Peter James, David Coon, C. Austin, N. Underwood, C. Meek, M. Chevalier, David Dean
The “Establishing AMR Structural Integrity Codes and Standards for UK GDA” (EASICS) project was established in 2019 to help support the acceptance of Advanced Modular Reactors, or AMRs, which are typically based on high temperature Generation IV reactors. The EASICS project is aiming to provide guidance on the requirements for codes and standards for the design of AMRs for use in the UK, to ensure that state-of-the art knowledge will be brought to bear on developing the required design and assessment methodologies. The EASICS project started in July 2019 and is looking to complete by December 2021. To support this aim, the work presented in this paper provides an overview of two interacting aspects of the programme. The first is to perform validation tests for high temperature creep-fatigue assessments of a plant relevant component. The second aspect is to use these results, to provide a comparison of internationally recognised approaches for the assessment of high temperature (creep regime) components. This approach will be repeated for two other case scenarios deemed to be plant relevant components. This paper builds upon the initial overview paper presented at the 2020 conference providing an update on progress. One of the cases presented herein, described as the Thin Walled Welded Pipe Test uses specialist testing of a plant relevant component under high temperature loading conditions is underway with some initial results available. The validation testing includes both fatigue tests and creep-fatigue tests on 316H welded thin section tubes. The tubes have been subjected to strain-controlled tension/compression (R-ratio of −1), with some including a displacement controlled dwell. The tests are being conducted at 525°C. An update to the progress of these tests is included herein. To help enhance interaction with the code bodies, and to understand the impact of differences in the approaches, comparative assessments have been performed when adopting R5, ASME Section III Div 5 and RCC-MRx. One comparison will be based around the tests detailed above (tube test). A further assessment comparison will consider the Evasion mock-up tests provided by CEA (sodium based thermal shock tests). The third assessment case is loosely based around a plant relevant assessment within one of the UK Advanced Gas Reactors (AGRs). This paper provides an overview of the results from all these cases using R5, ASME Section III Div 5 and RCC-MRx. The subsequent discussions covers results, differences and potential impact to the codes which will all help to inform a guidance document to support assessing AMRs within a UK regulatory framework.
“为英国GDA建立AMR结构完整性规范和标准”(EASICS)项目成立于2019年，旨在帮助支持先进模块化反应堆(AMR)的接受，这些反应堆通常基于高温第四代反应堆。EASICS项目旨在为英国使用的amr设计的规范和标准要求提供指导，以确保将最先进的知识用于开发所需的设计和评估方法。EASICS项目于2019年7月开始，预计将于2021年12月完成。为了支持这一目标，本文提出的工作概述了该方案的两个相互作用的方面。首先是对工厂相关部件的高温蠕变疲劳评估进行验证试验。第二个方面是利用这些结果，为高温(蠕变状态)组件的评估提供国际公认的方法的比较。此方法将重复用于被认为是工厂相关组件的另外两种情况。本文以2020年会议上提交的初步概述文件为基础，提供了进展的最新情况。本文所介绍的薄壁焊管试验的一个案例是，在高温载荷条件下对工厂相关部件进行了专家测试，目前已有一些初步结果。验证试验包括316H焊接薄壁管的疲劳试验和蠕变疲劳试验。这些管受到应变控制的拉伸/压缩(r -比为- 1)，其中一些包括位移控制的驻留。试验在525°C下进行。本文包括这些测试进展的最新情况。为了帮助加强与规范主体的互动，并了解方法差异的影响，在采用R5、ASME Section III Div 5和RCC-MRx时进行了比较评估。一个比较将基于上面详细的测试(试管测试)。进一步的评估比较将考虑CEA(钠基热冲击试验)提供的逃避模型试验。第三个评估案例松散地基于英国先进气体反应堆(agr)的一个工厂相关评估。本文概述了使用R5, ASME Section III Div 5和RCC-MRx的所有这些案例的结果。随后的讨论涵盖了结果、差异和对代码的潜在影响，这些都将有助于为指导文件提供信息，以支持在英国监管框架内评估amr。
{"title":"Progress of EASICS Validation Experiments and Code Comparison of R5, RCC-MRX and ASME III Division V","authors":"Peter James, David Coon, C. Austin, N. Underwood, C. Meek, M. Chevalier, David Dean","doi":"10.1115/pvp2021-62845","DOIUrl":"https://doi.org/10.1115/pvp2021-62845","url":null,"abstract":"\u0000 The “Establishing AMR Structural Integrity Codes and Standards for UK GDA” (EASICS) project was established in 2019 to help support the acceptance of Advanced Modular Reactors, or AMRs, which are typically based on high temperature Generation IV reactors. The EASICS project is aiming to provide guidance on the requirements for codes and standards for the design of AMRs for use in the UK, to ensure that state-of-the art knowledge will be brought to bear on developing the required design and assessment methodologies. The EASICS project started in July 2019 and is looking to complete by December 2021.