This thermal cycle is representative of the welding parameters. This thermal cycle is then prescribed to all nodes in a given sector of the weld. This simplified method is called “prescribed thermal cycle” method.

A comparison between displacements and stresses obtained by both methods is completed in order to validate the hypothesis of the simplified approach. The results show a good agreement between the transient “step by step” method and the simplified one. Furthermore, the simplified method speeds up the calculations by a factor of 10. These performances offer the possibility to simulate a multi-pass welding of Pressurized Water Reactor components in a limited calculation time, providing an efficient decision making tool for engineering purpose.

This work is the result of a fruitful collaboration between AREVA and ESI-FRANCE. All the computations are performed with SYSWELD™ software [3].

Commentary by Dr. Valentin Fuster
2016;():V06BT06A006. doi:10.1115/PVP2016-63533.

The present paper details a post-processing method to provide a local update of the material properties after a thorough manufacturing process simulation. This update degrades the hardening behavior model to fit in-service analyses requirements. It allows a drastic reduction of the level of complexity of the numerical model, while keeping the initial state of the component as accurate as necessary for service analyses. It makes the results of advanced best estimate numerical simulation available as direct inputs for regulatory analysis.

Commentary by Dr. Valentin Fuster

Materials and Fabrication: Materials and Technologies for Nuclear Power Plants

2016;():V06BT06A007. doi:10.1115/PVP2016-63045.

Transition welds represent a challenge for the assessment of structural integrity of nuclear plant due to the complexity of the microstructure, properties and local stress state. This paper presents the initial findings of a study aimed at characterising the local microstructure and properties of a transition weld between SA508-4N ferritic steel and SS316LN austenitic stainless steel using a nickel-base filler of Alloy 82. The local microstructures and local composition of the material interfaces are characterised using backscattered electron imaging and Energy-dispersive X-ray spectroscopy. The ferritic steel shows significant grain refinement in the heat affected zone compared to the base metal. This refinement is also observed in the heat affected zone of the austenitic stainless steel although not as significant. Micro-hardness testing has also been incorporated to provide an indication of the influence of local microstructure on flow properties across the weld region. The results indicate a hardness range of between 180–340HV across the weld with the highest value in the heat affected zone of the ferritic steel and the lowest in the austenitic stainless steel. Yield and flow properties derived from flat transweld tensile tests incorporating digital image correlation are related to the micro-hardness results and microstructural characterisation, and an initial assessment of the fracture mechanism performed using fractography.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A008. doi:10.1115/PVP2016-63130.

Structural materials (Alloy N) of molten salt reactor are used to fabricate the reactor vessel, loops, heat exchangers, control rod sleeves, which should have long-term stable properties and enough strength and ductility in multiple extreme environments e.g. high temperature, high pressure, corrosion and neutron irradiation.

The choice and evaluation of materials for different reactor components depend on the specific working conditions of the components. For thorium molten salts reactor-solid fueled one (TMSR-SF1), the reactor vessel is filled with argon gas and the outside parts are exposed to the air. It contains a primary loop of FLiBe coolant and a secondary loop of FLiNaK coolant. The outlet temperature is 700 °C, and the core operating pressure is less than 0.5MPa.The calculated total neutron fluence after 20 years’ service is 2.01×1017n/cm2 for the reactor vessel. Based on these working conditions, the different properties of structural materials are considered and evaluated.

The most important issues of TMSR-SF1 materials include the supply of materials (their preparation, processing, welding, for example) and their qualification in ASME Code. We review alloy N for TMSR-SF1 including alloy physical properties, creep and rupture life time, thermal stability, corrosion, irradiation performance, as well as reliability evaluation of the weldments and so on.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A009. doi:10.1115/PVP2016-63162.

The goal of the U.S. Department of Energy (DOE) Accident Tolerant Fuel Program (ATF) for light water reactors (LWR) is to identify alternative fuel system technologies to further enhance the safety of commercial nuclear power plants. An ATF fuel system would endure loss of cooling in the reactor for a considerably longer period of time than the current systems. The General Electric (GE) and Oak Ridge National Laboratory (ORNL) ATF design concept utilizes an iron-chromium-aluminum (FeCrAl) alloy material as fuel rod cladding in combination with uranium dioxide (UO2) fuel pellets currently in use, resulting in a fuel assembly that leverages the performance of existing/current LWR fuel assembly designs and infrastructure with improved accident tolerance. Significant testing was performed in the last three years to characterize FeCrAl alloys for cladding applications, both under normal operation conditions of the reactor and under accident conditions. This article is a state of the art description of the concept.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A010. doi:10.1115/PVP2016-63164.

There is a worldwide effort to develop nuclear fuels that are resistant to accidents such as loss of coolant in the reactor and the storage pools. In the United States, the Department of Energy is teaming with fuel vendors to develop accident tolerant fuels (ATF), which will resist the lack of cooling for longer periods of times than the current zirconium alloy - uranium dioxide system. General Electric (GE) and its partners is proposing to replace zirconium alloys cladding with an Iron-Chromium-Aluminum (FeCrAl) alloy such as APMT, since they are highly resistant to attack by steam up to the melting point of the alloy. FeCrAl alloys do not react with hydrogen to form stable hydrides as zirconium alloys do. Therefore, it is possible that more tritium may be released to the coolant with the use of FeCrAl cladding. This work discusses the formation of an alumina layer on the surface of APMT cladding as an effective barrier for tritium permeation from the fuel to the coolant across the cladding wall.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A011. doi:10.1115/PVP2016-63250.

GH3535 supperalloy, whose grade of ASME is UNS N10003, is currently considered as a candidate material for solid-fuel and fluid-fuel molten salt reactor in china. During the development of procedures for welding GH3535 superalloy, consideration should always be given to the possibility that repair welding may be necessary. This paper presents weld repairs of GH3535 alloy rolled plates using gas tungsten arc welding with filler metal. The purpose of this work is to evaluate the low heat input process for weld repair of GH3535 alloy plates about the microstructure features and mechanical properties. The results demonstrated that sound joints without defects could be obtained after weld repairs. Due to repair thermal cycles on the original weld seam, the size of carbide precipitate became large, but repair welding is found to cause no decrease in short-term time-independent strength.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A012. doi:10.1115/PVP2016-63307.

The power plant materials engineers are still tweaking the specification of grade 91 in order to prevent the improper production and construction. In this sense the chemistry, heat treatment conditions and design/fabrication rules for this steel should be highly optimized to achieve the safe elevated temperature properties and to avoid the abuse during the construction. This paper presents the recent findings on creep degradation behavior in grade 91 followed by the review and discussion about the effect of chemical composition and constituent on the creep strength/ductility in correlation to the chemical specification for this steel. Additionally the design factor for creep and creep-fatigue of grade 91 steels was considered to propose a new rule on the stand point of the changes in creep degradation mechanism at long-term region. As the results it was found that the martensitic lath structure in the grade 91 exposed at the creep loading significantly reduces the dislocation density to form the ferritic subgrain structure at long-term exposure which exhibits different creep mechanism from that of original martensitic structure. So that the multiplier of 0.67 applied to average stress for rupture in 100.000 h may not be used even below the temperature of 815°C. Therefore the F-factor is recommended to be applied at the temperature above 615°C for base metal and 545°C for welds considering the magnitude of the creep strength scatter band in the existing chemical specification ranges.

Topics: Creep , Steel , Design
Commentary by Dr. Valentin Fuster
2016;():V06BT06A013. doi:10.1115/PVP2016-63316.

Creep rupture data of welded joints of ASME Grade 91 type steel have been collected from Japanese plants, milling companies and institutes, and the long-term creep rupture strength of the material has been evaluated. This evaluation of welded joints of Grade 91 steel is the third one in Japan as similar studies were conducted in 2004 and 2010. The re-evaluation of the creep rupture strength was conducted with emphasis on the long-term creep rupture data obtained since the previous study, with durations of the new data of up to about 60000h. The new long-term data exhibited lower creep strength than that obtained from the master creep life equation for welded joints of Grade 91 steel determined in 2010, then the master creep life equation was again reviewed on the basis of the new data using the same regression method as that used in 2010. Furthermore, the weld strength reduction factors obtained from 100000h creep strength of welded joints and the base metals are given as a function of temperature, where the master creep equations of the base metals are also redetermined in this study.

Topics: Creep , Steel , Welded joints
Commentary by Dr. Valentin Fuster
2016;():V06BT06A014. doi:10.1115/PVP2016-63355.

Creep rupture strength of ASME Grades 91, 92, 122 and 23 type steels were evaluated by the SHC committee in 2004 and 2005, and the Assessment Committee on Creep Data of High Chromium Steels in 2010. According to the evaluation of creep rupture strength, allowable stress of the steels was revised and weld strength reduction factor (WSRF) was established. In 2015, the creep rupture data of those steels was collected from materials producers, power plant manufacturers and institutes in Japan and a review of long-term creep rupture strength of the steels was conducted by the Assessment Committee on Creep Data of High Chromium Steels in reference to the previous evaluation. It has been confirmed with the latest dataset that re-evaluation of long-term creep rupture strength is not required for Grades 92, 122 and 23 type steels. On the other hand, lower creep rupture strength compared with the previous evaluation was recognized on the new creep rupture data of Grade 91 steels, therefore, re-evaluation of creep rupture strength was conducted on Grade 91 steels. Creep rupture strength was assessed by means of region splitting analysis method in consideration of 50% of 0.2% offset yield strength, in the same way as the previous study. According to the evaluation of long-term creep strength of the steels, allowable tensile stress was reviewed and proposed revision was concluded.

Topics: Creep , Steel , Base metals
Commentary by Dr. Valentin Fuster
2016;():V06BT06A015. doi:10.1115/PVP2016-63356.

Creep deformation property of Grade 91 steels was analyzed on more than 370 creep curves over a wide range of time to rupture from about 10 hours to beyond 100,000 hours, in order to evaluate time to 1% total strain, time to minimum creep rate and time to initiation of tertiary creep. Time to initiation of tertiary creep was assessed as a 0.2% offset with a slope of minimum creep rate. It is difficult to determine time to minimum creep rate precisely, which is a basis of 0.2% offset, however, it has been confirmed that time to initiation of tertiary creep is not sensitive to the time when the creep rate indicates minimum value. Life ratio of 1% total strain time against creep rupture time increases up to about 60% with increase of temperature and decrease of stress. Life ratio of time to initiation of tertiary creep also tends to increase with decrease in stress. However, change of it is in a range of 50 to 60% of creep rupture life over a wide range of creep rupture life from 10 hours to 100,000 hours, and it is not sensitive to creep test temperature. Over a range of temperatures from 500 to 600°C and up to about 200,000 hours, a temperature and time-dependent stress intensity limit, St is controlled by 67% of minimum stress to rupture. However, a difference between 67% of minimum stress to rupture and 80% of minimum stress to initiation of tertiary creep decreases with increases in temperature and time, and both values approach each other in the long-term beyond about 100,000 hours at 600°C. In the long-term beyond about 10,000 hours at 650°C, St is controlled by 80% of minimum stress to initiation of tertiary. The stable life fraction of time to initiation of tertiary creep establish a reliability of a temperature and time-dependent stress intensity limit value.

Topics: Creep , Steel
Commentary by Dr. Valentin Fuster
2016;():V06BT06A016. doi:10.1115/PVP2016-63446.

