Abstract ASME Boiler Pressure Vessel Code, Section III, Division 1, Subsection NC, Class 2 components, and Subsection ND, Class 3 components, have significant technical and administrative similarities. The ASME BPV III Standards Committee has a long-standing goal of combining these two subsections (NC and ND). Consolidating Subsections NC and ND will simplify, reduce repetitions and make the Code easier to use. Additionally, a combined Subsection NC/ND will simplify Code maintenance. To facilitate this consolidation, the Subgroup on Component Design, under the BPV III Standards Committee assigned a Task Group to develop a strategy to combine the two subsections into a single subsection while maintaining both Class 2 and Class 3 as separate classes of construction. Both Subsections NC and ND of the Code have been completely reviewed, compared and the technical bases for the differences have been established. The conclusion of this review is that there are only a few major technical differences between the two code class rules; however, there are a significant number of editorial differences. Based on the review, the Task Group developed a strategy that completes the consolidation within two publishing cycles of Code edition. For the Code Edition 2019, two separate Subsections NC and ND books will be published to resolve editorial differences and otherwise align the two subsections. For the Code Edition 2021, a single merged subsection will be published. This paper provides the background for the proposed code change, discusses the detailed result of the NC/ND comparison, and provides the basis for the major technical differences. The paper will also update the status of the project and code actions needed to consolidate to a single subsection.

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 Mandatory Appendix XXVI of Section III of the ASME B&PV Code contains rules for the construction of Class 3 polyethylene pressure piping systems. The scope is limited to buried portions of Class 3 service water or buried portions of Class 3 cooling water systems, consisting of PE4710 High Density Polyethylene (HDPE) materials. The minimum Pennsylvania Notched Test (PENT) rating for the HDPE material is 2,000 hours. Appendix XXVI contains acceptance standards for the maximum allowable depths of gouges, cuts or other surface conditions that are characterized as indentations. The acceptance standards are considered to be very restrictive, in particular for large diameter HDPE pipes. Less restrictive maximum allowable indentation depths for pipes with a minimum PENT rating of 2,000 hours were developed based on use of results from tests performed on pressurized HDPE pipes containing flaws in the parent material. These maximum allowable indentation depths were implemented into the new Section III Code Case N-891 on alternative requirements to Appendix XXVI for inspection and repair. The technical basis for the maximum allowable indentation depths is described in this paper.

Abstract For several decades pressurized water reactors have experienced Primary Water Stress Corrosion Cracking (PWSCC) within Alloy 600 components and welds. The nuclear industry has developed several methods for mitigation of PWSCC to prevent costly repairs to pressurized water reactor (PWR) components including surface stress improvement by peening. Laser shock peening (LSP) is one method to effectively place the surface of a PWSCC susceptible component into compression and significantly reduce the potential for crack initiation during future operation. The Material Reliability Program (MRP) has issued MRP-335, which provides guidelines for effective mitigation of reactor vessel heads and nozzles constructed of Alloy 600 material. In addition, ASME Code Case N-729-6 provides performance requirements for peening processes applied to reactor vessel head penetrations in order to prevent degradation and take advantage of inspection relief, which will reduce operating costs for nuclear plants. LSP Technologies, Inc. (LSPT) has developed and utilized a proprietary LSP system called the Procudo ® 200 Laser Peening System. System specifications are laser energy of 10 J, pulse width of 20 ns, and repetition rate of 20 Hz. Scalable processing intensity is provided through automated focusing optics control. For the presented work, power densities of 4 to 9.5 GW/cm 2 and spot sizes of nominally 2 mm were selected. This system has been used effectively in many non-nuclear industries including aerospace, power generation, automotive, and oil and gas. The Procudo ® 200 Laser Peening System will be used to process reactor vessel heads in the United States for mitigation of PWSCC. The Procudo ® 200 Laser Peening System is a versatile and portable system that can be deployed in many variations. This paper presents test results used to evaluate the effectiveness of the Procudo ® 200 Laser Peening System on Alloy 600 material and welds. As a part of the qualification process, testing was performed to demonstrate compliance with industry requirements. The test results include surface stress measurements on laser peened Alloy 600, and Alloy 182 coupons using x-ray diffraction (XRD) and crack compliance (slitting) stress measurement techniques. The test results are compared to stress criteria developed based on the performance requirements documented in MRP-335 and Code Case N-729-6. Other test results include surface roughness measurements and percent of cold work induced by the peening process. The test results demonstrate the ability of the LSP process to induce the level and depth of compression required for mitigation of PWSCC and that the process does not result in adverse conditions within the material.