\u0000 To support this aim, the work presented in this paper provides an overview of two interacting aspects of the programme. The first is to perform validation tests for high temperature creep-fatigue assessments of a plant relevant component. The second aspect is to use these results, to provide a comparison of internationally recognised approaches for the assessment of high temperature (creep regime) components. This approach will be repeated for two other case scenarios deemed to be plant relevant components. This paper builds upon the initial overview paper presented at the 2020 conference providing an update on progress.\u0000 One of the cases presented herein, described as the Thin Walled Welded Pipe Test uses specialist testing of a plant relevant component under high temperature loading conditions is underway with some initial results available. The validation testing includes both fatigue tests and creep-fatigue tests on 316H welded thin section tubes. The tubes have been subjected to strain-controlled tension/compression (R-ratio of −1), with some including a displacement controlled dwell. The tests are being conducted at 525°C. An update to the progress of these tests is included herein.\u0000 To help enhance interaction with the code bodies, and to understand the impact of differences in the approaches, comparative assessments have been performed when adopting R5, ASME Section III Div 5 and RCC-MRx. One comparison will be based around the tests detailed above (tube test). A further assessment comparison will consider the Evasion mock-up tests provided by CEA (sodium based thermal shock tests). The third assessment case is loosely based around a plant relevant assessment within one of the UK Advanced Gas Reactors (AGRs). This paper provides an overview of the results from all these cases using R5, ASME Section III Div 5 and RCC-MRx.\u0000 The subsequent discussions covers results, differences and potential impact to the codes which will all help to inform a guidance document to support assessing AMRs within a UK regulatory framework.","PeriodicalId":402302,"journal":{"name":"Volume 4: Materials and Fabrication","volume":"28 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2021-07-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116627916","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Bill Press, Adam Dukes, David Poole, Jack Adams, L. Burling, J. Sulley
The Additive Manufacture (AM) of nuclear plant components, such as small-bore globe valves, offers opportunities to reduce costs and improve production lead-times. Cost reductions can be achieved by reducing raw material quantities, removing machining operations, and eliminating the welding of sub-assemblies. Furthermore, there is the opportunity to reduce production lead-times by simplifying the supply chain, e.g. reducing the number of parts to be sourced and eliminating special operations. Such opportunities are important against a backdrop of industry striving to reduce the cost of nuclear power generation in order to ensure viability with other forms of power generation. However, AM is a relatively new and innovative manufacturing technology, and although now seeing greater use in industry, there are still very few examples of where the technology has been applied to components used in safety critical applications. Furthermore, it is not covered by the American Society of Mechanical Engineers (ASME), Section III, nuclear design code. For nuclear plant applications, it is imperative a robust safety justification is provided. This paper presents Rolls-Royce’s approach to provision of a high integrity safety justification to enable the implementation of AM small-bore globe valves, up to a nominal bore size of 2” to nuclear plant. The material of construction is AM Laser Powder Bed Fusion (LPBF) 316LN stainless steel, with a Hot Isostatic Press (HIP) bonded LPBF Tristelle 5183 low cobalt hard facing seat. The paper describes the structure of the safety justification, which follows a multi-legged approach. It provides an overview of the innovative manufacturing process, which is, to the best of Rolls-Royce’s knowledge, the first of a kind application on nuclear pressure boundary components. The paper provides a summary of the suite of materials testing and metallurgical examinations conducted, and majors on prototype functional and performance testing where comparisons are made with the previous forged form. Pressure testing is covered which includes ultimate pressure testing to 2,000 bar, as well as: functional cyclic testing, hard facing bond strength tests, dynamic loading (shock), and cyclic thermal tests. In all cases the additive manufactured small-bore globe valves performed as well, and in some cases better than the forged material equivalent.
