Temper bead (TB) welding is routinely used as an alternative to post weld heat treatment (PWHT) when performing weld repairs on low alloy steels in the nuclear power industry. Historically, TB welding procedures have been qualified by Charpy V-notch (CVN) testing to demonstrate acceptable impact properties in the weld heat affected zone (HAZ). Part 1 of this work investigated a hardness drop criterion to demonstrate proper HAZ tempering using a vintage heat of SA-508 forged material (see PVP2019-93950). Optimization of hardness test parameters was also a major piece of the Part 1 scope of work. Part 2 of this work adds a more modern heat of SA-508 material and two heats of A-387 F22 material to the study. The overarching goal of this work is to support an alternative hardness test protocol for qualification of temper bead welding (see PVP2015-45663).

The objective of this work was to develop a procedure for evaluation and quantification of the tempering efficiency of corrosion resistant weld overlays used in the power generation and oil and gas industries. Three two-layer weld overlays of Alloy 625 on Grade 22 steel plates were produced using GTAW cold wire procedures. Typical welding parameters corresponding to low, medium, and high heat input were utilized. The overlays consisted of nine beads on the first layer and five to seven beads on the second layer. The weld thermal histories experienced in the coarse-grained heat affected zone (CGHAZ) were measured with Type K thermocouples and recorded with a 55 Hz sampling rate. Two rows of seven thermocouples were used in each overlay: one row located in a mid-bead position beneath the center bead of the overlay and the other row located in the nearest bead overlap position. Additionally, one Type C thermocouple was plunged into the weld pool of a second layer weld bead. The acquired thermal histories and the CGHAZ hardness at the thermocouple locations were evaluated to quantify the tempering efficiency in each welding procedure. The weld thermal histories with peak temperatures between 500°C, assumed as the minimum tempering temperature, and the base metal A C1 temperature were considered as tempering thermal cycles. The number of tempering thermal cycles and the sum of tempering cycle’s peak temperatures in each thermocouple location, as well as the corresponding hardness were used to quantify the tempering response efficiency for each of the three welding procedures. The results of this study will be used for validation of a computational model-based approach for prediction of tempering response and optimization of temper bead welding procedures.

Temper bead (TB) welding is often used as an alternative to post weld heat treatment (PWHT) for repair of pressure vessels and piping in the nuclear power industry. Historically, qualification of TB welding procedures has employed the Charpy V-notch test to ensure acceptable heat-affected-zone (HAZ) impact properties. The 2004 Edition of ASME Section IX provided a new provision in QW-290 that allows temper bead qualification using a peak hardness criterion. The peak hardness provision is appropriate for industries such as oil and gas, where peak allowable hardness is specified to ensure adequate resistance to sulfide stress cracking in sour service environments. However, a peak hardness criterion is not appropriate where impact properties are specified for resistance to brittle fracture during low temperature conditions that can occur during certain postulated accident scenarios at a nuclear power plant. Work at the Electric Power Research Institute (EPRI) and The Ohio State University (OSU) show that a hardness drop protocol can be used to demonstrate acceptable impact properties in the HAZ of a temper bead weld. This paper presents a quantitative correlation between hardness measurements and HAZ microstructures with presumed optimum impact properties using a hardness drop approach. The overarching goal is to develop a hardness test protocol for temper bead weld procedure qualification for applications where impact properties are specified.

Since 1982 the nuclear industry has employed weld overlay repairs to address intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWR) and primary water stress corrosion cracking (PWSCC) in pressurized water reactors (PWR). The American Society of Mechanical Engineers (ASME) has created several documents to provide rules and guidelines for weld overlay repair of nuclear components that have experienced stress corrosion cracking (SCC). These documents include ASME Code Case N-504-4 and ASME Section XI, Nonmandatory Appendix Q which specifically address weld overlay repair of stainless steel components. Recently, stainless steel components that have experienced thermal fatigue cracking at the inner diameter surfaces have been repaired with weld overlays using the methodology of Case N-504-4 and Appendix Q. This repair technique is appropriate to address thermal fatigue cracking because it provides structural reinforcement to the affected location and places the inside diameter (ID) surface into compression preventing, or significantly reducing, further flaw growth. However, the rules of Case N-504-4 and Appendix Q were not specifically written to address thermal fatigue cracking as the primary cause and may not adequately address design, analysis and examination requirements when thermal fatigue is the active mechanism because it is very different in nature than SCC. For example, SCC is driven by a combination of steady state operating stresses, residual stresses from welding and fabrication processes, and temperature, whereas thermal fatigue is driven by thermal stress cycles resulting from fluid thermal cycling or stratification. The source of the thermal events that result in cracking may not be as well understood or predicable as SCC degradation. Therefore, alternative rules are needed for repair of piping damaged by thermal fatigue. This paper provides a technical basis for weld overlay repair of components that have experienced thermal fatigue cracking. It addresses design, analysis and examination requirements considering the nature of thermal fatigue in nuclear piping systems. The basis begins with the industry accepted rules of Case N-504-4 and Appendix Q and discusses the appropriate modifications needed to address thermal fatigue cracking. These modifications include removing restrictions such as the delta ferrite limit that is only applicable to address SCC, and enhancements to the examination requirements to ensure that the repaired location is adequately monitored throughout the remaining service life of the plant. The purpose of this paper is to outline the basis for a new ASME Code Case that will contain the appropriate rules for weld overlay repair of Class 1, 2 and 3 stainless steel piping degraded by thermal fatigue cracking. The new Case is currently in draft form and the requirements and specific details are still evolving. Thus, it is envisioned that this technical basis will be revised to include updates and revisions to the Case.