Abstract Tube integrity is an important aspect for safe and reliable operation of nuclear power plant steam generators. As a U.S. industry and licensing requirement, all in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions and design basis accidents by meeting the structural integrity performance criterion (SIPC) as given in NEI 97-06. The SIPC margin shall be maintained during plant operation between tube examinations. The burst strength of tubes subjected to wall thinning will depend on the extent and mode of degradation, and the magnitude of design loads to include pressure differential across the tube wall during normal operation and postulated accident conditions. In addition, non-pressure loads that can occur during postulated accident events shall be evaluated and included in the assessment of tube integrity if determined to significantly reduce the tube burst strength. The EPRI Flaw Handbook provides burst pressure relationships for flaws which include a reduction factor that accounts for the effect of applied bending stress on circumferential degradation. However, this previous industry work was only for planar crack-like flaws and did not directly address uniform volumetric wall loss which can have both axial and circumferential extent. This paper describes a test program to determine the effect of bending loads on the burst pressure of a tube with uniform thinning over a given axial length. The uniform thinning geometry was selected since it represented a bounding case of general wall loss and is conservative for calculating a tube repair limit for volumetric degradation for a given steam generator design. Tube repair limits are required for defining an upper limit on in-service degradation for which a tube is to be removed from service. Tube repair limits are cited in the Plant Technical Specifications, which is an important part of the licensing basis.

Abstract In the Fall of 2018 Westinghouse designed a welded plug to repair a leak around a previously capped nozzle in the Boron Injection Tank at Indian Point Unit 3. The leak was caused by weld cracking around the nozzle. In order to complete the repair, the existing nozzle and weld needed to be removed by boring into the tank so that a custom plug could be installed. Due to site procedures, welding could not be performed inside the tank, so a single weld on the outside of the tank needed to be designed to meet ASME B&PV Code, Section III requirements. In addition, the tank was manufactured from carbon steel with a stainless steel cladding, so future corrosion of the carbon steel needed to be assessed. The leak was an outage related incident, which required the design to be expedited. This paper will discuss the technical challenges in designing the welded plug to meet the site and code requirements.

Abstract Post weld heat treatment (PWHT) is applied to welded pressure vessels and piping to relieve residual stress built up during welding. During the PWHT, pressure vessel made of carbon steel is heated up to a minimum temperature of 595 °C for a holding period as required based on the thickness (Refer ASME Section VIII, Division 1, UCS 56) [1]. However, for equipment susceptible to carbonate stress corrosion cracking, PWHT is required irrespective of thickness at a temperature range of 649 °C to 663 °C as per WRC 452 [2]. A process column built in 1983 and operating in carbonate stress corrosion cracking environment was observed to have widespread corrosion under insulation at reboiler supports. Material of construction for the column was SM41B (Japanese Industry Standards Carbon Steel). Four Insert plates of various sizes were welded at same circumferential band during turnaround. The insert plate repair sizes were relatively large. Local PWHT was required to be performed with the equipment in vertical condition. Due to risk of uneven temperature distribution and resultant local thermal stresses, spot PWHT or Bulls-eye method of PWHT was not considered. Full circumferential local PWHT was considered for this equipment. During the actual PWHT process, localized overheating occurred, and some areas of deformation were observed in the equipment. A multi discipline review was performed to understand the root cause of the localized overheating and subsequent deformation. This paper describes the methodology and results of the fitness-for-service (FFS) assessments that were performed based on API 579-1/ ASME FFS-1, 2016 [3] to assess the integrity of the column. Based on the assessment performed, the equipment was found to be fit for service and continued safe operations.

