Abstract Radiation-induced degradation of reactor pressure vessels and internals is a concern to the aging nuclear fleet and welding solutions will be required if repair of these irradiated components is deemed necessary. However, the weldability of highly irradiated austenitic materials is significantly diminished due to the presence of irradiation induced helium in the material matrix. Helium-induced weld cracking is a complex phenomenon that is related to the concentration of helium, the heat input from the welding process, and stresses generated during cooling of the weld. During conventional high heat input welding methods such as gas tungsten arc welding, helium bubbles can coalesce and grow on base metal grain boundaries within the heat-affected zone which subsequently causes intergranular cracking. The objective of this work was to obtain weldability data by characterizing welds made on highly activated, neutron irradiated Type 304 stainless steel containing both radiation-induced helium and microstructural damage such as void swelling. All irradiated materials welding was performed inside a Westinghouse hot cell utilizing a pulsed Nd:YAG laser with welds made on three rectangular samples of highly activated Type 304 stainless steel. The rectangular samples were cut and milled in-cell from sections previously obtained from two neutron reflector hex blocks. The hex blocks are U.S. Department of Energy owned material and were irradiated for approximately 13 years in the EBR-II sodium cooled fast reactor from 1982 until 1995. The three samples selected for welding have nominal damage doses of approximately 0.4, 11, and 28 dpa with corresponding estimated helium contents of 0.2, 3 and 8 appm helium, respectively. A number of different weld parameter sets were utilized and included variations of travel speed, wire feed speed and lens-to-work distances. The parameter sets allowed for a range of effective weld heat input levels to be compared. Single pass and multiple pass as well as wire fed and autogenous welds were made. This paper presents the results from post-weld evaluations performed on the three welded irradiated samples, focusing on the reduced tendency for cracks to form adjacent to the weld as a function of weld parameters, lens-to-work distance and helium content.

Abstract 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.

Abstract A typical Fluid Catalytic Cracking Unit (FCCU) generates high temperature flue gas in the process of regenerating the catalyst. This flue gas is diverted to a stack after removal of catalyst fines and excess heat using a Waste Heat Recovery Unit (WHRU) or CO boiler. This flue gas line is a large diameter (1.2 m /2 m) piping and is a combination of Hot Wall (bare SS304H piping with external insulation) upstream of Orifice Chamber and Cold Wall (Carbon steel piping with internal refractory lining) for the downstream side. In a major revamp project, large portion of flue gas line was replaced with some dimensional and design changes. A crack was noticed at the SS304H side of hot wall to cold wall transition joint downstream of Orifice Chamber after approximately 2 years in operation. The line operates around 700 °C and 0.15 Bar(g) at the location of the crack. The initial crack was measured to be approximately 250 mm to 300 mm and grew to a full circumference crack in a short time resulting in minor flue gas leaking with catalyst fines. This paper discusses the details on how the issue was addressed on site and a temporary repair (i.e. welding of a box on high temperature piping) was carried out online safely, while the unit remained in operation. Further, the paper presents the root cause assessment and design modifications implemented for hot wall to cold wall transition joint during a scheduled turnaround.

Abstract Coke drums are subjected to severe thermal cycling with the skirt to shell connection weld being vulnerable to fatigue cracking. It is essential this connection is well designed to ensure a long life before repairs are inevitably required. Much has been written on coke drum skirt design with the aim of reducing the thermal stresses and strains encountered at the skirt connection weld, some designs have removed the weld completely allowing the drum to sit in an “egg-in-cup” arrangement. This paper includes a short literature review discussing Coke drum skirt designs and explains skirt behaviour during the drum cycle that results in eventual skirt cracking. A case study is reviewed in detail for a new pair of coke drums, where the predicted fatigue life of the chosen welded connection is assessed using axisymmetric, quarter symmetry and half symmetry finite element analysis supported by thermocouple data. The optimised design focuses on a conventional tangential design where the effects of the essential variables such as skirt thickness, skirt connection location, skirt-to head-gap and slot design (length, location & spacing) have been modelled and optimised to obtain a skirt design that produces the longest fatigue life for the intended duty cycle. Coke drum skirts must be installed onto the shell to exacting tolerances during manufacture to ensure concentricity and minimal gap between the skirt and shell. A brief overview of how this is achieved will be presented.