The mechanical properties and irradiation embrittlement behavior of SA508 Gr.4N low alloy steel have been characterized systematically using SA508 Gr.4N model alloys. For an evaluation of neutron irradiation embrittlement behavior of model alloy, several irradiation tests were carried out at the research reactors, HANARO and HBWR, up to a fluence level of 1.5 × 1020n/cm2 (E>1MeV) at 290 ± 10°C. The master curve method according to ASTM E1921 was adopted to evaluate the fracture toughness in the transition region. Ni and Cr additions resulted in increasing the martensite fraction in low alloy steel by enhancing the hardenability of the steel. Thus, the predominant microstructure of SA508 Gr.4N model alloy is a mixture of tempered martensite and bainite, while SA508 Gr.3 steel shows a typical tempered upper bainitic structure. SA508 Gr. 4N model alloy shows excellent strength and transition behavior compared to commercial SA508 Gr.3 steel. After neutron irradiation, the yield strength and tensile strength of model alloy were increased with an increase in the neutron fluence level. The transition temperature shifts of SA508 Gr.4N model alloy obtained by both Charpy impact and fracture toughness tests were not significantly larger than those of commercial SA508 Gr.3 low alloy steel. It seems that the increased Ni content in the SA508 Gr.4N model alloy did not show significant effects on the irradiation embrittlement behavior owing to the controlled low Mn content. In addition, good fracture toughness of the SA508 Gr.4N model alloy was maintained even after neutron irradiation up to a level of ∼1020n/cm2.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A017. doi:10.1115/PVP2016-63637.

Ferritic steel 2 ¼ Cr is a candidate material for future pressure component in nuclear fields. In order to validate this choice, it is necessary, firstly to verify that it is able to withstand the planned environmental and operating conditions, and secondly to check if it is covered by the existing design codes, concerning its procurement, fabrication, welding, examination methods and mechanical design rules. A large R&D program on 2 ¼ Cr steel has been undertaken at CEA and Areva in order to characterize the behavior of this material and of its welded junctions.

In this frame, a new measurement system for tensile testing was developed in the LISN laboratory of the CEA (French atomic commission), in order to characterize the local behavior of the material during a whole tensile testing. Indeed, with the conventional measurement system (typically an extensometer), the local behavior of the material can only be determinate during the stable step of the testing. So, usually the behavior of the material during the necking step of the step is unknown.

This new measurement is based on the use of some laser micrometers which allow measuring the minimum diameter of the specimen and the curvature radius during the necking phase with a great precision. Thanks to the Bridgman formula, we can evaluate the local behavior of the material until the failure of the specimen.

This new system was used to characterize the tensile propriety of a bimetallic welded junction of 2 ¼ Cr steel and austenitic stainless steel 316L(N) realized with inconel filler metal.

These works lead to propose a tensile curve for each materials of the welded junction at room temperature and the effect of postweld heat treatment.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A018. doi:10.1115/PVP2016-63666.

Irradiation embrittlement is a limiting condition for the long-term safety operation of a nuclear Reactor Pressure Vessel (RPV). When a Boiling Water Reactor (BWR) is approaching its initial licensing, in order to operate the reactor for another 20 years and more, it should be demonstrated that the irradiation embrittlement of the reactor vessel materials will be adequately managed by ensuring that the fracture toughness properties are above a certain level of the required safety margin. In this work the Charpy specimens recovered from two surveillance capsules of two BWRs (fluence 3.58×1017 – 9.03×1017 n/cm2) were impact tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials. The measured transition temperature shifts (ΔRTNDT) and the Upper Shelf Energy (USE) for the plate and weld materials were compared to the predictions calculated according to Regulatory Guide 1.99 Rev.2. The credibility of surveillance data were analyzed according with the five criteria established in the Regulatory Guide 1.99, Revision 2. The Master Curve (MC) approach and the instrumented impact tests using pre-cracked Charpy specimens were implemented in order to fully validate this techniques that can be used for embrittlement monitoring during life extension periods.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A019. doi:10.1115/PVP2016-63686.

Current and future nuclear technologies such as fission and fusion reactor systems depend on well characterized structural materials, underpinned by reliable material models. ANSTO is pursuing research into the nuclear power cycle on many fronts, including: modelling and measurement of weld residual stresses; simulation of radiation damage by molecular dynamics modelling; assessment of radiation, cycling and aging effects in power plant structural materials; and characterization of materials for the next generation of nuclear power plants. Several examples of past and current research activities are used to highlight the potential of ANSTO facilities, techniques and capabilities available for collaborative research in the nuclear space.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A020. doi:10.1115/PVP2016-63781.

The commonly accepted approach to dealing with material damage as the cause of structural failure is to treat the most highly distressed location in the structure as an equivalent simple test and to define failure of the structure as a whole as being failure at that point location.

The exception to this rule is plastic deformation. Yielding at a point was recognized several decades ago as being an excessively conservative definition of component failure and it is now standard design practice to accept failure as being the limit load, which is only reached, sometimes after extensive propagation of a plastic zone.

Other material failure mechanisms also occur after a finite period of damage propagation, but this additional strength, or life, is not usually taken into account, partly because the damage mechanisms themselves are not always well defined, and partly because of the computational difficulty involved in assessing the propagation of damage.

Creep rupture falls into the category of a mechanism which can enjoy an extensive period of damage propagation before structural failure occurs, but the difficulty of evaluating it quantitatively has meant that it continues to be dealt with as essentially a point failure phenomenon.

Relatively recently, many of the problems associated with assessing creep damage have been resolved, on the material side by increased use of so-called “continuum damage mechanics” based models such as Kachanov and Omega and, on the computational side, by the exponential growth in the capabilities of advanced Finite Element Analysis. It is now possible in principle to trace the entire life of a complex component, down to final disintegration. However, this capability still comes at a significant cost, and there is still room for simplification in order to bring this capability to a wider range of potential users.

This paper describes a process for evaluating the propagation of creep damage, down to the point of total disintegration, using approximations which exist within the standard capabilities of a typical FE design package. This innovation does not do anything that cannot be done today using the full repertoire of computational tools that exist, notably user subroutines, but provides a simpler platform which can be used to push damage evaluation further into the activities of day-to-day design with a significant reduction in the resource allocation currently required to do the job.

Results are compared with creep experiments on notched bars.

Topics: Creep , Computation , Damage
Commentary by Dr. Valentin Fuster
2016;():V06BT06A021. doi:10.1115/PVP2016-63846.

Ductility-dip cracking (DDC) in high-chromium nickel-base weld metals has been an issue during fabrication and repair of nuclear power plant components for many years. DDC is a solid-state cracking phenomenon and several theories have been proposed for the mechanism. The research conducted to develop DDC theories has primarily been performed using test methods involving small-scale specimens that may not replicate all the welding conditions and factors that cause DDC. Due to the complexities of welding, there are potentially significant differences in the applied strain, strain rates, stresses, and thermal cycles that occur with small scale test methods and actual multi-pass welding conditions. EPRI is working to devise a method to predict DDC susceptibility in multi-pass high-chromium nickel-base welds and to develop procedures and techniques that minimize the occurrence of DDC — a key issue in the nuclear welding industry that has yet to be fully resolved. The primary aim is to design a weld mockup that replicates strain, strain-rates, stresses, and thermal cycles that occur in multi-pass field welds and which produces DDC in predicted weld regions. If successful, the data from this work will be used to assist in development of a simplified field deployable test which can effectively screen for DDC susceptibility. For the first phase of this work multi-pass narrow groove mockups using GTAW and filler metals 52 and 52M were made with precise heat input and bead placement controls to isolate the occurrence of DDC to a known region in the weld deposit. To assist with understanding the correlation between strain accumulation and the occurrence of DDC, computer modeling using SysWeld™, with validated weld parameter inputs, was used to simulate the narrow groove mockup weld. Comparison of DDC occurrence to the model results suggests that multiple reheat cycles in the ductility-dip temperature range accumulate plastic strains during both the on-heating and on-cooling portions of the reheat cycles, resulting in a strain ratcheting effect. It is postulated that this strain ratcheting exhausts the strength of the grain boundaries within the excessively reheated weld regions, which, when combined with the shear stress induced at the grain boundaries during the ratcheting events, promotes DDC.

Commentary by Dr. Valentin Fuster

Materials and Fabrication: Materials for Hydrogen Service (Joint With CS-8)

2016;():V06BT06A022. doi:10.1115/PVP2016-63179.

In recent years, hydrogen has been increasingly used as a clean energy source. Accordingly, the manufacturing and safety detection of hydrogen storage and transportation equipment is also becoming more and more important. To evaluate the hydrogen compatibility and suitability of 4130X seamless steel tube, CT and WOL specimen are used to evaluate its resistance of the hydrogen induced cracking by referring ISO 11114-4: 2005. The specimens were placed at 90 °C in 90 MPa high-pressure hydrogen environment for 1000 h after loading. The CT test results show that the 4130X materials have a good resistance to hydrogen-induced cracking under the standard required loading condition. The WOL test show that no crack occurs when the stress intensity factor KI of 4130X-A, 4130X-B is about 100–120 MPa•m1/2. The 4130X-C material exhibit the hydrogen-induced crack in the high-pressure hydrogen environment, the critical stress intensity factor KIH is about 87.81 MPa•m1/2. The study proves that the 4130X seamless steel tube has a good resistance to hydrogen embrittlement.

Topics: Steel , Hydrogen
Commentary by Dr. Valentin Fuster
2016;():V06BT06A023. doi:10.1115/PVP2016-63198.

Although employing high strength steels in pipelines provides many benefits, it is difficult to satisfy all required mechanical properties simultaneously because some are potentially at odds with each other. Additionally, when new natural gas pipelines are constructed for severe sour service, the hardness must be below 248 Vickers to avoid sulfide stress cracking (SSC) regardless of pipe grades, and this has been standardized by NACE and applied for approximately five decades. On the other hand, the relevance of this hardness criterion has been controversial. This paper proposes three possible methods to improve SSC resistance for weld metals; 1) reducing impurities, 2) producing fine and homogeneous microstructure, 3) controlling microstructures that characterize high hydrogen permeability, solubility, and low diffusivity. This paper states that reducing impurities and producing fine and homogeneous microstructure would reduce SSC susceptibility and an acicular ferrite would be the effective microstructure to increase SSC resistance for weld metals.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A024. doi:10.1115/PVP2016-63213.

In 2012, the Japanese regulation for selecting SUS316 austenitic stainless steel with a specific Ni equivalent (SUS316 and SUS316L can be used in the temperature ranges between −45 and 250 °C for a Ni equivalent of ≧28.5%, between −10 and 250 °C for a Ni equivalent of ≧ 27.4%, and between 20 and 250 °C for a Ni equivalent of ≧ 26.3%) as an appropriate material available in hydrogen refueling stations (HRSs) that provide 70 MPa fueling to fuel cell vehicles (FCVs) was updated with the support of NEDO (New Energy and Industrial Technology Development Organization) Program Phase 1 [1].

Topics: Alloy steel , Hydrogen
Commentary by Dr. Valentin Fuster
2016;():V06BT06A025. doi:10.1115/PVP2016-63336.

Evaluations of mechanical properties under high pressure hydrogen gas environments are difficult because special testing equipment which can obtain data under high pressure hydrogen gas is necessary and the cost of installing those machines is very high. Therefore, a method for predicting mechanical properties under high pressure hydrogen gas without special equipment was investigated in this study. A JIS SCM435 steel, of which the typical chemical composition is Fe-0.35%C-0.2%Si-0.75%Mn-1.1%Cr-0.2%Mo, was used. Material with the size of 22mm in square and 200mm in length was heat-treated by quenching in oil and tempering, followed by water cooling. The material microstructure was tempered martensite, and the tensile strength of the steel was about 1000MPa. The fatigue property was evaluated in cathodic hydrogen charging service and high pressure hydrogen gas service. The fatigue property obtained by the cathodic charging method was almost the same as the property under the high pressure hydrogen gas condition. The reason is conjectured to be that the hydrogen fugacity on the specimen surface and the surface state of the specimen under those test services was substantially the same.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A026. doi:10.1115/PVP2016-63338.