{"title":"Safety Justification Strategy for the Implementation of Additive Manufacture Small-Bore Globe Valves for Nuclear Plant","authors":"Bill Press, Adam Dukes, David Poole, Jack Adams, L. Burling, J. Sulley","doi":"10.1115/pvp2021-62614","DOIUrl":"https://doi.org/10.1115/pvp2021-62614","url":null,"abstract":"\u0000 The Additive Manufacture (AM) of nuclear plant components, such as small-bore globe valves, offers opportunities to reduce costs and improve production lead-times. Cost reductions can be achieved by reducing raw material quantities, removing machining operations, and eliminating the welding of sub-assemblies. Furthermore, there is the opportunity to reduce production lead-times by simplifying the supply chain, e.g. reducing the number of parts to be sourced and eliminating special operations. Such opportunities are important against a backdrop of industry striving to reduce the cost of nuclear power generation in order to ensure viability with other forms of power generation.\u0000 However, AM is a relatively new and innovative manufacturing technology, and although now seeing greater use in industry, there are still very few examples of where the technology has been applied to components used in safety critical applications. Furthermore, it is not covered by the American Society of Mechanical Engineers (ASME), Section III, nuclear design code. For nuclear plant applications, it is imperative a robust safety justification is provided.\u0000 This paper presents Rolls-Royce’s approach to provision of a high integrity safety justification to enable the implementation of AM small-bore globe valves, up to a nominal bore size of 2” to nuclear plant. The material of construction is AM Laser Powder Bed Fusion (LPBF) 316LN stainless steel, with a Hot Isostatic Press (HIP) bonded LPBF Tristelle 5183 low cobalt hard facing seat.\u0000 The paper describes the structure of the safety justification, which follows a multi-legged approach. It provides an overview of the innovative manufacturing process, which is, to the best of Rolls-Royce’s knowledge, the first of a kind application on nuclear pressure boundary components.\u0000 The paper provides a summary of the suite of materials testing and metallurgical examinations conducted, and majors on prototype functional and performance testing where comparisons are made with the previous forged form. Pressure testing is covered which includes ultimate pressure testing to 2,000 bar, as well as: functional cyclic testing, hard facing bond strength tests, dynamic loading (shock), and cyclic thermal tests. In all cases the additive manufactured small-bore globe valves performed as well, and in some cases better than the forged material equivalent.","PeriodicalId":402302,"journal":{"name":"Volume 4: Materials and Fabrication","volume":"152 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2021-07-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116392383","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
C. Huotilainen, Heikki Keinänen, Juha Kuutti, P. Nevasmaa, Henrik Sirén, I. Virkkunen
Extending the lifetime of existing nuclear power reactors is an increasingly important topic. As the existing fleet of nuclear power reactors ages and approaches the end of their design lifetimes or enters periods of lifetime extension, there is an increased probability for defect repairs due to extended exposure to the operating environment (e.g. high temperature, high pressure, corrosion environment, neutron irradiation, etc.). Concerning repair welding, should a critical need for repair arise, qualified and validated solutions must be readily available for rapid deployment. A proposed method using robotized gas metal arc welding-cold metal transfer to repair a “worst-case” scenario, linear crack like defect beneath the cladding, which extended into the reactor pressure vessel steel, was evaluated on laboratory scale in previous works (PVP2020-21233, PVP2020-21236). These previous studies demonstrated that cold metal transfer has the potential to produce high quality welds in the case of a reactor pressure repair. In the current study, the lessons learned from the previous work were applied to repair a postulated surface crack on a thermally embrittled and cladded low alloy steel plate using a nickel base Alloy 52 filler metal. Two excavations were filled using different weld bead arrangements — a traditional pattern (92 weld beads, Q = 0.6 kJ/min) and a 45°-hatch pattern (184 weld beads, Q = 0.9 kJ/min) — by gas metal arc welding-cold metal transfer. No pre-heating or post-weld heat treatment were applied, to remain in line with what can be expected in a real pressure vessel repair situation. The 0° angle pattern acts as a reference for previous studies, while the 45°-hatch pattern, aims to minimize the residual stresses caused by repair welding. Finite element modeling was used to predict the initial (cladded, embrittled and excavated) condition of the steel plate, followed by simulating the welding using the actual welding conditions and material constants for both bead patterns as input parameters. The resulting deformation, strains and stresses created in the material due to repair welding were predicted and the welding’s effectiveness was estimated. In addition, the post-repair weld mechanical properties and microstructure, specifically focusing on the fusion boundary and heat-affected zone, were evaluated using various microscopy techniques and hardness measurements. The outcomes of the performed simulations, corresponding characterizations and lessons learned are presented in this study.
