Plasma Arc Welding (PAW) has been used for many critical applications due to its flexibility, reliability and high weld quality. In this paper, two code compliance plasma arc welding applications in the nuclear industry are discussed. The first application is an innovative welding process using PAW with specially designed dual hot wire feeding system, namely Dual Hot Wire Gas Metal Plasma Arc Welding (GMPAW). The GMPAW process offers unique advantages for high deposition and low dilution weld overlay application. The second application is a remote weld overlay repair from pipe inside diameter (ID) for a highly radiated nuclear component using PAW process with remote machining and NDE capability. In this paper, the benefits and advantages are provided for the aforementioned PAW applications. The versatility of plasma arc welding system configuration, as well as high quality and productivity can make plasma arc welding a good candidate for many critical code compliance applications.

Preventing failure due to corrosion poses a challenge to the oil and gas industry. A cost-effective way to prevent such failures is the application of corrosion-resistant nickel-based weld overlays using arc welding processes. Previous research performed at The Ohio State University indicates low heat input GMAW processes, such as cold metal transfer (CMT), produce weld overlays which corrode up to ten times slower than overlays produced with cold wire GTAW [1, 2], with up to ten times higher deposition rates [3]. However, formation of lack of fusion and lack of penetration defects has been a major concern related to the widespread application of low heat input GMAW processes in the industry. In this study, optimal windows of CMT welding parameters for producing defect-free welds were established using a design of experiment approach. CMT weld overlays were compared with hot wire (HW)-GTAW overlays currently used in industry with respect to bead characteristics, microstructure, and process capability. As compared with the HW-GTAW process, the CMT process produced weld overlays with up to four times lower dilution, seven times smaller interdendritic arm spacing, and four times higher deposition rates. Additionally, average heat affected zone and fusion boundary hardness values in the CMT overlays were below 248 HV 0.1 and may not require the post weld heat treatment specified by NACE MR0175.

The existing Class A metallic materials qualified in the ASME Section III, Division 5 rules for high temperature nuclear reactors do not have optimal corrosion resistance for some reactor coolants such as liquid lead, lead/bismuth eutectic, and molten salts which is a major constraint on long life designs. A near term solution to this limitation is the use of cladded components — overlay the Class A materials with a thin layer of some corrosion resistant material. However, this necessitates the development of a design method for cladded components without requiring long-term testing of clad materials in order to support the near-term deployment of these advanced reactor systems. In two other PVP papers [PVP2020-21469, PVP2020-21493], the development of such design method along with a complete set of design rules and clad selection criteria are presented. However, the developed design rules and clad selection criteria are based on a priori assumption of perfect bonding between the clad and base materials. In practice the clad/base debonding may occur before the end of the design life of the component. Therefore, this paper focuses on developing a general acceptance test for checking whether the mechanical integrity of the clad/ base metal interface will be retained till the end of the design life. This work proposes to perform temperature cycling tests on cladded buttons of 12.7 mm diameter and a few mm thickness. A simple analytical expression is provided to determine the temperature range for the temperature cycling tests so that the clad/base interface shear stress intensity at the edge of the cladded button mimics the maximum interface shear stress intensity experienced by the component during the design transients. The paper also provides a method to determine the clad/base interfacial shear stress from finite element structural analysis of components.

The ASME Code, Section XI is working on guidance for application of weld overlay repairs to repair thermal fatigue cracking in nuclear piping systems. This new guidance will eventually be published as Code Case N-894 with the original technical basis in PVP2019-93360. Weld overlays have been extensively used in boiling water reactors (BWRs) and pressurized water reactors (PWRs) to mitigate stress corrosion cracking (SCC). The weld overlays applied to date mitigate SCC by putting the flaw into compression and they use materials that are resistant to SCC. Mitigation of thermal fatigue requires the crack to be in compression so that it does not achieve tensile cycling under the thermal fatigue loading condition. Code Case N-894 allows for the use of either stainless steel or nickel alloy filler metals to repair thermal fatigue flaws. This paper will evaluate the use of both filler metals for the weld overlay process to assess the performance difference between the two filler metals such that the welding advantages of stainless steel over nickel alloys can be quantified. Specifically, this paper assesses the residual stress state difference for nominal sized weld overlays on a six-inch pipe. Various cyclic thermal loading conditions are postulated, and the stress intensity factors are determined for both filler metals to assess the difference in mitigation of thermal fatigue flaws.