Hydrogen generally results in a degradation of fatigue crack growth resistance in metals depending on the nature of the hydrogen interaction with crystalline defects developed during a cyclic loading. In this paper we present the first results regarding the characterization of the microstructure and the mechanical properties of Armco Iron with a ferritic microstructure, and the investigation of embrittlement under gaseous hydrogen environment. For this purpose, fatigue crack propagation tests were performed on CT40 specimens under high pressure gaseous hydrogen using Hycomat test bench, at the Pprime Institute in Poitiers.

The fatigue crack growth rate data obtained so far at the 35 MPa of hydrogen pressure and at loading frequency of 20 Hz indicate a sharp increase in crack growth rates in a narrow range of stress intensity factor amplitudes. Also, it was observed that by decreasing the loading frequency to 2 or 0.2 Hz at the above mentioned hydrogen pressure no significant change in this transition regime happens. Scanning electron microscope observations of the fracture surfaces are used to support the explanations proposed for the hydrogen inducted intergranular failure in this material.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A027. doi:10.1115/PVP2016-63375.

Tension-compression fatigue tests using smooth specimens of low carbon steel JIS-SM490B were carried out in air and hydrogen gas environment under the pressure of 0.7 and 115 MPa at room temperature. In 0.7 MPa hydrogen gas, fatigue life curve was nearly equivalent to that in air. On the other hand, in 115 MPa hydrogen gas, fatigue life was significantly degraded in the relatively short fatigue life regime (e.g. Nf < 105). To clarify the effect of hydrogen environment on fracture process, fracture surfaces of these specimens were observed. In general, fatigue fracture process of steels with low or moderate strength is macroscopically divided into 3 stages. In the first stage (stage I), fatigue cracks initiate in some crystalline grains. In the second stage (stage II), the cracks propagate stably on a cycle-by-cycle basis. In the final stage (stage III), a tilted fracture surface, shear-lip, is formed by ductile tearing. In SM490B steel, this general fracture process was confirmed in air and 0.7 MPa hydrogen gas. In contrast, in 115 MPa hydrogen gas, there was no tilted portion in the stage III region, and the fracture surface was totally flat. Observation with scanning electron microscope revealed that dimples were formed by ductile tearing in the tilted fracture region in air and 0.7 MPa hydrogen gas. On the other hand, only a quasi-cleavage fracture surface existed in the final fracture region in 115 MPa hydrogen gas. To understand the cause of this peculiar fracture morphology, we conducted elasto-plastic fracture toughness tests in each environment, and investigated the fracture morphology.

As a result of fracture toughness tests, crack growth rate in air and 0.7 MPa hydrogen gas was approximately equal to each other, and both the fracture surfaces were covered by dimples. This fracture morphology was in accordance with that of stage III morphology in fatigue specimen tested in air and 0.7 MPa hydrogen gas. However, in 115 MPa hydrogen gas, the crack growth was significantly accelerated, and the whole fracture surface was covered by quasi-cleavage.

In this paper, firstly, the similarity of fracture surface between two test methods, i.e. fatigue test and fracture toughness test, is investigated. And then, the formation mechanism of the flat fracture surface is discussed by paying attention to the crack-growth acceleration in high-pressure hydrogen gas.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A028. doi:10.1115/PVP2016-63390.

Understanding of hydrogen effect on local mechanical properties of metals is important for understanding hydrogen embrittlement mechanisms. The effect of thermal gaseous hydrogen precharging on the nanomechanics of SUS310S and SUS304 austenitic stainless steels has been investigated using a combination of nanoindentation and atomic force microscopy (AFM). It is observed that hydrogen precharging decreases the first excursion load in load versus displacement curves and enhances the slip steps around indentations for both the materials, which experimentally support the hydrogen-enhanced localized plasticity (HELP) mechanism. The nanohardness in SUS310S stable austenitic stainless steel is increased by hydrogen precharging while that in SUS304 metastable austenitic stainless steel is decreased by hydrogen precharging. The hydrogen-induced hardening in SUS310S and softening in SUS304 are discussed in terms of the hydrogen/deformation interaction and the effect of hydrogen on strain-induced martensite transformation.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A029. doi:10.1115/PVP2016-63394.

Several criteria based on reduction in area (RA) or relative RA (RRA) are proposed for determining the hydrogen compatibility of austenitic stainless steels; however, the mechanism of hydrogen-induced degradation in RA and RRA is not necessarily clear. The degradation in the RA and RRA of the austenitic stainless steels is attributed to hydrogen-assisted surface crack growth (HASCG) accompanied by quasi-cleavages; therefore, a mechanism of the HASCG should be elucidated to establish novel criteria for authorizing various austenitic stainless steels for use in high-pressure gaseous hydrogen. To elucidate the HASCG mechanism, this study performed slow strain rate tensile (SSRT), elasto-plastic fracture toughness (JIC), fatigue crack growth (FCG) and fatigue life tests on Types 304, 316 and 316L in high-pressure hydrogen gas. Experimental results of Type 304 were provided in this paper as a representative of Types 304, 316 and 316L. The results demonstrated that the SSRT surface crack grew via the same mechanism as for the JIC and fatigue cracks, i.e., these crack growths could be uniformly explained on the basis of the hydrogen-induced successive crack growth (HISCG) model, which considers that cracks successively grow with a sharp shape under the loading process, due to local slip deformations near the crack tip by hydrogen. Accordingly, the HIS crack is ductile, not brittle.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A030. doi:10.1115/PVP2016-63403.

Structural materials employed for high-temperature and high-hydrogen-pressure service require excellent strength, and 2.25Cr-1Mo steel is widely used for such purposes. For instance, it is used in pressure vessels employed for hydro desulfurization, hydrocracking, and coal-liquefaction units. However, instances of hydrogen-related cracks occurring during a shutdown period have been reported when 2.25Cr-1Mo steel is used for long periods at high temperatures and high hydrogen pressures. As the risk of internal hydrogen embrittlement during shutdown, along with the detrimental effect of long-term temper-embrittlement, is a particular concern for the integrity and safe operation of pressure vessels, the combined effect of temper-embrittlement and hydrogen embrittlement in Cr-Mo steels was previously investigated using a step-cooling treatment, which is an accelerated temper-embrittlement treatment. In practical usage, it is important to ascertain the threshold stress-intensity factor at the onset of hydrogen-enhanced crack growth, KIH, of the material following long-term use. In the present study, we employed the offset potential-drop method to clarify the threshold stress-intensity factor, KIH for 2.25Cr-1Mo steel that had been used in actual equipment for a period of approximately 64,000 h at service temperatures (360–440 °C).

Commentary by Dr. Valentin Fuster
2016;():V06BT06A031. doi:10.1115/PVP2016-63536.

The effect of hydrogen gas environment (external hydrogen) and hydrogen-charging (internal hydrogen) on the fatigue crack growth (FCG) in two materials, austenitic stainless steel Type 304 and ductile cast iron, was investigated at various test frequencies. The pressure of hydrogen gas was 0.7 MPa. Both in the tests of external hydrogen and internal hydrogen, ratio of hydrogen-induced FCG acceleration was gradually increased with a decrease in test frequency in the range of 10 ∼ 0.1 Hz, and then peaked out at 0.1 ∼ 0.01 Hz. The frequency at the maximum acceleration was dependent on materials and test types (i.e. external hydrogen or internal hydrogen). It has been pointed out that, in the test of external hydrogen, a small amount of oxygen impurity contained in hydrogen gas, if any, adsorbs on newly-created crack surface, which inhibits hydrogen penetration into the material near crack tip. Lower frequency allows longer time for oxygen adsorption, and consequently, hydrogen-induced acceleration cannot be prominent at very low frequencies (e.g. 0.001 Hz). However, in this study, similar frequency dependence of hydrogen-induced FCG acceleration was also observed in the case of internal hydrogen. The results inferred the presence of another mechanism producing the frequency dependence of hydrogen-induced FCG acceleration, i.e. hydrogen-induced slip localization dominated by the gradient of hydrogen concentration ahead of crack tip.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A032. doi:10.1115/PVP2016-63555.

ASTM A 297 grade HP steels are widely employed for radiant tubes in reforming furnaces: this class of heat resistant alloys shows high creep and corrosion resistance, ensuring good performances in extreme pressure and temperature conditions. The typical microstructure of such materials is an austenitic matrix surrounded by a network of interdendritic carbides, which contain chromium and other carbide forming elements, namely Nb, Ti, W, Zr and Y.

During long service life, these high strength materials may suffer aging or even severe damage, especially when process conditions allow coke deposition, or maintenance procedures are not carried properly.

Service aging can be summarized, for HP steels, in terms of microstructure degradation: coalescence and coarsening of interdendritic precipitates, precipitation of secondary carbides in the austenite matrix and transformation of niobium-rich carbides in the G-phase silicide are the typical phenomena occurring on the microstructure of these alloys during service. Carburization can also occur in radiant tubes, since their inner wall side is exposed to hydrocarbon-rich process fluids: carbon diffuses into the metal matrix, causing massive precipitation of chromium-rich carbides. The alloy corrosion resistance is then reduced, resulting in surface attack, cracks development and a general wastage of the material. Furthermore, the high temperatures, which tubes are exposed to, can also induce creep, especially if a local tube overheating occurs: cavities and microcracks, mainly localized at precipitates, are the typical evidences of creep damage on HP steels.

The present work is aimed on the damage characterization of several radiant tubes in HP alloys, after long term service aging in reforming plants. We employed optical and electron microscopy, EDX elements mapping and mechanical tests, in order to characterize and evaluate the various damages affecting the alloys.

Microstructure evolution has been detected in all the analyzed tubes, but we found that such a phenomenon was strictly influenced by the chemical composition of each alloy, so that in presence of small amounts of titanium and tungsten, the chemical evolution of the secondary phases was appreciably contained. Creep also was observed in all the investigated tubes and its extent was found to be related to both alloy composition and process conditions. These latter have assumed to be the main driving factor for carburization, since we observed that slight differences in temperature, pressure, chemical composition of the process fluid and tube maintenance dramatically conditioned the performances of each tube. Massive precipitation and material degradation, in fact, were found in some cases, but, on the other side, no appreciable evidence of carburization damage was observed on other cases.

Topics: Heat , Steel , Damage
Commentary by Dr. Valentin Fuster
2016;():V06BT06A033. doi:10.1115/PVP2016-63563.

The degradation of stress-controlled fatigue-life (stress-life) of notched specimens was measured in the presence of internal and in external hydrogen for two strain-hardened austenitic stainless steels: 316L and 21Cr-6Ni-9Mn. To assess the sensitivity of fatigue performance to various hydrogen conditions fatigue tests were performed in four environments: (1) in air with no added hydrogen, (2) in air after hydrogen pre-charging to saturate the steel with internal hydrogen, and in external gaseous hydrogen at pressure of (3)10 MPa (1.45 ksi), or (4) 103 MPa (15 ksi). The fatigue performance of the strain-hardened 316L and 21Cr-6Ni-9Mn steels in air was indistinguishable for the tested conditions. Decreases in the fatigue-life at a given stress level were measured in the presence of hydrogen and depended on the hydrogen environment. Testing in 103 MPa (15 ksi) external gaseous hydrogen always resulted in a clear decrease in the fatigue-life at a given maximum stress. Alloy dependent reductions in the observed life at a given maximum stress were observed in the presence of internal hydrogen or in gaseous hydrogen at a pressure of 10 MPa (1.45 ksi). The measured fatigue-life of hydrogen pre-charged specimens was comparable to the life with no intentional hydrogen additions. Accounting for the increased flow stress resulting from the supersaturation of hydrogen after pre-charging results in consistency between the measured fatigue-life of the pre-charged condition and measurements in 103 MPa (15 ksi) external hydrogen. The current results indicate that internal hydrogen may be an efficient method to infer hydrogen-assisted fatigue degradation of stainless steels in high-pressure gaseous hydrogen.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A034. doi:10.1115/PVP2016-63609.