{"title":"Evaluation of an Alloy 52 / Cladded Carbon Steel Repair Weld by Cold Metal Transfer","authors":"C. Huotilainen, Heikki Keinänen, Juha Kuutti, P. Nevasmaa, Henrik Sirén, I. Virkkunen","doi":"10.1115/pvp2021-61981","DOIUrl":"https://doi.org/10.1115/pvp2021-61981","url":null,"abstract":"\u0000 Extending the lifetime of existing nuclear power reactors is an increasingly important topic. As the existing fleet of nuclear power reactors ages and approaches the end of their design lifetimes or enters periods of lifetime extension, there is an increased probability for defect repairs due to extended exposure to the operating environment (e.g. high temperature, high pressure, corrosion environment, neutron irradiation, etc.). Concerning repair welding, should a critical need for repair arise, qualified and validated solutions must be readily available for rapid deployment. A proposed method using robotized gas metal arc welding-cold metal transfer to repair a “worst-case” scenario, linear crack like defect beneath the cladding, which extended into the reactor pressure vessel steel, was evaluated on laboratory scale in previous works (PVP2020-21233, PVP2020-21236). These previous studies demonstrated that cold metal transfer has the potential to produce high quality welds in the case of a reactor pressure repair.\u0000 In the current study, the lessons learned from the previous work were applied to repair a postulated surface crack on a thermally embrittled and cladded low alloy steel plate using a nickel base Alloy 52 filler metal. Two excavations were filled using different weld bead arrangements — a traditional pattern (92 weld beads, Q = 0.6 kJ/min) and a 45°-hatch pattern (184 weld beads, Q = 0.9 kJ/min) — by gas metal arc welding-cold metal transfer. No pre-heating or post-weld heat treatment were applied, to remain in line with what can be expected in a real pressure vessel repair situation. The 0° angle pattern acts as a reference for previous studies, while the 45°-hatch pattern, aims to minimize the residual stresses caused by repair welding. Finite element modeling was used to predict the initial (cladded, embrittled and excavated) condition of the steel plate, followed by simulating the welding using the actual welding conditions and material constants for both bead patterns as input parameters. The resulting deformation, strains and stresses created in the material due to repair welding were predicted and the welding’s effectiveness was estimated. In addition, the post-repair weld mechanical properties and microstructure, specifically focusing on the fusion boundary and heat-affected zone, were evaluated using various microscopy techniques and hardness measurements. The outcomes of the performed simulations, corresponding characterizations and lessons learned are presented in this study.","PeriodicalId":402302,"journal":{"name":"Volume 4: Materials and Fabrication","volume":"6 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2021-07-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129335940","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper discusses high energy piping (HEP) system walkdown requirements and guidelines in compliance with the American Society of Mechanical Engineers (ASME) B31.1 Code. Chapter VII states that the Operating Company shall develop and implement a program requiring documentation of piping support readings and recorded piping system displacements. Guidelines for this program are provided in Nonmandatory Appendix V, para. V-7. The Code also requires that the Operating Company shall evaluate the effects of unexpected piping position changes, significant vibrations, and malfunctioning supports on the piping system’s integrity and safety. These requirements and guidelines have been developed for personnel safety and piping system reliability. The HEP system should be maintained to behave as expected in the original design analysis unless a field change is justified by qualified personnel. The walkdown program should be an integral part of an asset integrity management program, including observations, documentation, evaluations, corrective actions, and countermeasures. A thorough HEP system walkdown includes more than documented hanger readings. It should include visual assessments of possible sagging pipe, unusual pipe slopes, building structure damage, lagging/insulation damage, locked spring hangers, piping interferences, damaged spring coils, loose/missing support fasteners, unloaded rigid supports, bent struts, insufficient hydraulic fluid in snubbers, detached Teflon strips on sliding supports, and confirmation that the current supports are consistent with the original design specifications. If