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 structural weld overlays (SWOL) using the methodology of Code Cases N-504-4 and N-740-2. The SWOL is a good choice for repair of thermal fatigue cracks in piping 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 N-740 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 environment, steady state operating stresses, residual stresses from welding and fabrication processes, and operating 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 predictable as SCC degradation. In addition, weld overlays applied to address SCC are constructed of SCC resistant material but are not resistant to thermal fatigue. Therefore, ASME Section XI recognized that alternative rules were 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 Code Case was originally drafted based on the industry accepted rules of Case N-504-4 and Appendix Q but includes appropriate modifications needed to address thermal fatigue cracking. These modifications include removing restrictions such as the delta ferrite limit for PWRs that is only applicable to address SCC in BWR environments, 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 document the technical basis for Code Case N-894, which is currently still under development by ASME Section XI.

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.

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.

Weld overlay is a repair method that has been used over the last 10 years to limit the growth of bulges and to extend the remaining life of delayed coking drums. Different refinery operators have used varied approaches, ranging from localized patches on specific regions of concern, to bands along circumferential welds, to large sections of structural repair that completely cover a bulged area. The authors have observed the evolution of regions repaired with internal weld overlay on 18 drums over periods of 5 and 8 years. A comparison of bulge sharpness and bulge depth on a year-over-year basis is presented to measure the effectiveness of metal weld overlay and how the overlay impacts continued distortion of the vessel. Furthermore, factors such as the taper ratio in the transition zones, and the distance between the peak of a bulged area and the edges of the weld overlay, are presented as key parameters that affect the likelihood of cracks developing along the transition zones at the upper and lower edges of the repair.

Cracking on metal surface including Stress Corrosion Cracking (SCC) is a concern for nuclear power plant reactor system piping and components. Several repair techniques have been developed over the years that help address the cracking issues. In the United States, some crack remaining repair technologies such as Weld Overlay (WOL) technique have already been developed and also some technologies such as Excavate and Weld Repair (EWR) are still being actively discussed. Fitness-for-Service (FFS) codes in Japan also define some repair methods which allow repairing the components without removing the cracks. This paper introduces three repair methods which are described in the Fitness-for Service code in Japan, the WOL, the cap repair technique for Bottom Mounted Instrumentation (BMI) nozzles and seal welding. WOL is a repair method that maintains structural integrity by depositing Intergranular Stress Corrosion Cracking (IGSCC) resistant weld to the outer surface of cracked circumferential welded joints in primary piping made of austenitic stainless steel. The cap repair technique for BMI nozzles is a method of repairing from the outer surface when leakage due to SCC occurs. The cap develops a new boundary outside the BMI nozzle to prevent the leakage by retaining the leakage inside the cap. The seal welding technique is also a repair method to cover the cracks by welding to prevent propagation of the cracks and to prevent leakage of reactor water. Provided the cracks are adequately managed and evaluated, these crack remaining repair methods are very beneficial.

Coke drums are massive pressure vessels used in the oil and gas industry, which are subjected to demanding cyclic thermomechanical loading. Such conditions generate severe plastic deformation that leads to bulging and cracking during service due to low-cycle fatigue. The cracks are often repaired in programed maintenance shutdowns, and the repair procedures can be significantly different depending on the organization and failure characteristics. In this work, two types of weld repairs are evaluated after six months of service: (1) a full-excavation crack weld repair, and (2) a partial-excavation crack weld overlay repair. The repair welds were executed on a 1.25Cr-0.5Mo steel coke drum after 20 years of service. This work evaluates the microstructure of the base material, the fabrication circumferential welds, and the weld repairs. The results show that, after prolonged exposure to elevated temperatures, the originally normalized and tempered heat-treated steel microstructure has undergone severe aging. Thermodynamic equilibrium and para-equilibrium calculations were used to determine the carbide formation sequence, and SEM analysis was used to characterize the carbides present on the as-fabricated and aged microstructures. Analysis of the repaired regions did not reveal new cracks.