ASME Section XI Nonmandatory Appendix Q and Code Case N-504-4 are routinely used to install full structural weld overlays in the nuclear power industry for repair or mitigation of stress corrosion cracking in austenitic stainless steel weldments. Both Appendix Q and N-504-4 specify a Ferrite Number (FN) and carbon content requirement for the stainless steel weld metal used for the weld overlay to ensure acceptable resistance to stress corrosion cracking. The Ferrite Number (FN) is used in the ASME Code for establishing the delta ferrite content in the deposited weld metal. Field experience indicates there is often confusion and differing opinion concerning how the Ferrite Number and carbon content requirements of Appendix Q and N-504-4 are satisfied. This is in part due to unavailability of the original technical basis for these requirements. This paper provides a background for the delta ferrite and carbon content requirements, information on influence of delta ferrite and carbon content on stress corrosion cracking and U.S. Nuclear Regulatory Commission (NRC) guidance on the issue. Finally, this paper details a proposed revision of Nonmandatory Appendix Q and N-504-4 to clarify the FN and carbon content requirements.

A weld overlay (WOL) following the general guidelines of ASME Code Case N-740-2 was successfully installed in March of 2018 on a dissimilar metal weld joining the super emergency feedwater 1 piping to steam generator (SG) at the Dukovany nuclear power plant (NPP) in the Czech Republic. The repair was necessary due to stress corrosion cracking detected in the super emergency feedwater nozzle to safe-end dissimilar metal weld. This was the first WOL installed in the Czech Republic and represents a significant step towards further acceptance of this proven repair technology in Europe. The WOL repair approach was accepted by the Czech regulator, and two different inspection agencies, following successful mockup demonstrations, welding procedure qualification, nondestructive examination demonstrations and weld residual stress analyses. This paper describes the preparatory work as well as field deployment of WOLs in the Czech Republic.

Cathodic protection (CP) is one of the primary methods to protect buried piping and pressure components from corrosion and is a critical element in asset management of buried piping at nuclear power plants. Implementation of cathodic protection requires non-structural attachments to the buried piping for electrical leads and connections. The method of attaching copper-copper alloy CP leads to carbon steel piping and components using traditional arc welding processes can be difficult and time consuming. A two-step process is frequently used where a carbon steel weld tab is first welded to the pipe or component by a traditional arc welding process. The copper-copper alloy CP lead is then joined to the carbon steel weld tab by the exothermic welding process. An alternative to this cumbersome two-step process is pin brazing which is an automatic brazing process that uses electric current resistance to heat the interface between a pin capsule and the component. An arc between the pin capsule and the outside surface of the electrical connector is then used to melt the capsule or pin that contains the brazing filler metal. The process is similar to stud welding in that the brazing pin is loaded into an automatic pistol and the brazement is made when the trigger is pulled. ASME Section XI Code Case N-882 delineates rules and requirements for application of pin brazing on Class 2 and 3 pressure boundary components. This paper provides the background and description of the pin brazing process with a summary of the technical basis for Case N-882.

Weld overlay (WOL) is a popular repair technique to mitigate stress corrosion cracking (SCC) in dissimilar metal weld (DMW) in U.S. pressurized water reactor (PWR) design. The WOL technique is being considered as a SCC mitigation technique for DMW in Russian water-water energetic reactor (WWER or VVER) design. A WOL mockup on a VVER super emergency feedwater nozzle DMW has been fabricated, which represents the first WOL on VVER with the goal to mitigate SCC and the first WOL in Czech Republic civilian nuclear power plants. This paper presents the two- and three-dimensional finite element analyses performed to assess the weld residual stresses in the WOL mockup. The analysis evaluates the stress distribution and changes in the DMW before and after the WOL application, as well as compares the results to established industry guidelines and comparable WOLs on U.S. PWR.