International standards and codes dedicated to design of pressure vessels are still unable to competitively ensure safe design and fitness for service of steel vessels for high pressure gaseous hydrogen. Emptying and shallow pressure cycles subject the material to hydrogen enhanced fatigue. A pre-normative project, MATHRYCE under the EU joint research program focused in this subject through material and component testing, analytical work, review of design methodologies and international collaboration.

An easy to implement, safe and economically competitive vessel design methodology is targeted. Steps towards this goal were taken by deepening our understanding on hydrogen enhanced fatigue in different kinds of laboratory specimens and real vessels designed for hydrogen service at maximum 45 MPa pressure. This included cyclic pressure testing of artificially notched vessels both in hydrogen and inert environment.

The effect of hydrogen pressure, frequency and mechanical loading parameters (ΔK, Sa) on fatigue crack initiation and propagation was analyzed. Attention was paid on the definition of “initiation” and influence of hydrogen on the relative parts of initiation and propagation on the fatigue life of a component. A good correlation between results with various test types was found. Particularly promising was the match between the measured — and estimated — crack growth rates in laboratory specimens and vessels. This supports our proposal for a safe design procedure based on crack growth and defect tolerant approach. Recommendations for implementation in a new international standard, on how to properly address hydrogen enhanced fatigue based on laboratory tests, were given and will be summarized in this presentation.

Our results indicate that crack initiation from inclusions or other small microstructural features is not necessarily affected by hydrogen to a similar extent as crack growth, but when initiated, the remaining life may be short due to fast growth. This is challenging for design and inspection rules to allow economically competitive construction of hydrogen equipment without compromising safety.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A035. doi:10.1115/PVP2016-63669.

Fatigue crack growth rate (da/dN) versus stress intensity factor range (ΔK) relationships were measured for various grades of pipeline steel along with their respective welds in high pressure hydrogen. Tests were conducted in both 21 MPa hydrogen gas and a reference environment (e.g. air) at room temperature. Girth welds fabricated by arc welding and friction stir welding processes were examined in X65 and X52 pipeline grades, respectively. Results showed accelerated fatigue crack growth rates for all tests in hydrogen as compared to tests in air. Modestly higher hydrogen-assisted crack growth rates were observed in the welds as compared to their respective base metals. The arc weld and friction stir weld exhibited similar fatigue crack growth behavior suggesting similar sensitivity to hydrogen. A detailed study of microstructure and fractography was performed to identify relationships between microstructure constituents and hydrogen accelerated fatigue crack growth.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A036. doi:10.1115/PVP2016-63683.

Pipelines are a practicable means for delivering large quantities of gaseous hydrogen over long distances and for distributing it as a transportation fuel at fueling stations in urban and rural settings. Glass-fiber-reinforced polymer (GFRP) pipelines are a promising alternative to the present-day use of low-alloy steel in pipelines for hydrogen transmission. GFRP pipelines offer advantages of lower capital cost and improved lifecycle performance, compared to steel pipelines. The technical challenges for adapting GRFP pipeline technology from oil and natural gas transmission, where it is in extensive service worldwide, to hydrogen transmission consists of evaluating the hydrogen compatibility of the constituent materials and composite construction, identifying the advantages and challenges of the various manufacturing methods, testing polymeric liners and pipelines to determine hydrogen permeability and leak rates, selecting options for pipeline joining technologies, establishing the necessary modifications to existing codes and standards to validate the safe and reliable implementation of the pipeline.

We performed examined the technical feasibility of using a commercially available spoolable glass-fiber-reinforced polymer (GFRP) pipeline for hydrogen transmission. We used an accelerated aging process based on the Arrhenius model to screen for hydrogen-induced damage in the pipeline and in the pipeline’s constituent materials. We also measured hydrogen leakage rates in short lengths of the pipeline. The accelerated aging process involved immersing GRFP pipeline specimens in pipeline-pressure hydrogen (6.9 MPa/1000 psi) at an elevated temperature (60°C) to promote an accelerated interaction of hydrogen with the pipeline structure. To assess specific effects on the constituent materials in the pipeline, specimens of fiberglass rovings, resin matrix and liner materials were immersed together with the pipeline specimens, and specimens of all types were subjected to either a one-month or an eight-month exposure to hydrogen at the elevated temperature. At the conclusion of each exposure interval the pipeline specimens were evaluated for degradation using hydrostatic burst pressure tests to assess the overall integrity of the structure, compression tests to assess the integrity of the polymer matrix, and bend testing to assess the integrity of the laminate. The results of these tests were compared to the results obtained from identical tests performed on un-conditioned specimens from the same manufacturing run. Tensile tests and dynamic mechanical analysis were performed on multiple specimens of constituent materials.

We measured the hydrogen leak rate in GFRP pipeline lined with pipeline-grade high-density polyethylene (PE-3408). The thickness of the liner was 0.526 cm and its inside diameter was 10.1 cm. The hydrogen pressurization during the leak rate measurements was 10.3 MPa (1500 psia) — the maximum recommended pressure — and all measurements were done at ambient temperatures in an air-conditioned laboratory. The pipeline was closed on each end using a steel cap with elastomer (O-ring) seals. The leak rate was calculated from the temperature-compensated pressure decay curve. Changes in pipeline volume that occurred due to pressure-induced dimensional changes in the pipeline length and circumference were measured using strain gauge sensors. These volumetric changes occurred at the earliest measurement times and diminished to near zero at the long measurement times during which the steady-state leak rate was determined. Leak rate measurements in three different lengths of pipeline yielded a leak rate was significantly lower than the predicted rate from the standard analytical model for a cylindrical vessel.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A037. doi:10.1115/PVP2016-63713.

Polymeric materials have played a significant role in the adoption of a multi-materials approach towards the development of a safe and cost-effective solution for hydrogen fuel storage in Fuel Cell Vehicles (FCVs). Numerous studies exist with regards to the exposure of polymeric materials to gaseous hydrogen as applicable to the hydrogen infrastructure and related compression, storage, delivery, and dispensing operations of hydrogen at fueling stations. However, the behavior of these soft materials under high pressure hydrogen environments has not been well understood. This study involves exposure of select thermoplastic and elastomeric polymers to high pressure hydrogen (70–100 MPa) under static, isothermal, and isobaric conditions followed by characterization of physical properties and mechanical performance. Special attempt has been made to explain hydrogen effects on polymer properties in terms of polymer structure-property relationships, and also understand the influential role played by additives such as fillers, plasticizers, and processing aids in polymers exposed to hydrogen. Efforts have also been focused on deriving suitable conditions of static testing in high pressure hydrogen environments as a valuable part of developing a suitable test methodology for such systems. Understanding the relationships between polymer composition and microstructure, time of exposure, rate of depressurization, purge and exposure conditions, etc. in this simple study will help better define the test parameters for upcoming high pressure cycling experiments in hydrogen.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A038. doi:10.1115/PVP2016-63927.

Duplex stainless steel (DSS) is one material choice to fabricate the reactor effluent air cooler (REAC) of hydrocracker units in order to improve the performance and service lifetime of these units. Unfortunately, several failures from around the world have been reported in REAC units constructed of DSS, some within five years of service. Based on failure analysis reports, the failures were generally associated with welded joints and were caused by crevice/pitting corrosion and stress corrosion cracking. Given the condition of hydrogen-rich environment, high-pressure process fluid, and service temperature, this type of cracking is most likely a form of hydrogen assisted cracking (HAC). It is highly influenced by phase balance (ferrite/austenite) after welding and welding procedures, with high levels of ferrite in the weld metal or HAZ increasing the susceptibility to HAC. In this study, different weld metal phase balances were prepared by autogenous gas tungsten arc welding (GTAW). The delayed hydrogen cracking test (DHCT) was used to evaluate the effects of the weld phase balance on the susceptibility to HAC in DSS 2205 welds. Using this approach, weld metal ferrite levels on the order of 90 vol% ferrite led to very rapid failure, while reducing the ferrite level to approximately 60 vol% greatly increased resistance to HAC.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A039. doi:10.1115/PVP2016-64033.

In this research paper, the focus is on SUS316 and SUS316L, types of austenitic stainless steel, as materials to be used for hydrogen refueling station (HRS) equipment from which hydrogen will be filled into 70 MPa onboard containers. Data on its characteristics in high-pressure-hydrogen environments, as well as methods of quantitative evaluation, were examined. On the basis of data, criteria for selecting materials and the standards to be used for HRS equipment were established.

The results indicate that the temperature ranges in which SUS316 and SUS316L can be used for HRS equipment with a hydrogen pressure of ≤82 MPa are as follows.

If the actual reduction of area of 75% is ensured, SUS316 and SUS316L can be used in the temperature ranges between −45 and 250 °C for a Ni equivalent of ≥28.5%, between −10 and 250 °C for a Ni equivalent of ≥27.4%, and between 20 and 250 °C for a Ni equivalent of ≥26.3%.

Topics: Hydrogen
Commentary by Dr. Valentin Fuster

Materials and Fabrication: Mechanistic Modelling of Materials

2016;():V06BT06A040. doi:10.1115/PVP2016-63028.

This paper presents a basic procedure for the integrity assessment of structural steels containing notches. It is based on the work developed by the authors in the last five years analyzing the notch effect in structural steels, with the Theory of Critical Distances as the main theoretical framework. The procedure combines the notch effect corrections provided by this theory with a basic Failure Assessment Diagram, and has been successfully validated through its application to 394 fracture tests performed on 4 different steels working at different temperatures.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A041. doi:10.1115/PVP2016-63277.

Loss of creep resistance in post-weld P91 alloy occurs mainly due to the change in microstructure particularly in the heat-affected zone under actual service operating conditions as well as residual stress from the welding process that is not often properly addressed in many damage models. In this paper, a validated deformation mechanisms map (DMM) using low temperature creep strain accommodation processes i.e. GBS, is used for the P91 alloy that predicts the creep rates over a wide range of temperature and stress including those arising under in the actual service conditions. These creep rates are further utilized into a microstructure-based creep damage model for accurate life prediction. A 3D transient computational welding mechanics (CWM) modeling of a pipe in a super-critical water loop, predicts the thermal, microstructure and stress state from welding. It also determines the coarse and fine grain heat affected zone (CGHAZ & FGHAZ). The CWM results are coupled with physics-based creep damage modeling to practically predict the creep life under the actual service conditions considering the welding residual stress and microstructure states.

Topics: Creep , Alloys , Welding , Damage
Commentary by Dr. Valentin Fuster
2016;():V06BT06A042. doi:10.1115/PVP2016-63464.

The need to predict changes in fracture toughness for materials where the tensile properties change through life, such as with irradiation, whilst accounting for geometric constraint effects, such as crack size, are clearly important. Currently one of the most likely approaches by which to develop such ability are through application of local approach models. These approaches appear to be sufficient in predicting lower shelf toughness under high constraint conditions, but may fail when attempting to predict toughness in the transition region, for low constraint geometries or for different irradiation states, when using the same parameters, making reliable predictions impossible. Cleavage toughness predictions in the transition regime are here made with a stochastic, Monte Carlo implementation of the recently proposed James-Ford-Jivkov model. This implementation is based around the creation of individual initiators following the experimentally observed distribution for specific reactor pressure vessel steel, and determining if these initiators form voids or cause cleavage failure using the model’s improved criterion for particle failure. This implementation has been presented previously in PVP2015-45905, where it was successfully applied across different constraint conditions; in the work presented here it is applied across different irradiation conditions for a second type of steel. The model predicts the fracture toughness in a large part of the transition region, demonstrates an ability to predict the irradiation shift and shows a level of scatter similar to that observed experimentally. All results presented, for a given material, are obtained without changes in the model parameters. This suggests that the model can be used predicatively for assessing toughness changes due to constraint-, irradiation- and temperature-driven plasticity changes.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A043. doi:10.1115/PVP2016-63513.