Delayed coker drums are unique in hydrocarbon processing facilities in that estimating their true design and service life has been problematic. Generally, pressure containing equipment in these facilities is designed using the notion of design life based on required pressure thickness and corrosion allowance considerations. Hence, pressure containing equipment is routinely monitored by facility inspectors for wall thickness. Although many analysts have ascribed coke drum failure to “thermal stress cycling”, the difficulty posed by the operation of coke drums results in an inability to measure or calculate the magnitude of the thermo-mechanical “stresses” and the actual number of significant exposures, that is, cycles causing fatigue damage. As well, the use of Code construction practices has been generally misapplied, for this specific equipment, as the practices are intended to define a safe design life rather than a service life. Indirect measures of service life based on shell bulge severity have fallen from favor by being ineffective. A trend to use a strain index method is somewhat more appealing but is based on static load and monotonic material property considerations rather than those properties indicative of thermal cyclic operation. Recent work has shown that thermo-mechanical strain cycling can be characterized quantitatively and used to determine a cyclic service life for both undamaged and damaged coke drums. This paper discusses some of the engineering specifics to generate a high probability estimate of coke drum fatigue service life for a new drum, a damaged-stable drum, drums with weld overlay and for drums exhibiting incremental damage.

For several decades the nuclear industry has used structural weld overlays (SWOL) to repair and mitigate cracking within pressurized water reactor (PWR) components such as nozzles, pipes and elbows. There are two known primary mechanisms that have led to cracking within PWR components. One source of cracking has been primary water stress corrosion cracking (PWSCC). Numerous SWOL repairs and mitigations were installed in the early 2000s to address PWSCC in components such as pressurizer nozzles. However, nearly all of the likely candidate components for SWOL repairs have now been addressed in the industry. The other cause for cracking has been by fatigue, which usually results from thermal cycling events such as leakage caused by a faulty valve close to the component. The PWR components of most concern for fatigue cracking are mainly stainless steel. Thus, ASME Section XI Code Case N-504-4 would be a likely basis for SWOL repairs of these components, although this Code Case was originally drafted to address stress corrosion cracking (SCC) in boiling water reactors (BWR). N-504-4 includes the requirements for the SWOL design and subsequent analyses to establish the design life for the overlay based on predicted crack growth after the repair. This paper presents analysis work performed using Code Case N-504-4 to establish the design life of a SWOL repair applied to a boron injection tank (BIT) line nozzle attached to the cold leg of an operating PWR. The overlay was applied to the nozzle to address flaws found within the stainless steel base metal during inservice examination. Analyses were performed to calculate the residual stresses resulting from the original fabrication and the subsequent SWOL repair. In addition, post-SWOL operating stresses were calculated to demonstrate that the overlay does not invalidate the ASME Section III design basis for the nozzle and attached pipe. The operating and residual stresses were also used for input to a fatigue crack growth (FCG) analysis in order to establish the design life of the overlay. Lastly, the weld shrinkage from the application of overlay was evaluated for potential impact on the attached piping, restraints and valves within the BIT line. The combined analyses of the installed SWOL provide a basis for continued operation for the remaining life of the plant.

An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature ( T o ) of specimens that were aligned with the middle section of a cladding strip (HAZ MCS ) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZ OCS ). T o values of partial area in the HAZ were obtained using Mini-C(T) specimen. The T o values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. T o values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.

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.

Several studies recognized that cracks in delayed coke drums resulted from low cycle fatigue induced by cyclic thermal stress [1], [2], [3]. According to a coke drum survey coordinated by API in 1996 [1], there are two different areas where cracks are produced. The first zone is located at the shell to skirt weld, and the second at the bulged areas found in the cylindrical section. In the second case, from 145 coke drums 57% reported that had shell bulging problems. Of the drums that bulged, 87 % also showed cracks. Recently, it has been reported the use of a novel weld repair procedure on bulged sections of a drum. In this repair, the bulge is overlaid with weld metal on the inside or outside on the bulge depending on the bulge shape. It has been reported that this repair procedure can stop further bulging on the shell, but detailed information about its influence has not yet been published. Finite element analysis of several bulged patterns that were identified from some laser mappings are used to compare the level of the stress after a weld overlay repair is made. The study was carried out running a sequentially-coupled thermo-mechanical analysis. The assessment shows the influence of the thickness and the extent of the weld overlay on the level of stress on bulged coke drums. The results indicated that depending on the initial bulged shape this repair method either reduces or increases the level of the stress. When an inward deformation pattern is observed, an external reinforcement is recommended; however, when an outward deformation pattern is developed in a coke drum, an external weld overlay repair is not recommended.