accessible, it should be confirmed that there are no gaps in the sliding supports. This paper illustrates that it is now possible to photographically document spring support position indicator readings from distances up to 30 feet (9.1 meters). Photographic documentation provides higher confidence in the position indicator readings and can resolve many visual documentation discrepancies, such as incorrect support readings, readings from opposite position indicator sides, and parallax issues. If accessible, closer inspections may confirm if a spring support is in fact internally bottomed-out or topped-out. Nonmandatory Appendix V provides recommended hot walkdown and cold walkdown forms. These forms provide additional space for applicable notes. Example photographs of many piping system anomalies and associated documentation are provided in this paper. ASME B31.1 requires that significant displacement variations from the expected design displacements shall be considered to assess the piping system’s integrity.
{"title":"High Energy Piping Walkdowns in Compliance With ASME B31.1","authors":"M. Cohn, Robert J. Gialdini, Osborne B. Nye","doi":"10.1115/pvp2021-62533","DOIUrl":"https://doi.org/10.1115/pvp2021-62533","url":null,"abstract":"\u0000 This paper discusses high energy piping (HEP) system walkdown requirements and guidelines in compliance with the American Society of Mechanical Engineers (ASME) B31.1 Code. Chapter VII states that the Operating Company shall develop and implement a program requiring documentation of piping support readings and recorded piping system displacements. Guidelines for this program are provided in Nonmandatory Appendix V, para. V-7. The Code also requires that the Operating Company shall evaluate the effects of unexpected piping position changes, significant vibrations, and malfunctioning supports on the piping system’s integrity and safety. These requirements and guidelines have been developed for personnel safety and piping system reliability.\u0000 The HEP system should be maintained to behave as expected in the original design analysis unless a field change is justified by qualified personnel. The walkdown program should be an integral part of an asset integrity management program, including observations, documentation, evaluations, corrective actions, and countermeasures.\u0000 A thorough HEP system walkdown includes more than documented hanger readings. It should include visual assessments of possible sagging pipe, unusual pipe slopes, building structure damage, lagging/insulation damage, locked spring hangers, piping interferences, damaged spring coils, loose/missing support fasteners, unloaded rigid supports, bent struts, insufficient hydraulic fluid in snubbers, detached Teflon strips on sliding supports, and confirmation that the current supports are consistent with the original design specifications. If accessible, it should be confirmed that there are no gaps in the sliding supports.\u0000 This paper illustrates that it is now possible to photographically document spring support position indicator readings from distances up to 30 feet (9.1 meters). Photographic documentation provides higher confidence in the position indicator readings and can resolve many visual documentation discrepancies, such as incorrect support readings, readings from opposite position indicator sides, and parallax issues. If accessible, closer inspections may confirm if a spring support is in fact internally bottomed-out or topped-out.\u0000 Nonmandatory Appendix V provides recommended hot walkdown and cold walkdown forms. These forms provide additional space for applicable notes. Example photographs of many piping system anomalies and associated documentation are provided in this paper. ASME B31.1 requires that significant displacement variations from the expected design displacements shall be considered to assess the piping system’s integrity.","PeriodicalId":402302,"journal":{"name":"Volume 4: Materials and Fabrication","volume":"31 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2021-07-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129690896","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This paper describes a new stress-intensity factor (SIF) solution for an external surface crack in a sphere that expands capabilities previously available for this common pressure vessel geometry. The SIF solution employs the weight function (WF) methodology that enables rapid calculations of SIF values. The WF methodology determines SIF values from the nonlinear stress variations computed for the uncracked geometry, e.g., from service stresses and/or residual stresses. The current approach supports two degrees