Abstract INTRODUCTION : Managing operational assets outside the Refinery, in unfenced/unmanned areas is of paramount importance for the successful and safe operations of the company. Operational assets include gas, oil and utility pipelines, distribution and metering stations, valve stations, power distribution cables and such other assets. These assets are ‘on land’ inside populated urban areas, desert locations and sub-sea installations. Primarily, company’s operational assets are protected by constant patrolling by the nation’s public security agencies and also by the company security department. However risks to the company’s operational assets arise from Contractors engaged by the company to perform any work on the asset or near the asset or other agencies including ministries or utility companies or authorities engaging their Contractors in carrying out new installation or maintenance work of their own assets located in and around company assets. In the context of Bahrain as a nation, the oil and gas industry contributes substantially to the economy and hence it is very important that all necessary processes are established to protect operational assets. When work is performed by Contractors not directly within operational control i.e. taking place in remote, normally unmanned or minimally manned facilities the inherent risk is high and has to be managed through legally mandated processes. SITUATIONAL ANALYSIS : Bapco has an established ‘Process Safety Management’ (PSM) system accepted and implemented in the company. The ‘Safe Work Permit’ system is a very critical control system/measure to ensure Contractor compliance when performing any work. Coupled with the effectiveness of PSM and Contractor Safety process the Government of Bahrain mandated legally binding process of the ‘Wayleave’ system along with other processes such as Planning Permission, helps in enforcing control of company’s operational assets in company’s unfenced or remote areas with minimal or no manning. A procedure for ‘Processing of Planning Permissions and Wayleave pertaining to BAPCO’s Unfenced Areas’ has been developed based on engineering design requirements basis international design codes and best professional practices. Details of these practices will be described and illustrated in the detailed presentation. PRESENTATION/CONCLUSION : Bapco has over the past years developed procedures and processes to manage operational assets/ areas within the Oil and Gas Fields, Oil transmission pipelines and the ‘Right-of-Way’, Gas Distribution Network and the ‘Right-of-Way’, Refinery, Tank Farms and its own Shipping Wharf Operations in keeping with the highest engineering standards, safety and quality requirements. Purpose of this presentation is to list and explain the adaptation of these procedures and processes coupled with pragmatic solutions that forms the basic architecture that aids in controlling the safety and operability of assets in ‘Unfenced/ unmanned Operating Areas’. This ensures generating positive economics and aiding in the growth of the nation’s economy.

Abstract The functional requirements of air-operated valves provided in nuclear power plants during an earthquake have been previously evaluated via seismic tests and so forth; however, since the response acceleration has increased in line with a recent reassessment of standard earthquake ground motions, it is necessary to evaluate functional maintenance with respect to high acceleration. It is also necessary to determine practical functionality from the viewpoint of equipment fragility in seismic Probability Risk Assessment (PRA) Here, we used a resonant shaking table (see Ref. [1]) in the Central Research Institute of the Electric Power Industry (CRIEPI), which is capable of seismic tests at acceleration levels that have been unachievable until now. From the results of seismic tests on air-operated valve actuators, the operability of active components used with an existing design was confirmed to be more than 15 × 9.8 m/s 2 (horizontal directions (X and Y)) and more than 20 × 9.8 m/s 2 (vertical direction (Z)).