A user-friendly creep-fatigue damage calculation tool is developed in Visual Basic for Applications (VBA) with the familiar Microsoft Excel® user interface for power plant operators. Operational pressure and temperatures (steam and pipe exterior) are directly input and automatically converted to stress-time histories based on the summation of thermally- and mechanically-induced stresses. The stress history is automatically analysed and segregated into periods of sustained stress levels (creep range) and periods of fluctuating stress (fatigue range). Total damage is determined by summing the creep damage fraction (via Larson-Miller equation and Robinson’s rule) and fatigue damage fraction (via a rainflow cycle counting subroutine, the Smith-Watson-Topper fatigue parameter and Miner’s rule). Pre-existing damage fraction can be incorporated into the calculations. The remaining life estimates based on repetition of the load profile are outputted for the user. Finally, the estimated damage and remaining life are compared to that determined via the ASME, EN and TRD codes, where applicable.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A044. doi:10.1115/PVP2016-63537.

Structural integrity assessment codes such as R6 [1] and BS7910 [2] provide guidance on the assessment of flaws that are assumed to be infinitely sharp using the Failure Assessment Diagram (FAD). In many cases, such as fatigue cracks, this assumption is appropriate, however it can be pessimistic for flaws that do not have sharp tips such as lack of fusion, porosity or mechanical damage. Several Notch Failure Assessment Diagram (NFAD) methods have been proposed in the literature to quantify the additional margins that may be present for non-sharp defects compared to the margins that would be calculated if the defect were assumed to be a sharp crack. This paper presents the first stage of on-going work to validate an NFAD method and to develop guidance for its application in safety assessments. The work uses 3D Finite Element (FE) Analysis in conjunction with a wide range of test data on non-sharp defects as a basis for validation. The paper also develops some practical guidance on the treatment of Lüders strain in the FE analysis of specimens containing notches instead of fatigue pre-cracks.

Topics: Failure
Commentary by Dr. Valentin Fuster

Materials and Fabrication: Pipeline Integrity

2016;():V06BT06A045. doi:10.1115/PVP2016-63079.

A series of single-edge notched tension (SENT or SE(T)) and single-edge notched bend (SENB or SE(B)) testing was carried out at −15°C using BxB specimens machined from two API X70 large diameter pipeline girth welds. An initial notch was placed either on the heat-affected zone (HAZ) or the weld metal center from the outer diameter side of pipe to simulate a circumferential surface flaw. SE(T) and SE(B) tests were performed according to the CANMET procedure and ASTM E1820, respectively. For all HAZ SE(B) specimens machined from one pipe, ductile cracks initially propagated away from the fusion line and toward the base metal side due to asymmetric deformation, and then pop-in (i.e., the initiation and arrest of a brittle crack) occurred after ductile crack growth of approximately 1 mm where the crack reached around the inter-critical heat-affected zone. HAZ SE(T) specimens also showed that the ductile crack propagation deviated toward the base metal side, but an unstable brittle crack extension was not observed from any SE(T) specimens as opposed to SE(B) specimens. None of the weld metal SE(T) and SE(B) specimens showed pop-in or brittle fracture at −15°C or room temperature. The difference in test results, for the same material, is likely due to the different constraint levels in the two loading modes. While pop-ins were triggered in high-constraint SE(B) tests, it was not the case for the low-constraint SE(T) tests. This paper presents these observations and discussion.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A046. doi:10.1115/PVP2016-63662.

Fracture toughness is often described by the J-integral or crack-tip opening displacement (CTOD) for ductile materials. ASTM, BSI and ISO have developed their own standard test methods for measuring fracture initiation toughness and resistance curves in terms of the J and CTOD using bending dominant specimens in high constraint conditions. However, most actual cracks are in low constraint conditions, and the standard resistance curves may be overly conservative.

To obtain more realistic fracture toughness for actual cracks in low-constraint conditions, different fracture test methods have been developed in the past decades. To facilitate understanding and use the test standards, this paper presents a critical review on commonly used fracture toughness test methods using standard and non-standard specimens in reference to the fracture parameters J and CTOD, including (1) ASTM, BSI and ISO standard test methods, (2) constraint correction methods for formulating a constraint-dependent resistance curve, and (3) direct test methods using the single edge-notched tension (SENT) specimen. This review discusses basic concepts, basic methods, estimation equations, test procedures, historical efforts and recent progresses.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A047. doi:10.1115/PVP2016-63851.

In this paper, the loading path effect on the elastic-plastic crack-front stress field in a thin plate is investigated. There different loading sequences include one proportional loading and two non-proportional loading paths are applied to the 3-D modified boundary layer (MBL) model under small-scale yielding conditions. For the same external displacement field applied at the outer boundary of the 3-D MBL model, the mode I K field and T-stress field combined as the different loading path is applied to investigate the influence of the nonproportional loading. The results show that for either the compressive or tensional T-stress, the loading path which applied K field followed by T field generates the lower crack-tip constraint. There is only slightly difference between the proportional loading path and the T-stress field following by K field loading path. The results show that it is very important to include the load sequence effects in fracture analysis when dealing with nonproportional loading conditions.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A048. doi:10.1115/PVP2016-63911.

In this work a phenomenological approach is employed for fracture initation and propagation in SENT specimens. Modified Mohr-Coulomb (MMC) fracture criterion establish the initiation of damage/fracture at the critical material point and the post-initation softening controls its rate of propagation. By means of a tridimensional mixed stress state (equivalent plastic strain, εp; Triaxiality, η; and Lode angle, θ) the nature of the growing crack front can be explained.

Commentary by Dr. Valentin Fuster

Materials and Fabrication: Plastic Pipe

2016;():V06BT06A049. doi:10.1115/PVP2016-63432.

A study was conducted to develop a new approach for characterizing environmental stress cracking resistance (ESCR) of polyethylene (PE). The main objective is to reduce time for the testing, to be shorter than that required for the current standard ESCR tests. The new approach applies indentation to generate deflection in the central region of a PE plate, and uses time for crack generation under constant displacement, during the exposure to an aggressive agent, to characterize ESCR. The indentation introduces stretch to a central annular region around the indenter, in which PE molecules are increasingly oriented in the stretch direction. Since this annular region is subjected to bi-axial tension, exposure to aggressive agent increases its vulnerability to crack formation in the stretch direction. ESCR is characterized by the time needed for crack formation during the exposure to an aggressive agent (10% Igepal CO-630 solution). This paper presents the test set-up for the new approach, and compares time for the crack generation using this new approach with that from ASTM D1693, condition A.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A050. doi:10.1115/PVP2016-63595.

The use of non-destructive examination (NDE) for assessing the quality of butt fusion joints in polyethylene (PE) pipes has been included in the draft Mandatory Appendix XXVI to Section III of the ASME Boiler and Pressure Vessel Code (Rules for construction of Class 3 buried polyethylene pressure piping). However, currently, there are no acceptance criteria for flaws in butt fusion joints in PE pipes. There is an ASME Task Group on flaw evaluation for PE pipe, which is developing a code case using linear elastic fracture mechanics (LEFM) to determine critical flaw sizes. However, the initial experimental crack growth data generated suggests that linear elastic fracture mechanics is not able to adequately describe slow crack growth in PE materials. In addition, this work is only considering planar lack of fusion flaws in the joint; it is not considering other critical flaw types that can occur in butt fusion joints, such as particulate contamination and cold fusion.

TWI has developed procedures using mechanical testing to develop flaw acceptance criteria for butt fusion joints in PE pipes. This is based on inserting lack of fusion flaws of known size and particulate contamination flaws of known concentrations into butt fusion joints and determining the effect of these flaws on both the short-term and long-term integrity of the joints. An important aspect of this work is to determine which of the wide array of mechanical tests available for assessing the integrity of butt fusion joints in PE pipes are the most discriminating.

This paper describes the procedures developed for inserting simulated flaws into butt fusion joints in PE pipes, the experimental work to compare the results from different standard short-term and long-term tests on flawed and unflawed joints and the procedures developed to determine flaw acceptance criteria.

Results have shown that the most discriminating short-term test for butt fusion joints in PE pipes is a tensile test using a waisted specimen, such as those defined in EN 12814-2, EN 12814-7 and ISO 13953, and the most discriminating property is the energy to break the specimen. The most appropriate long-term test for butt fusion joints in PE pipes is the whole pipe tensile creep rupture test, as defined in EN 12814-3; this is the only long-term whole pipe test that consistently generates slow crack growth in the fused joint, even if it contains no flaws.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A051. doi:10.1115/PVP2016-63688.

The desire to use high-density polyethylene (HDPE) piping in buried Class 3 service and cooling water systems in nuclear power plants is primarily motivated by the material’s high resistance to corrosion relative to that of steel alloys. The rules for construction of Class 3 HDPE pressure piping systems were originally published as an alternative to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC) in ASME Code Case N-755 and were recently incorporated into the ASME BPVC Section III as Mandatory Appendix XXVI (2015 Edition). The requirements for HDPE examination are guided by criteria developed for metal pipe and are based on industry-led HDPE research and conservative calculations.

Before HDPE piping will be generically approved for use in U.S. nuclear power plants, the U.S. Nuclear Regulatory Commission (NRC) must have independent verification of industry-led research used to develop ASME BPVC rules for HDPE piping. With regard to examination, the reliability of volumetric inspection techniques in detecting fusion joint fabrication flaws against Code requirements needs to be confirmed. As such, confirmatory research was performed at the Pacific Northwest National Laboratory (PNNL) from 2012 to 2015 to assess the ability of phased-array ultrasonic testing (PAUT) as a nondestructive evaluation (NDE) technique to detect planar flaws, represented by implanted stainless steel discs, within HDPE thermal butt-fusion joints. All HDPE material used in this study was commercially dedicated, 305 mm (12.0 in.) nominal diameter, dimension ratio (DR) 11, PE4710 pipe manufactured with Code-conforming resins, and fused by a qualified and experienced operator. Thermal butt-fusion joints were fabricated in accordance with or intentionally outside the standard fusing procedure specified in ASME BPVC. The implanted disc diameters ranged from 0.8–2.2 mm (0.03–0.09 in.) and the post-fabrication positions of the discs within the fusion joints were verified using normal- and angled-incidence X-ray radiography. Ultrasonic volumetric examinations were performed with the weld beads intact and the PA-UT probes operating in the standard transmit-receive longitudinal (TRL) configuration. The effects of probe aperture on the ability to detect the discs were evaluated using 128-, 64-, and 32-element PA-UT probe configurations. Results of the examinations for each of the three apertures used in this study will be discussed and compared based on disc detection using standard amplitude-based signal analysis that would typically be used with the ultrasonic volumetric examination methods found in ASME BPVC.

Topics: Density
Commentary by Dr. Valentin Fuster

Materials and Fabrication: Probabilistic Assessment of Failure

2016;():V06BT06A052. doi:10.1115/PVP2016-63112.