of freedom that denote the two crack tips located normal to the surface and the surface of the sphere. The geometric formulation of this solution enforces an elliptical crack front, maintains normality of the crack front with the free surface, and supports two degrees of freedom for fatigue crack growth from an internal crack tip and a surface crack tip. The new SIF solution accommodates spherical geometries with an exterior diameter greater than or equal to four times the thickness. This WF SIF solution has been combined with stress variations common for spherical pressure vessels: uniform internal pressure on the interior surface, uniform tension on the crack plane, and uniform bending on the crack plane. This paper provides a complete overview of this solution. We present for the first time the geometric formulation of the crack front that enables the new functionality and set the geometric limits of the solution, e.g., the maximum size and shape of the crack front. The paper discusses the bivariant WF formulation used to define the SIF solution and details the finite element analyses employed to calibrate terms in the WF formulation. A summary of preliminary verification efforts demonstrates the credibility of this solution against independent results from finite element analyses. We also compare results of this new solution against independent SIFs computed by finite element analyses, legacy SIF solutions, API 579, and FITNET. These comparisons indicate that the new WF solution compares favorably with results from finite element analyses. This paper summarizes ongoing efforts to improve and extend this solution, including formal verification and development of an internal surface crack model. Finally, we discuss the capabilities of this solution’s implementation in NASGRO® v10.0.
{"title":"A New Stress-Intensity Factor Solution for an External Surface Crack in Spheres","authors":"J. Sobotka, Yi-der Lee, J. Cardinal, R. Mcclung","doi":"10.1115/pvp2021-61397","DOIUrl":"https://doi.org/10.1115/pvp2021-61397","url":null,"abstract":"\u0000 This paper describes a new stress-intensity factor (SIF) solution for an external surface crack in a sphere that expands capabilities previously available for this common pressure vessel geometry. The SIF solution employs the weight function (WF) methodology that enables rapid calculations of SIF values. The WF methodology determines SIF values from the nonlinear stress variations computed for the uncracked geometry, e.g., from service stresses and/or residual stresses. The current approach supports two degrees of freedom that denote the two crack tips located normal to the surface and the surface of the sphere. The geometric formulation of this solution enforces an elliptical crack front, maintains normality of the crack front with the free surface, and supports two degrees of freedom for fatigue crack growth from an internal crack tip and a surface crack tip. The new SIF solution accommodates spherical geometries with an exterior diameter greater than or equal to four times the thickness. This WF SIF solution has been combined with stress variations common for spherical pressure vessels: uniform internal pressure on the interior surface, uniform tension on the crack plane, and uniform bending on the crack plane.\u0000 This paper provides a complete overview of this solution. We present for the first time the geometric formulation of the crack front that enables the new functionality and set the geometric limits of the solution, e.g., the maximum size and shape of the crack front. The paper discusses the bivariant WF formulation used to define the SIF solution and details the finite element analyses employed to calibrate terms in the WF formulation. A summary of preliminary verification efforts demonstrates the credibility of this solution against independent results from finite element analyses. We also compare results of this new solution against independent SIFs computed by finite element analyses, legacy SIF solutions, API 579, and FITNET. These comparisons indicate that the new WF solution compares favorably with results from finite element analyses. This paper summarizes ongoing efforts to improve and extend this solution, including formal verification and development of an internal surface crack model. Finally, we discuss the capabilities of this solution’s implementation in NASGRO® v10.0.","PeriodicalId":402302,"journal":{"name":"Volume 4: Materials and Fabrication","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2021-07-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125550668","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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