Abstract History of earthquake’s damages have illustrated the high vulnerability and risks associated with failure of water transfer and distribution systems. Adequate mitigation plans to reduce such seismic risks are required for sustainable development. The first step in developing a mitigation plan is prioritizing the limited available budget to address the most critical mitigation measures. This paper presents an optimization model that can be utilized for financial resource allocation towards earthquake risk mitigation measures for water pipelines. It presents a framework that can be used by decision-makers (authorities, stockholders, owners and contractors) to structure budget allocation strategy for seismic risk mitigation measures such as repair, retrofit, and/or replacement of steel and concrete pipelines. A stochastic model is presented to establish optimal mitigation measures based on minimizing repair and retrofit costs, post-earthquake replacement costs, and especially earthquake-induced large losses. To consider the earthquake induced loss on pipelines, the indirect loss due to water shortage and business interruption in the industries which needs water is also considered. The model is applied to a pilot area to demonstrate the practical application aspects of the proposed model. Pipeline exposure database, built environment occupancy type, pipeline vulnerability functions, and regional seismic hazard characteristics are used to calculate a probabilistic seismic risk for the pilot area. The Global Earthquake Model’s (GEM) OpenQuake software is used to run various seismic risk analysis. Event-based seismic hazard and risk analyses are used to develop the hazard curves and maps in terms of peak ground velocity (PGV) for the study area. The results of this study show the variation of seismic losses and mitigation costs for pipelines located within the study area based on their location and the types of repair. Performing seismic risk analysis analyses using the proposed model provides a valuable tool for determining the risk associated with a network of pipelines in a region, and the costs of repair based on acceptable risk level. It can be used for decision making and to establish type and budgets for most critical repairs for a specific region.

Abstract A challenge for owners operating nuclear power plants and spent fuel canisters is to provide sound technical basis for the safety and security of long-term operation and storage, respectively. Two code cases N-729-6 and N-770-5 provide performance requirements for mitigation of Primary Water Stress Corrosion Cracking (PWSCC) in ASME Section XI and Code Case N-860 “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems” is under development in ASME Section XI. The primary degradation concern being addressed in N-860 is Chloride Induced Stress Corrosion Cracking (CISCC). ASME Section III has formed two Task Groups with the intent to reduce PWSCC and CISCC in operation by mitigating residual stresses during new construction of components. ASME Section III formed Task Group Weld Residual Stress (TG-WRS) to incorporate lessons learned into the Code from operating plants regarding their experience with stress corrosion cracking. An example of expected guidance from TG-WRS is to consider residual stresses after weld repair of a susceptible material when the repair surface is exposed to the wetted environment. To address this concern, one form of mitigation could be the use of SSI which has been used in operating plants. However, there is a gap because no acceptance or performance criteria exists for mitigation of SCC by surface stress improvement for ASME Section III components. The executive committee of ASME requested that a new Task Group named “Advanced Surface Stress Improvement Technology” (ASSIT) be formed by the Korean International Working Group (KIWG) and Materials Fabrication and Examination (MF&E) to address this gap by developing a code case for mitigation of PWSCC and CISCC for new ASME Section III components by SSI techniques. This paper will discuss the background and current approach being taken by Task Group ASSIT in developing SSI acceptance or performance criteria for mitigation of SCC. The necessary technical reports supporting this code case are also under development as part of joint projects between Doosan Heavy Industries and Construction (DOOSAN), Electric Power Research Institute (EPRI), and Sun Moon University (SMU). Interim reports summarizing the status of these joint projects will be presented as separate papers in the same session.

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 The Primary Water Stress Corrosion Cracking causes existing ASME BPV Code Section XI inspection and repair methodology to need enhancement. This forms the basis for Code Case N-770 and revisions through 6. The paper includes the bases for examination requirements, examination schedule, and repair criteria.