Jagged cracks were observed in SA240 Type 405 stainless steel cladding of Inconel 625 overlay repaired coke drums. It is found that intergranular cracking is the dominant fracture mode in the fine-grained heat-affected zone (FGHAZ) of the boat specimens. The sensitization effect from the operation and welding thermal cycles leads to the depletion of Cr with the preferential precipitation of Cr-rich M 23 C 6 carbides along the grain boundaries. The cladding FGHAZ has the largest frequency of grain boundaries with higher local strain levels and the highest fraction of grain boundary Cr-rich M 23 C 6 carbides. Thermal stress distributions predicted by finite element analysis clearly show the maximum shear stress to exhibit the typical “jagged” pattern near the cracked regions. Thermal expansion coefficient and strength mismatch among the shell base metal, cladding, and overlay is believed to have caused the unique jagged maximum shear stress distribution in the cladding HAZ of Inconel 625 overlay. The magnitude of this thermal stress can reach the yield strength of the cladding at 900 °F (482 °C) service temperature, therefore, provides the driving force for the jagged cracking formation in the sensitized HAZ.

In the present study, the plastic limit loads of the complex-cracked pipes, which have 360° internal surface crack along the circumference of pipe together with partial through-wall crack, are investigated via detailed 3-dimensional (3-D) finite element (FE) limit analyses. In terms of geometry of the complex-cracked pipe, the Alloys 82/182 butt weld that has welds overlay using the Alloys 52/152 for mitigating primary water stress corrosion cracking (PWSCC) in the nuclear pipe is considered. In the present FE limit analyses on complex-cracked pipes with welds overlay, the effect of strength mismatch between Alloys 82/182 and Alloys 52/152, depth of 360° internal surface crack and length of partial through-wall crack on plastic limit loads of complex-cracked pipes is systematically evaluated. As for loading conditions, axial tension, global bending and internal pressure are employed. The resulting FE plastic limit loads of complex-cracked pipes with welds overlay are also compared with the existing solutions, and finally the applicability of simple engineering estimate using the single material property based on the weighted average concept instead of using the properties of Alloys 52/152 and Alloys 82/182 to the prediction of the plastic limit loads of complex-cracked pipes with welds overlay is pursued.

9Cr-1Mo-V steel, which was developed for application as a steam generator for fast breed reactors in the 1970s, has a higher strength at high temperatures and has been used for equipment and piping systems in the fossil power industries. ASME, Section VIII, Division 2 [1] gives 9Cr-1Mo-V steel a maximum design temperature of 649°C and an operating temperature of 500°C. And it has higher allowable stresses at 450°C or over, compared to 2¼Cr-1Mo-V steel. Therefore, if this material can be used, more economical pressure vessels operating at 454–500°C can be designed and manufactured. In our previous study for base metal, a large forged shell ring of 9Cr-1Mo-V steel was manufactured and for base metal welding, cracking susceptibilities and weldability were investigated. For overlay, welding consumables with high resistance to sigma phase embrittlement were developed [2]. In this study, highly efficient welding consumables for tandem SAW designed for circumferential welding of heavy wall shells were developed and welding using the full-scaled shell ring was demonstrated, and then the mechanical properties of the weld metal were evaluated. Results indicated that, regardless of the weld thickness, a minimum of 8 hours postweld heat treatment (PWHT) at 745°C was required to meet hardness and toughness requirements for conventional reactors. The strength of the materials can comply with the Code requirements after 3 cycles of PWHT considering the PWHT in fabrication and after weld repair. Furthermore, the following new Code Cases and Code revision were proposed and approved in order that pressure vessels can be designed in accordance with ASME, Section VIII, Division 2. • New Code Case for Fatigue Evaluation • New Code Case to apply SA-336-F91 in ASME Section VIII, Division 2 • Revision of Table 5A to add SA-336-F91 As a result, it has become possible to design and manufacture refining reactors to operate at 454–500°C.