This paper presents a finite element analysis (FEA) based approach to perform crack growth evaluation of remnant cracks in a mockup with a partial arc excavate and weld repair (EWR). 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). To create a representative partial arc EWR application, the outer half of the DMW is excavated and repaired with Alloy 52M weld metal. The crack growth evaluation process presented herein represents an advanced method to evaluate the Alloy 82/182 remnant crack growth 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. After the repair, any crack that remains in the Alloy 82/182 remnant and underneath the EWR needs to be evaluated for stress corrosion cracking (SCC) to assess its potential to grow beyond the EWR coverage area. Conventional fracture mechanics approach may not be suitable to evaluate such a remnant crack because of its close proximity to multiple materials of different mechanical properties and unconventional crack shape. In the crack growth evaluation, a crack that is reminiscent of a circumferential crack in a pipe, and a crack that is reminiscent of a laminar crack in a pipe are evaluated to predict the time for each of them to grow beyond the partial arc EWR coverage arc length. It is expected that the approach, analysis steps, calculation procedures presented in this paper will be applicable to analyzing a pipe geometry using realistic residual stresses and operating stresses for an EWR.

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.

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.

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.

In 2004, ASME Section IX added maximum hardness essential variables for temper bead procedure qualification when impact testing is not specified or required by the applicable book section. The assumption with specifying a maximum hardness criterion is that high hardness after temper bead welding indicates inadequate tempering. As discussed in PVP2013-97793 [2], imposing a maximum hardness criterion for temper bead qualification can actually lead to acceptance of a heat affected zone (HAZ) microstructure with less than optimum impact properties. In fact, depending on the weld HAZ microstructure the impact properties can vary widely from very low to very high at the same hardness. This paper describes an alternative hardness test protocol for temper bead procedure qualification. Rather than using a single maximum hardness acceptance threshold, this new test protocol characterizes the base material response to temper bead welding by determining the maximum achievable hardness with a bead-on-plate test and with a hardness calculation. Research shows that a high hardness in the HAZ prior to depositing the tempering weld layers provides the optimum microstructure for achieving desired HAZ impact properties. With proper tempering the HAZ hardness is reduced below the maximum achievable hardness. Temper bead procedure acceptance is thus determined by the drop in HAZ hardness after depositing the temper bead weld layers. Application of this new hardness test protocol for temper bead qualification is proposed as an alternative to a single maximum hardness acceptance criterion.

Predicting weld dilution for machine gas tungsten arc welding (GTAW) is a challenge due to the number of variables associated with the welding process. Proper heat input and power ratio controls are critical in many welding applications to control weld dilution, such as for dissimilar metal welds where low weld dilution is necessary to prevent solidification cracking or for cladding where weld dilution is minimized to maintain corrosion resistance of the clad material. This paper discusses the preliminary development and validation of improved weld dilution, heat input, and power ratio equations for the GTAW process. The new equations incorporate power added for the hot wire GTAW process, filler metal material properties, and the heat used to melt the filler metal when added to the GTAW process. The weld dilution equation was validated by comparing calculated dilution values to measured values from bead-on-plate weld trials performed on a variety of filler metals and substrates. Results of the testing and validation along with limitations of the new equations are discussed.

Temper bead welding is used routinely for weld repair on low alloy pressure vessel steels in the nuclear power industry. Temper bead procedure qualification is contingent on demonstration that the welding process does not degrade the mechanical properties, including fracture toughness, in the base metal weld heat affected zone (HAZ). Historically for temper bead qualification acceptance, adequate HAZ properties have been verified by tensile, impact, and bend testing, while hardness criteria has not been specified. In fact, temper bead welding has been successfully applied for welding on low alloy steels without any hardness criteria for many years. In 2004, ASME Section IX added hardness testing for temper bead procedure qualification when impact testing is not required. The Eurocode, ISO standards, and numerous other European specifications include maximum hardness criteria for general welding procedure qualification and have invoked these same criteria for temper bead procedures. Test results indicate that imposing maximum hardness criteria can actually lead to acceptance of less than optimum fracture toughness in the temper bead weld HAZ due to rapidly changing microstructures in low alloy steels. Impact properties for such microstructures can vary widely even though similar levels of hardness are exhibited. This paper investigates the legitimacy of using maximum hardness criteria to demonstrate acceptable HAZ fracture toughness in low alloy pressure vessel steels.