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 At ageing power plants, local thinning of pipework or vessel is unavoidable due to erosion/corrosion or other reasons such as flow accelerated corrosion (FAC) — one of the common degradation mechanisms in pipework of nuclear power plant. Local thinning reduces the structure strength, resulting in crack initiation from the corrosion pit or welding defect when subject to cyclic loading. General practice is to use the minimum thickness of the thinned area to calculate both limit load and stress intensity factor (SIF) in performing Engineering Critical Assessment (ECA) using Failure Assessment Diagram (FAD). Using the minimum thickness is normally overly conservative as it assumes that thinning occurs grossly instead of locally, leading to unnecessary early repair/replacement and cost. Performing cracked body finite element analysis (FEA) can provide accurate values of limit load and SIF, but it is time consuming and impractical for daily maintenance and emergent support. To minimise the conservatisms and provide a guidance for the assessment of locally thinned pipework or vessel using existing handbook solutions, a study was carried out by the authors on the effect of local thinning on limit loads. The study demonstrates that local thinning has significant effect on limit load if the thinning ratio of thinning depth to original thickness is larger than 25%. It concluded that the limit load solutions given in handbooks (such as R6 or the net section method) are overly conservative if using the minimum local thickness and non-conservative if using the nominal thickness. This paper discusses the effect of local thinning on SIFs of internal/external defects using cracked body finite element method (FEM). The results are compared with R6 weight function SIF solutions for a cylinder. A modified R6 SIF solution is proposed to count for the effect of local thinning profile. Along with the previous published paper on limit load it provides comprehensive understanding and guidance for fracture assessment of the local thinned pipework and vessel.

Abstract This paper describes the use of Carbon Fiber Reinforced Polymer (CFRP) for the repair of leaking Fiber Reinforced Polymer (FRP) pipe. The existing piping system, which is used for slurry transport, consisted of straight pipe, reducers, and elbows connected by bell and spigot joints. The piping system was evaluated for pressure, deadweight, and fluid thrust loads in accordance with B31.1 Power Piping Code and ASME BPVC Section III, Code Case N-155. The pipe leaks and joint repairs were in accordance with ASME PCC-2 and ASME BPVC, Section III, Code Case N-155. The CFRP repair was done with Aegion TYFO ® Fibrwrap ® carbon fiber reinforced fiber wrap (TYFO SCH-41-2X) saturated with epoxy. The repair uses the unidirectional carbon fiber installed in layers 90° to each layer. The circumferential layers are used for hoop pressure stress while the axial layers restrain axial pressure, deadweight, and fluid thrust loadings on the pipe.

Abstract 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.

Abstract In recent years the understanding of the relationship between drum damage and bulge sharpness has improved significantly. The authors of this paper developed a new parameter called bulge sharpness and have previously shown the relationship between sharpness and observed damage. Further to this study, the authors have exhaustively studied the evolution of stress cracking (elephant skin) on mid-course bulges and have estimated the likelihood of finding a particular type of surface damage based on the observed sharpness levels. This correlation has led to a proposed scale to categorize stress cracking into three levels: minor, intermediate, and significant. In addition, the progression of bulge sharpness over time was analyzed and it was determined through statistical modeling that bulge sharpness can have a range of rates of change or sharpness growth rates: low, medium, and high. These sharpness growth rates were subsequently studied and their relationship with overall cycle times analyzed. The study also shows that individual coke drums can experience different sharpness growth rates and there can be a distribution of these rates. To determine when repairs should be conducted, coke drum operators must consider the expected operational run. While the random nature of coke drum damage can defy such targets, bulge sharpness growth assessments can be used to better define when repairs should be conducted. Understanding current bulge sharpness levels, year-over-year sharpness growth rates and their distribution, can significantly assist in targeting areas of concern for optimized repair strategies and can also be used to avoid unnecessary repairs.

Abstract Flow-induced vibrations of Steam Generator tube bundles are a major concern for the operators and designers of nuclear power plants. In order to predict damages due to such vibrations, predictive calculations are performed, both by designers and by operators, which allow one to evaluate safety margins and thereafter to optimize the SG maintenance policy. In cases when semi analytical models of fluid-dynamic forces are used, together with corresponding dimensionless fluid force coefficients, closures are assessed by experiment. The database of dimensionless coefficients should then cover relevant tube bundle configurations. Within this framework, a test rig which was available at EDF was modified in order to arrange different pitch-to-diameter ratio bundles and different tube diameters. This paper discusses the fluid-induced forces measured on a set of flexible tubes placed in a triangular pitch bundle both in the lift and drag directions. Two-phase flow excitation spectra are presented as well.