This paper describes the effect of variability of fracture toughness of nuclear pressure vessels during a PTS event. The model used in this paper is based on the NESC-1 experiment. To determine the behavior of the surface breaking defect NRG performed three dimensional finite element calculations and subsequently extended these calculations to the probabilistic calculations. Three-dimensional finite-element model of the cladded cylinder was generated using ANSYS with semi-elliptical surface crack having a crack depth of 75 mm and a crack length of 205 mm. The cylinder specimen was subjected to thermal-shock and centrifugal loading conditions and analyzed with a themo-elastic-plastic material model and subsequently determined the fracture mechanics parameters (J and K) along the elliptical crack front as a function of time and temperature. The determined stress intensity factor K has been compared with the available cleavage fracture toughness (KJC) data with 50% fracture probability which has been obtained from the Master Curve according to BS7910. The comparison has been performed for the locations in the base metal as well as the locations in the heat affected zone.

Deterministic analysis has been extended to probabilistic analysis to calculate the failure probability for the crack initiation at the locations in the base metal as well as the locations in the heat affected zone along the crack front by considering probabilistic distributions from Master Curve and FAVOR. Master Curve analysis through the ASME code case N-629 has been applied to the material. Results obtained from these two methods have been compared and also the results are used to compare the inherent safety factors in the deterministic analysis using RTNDT and Master Curve.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A053. doi:10.1115/PVP2016-63771.

In the past, the US Nuclear Regulatory Commission (NRC) has typically regulated the use of nuclear structural materials on a deterministic basis. Safety factors, margins, and conservatisms were used to account for model and input uncertainty. However, in the mid-1990s, the NRC issued a policy statement that encouraged the use of Probabilistic Risk Assessments (PRA) to improve safety decision making and improve regulatory efficiency. Since that time, the NRC has made progress in its efforts to implement risk-informed and performance-based approaches into its regulation and continues to revisit and update the approaches on a regular basis.

A major component to the overall safety of nuclear structures is the fracture behavior of the materials. Consensus codes and standards responsible for the design and analysis of such structures, such as ASME Boiler and Pressure Vessel code, typically rely on conservative fracture models with applied safety factors and conservative bounding inputs to account for the numerous uncertainties that may be present. Improving the reliability of such models by truly understanding the impacts of the assumptions and uncertainties becomes difficult because of the conservative nature of the models and inputs and the inadequate documentation of the basis for safety factors.

As a subset of PRA, probabilistic fracture mechanics (PFM) is an analysis technique that allows greater insight into the structural integrity of components than similar deterministic analyses. PFM allows the direct representation of uncertainties through the use of best-estimate models and distributed inputs. This analysis methodology permits determination of the direct impact of uncertainties on the results, which gives the user the ability to determine and possibly refine the specific drivers to the problem. However, PFM analyses can be more complicated and difficult to conduct than deterministic analyses. Determining validated best-estimate models, developing input distributions with limited data, characterizing and propagating input and model uncertainty, and understanding the impacts of problem assumptions on the adequacy of the results, can complicate the development and approval of PFM analyses in a regulatory application.

This paper provides some thoughts on how to improve confidence in structural analyses performed using PFM, by focusing on topics such as solution convergence, input distribution determination, uncertainty analyses, sensitivity analyses (to determine impact of uncertainties on result) and sensitivity studies (to determine impact of mean values on the results). By determining the main drivers to the probabilistic results and investigating the impacts of the assumption made to develop those drivers, the confidence in the overall results can be greatly improved.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A054. doi:10.1115/PVP2016-63801.

Several nuclear power plants in Japan have been operating for more than 30 years and cracks due to age-related degradations have been detected in some piping systems during in-service inspections. Furthermore, several of them have experienced severe earthquakes in recent years. Therefore, failure probability analysis and fragility evaluation for piping systems, taking both age-related degradations and seismic loads into consideration, has become increasingly important for the structural integrity evaluation and the seismic probabilistic risk assessment.

Probabilistic fracture mechanics (PFM) is recognized as a rational methodology for failure probability analysis and fragility evaluation of aged piping, because it can take the scatters and uncertainties of influence parameters into account. In our Japan Atomic Energy Agency (JAEA), a PFM analysis code PASCAL-SP was developed for aged piping considering age-related degradations. In this study, we improved PASCAL-SP for the fragility evaluation taking both age-related degradations and seismic loads into account. The details of the improvement of PASCAL-SP are explained and some example analysis results of failure probabilities, fragility curves and a preliminary investigation on seismic safety margin are presented in this paper.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A055. doi:10.1115/PVP2016-63825.

The Highly Active Liquid Effluent and Storage plant at Sellafield, UK, currently uses three evaporators to reduce the volume of active liquor stored within the facility before being vitrified for long term storage. This liquor is highly corrosive and the lifetime of the evaporators is potentially limited by the corrosion loss from the heating elements, comprising an external jacket and a number of internal coils, all heated by low pressure steam. Inspection of the heating coils inside the evaporators is possible and measurement data is available of their thicknesses by depth at various inspection intervals. This inspection data has been combined with operational data and thermal models for the heating elements. Our theoretical understanding from laboratory measurements suggests that corrosion is related to temperature through an Arrhenius relationship. As such we have been able to develop a predictive model for the thickness profiles and remaining useful life of the uninspected components. This model is a non-linear mixed effects (multilevel) model and has undergone significant developmental work to account for a number of practical data issues. This paper will briefly outline the various components of the model, whilst discussing issues relevant to any statistical model such as complexities of data collection, approaches to handling correlated data, selecting appropriate model formulations and data transformations. The inclusion of uncertainties in prediction via Monte-Carlo simulation will also be discussed.

Topics: Vessels , Heating
Commentary by Dr. Valentin Fuster
2016;():V06BT06A056. doi:10.1115/PVP2016-63919.

Product durability and reliability validation requires fatigue testing of limited samples and the corresponding statistical and probabilistic analyses of the test data. The uncertainties introduced into the tests with limited sample size and the assumptions made about the underlying probabilistic distribution function will significantly impact the analysis results and the result interpretation. In this paper, a large amount of fatigue life data collected from several programs are analyzed. The values of the parameters estimated from the two-parameter and three-parameter probabilistic distribution functions, Weibull in particular, are compared. Finally, the observations are summarized and some recommendations are provided.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A057. doi:10.1115/PVP2016-63962.

NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals.

US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR), defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations [1].

Using the lessons learned from the pilot study [2] the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB [3]. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties.

One of the most influential sources of uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system.

Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. This paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. The current WRS sampling scheme employs correlation in order to smooth the shape of the WRS fields through the thickness of a dissimilar metal weld. This method presents an enrichment of the Cholesky decomposition on the correlation matrix, in order to satisfy the other two requirements.

Topics: Stress , Uncertainty
Commentary by Dr. Valentin Fuster
2016;():V06BT06A058. doi:10.1115/PVP2016-63963.

NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals.

US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR) [1], defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations.

Using the lessons learned from the pilot study, the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties.

One of the most influential uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system.

Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. A companion paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. This paper illustrate with examples the implementation of the described methods into xLPR v2.0.

Topics: Stress , Uncertainty
Commentary by Dr. Valentin Fuster

Materials and Fabrication: Small-Scale and Miniature Mechanical Testing

2016;():V06BT06A059. doi:10.1115/PVP2016-63502.

This work is about evaluating the behavior facing HIC of high strength steels by means of the Small Punch Test (SPT). It can be considered as a quasi-non-destructive test in comparison to structural integrity analysis of large components. It was developed during the 80’s with the purpose of estimating the embrittlement grade of nuclear components reducing the amount of material employed. During the last years it has been successfully employed in the evaluation of mechanical properties of different materials and creep behavior. Also approximations for the fracture properties estimations have been carried out using this method. Although a reference standard that includes the tensile and fracture parameters estimations by SPT does not exist, a European Code of Practice (CWA 15627:2008) was recently developed. In addition a European standard is in preparation, including the ultimate research and the backup of the most relevant groups.

In this work, high strength steels behavior facing stress corrosion cracking (SCC) or hydrogen Embrittlement (HE) processes are analyzed by means of the Small Punch Test (SPT). The evaluation of the response of materials facing environmental damage processes requires a different consideration if cracks are present on the material or not. In a first stage the study carried out tries to analyze the behavior without cracks, using the threshold stress (σscc) parameter. The aforementioned parameter is obtained from slow strain rate tensile tests (SSRT), which involves its own particular disadvantages. Thus the aptitude of the SPT to obtain the threshold stress is studied, evaluating the influence of variables such as the solicitation rate.

In the second part of the work, specimen geometry and test conditions are proposed for the SPT, in order to evaluate the susceptibility facing SCC and HE in presence of cracks for the materials studied. In this case, the fracture toughness parameter that describes the crack initiation process (Khe) will be evaluated and validated by conventional tests based on fracture mechanics. The influence of variables, such as test solicitation rate on the results, is analyzed in order to obtain a qualitative methodology to evaluate mechanical-environmental damage processes by SPT means. For the SPT tests carried out, common Small Punch specimens of 10×10 mm of section and 0,5 mm of thickness are used for σscc determination.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A060. doi:10.1115/PVP2016-63652.

The Small Punch test is a miniaturized non-intrusive methodology that allows performing creep tests using very small specimens. It can be used for the residual lifetime assessment of in service components.

In the last years, numerous applications of Small Punch Testing as well as related pre-normative work and Round Robin exercises have led to new insights which motivate a revision of the current Code of Practice for Small Punch Testing (CWA 15627, Dec. 2007) and/or its transfer in an EN norm.

In this paper we present the state-of-art of the small punch technique: its performances, usefulness and critical aspects have been discussed and some examples have been reported. Some measurements carried out during the round-robin exercises have been reported: several testing laboratories are cooperating in a round-robin on a virgin P92 material. The tests are still in progress, so only preliminary data will be shown in this paper.

A series of measurements performed during the pre-normative time will also be shown: a serviced (116,000 hours) ASME A213 T91 tube, installed in a petrochemical plant has been investigated. Small punch tests were performed and the residual life was then estimated by Omega Method. A microstructure characterization has been also reported.

The results obtained encourage new efforts for material testing by SPT; sampling, however, was found as a critical step if the analyzed material may suffer from localized damage.

Topics: Steel , Testing
Commentary by Dr. Valentin Fuster
2016;():V06BT06A061. doi:10.1115/PVP2016-63891.

Fracture toughness of reactor pressure vessel materials appears obvious scatter. It is desirable in assessment codes to characterize fracture toughness by a low fractile of its distribution. This low fractile is known as a characteristic value. However, the real distribution type is unknown, and usually assumed to be normal, lognormal or Weibull. In this paper, the characteristic values with given confidence level and probability are obtained by one-sided tolerance factors for normal, lognormal and Weibull distribution. These characteristic values are compared with that obtained with minimum of three equivalent.

Commentary by Dr. Valentin Fuster

Materials and Fabrication: Structural Integrity for Spent Fuel Canisters

2016;():V06BT06A062. doi:10.1115/PVP2016-63311.

Ultrasonic inspection methodologies for welded stainless steel canisters in spent nuclear fuel dry storage casks are investigated for detection of stress corrosion cracking (SCC) in the heat affected zone (HAZ) of a weld. Shear horizontal (SH) waves having 3D wave motion, which are actuated and received by periodic permanent magnet electromagnetic acoustic transducers (EMATs), are used to detect cracks perpendicular or parallel to the weld line. Due to the limited accessibility to the welds in the stainless steel canister, three different types of sensor layouts amenable to robotic delivery are proposed that use either through-transmission or pulse-echo sensor configuration. A stainless steel welded plate having artificial surface breaking notches near the weld is inspected to demonstrate the performance of the methods. Two dimensional ultrasonic images showing the notch locations are obtained by scanning a pair of EMATs in the direction parallel to the weld line. The feasibility of detecting cracks normal to, and along, the weld line using pulse-echo mode is demonstrated. Thus, robotically delivered SH-wave EMATs will be effective for inspecting the entirety of the weld lines in the stainless steel canister, even in the presence of guide channels.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A063. doi:10.1115/PVP2016-63312.