Abstract Bottom head flange leaks on coke drums are a too common occurrence for coke drum operators. This paper discusses why these complex joints leak, exploring phenomena such as bolt relaxation, flange rotation, bolt hole cracking, plastic deformation of a flange face due to cyclic thermal transients, as well as investigating the effects of gasket stress variation which can lead to gasket movement and distortion during successive drum cycles. The industry trend is to install automatic bottom un-heading devices (BUDs) to facilitate safe, reliable coke removal and to increase production through shortening un-heading operations. A case study is reviewed which shows how a new drum flange, coupled to a BUD, has been optimized using Finite Element Analysis (FEA). The findings show adequate flange thickness and optimized hub dimensions are required to combat plastic deformation of the flange ring and reduce gasket stress variation. Five designs are modelled, varying the flange thickness, outside diameter and hub geometry. Due to the close proximity of a new side nozzle, the full lower section of the drum has been modelled using quarter and half symmetry FEA with applied temperature distributions from each phase of a typical coke drum cycle; heating, coking and quench. This has allowed for nozzle loadings to be evaluated and the location of the flange weld to be optimized to give the greatest fatigue life. An explanation into why periodic re-tightening is required to keep these joints tight is provided along with recommendations on suitable joint assembly techniques using a combination of bolt load verification and alternative bolting patterns from ASME PCC-1.

Abstract A recent increase in operating experience (OE) related to pipe cracking in non-isolable auxiliary piping systems has been realized in the Pressurized Water Reactor (PWR) nuclear power industry. The majority of PWR auxiliary piping systems are comprised of welded stainless steel pipe and piping components. The susceptible piping systems are Class 1 pressure boundary and typically non-isolable from the primary loop. Since they are non-isolable, when a pipe crack or crack indication is identified, an emergent flaw evaluation and/or repair is required. Typically, the evaluations begin with an ASME Section XI IWB-3640 flaw evaluation to determine acceptability of the as-found flaw at the time of shutdown. Subsequent flaw evaluations are performed to demonstrate the possibility of continued operation of the piping component by leaving the flaw as-is without repair. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth evaluations, and the detection capabilities of the non-destructive examination technique. If evaluation of the as-found indication does not produce acceptable results, then a repair/replacement activity per ASME Section XI is considered. Possible repair scenarios include replacement of the piping section or component, or structural weld overlay. The results of the flaw evaluations or repairs must ensure that the auxiliary piping system will continue to operate safely. This paper will discuss the recent experiences of two different piping systems (boron injection tank line and drain line) that experienced cracking, the potential causes for the cracking in the absence of evidence, and the ASME Code flaw evaluations and/or repairs performed to support continued operation of the plant.

Abstract The operation of autoclave vessels at elevated temperature and pressure is common across a variety of industries. Many of the older operating vessels were fabricated from multiple forgings by welding. Austenitic stainless steels are the material of choice for many of these vessels, and precautions to prevent sensitization in the weld heat affected zone may not have been implemented during construction, which can increase susceptibility to environmental cracking conditions. Recently a vessel developed a through wall leak during the operating cycle. Three complimentary vessels at the same facility were inspected resulting in one additional vessel not returning to service and two with similar flaws. A multi-disciplinary approach was used to evaluate the failure, perform a fitness for service determination, and nondestructive examinations to monitor the flaw growth and validate the fitness for service inputs. The overall approach is discussed.

Abstract The integrity of transmission oil pipelines are often managed through in-line inspections (ILI) at regular intervals. For the last two decades, such ILI-based integrity programs along with excavations and field non-destructive testing (NDE) have proven their effectiveness in terms of reliability. In a few cases, some pipes contain; for example, a unique cracking mechanism exhibited by short, deep axial cracks located in the vicinity of girth welds. These attributes pose sizing difficulties for ultrasonic crack ILI tools. Accordingly, operators may lean on supplemental integrity activities to prove the safety of the pipelines such as; but not limited to, hydrostatic testing, laboratory testing of cut-outs, qualitative ranking of features, borehole leak detection analysis, Just-Missed-Flaw (JMF) or Just Surviving Flaw (JSF) analysis, discharge and/or point pressure restrictions, and/or a mix between all the previous techniques. Moreover, it is the operators’ responsibility to evaluate the risk associated with their integrity plans. Hence, it is important to be able to analyze the reliability of such integrity activities quantitatively. This paper presents an event-tree approach which can augment standard ILI or hydrostatic test results and probabilistic analysis with non-ILI integrity measures under one umbrella. In this approach, the likelihood of failure for both leak and rupture modes can be comprehensively estimated. The event tree approach is used herein as an inductive analytical diagram in which failure events are analyzed using Boolean logic to examine a chronological series of subsequent integrity actions and consequences. The proposed approach is also designed to capture subject matter experts’ opinion into the analysis as part of the integrity management program. The work discusses a real practical application along with verification and validation elements of the proposed integrated approach.