Temperbead welding is common practice in the nuclear power industry for in-situ repair of quenched and tempered low alloy steels where post weld heat treatment is impractical. The temperbead process controls the heat input such that the weld heat-affected-zone (HAZ) in the low alloy steel is tempered by the welding heat of subsequent layers. This tempering eliminates the need for post weld heat treatment (PWHT). Unfortunately, repair organizations in the nuclear power industry are experiencing difficulty when attempting to qualify temperbead welding procedures on new quenched and tempered low alloy steel base materials manufactured to modern melting and deoxidation practices. The current ASME Code methodology and protocol for verification of adequate fracture toughness in materials was developed in the early 1970s. This paper reviews typical temperbead qualification results for vintage heats of quenched and tempered low alloy steels and compares them to similar test results obtained with modern materials of the same specification exhibiting superior fracture toughness.

Alloy 690 is a 30 wt% chromium base metal with excellent resistance to PWSCC. Alloy 52M (ERNiCr-7A) is weld filler metal that closely matches the composition of Alloy 690 and also exhibits excellent resistance to PWSCC [1]. Alloy 52M is currently used for new nuclear component fabrication and for repair and mitigation of PWSCC in operating nuclear power plants. Unfortunately, industry experience and laboratory testing has shown that Alloy 52M is susceptible to hot cracking when welding over some austenitic stainless steel base material compositions. Currently there is no clear understanding of the specific composition levels that increase the susceptibility to hot cracking. This paper summarizes preliminary results of an EPRI study on hot cracking susceptibility Alloy 52M when welded on a range of austenitic stainless steel samples [2]. The focus of the EPRI study was to identify high risk base metals with compositions that promote hot cracking and provide counter measures to mitigate or lessen the propensity for hot cracking.

Dissimilar metal welds of filler metal 182 (ENiCrFe-3) in the primary loop of pressurized water reactor (PWR) nuclear plants are susceptible to primary water stress corrosion cracking (PWSCC) after decades of service. Repair or mitigation has been routinely accomplished by installing a structural weld overlay (SWOL) on the filler metal 182 weld joint with the more PWSCC resistant filler metal 52M (ERNiCrFe-7A). The typical dissimilar metal joint consists of a low alloy steel vessel nozzle welded to an austenitic stainless steel safe end. The SWOL extends from the low alloy steel nozzle over the safe end and most often onto the adjoining wrought or cast stainless steel pipe. Field experience shows that filler metal 52M is susceptible to hot cracking when welding on certain heats of centrifugally cast stainless steel piping. This report evaluates 52M hot cracking when welding on CASS piping and provides the likely cause and mechanism for the cracking. The synergistic influence of silicon (Si) and sulfur (S) elements on the weld bead shape and dilution that leads to hot cracking is investigated. In addition, studies on the influence and use of the gas tungsten arc welding (GTAW) power ratio parameter for 52M overlays are presented.

Nickel alloys with high chromium content provide optimum resistant to stress corrosion cracking for service in the reactor coolant system of commercial nuclear power plants. High chromium nickel-base alloys however present many challenges, such as less than ideal weldability and susceptibility to solidification cracking or solid-state cracking depending on welding conditions and dilution effects with dissimilar metals. Moreover, the presence of large solidification grains, typical of nickel alloy weld metals, makes ultrasonic examination of the weldment difficult. Magnetic stirring of the nickel alloy weld pool has the potential to address these challenges and improve joining, overlay welding, cladding, and repair of critical components in commercial nuclear power plants. This study evaluates use of magnetic arc stirring to modify weld pool solidification conditions in order to promote a fine solidification grain structure in nickel alloy welds.

Section III of the ASME Boiler and Pressure Vessel Code requires radiographic testing (RT) of pressure boundary welds. RT is performed to detect flaws that might be created in welds as they are fabricated. Current Section III acceptance standards require rejection and repair of flaw indications characterized as cracks, lack of fusion, or incomplete penetration regardless of the size of the indication or the structural significance of such indications on fitness for service (FFS). The current Section III requirements have been effective in meeting the design objective of preventing pressure boundary failures. However, the rules are sufficiently conservative that not only are structurally significant flaws excluded, but they also exclude more benign indications that have no impact on structural integrity. This approach has resulted in repairs for even minor flaws that have no FFS impact. In addition to the cost of performing these unnecessary repairs, the repairs may have contributed to service induced cracking because of the higher residual stresses from the repair. Clearly, there is a need to revisit the Section III inspection and repair rules so as to distinguish between structurally unacceptable flaws and benign flaws that have no FFS impact. This paper describes the technical basis for the proposed Section III Code Case that uses the FFS approach to eliminate the need for weld repairs for minor flaws that have been shown to have no structural impact. Specifically, the Code Case will provide the option to use qualified volumetric inspection to size the flaw indications accurately and define acceptance criteria to determine flaw sizes that are judged to have little structural significance. In addition to describing the requirements of the proposed Code Case, this paper also describes the technical basis for the flaw acceptance criteria and the results of ultrasonic (UT) qualification testing to demonstrate the capability to detect and characterize fabrication flaw indications.