Extended dry storage of spent nuclear fuel makes it desirable to assess the structural integrity of the storage canisters. Stress corrosion cracking of the stainless steel canister is a potential degradation mode especially in marine environments. Sensing technologies are being developed with the aim of detecting the presence of chloride-bearing salts on the surface of the canister as well as whether cracks exist. Laser induced breakdown spectroscopy (LIBS) methods for the detection of Chlorine are presented. Detection of a notch oriented either parallel or perpendicular to the shear horizontal wave vector is demonstrated using the pulse-echo mode, which greatly simplifies the robotic delivery of the noncontact electromagnetic acoustic transducers (EMATs). Robotic delivery of both EMATs and the LIBS system is necessary due to the high temperature and radiation environment inside the cask where the measurements need to be made. Furthermore, the space to make the measurement is very constrained and maneuverability is confined by the geometry of the storage cask. In fact, a large portion of the canister surface is inaccessible due to the presence of guide channels on the inside of the cask’s overpack, which is strong motivation for using guided waves for crack detection. Among the design requirements for the robotic system are: to localize and track where sensor measurements are made to enable return to those locations, to avoid wedging or jamming of the robot, and to tolerate high temperatures and radiation levels.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A064. doi:10.1115/PVP2016-63634.

This paper describes the derivation and experimental validation of geometric equations that govern insertion and extraction of a robotic inspection system that operates in gaps around vertical dry storage casks. During insertion, a robotic system may become jammed due to unbalanced forces acting on the robot, or wedged due to over-sized robot geometry. The robot must be removable by a tether in the event of power loss. Assuming simplified geometry and a quasi-static approach, the problem is modeled using a two-dimensional representation in which the robot is assumed to be rigid with equal weight distribution and a constant friction coefficient between surfaces. Equilibrium equations are derived from a modified peg-insertion formulation, allowing calculation of the maximum size of the robot and angle of insertion as a function of inspection gap geometry and friction. Experimentation tested the derived relationships using varying robot dimensions in a 1:1 scale mock-up of the overpack-to-canister gap space of a nuclear dry storage container. Experimental data confirmed that the modifications of the typical peg-insertion predicted successful insertion and extraction better than unmodified equations. The error between the model and experimentation had a mean and standard deviation of 4.4 and +/− 0.53 degrees.

Topics: Robotics , Storage
Commentary by Dr. Valentin Fuster
2016;():V06BT06A065. doi:10.1115/PVP2016-63884.

The conditions of continued dry storage of the spent nuclear fuel in multipurpose canisters render the canisters, a component for confinement in dry storage cask systems, susceptible to chloride-induced stress corrosion cracking (SCC). The requisite conditions involve deposits of chloride-bearing marine salts and/or dust that deliquesce on the external surface of the cooling canister to create brine at weld residual stress regions. The subcritical crack growth rate at this “dry salt” condition, investigated by several researchers, has shown a relatively slow growth rate compared to chloride-cracking under aqueous conditions. A new SCC growth rate test specimen configuration has been developed to enable an initially dried salt assemblage to deliquesce under temperature and humidity conditions to load the fatigue pre-cracked, wedge-opening-loaded (WOL) specimen with the brine and enable measurements of crack growth rate (da/dt) under falling stress intensity factor, KJ, conditions. The application of the results to a canister weldment with a residual stress profile to predict crack extension in time is described. The results are evaluated in terms of development of acceptance standards for this type of flaw, should SCC be identified and characterized through inservice inspection (ISI).

Commentary by Dr. Valentin Fuster
2016;():V06BT06A066. doi:10.1115/PVP2016-63887.

Many stainless steel canisters for the dry storage of spent nuclear fuel are located in coastal regions. Because the heat treatment for relieving the welding residual stress is not required during fabrication, these canisters may be susceptible to chloride induced stress corrosion cracking due to the deliquescence of chloride-bearing marine salts or dust that enter the overpack system and deposit on the canister external surface. The NDE techniques and the associated delivery system are being developed to conduct periodic inservice inspections. The acceptance standards are needed to disposition findings should flaw-like indications be found. The instability crack lengths and depths for these flaws in the form of semi-elliptical shape near the welds are determined with R6 procedure. The cracks are subject to the canister design pressure and handling loads as well as the estimated welding residual stresses.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A067. doi:10.1115/PVP2016-64037.

Paragraph IWB-3641 of Section XI [1] permits the use of Nonmandatory Appendix C or H to perform the analytical evaluation of flaws in austenitic and ferritic piping. When considering welded austenitic material, Appendix C does not include the welding residual stress as an input to the calculations for maximum allowable flaw length and depth. In contrast, Appendix H does include the welding residual stress when the K′r coordinate on the failure assessment diagram curve is calculated for austenitic materials. This paper calculates the maximum allowable axial and circumferential flaws in a welded austenitic component subject to internal pressure loads using Appendix C procedures and using Appendix H procedures. The results are compared to investigate cases where a flaw may be acceptable using one set of procedures and not acceptable using the other.

Commentary by Dr. Valentin Fuster

Materials and Fabrication: Welding Residual Stress and Distortion Simulation and Measurement

2016;():V06BT06A068. doi:10.1115/PVP2016-63197.

This paper describes a sequence of residual stress measurements made to determine a two-dimensional map of biaxial residual stress (weld direction and transverse to the weld direction) in a mockup with a partial arc excavation and weld repair (EWR), as well as three additional maps of one component of residual stress. The mockup joins two dissimilar metal plates (SA-508 low alloy steel and Type 316L stainless steel) with a nickel alloy weld metal (Alloy 82/182). A partial groove is then excavated and filled in with SCC resistant Alloy 52M weld metal. The mockup was fabricated to investigate the effectiveness of the EWR mitigation methodology being investigated through the development of ASME Code Case N-847 to address stress corrosion cracking problems in reactor coolant system butt welds. The biaxial stress map is determined using a newly developed technique called primary slice removal (PSR) mapping, which uses both contour method and slitting measurements. In this case, the technique requires measuring the longitudinal stress along a plane and the long transverse stress remaining in a slice removed adjacent to that plane. This paper includes descriptions of the experiments and data analysis. The measured residual stresses follow expected trends and compare favorably to the results of computational weld residual stress modeling.

Topics: Maintenance , Stress
Commentary by Dr. Valentin Fuster
2016;():V06BT06A069. doi:10.1115/PVP2016-63237.

Residual stress prediction contributes to nuclear safety by enabling engineering estimates of component service lifetimes. Subcritical crack growth mechanisms, in particular, require residual stress assumptions in order to accurately model the degradation phenomena. In many cases encountered in nuclear power plant operations, the component geometry permits two-dimensional (i.e., axisymmetric) modeling. Two recent examples, however, required three-dimensional modeling for a complete understanding of the weld residual stress distribution in the component. This paper describes three-dimensional weld residual stress modeling for two cases: (1) branch connection welds off reactor coolant loop piping and (2) a mockup to demonstrate the effectiveness of the excavate and weld repair process.

Topics: Stress , Nuclear power
Commentary by Dr. Valentin Fuster
2016;():V06BT06A070. doi:10.1115/PVP2016-63269.

In temper bead welding, the heat input from welding is purposefully utilized to temper the hard microstructure for improving toughness. An optimal temper bead welding requires careful control of heat input and bead placement. In this study, the effect of linear heat input on heat-affected zone (HAZ) tempering was studied by a combination of experimental testing and numerical modeling. Temper bead welding experiments were performed on SA-533 steel using three different heat inputs while keeping the power ratio constant. The extent of tempering in the HAZ was quantified using micro-hardness mapping. A 2-D weld heat transfer model, using the double-ellipsoidal heat flux equation, was developed to calculate the temperature evolution. The peak temperatures experienced in the substrate’s HAZ was correlated to the hardness distribution. The results indicate that the linear heat input can have a significant influence on the extent of tempering in temper bead welding.

Topics: Heat , Welding
Commentary by Dr. Valentin Fuster
2016;():V06BT06A071. doi:10.1115/PVP2016-63358.

Historically, weld residual stresses (WRS) have been used as the primary validation parameter for welding simulations, largely due to the importance of predicting WRS for structural integrity assessments. However, the extent of welding-induced plasticity (WIP) caused by the plastic flow of near-weld material is also an important characteristic affecting weld performance. WIP has been shown to negatively affect weld integrity, since the associated accumulation of defects (dislocations) in the material will accelerate the nucleation of macro-scale defects that lead to component failure. Information on WIP is particularly important when attempting to validate the constitutive models used for weld simulation, and can assist with the proper definition of material yield strength.

The present study highlights two approaches to assess WIP in welded structures. The first approach involves the development of a micro-hardness correlation to infer the level of WIP across the near-weld region. The second approach uses electron backscatter diffraction (EBSD) data to directly calculate the average crystal misorientation in the region of interest, which is proportional to the amount of geometrically necessary dislocations present. The dissimilar approach to determine WIP between the two characterization methods allows a degree of confidence in the results obtained, therefore providing an accurate dataset for weld model validation. To exemplify this point, the two approaches are used to characterize WIP across a three-pass slot weld in AISI 316 steel (NeT TG4 specimen), and the results are compared to weld modelling predictions.

Topics: Plasticity , Steel , Welding
Commentary by Dr. Valentin Fuster
2016;():V06BT06A072. doi:10.1115/PVP2016-63378.

This research investigated the effects of global (in other words, furnace-based) and local post weld heat treatment (PWHT) on residual stress (RS) relaxation in API 5L X65 pipe girth welds. Two pipe spools were fabricated using identical pipeline production procedures for manufacturing multi-pass narrow gap welds. Non-destructive neutron diffraction strain scanning was carried out on girth welded pipe spools and stress-free comb samples and also thin slices for the determination of lattice spacing. All residual stress measurements were carried out at the KOWARI strain scanning instrument at the Australian Nuclear Science and Technology Organization (ANSTO).

Residual stresses of two pipe spools (in the as-welded condition) were measured through the thickness in the weld material and adjacent parent metal starting from the weld toe. Three line-scans were completed 3mm below outer surface, at mid thickness and 3mm above the inner surface. PWHT was adopted for stress relaxation; one pipe was conventionally heat treated entirely in an enclosed furnace and the other was locally heated by a flexible ceramic heating pad. Residual stresses were measured after PWHT at exactly the same locations as those used for the as-welded condition. Residual stress states of the two pipe spools in as-welded condition and after PWHT were compared and the results were presented in full stress maps. Additionally, through thickness residual stress profiles and the results of one line scan (3mm below outer surface) were compared with the respective residual stress profiles advised in British Standard BS 7910 “Guide to methods for assessing the acceptability of flaws in metallic structures” and the UK nuclear industry’s R6 procedure. The residual stress states of the two pipe spools measured in the as-welded condition were similar. With the given parameters, local PWHT has effectively reduced residual stresses in the pipe spool to such a level that it prompted the thought that local PWHT can be considered a substitute for global PWHT.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A073. doi:10.1115/PVP2016-63613.

The integrity of structures in nuclear power plants has to be assessed to meet given safety criteria. For a better understanding of the in-service loads in welded areas of PWRs components, the residual stresses resulting from the welding process have to be evaluated. For that purpose, numerical simulation of welding has proved efficient. However, in an industrial context, simple models have to be used to stay reliable, brief and easy to use.

Among all the steps required to run a suitable computation of the welding, this paper focuses on the calibration of the heat source. Two complementary approaches are applied to define an equivalent heat input, which coarsely include complex phenomena in the weld pool. The engineering practice is driven by a know-how: simplified method are used to get an efficient result in short modelling time. On the other hand, EDF R&D develops tools which quantify the difference between models and lead to an objective choice. The mix of both approaches enables to consolidate the engineering practice.