Abstract High density polyethylene (HDPE) is widely used in city water and gas piping system and even in nuclear engineering for its excellent properties in flexibility, impact resistance, aging and corrosion resistance, reliable connection, long service life. It is generally believed that the typical failure mode of PE pipe under long-term static pressure load is the slow crack growth (SCG) caused by creep crack initiation (CCI) and creep crack growth (CCG), which is a kind of quasi-brittle failure. Scratching often occurs during PE pipe installation due to the drag of pipe without any protection, which would leave axial cracks as the initial defect. This would harm the long term performance of pressurized HDPE pipe. In the present work, the creep crack kinetics was experimentally determined in term of C * integral, and then CCG behavior of PE pipe with axial semi-elliptical crack was discussed. With the help of a CARE electronic universal testing machine and DIC displacement measuring system, creep tests are performed with smooth bar specimen of PE100 pipe parent material. And the stationary creep law parameters were determined. The crack opening displacement ( COD ) rates were determined by creep cracking test with cracked round bar specimen. With the help of compliance calibration tests, the COD curves were changed into crack propagation curves and the creep crack kinetics was eventually determined in term of C * integral. By assuming an semi-elliptical crack front with nearly the same C * integral, the creep crack propagation was predicted with the determined crack kinetics. And the allowable initial crack size was also determined for the PE pipe discussed. Compared with that determined in terms of SIF, current study yields much margin for the allowable initial crack.

Abstract A hydrogen manufacturing plant experienced circumferential cracking at the dissimilar weld on the outlet header. The outlet header was a cold wall design and the dissimilar weld was between HP40 modified and carbon ½ Mo steels. The resultant failure investigation found the cause to be hydrogen induced cracking of the dissimilar weld at the fusion boundary zone. The hydrogen was generated from the CO 2 corrosion which occurred due to operating the tubes below the dew point and the hydrogen was trapped in the steel by the CO chemisorption onto the steel. The following paper outlines the failure investigation and the fitness for service conducted to maintain the running of the plant.

Abstract Making light water reactor (LWR) components using additive manufacturing (AM) provides a high degree of design freedom to create complex near net shape geometries, with significant reduction in the deployment time. Since most of the current AM research focus on the refining of the fabrication variables, little information exists on the actual corrosion behavior (including general corrosion, localized corrosion and environmentally assisted cracking - EAC) of AM components. The most frequent operational material failure modes in LWR is EAC and debris fretting of fuel rods, therefore the objective of this work was to evaluate the environmental performance of AM type 316L SS in relation to the traditionally melted and forged wrought material. Stress corrosion cracking results show similar behavior between AM and wrought 316L SS in high temperature water. Ambient temperature electrochemical testing showed that the AM 316L SS was slightly more resistant to localized corrosion than the traditional wrought material and that AM material did not suffer sensitization during fabrication or heat treatment processes.

Abstract This paper provides a technical basis for a crack growth rate (CGR) for use in performing evaluations of cracking in stainless steel canister materials. The source of crack initiation and growth is deposition of chloride aerosols on the canister surface followed by deliquescence leading to a brine solution. The brine solution attacks the stainless steel surface, leading to pitting; in the presence of tensile stress (such as residual tensile stress due to welding), stress corrosion cracking can occur. The CGR will be used for evaluating flaw growth under a proposed ASME Boiler and Pressure Vessel Code Case covering stainless steel canisters.

Abstract The welding task focuses on development of advanced welding technologies for repair and maintenance of nuclear reactor structural components to safely and cost-effectively extend the service life of nuclear power reactors. This paper presents an integrated research and development effort by the Department of Energy Light Water Reactor Sustainability Program through the Oak Ridge National Laboratory (ORNL) and Electric Power Research Institute (EPRI) to develop a patent-pending technology, Auxiliary Beam Stress Improved Laser Welding Technique, that proactively manages the stresses during laser repair welding of highly irradiated reactor internals without helium induced cracking (HeIC). Finite element numerical simulations and in-situ temperature and strain experimental validation have been utilized to identify candidate welding conditions to achieve significant stress compression near the weld pool during cooling. Preliminary welding experiments were performed on irradiated stainless-steel plates (Type 304L). Post-weld characterization reveals that no macroscopic HeIC was observed.

Abstract 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.

In this present paper, the effect of specimen thickness on carburized layer thickness and surface residual stress of low temperature gaseous carburized AISI316L austenitic stainless steel was investigated by using specimen with thicknesses from ∼0.1 to ∼3 mm. After 15 and 30 hrs Low Temperature Gaseous Carburization (LTGC) treatment, the carburized layer thickness, surface residual stress and surface morphology were studied by optical microscope (OM), X-ray residual stress analyzer and scanning electron microscope (SEM). The results show that the specimen original thickness has no effect on the thickness of carburized layer. Surface compressive residual stresses are constant as about −1.6 and −2.1 GPa when the specimen thicknesses are not less than 0.485 mm for 15 hrs and 0.926 mm for 30 hrs LTGC treatment respectively. With the reduction of specimen thicknesses from 0.485 to 0.081 mm for 15 hrs LTGC treatment and 0.926 to 0.082 mm for 30 hrs LTGC treatment, the compressive residual stresses declined and finally reached about +0.4 and +1.0 GPa, respectively. Surface inter-granular cracking occurred on 0.082 mm specimen after 30 hrs LTGC treatment.

Grade 91 and other creep strength enhanced ferritic steels are widely used in fossil power plants as the structural materials of high temperature piping and components. As the operating temperature of these plants is as high as 600°C, creep deformation and accompanying damage cannot be neglected in evaluating their integrity. In particular, creep damage tends to concentrate in the softened heat-affected zone in welded joints and often caused so-called type-IV cracking, in advance of failure in base metal or weld metal. In evaluating the likelihood of such a failure, it is important to understand the mechanical property of such region in comparison with base metal and weld metal. Previous studies have shown that considerable degradation of creep strength brought by thermal cycles simulating the temperature history experienced by such a zone. In order to strengthen the understanding, thermal cycle was given to six Grade 91 heats encompassing a wide range of creep properties and their creep behavior was evaluated in this study. As a result, it was found that resistance for creep deformation and rupture life of six heats showed different trend than what was observed in the original metal but the heat-dependency of the ductility seems to be inherited.

This report introduces techniques involving novel calculation methods developed by the Electric Power Research Institute and Bechtel Marine Propulsion Corporation [1] that can be used to develop an improved understanding of welding factors affecting substrate weld heat input and dilution. These improved calculation methods are especially warranted in ASME BPVC Section XI repair and replacement welding activities such as mitigation of hot cracking (dilution control) and optimization of temper bead welding (heat input control). Various parameter sets are tested to demonstrate the effectiveness of the new calculation methods and how they can be refined in future studies to further improve accuracy. Additionally, this report highlights the limitations associated with the ASME BPVC Section IX theoretical heat input and power ratio calculation methods widely used in the welding industry for control of substrate heat input and weld dilution. Weld trials are performed in this study to demonstrate the limitations and determine proper precautions to take when using the calculation methods.

During the final post welding heat treatment (PWHT), residual stress relieves gradually with the accumulation of creep strain. However, reheat cracking with intergranular characteristic will occur when grain boundary cannot accommodate this kind of strain for some special steel welding, such as the welding coarse grained heat affected zone (CGHAZ) of 2.25Cr1Mo0.25V steel. Based on the principle of stress relaxation similar to the process of PWHT, two methods are applied to study the strain criterion of reheat cracking. Stress relaxation testing is performed on CGHAZ materials prepared by Gleeble thermomechanical simulator. The critical strain is calculated using the relationship between stress reduction and creep deformation. Self-loaded notched C-ring specimens are tested taken from the welding structure, coupled with finite element modeling and multiaxial creep coefficient to determine the critical strain. The results show that there is a large numerical difference between the critical strains from two methods. The possible reasons for the difference are given. Regarding the PWHT as a service process, whether the critical strain values obtained exceed the strain limits in ASME-NH is discussed.

Seismic responses from linear and nonlinear dynamic analyses of reinforced concrete (RC) shear walls are compared to see how the damping ratio and cracking behavior affect the dynamic response of the RC structures used in the nuclear power plant. The nonlinear dynamic analyses are conducted based on the numerical model which is developed to simulate the nonlinear hysteretic behavior of RC structures subjected to in-plane shear. Through comparison of the obtained numerical results with experimental data such as load-displacement relationships and response time-histories, the developed numerical model is validated. The acceleration response spectra from the nonlinear dynamic analysis results of selected RC shear wall and those from linear dynamic analysis with combinations of the damping ratio and concrete stiffness considerations according to the level of earthquake loads and the resultant stresses are addressed.

Coke drums undergo cyclic operations typically in temperatures ranging from the ambient temperature to about 500°C. Generally, coke drums are fabricated from clad plates consisting of chromium-molybdenum low alloy steel as the base metal and ferritic stainless steel typically Type 410S stainless steel (TP.410S) as cladding material. TP.410S cladding material is used for protection against sulfide corrosion. Coke drums frequently display major cracking on the inside surface of their circumferential weld along the fusion line between restoration weld and the TP.410S cladding. Moreover, an array of cracks commonly referred to as “elephant skin” are typically observed proximity to the major cracks. These cracks have been reported by multiple papers, particularly in the API Technical Report 934-G, up to the present and primarily attributed to the thermal fatigue which occurs from the repeated thermal loads experienced during each operating cycle. When the existence of severe cracks is recognized, users are forced to schedule extensive repairs. The authors conducted fatigue tests employing coupons made from two kinds of cladding material, TP.410S and Alloy 625, which simulate the circumferential weld seam of coke drums. The paper summarizes the essential properties for the cladding material in order to delay cracking on the inside surface and proposes new common cladding materials which could be employed for coke drums based on test results.