Topics: Welding , Computation
Commentary by Dr. Valentin Fuster
2016;():V06BT06A074. doi:10.1115/PVP2016-63733.

A catastrophic furnace roll failure was observed in a continuous hot dip line. The failure occurred in the weld joining an end bell to a roll shell and resulted in the complete separation of the end bell from the roll shell. The roll had been in service for approximately 7 months before failure. Typically, rolls in this furnace roll position have a short life, less than 1 year. The roll shell and the end bell were made of high-temperature alloy. The journal was made of stainless steel. Inconel welding electrode was used as the weld filler material. The roll shell and the end bells were shrink fitted before welding. The welding process was flux-cored arc welding (FCAW). Upon completion of welding post-weld heat treatment (PWHT) was applied for 3 hours at 1093°C to the entire roll. To extend the roll life, a new design and a new welding method, electron beam welding (EBW) without filler metal, were proposed. To evaluate the effectiveness of the new design and the new welding process, four kinds of numerical analyses were conducted on the new design for both FCAW and EBW which included a weld residual stress analysis, a PWHT analysis, a heat transfer analysis, and a creep-fatigue analysis. Analysis results showed that the new design with EBW has lower stress and creep strain than with FCAW, which could improve the roll creep-fatigue life.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A075. doi:10.1115/PVP2016-63815.

This paper presents predictions of weld residual stresses in a mockup with a partial arc excavate and weld repair (EWR) utilizing finite element analysis (FEA). The partial arc EWR is a mitigation option to address stress corrosion cracking (SCC) in nuclear power plant piping systems. The mockup is a dissimilar metal weld (DMW) consisting of an SA-508 Class 3 low alloy steel forging buttered with Alloy 182 welded to a Type 316L stainless steel plate with Alloy 82/182 weld metal. This material configuration represents a typical DMW of original construction in a pressurized water reactor (PWR). After simulating the original construction piping joint, the outer half of the DMW is excavated and repaired with Alloy 52M weld metal to simulate a partial arc EWR. The FEA performed simulates the EWR weld bead sequence and applies three-dimensional (3D) modeling to evaluate the weld residual stresses. Bi-directional weld residual stresses are also assessed for impacts on the original construction DMW. The FEA predicted residual stresses follow expected trends and compare favorably to the results of experimental measurements performed on the mockup. The 3D FEA process presented herein represents a validated method to evaluate weld residual stresses as required by ASME Code Case N-847 for implementing a partial arc EWR, which is currently being considered via letter ballot at ASME BPV Standards Committee XI.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A076. doi:10.1115/PVP2016-63849.

Weld residual stress simulations have become an essential tool in structural integrity assessments. In piping, two dimensional axisymmetric simulations generally give good estimations of residual stresses but clearly cannot capture the three dimensional nature of the welding process: the start/stop effects and the constant change in mechanical restraint during a weld pass. In this study, three dimensional welding simulations have been carried out for piping butt welds, first on a dissimilar metal weld in a thin-walled pipe and second on a narrow gap weld in a thicker stainless steel pipe. The effects of mechanical boundary conditions and start/stop positions have been investigated and stress fields are shown to markedly deviate from axisymmetry. As an illustration, a fracture mechanical analysis of a partial internal surface crack reveals noticeable changes in critical crack sizes.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A077. doi:10.1115/PVP2016-63879.

The presence of high magnitude residual stresses in welded components causes material degradation, local yielding and plastic deformation. Their presence provides the potential for premature failure and compromises the integrity of a structure. This paper presents a review of work carried out to ascertain the residual stresses present within T-section specimens, made from ferritic steel, in their as-welded condition. The standard and incremental deep hole drilling (DHD and iDHD) techniques, the neutron diffraction (ND) and the contour method were applied to characterise the residual stresses in the regions in and around the two fillet welds of the specimens and the surrounding parent material within which the balancing residual stresses needed to be measured. The results of these measurements are presented and compared to highlight agreements and discrepancies in the measured residual stress distributions using these different techniques. A compendium of measurements at a similar location in various T-sections and their comparison with the BS7910 standard show that the measured longitudinal distributions are similar despite the observed scatter. Finally, this paper briefly attempts to investigate and discuss the technical challenges identified when applying the contour method to complex geometry components. The constraint of the specimen during the wire electro-discharge machining (EDM) process, the quality of the wire EDM cut made and the analysis of the raw data for the conversion into residual stresses directly affect the accuracy of the contour method results. The identification and investigation of these challenges lead to continuous improvements of the contour method procedure and reduce uncertainties of the measurement.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A078. doi:10.1115/PVP2016-63880.

Girth welded pipes such as those located offshore on platforms in the North Sea are subjected to highly corrosive environment. The need to consider welding residual stresses in the assessment of the fitness for service and damages to these pipes when investigating local corrosion damages across a welded region is therefore important for the operators of the platforms and the manufacturers of the pipes.

This paper presents a review of work carried out to ascertain the welding residual stresses present within a thin-walled girth welded pipe mock-up made from steel API 5LX Grade 52. The mock-up was manufactured to replicate typical pipes used to convey gas, oil and water through the platforms. The mock-up was of diameter 762mm and of thickness 19mm. The incremental deep hole drilling (iDHD) technique and the contour method were applied to characterize the residual stresses in the weld and heat affected zone of the specimen. The results of these measurements are presented and compared to highlight agreements and discrepancies in the measured residual stress distributions using these different techniques.

Most residual stress measurement methods are limited in terms of their stress and spatial resolution, the number of measurable stress tensor components and their quantifiable measurement uncertainty. In contrast, finite element simulations of welding processes provide full field distributions of residual stresses, with results dependent on the quality of the input conditions available. As measurements and predictions are not often the same, the true residual stress state is therefore difficult to determine. In this paper, through-thickness residual stress measurements are made using the contour and iDHD methods and these residual stresses measured using the iDHD technique are then used as input to a residual stress mapping technique provided within a finite element analysis to reconstruct the residual stress field in the whole specimen. The technique is applied iteratively to converge to a balanced solution which is not necessarily unique. The solution can then be reused for further simulations and residual stress analyses, such as corrosion simulation. Results of the reconstruction are presented here.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A079. doi:10.1115/PVP2016-63940.

Low pressure electron beam welding offers the prospect of large increases in productivity for thick section welds in RPV steels. However, it is important to understand how this welding process affects the structural performance of the completed weld. This paper reviews and presents key results from a programme of weldment manufacture, materials characterisation, residual stress measurements, and finite element modelling of EB welds made in plate of three thicknesses, 30mm, 130mm, and 200mm, and in three steels: SA508 Gr 2, SA508 Gr 3 C1 1, and SA516 Gr 70.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A080. doi:10.1115/PVP2016-63941.

The mission of the NeT European collaborative network is to develop experimental and numerical techniques and standards for the reliable characterisation of residual stresses in structural welds. NeT achieves this by conducting parallel measurement and prediction round robins on closely controlled and well characterised benchmark weldments. NeT TG6 follows on from the successful TG1 and TG4 benchmarks, which both examined welds in AISI 316L material. NeT TG6 examines an Alloy 600 plate containing a three pass “slot” weld made with Alloy 82 consumables. This paper describes the NeT TG6 project as a whole, and presents preliminary materials characterisation, residual stress measurement, and residual stress modelling results.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A081. doi:10.1115/PVP2016-63951.

A bottom mounted instrument (BMI) nozzle in a European nuclear pressure vessel experienced some cracking caused by primary water stress corrosion cracking (PWSCC). Rather than repair the nozzle a decision was made to place a plug over the nozzle hole since this particular instrument panel could be removed in the reactor. The plug repair will be installed remotely during an outage and consists of performing the welding from a location near the top of the vessel which is more than 20 meters away. This makes the design of the repair critical.

This paper uses 3D computational weld analyses to design the repair. Considerations for the design include the possibility of repair welds, proper seating of the plug (distortion control), weld residual stress control, and load applications to the plug during the repair process. Computational weld modeling permitted a proper design to be realized and helped in the regulatory certification process for the plug repair. These are discussed in this paper.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A082. doi:10.1115/PVP2016-63957.

The computer simulation of multiple layering of welds is necessary to determine the distortion and residual stresses arising from the welding process. The welding simulation requires thermal and structural solutions, which are usually carried out in two simulations. Once solved, the thermal transient model temperature results are read in to the structural model to solve for component stresses.

This paper describes the application of the Abaqus Weld Interface (AWI) plug-in for 2D axisymmetric simulation of the residual stresses generated in a Dissimilar Metal Weld (DMW) nozzle to pipe joint comprising of an Alloy 600 nozzle girth weld to a 316 LN Stainless Steel (SS) safe-end pipe. The test piece was manufactured for an ongoing programme within Rolls-Royce PLC. A mechanised Tungsten Inert Gas (TIG) welding process was employed depositing 83 weld bead passes. The weld filler material was Alloy 82.

The AWI Graphical User Interface (GUI) simplifies and saves a large amount of time towards generating the Finite Element Models (FEMs). By using the AWI plug-in within Abaqus/CAE, the FEMs take approximately a month to generate and solve with significant time savings associated with setting up the surfaces for the welding bead sequences and matching the heat input to the actual specimen. The GUI rapidly creates both the thermal and structural input files for the Abaqus/Standard solver. However, modifications were made to the thermal and structural FEM model input files to suit the analysis pre-processing requirements for idealised conditions to match the test piece pipework conditions.

FEM predictions captured the characteristic through-wall Weld Residual Stress (WRS) profiles measured by Deep Hole Drilling (DHD). The weld shrinkage was under estimated.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A083. doi:10.1115/PVP2016-64003.

A 3-D sequential coupling finite element simulation is performed to investigate the temperature field and residual stress in the dissimilar metal weld of a PWR safe-end and nozzle. Chemical compositions and welding residual stress of the dissimilar metal weld are measured. And residual stress of the welded joint of nozzle and safe-end has been studied, aiming to provide a reference for the fabrication and operation of safe-end and nozzle. The testing results show that the experiment results are consistent with FE results. The FE simulation method can be used for the welding residual analysis of the welded joint. And, the calculating results show that large hoop (S33) and axial (S22) of welding residual stresses are generated in the weld metal. The maximum tensile and compressive stresses of S22 and S33 are all in the weld metal or at the interface of the nozzle and weld metal. Due to the difference in mechanical properties and chemical compositions between the base metal and weld metal, a discontinuous stress distribution is generated across the interface between the weld metal and nozzle.

Commentary by Dr. Valentin Fuster
2016;():V06BT06A084. doi:10.1115/PVP2016-64035.

Computational weld residual stress analyses are commonly evaluated at room temperature in order to validate against weld residual stress measurements, which are conducted at room temperature. However, in addition to weld residual stress produced in the course of manufacturing, plant components are subject to internal water pressure and elevated temperature during operation. The current work explores the changes in weld residual stress state due to the presence of internal pressure and temperature at operating conditions. This paper is a follow-up to earlier work, which presented a numerical finite element simulation of the weld residual stress in a pressurizer surge nozzle full-scale mockup as a part of a broader program of cooperative work on weld residual stress organized by the U.S. Nuclear Regulatory Commission and the Electric Power Research Institute. The analysis is performed using two different constitutive hardening models (isotropic and nonlinear kinematic). Two main effects on the weld residual stress field result from the application of internal pressure and temperature: one is elastic, and reversible, and the other is plastic, which is irreversible. The results indicate that the majority of the change in plasticity, and hence the change in stress, occurs during the initial increase in internal pressure and temperature. Furthermore, the results demonstrate that the additional stress due to operating conditions is largely due to the thermal expansion between the ferritic steel, stainless steel, and nickel-based alloy weld.

Topics: Stress , Nozzles , Surges
Commentary by Dr. Valentin Fuster

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