The ASME Boiler & Pressure Vessel Code Section XI Committee is currently developing a new Division 2 nuclear code entitled the “Reliability and Integrity Management (RIM) program,” with which one is able to arrive at a risk-informed, NDE-based engineering maintenance decision by estimating and managing all uncertainties for the entire life cycle including design, material selection, degradation processes, operation and non-destructive examination (NDE). This paper focuses on the uncertainty of the NDE methods employed for preservice and inservice inspections due to a large number of factors such as the NDE equipment type and age, the operator’s level and years of experience, the angle of probe, the flaw type, etc. In this paper, we describe three approaches with which uncertainty in NDE-risk-informed decision making can be quantified: (1) A regression model approach in analyzing round-robin experimental data such as the 1981–82 Piping Inspection Round Robin (PIRR), the 1986 Mini-Round Robin (MRR) on intergranular stress corrosion cracking (IGSCC) detection and sizing, and the 1989–90 international Programme for the Inspection of Steel Components III-Austenitic Steel Testing (PISC-AST). (2) A statistical design of experiments approach. (3) An expert knowledge elicitation approach. Based on a 2003 Pacific Northwest National Laboratory (PNNL) report by Heasler and Doctor (NUREG/CR-6795), we observe that the first approach utilized round robin studies that gave NDE uncertainty information on the state of the art of the NDE technology employed from the early 1980s to the early 1990s. This approach is very time-consuming and expensive to implement. The second approach is based on a design-of-experiments (DEX) of eight field inspection exercises for finding the length of a subsurface crack in a pressure vessel head using ultrasonic testing (UT), where five factors (operator’s service experience, UT machine age, cable length, probe angle, and plastic shim thickness), were chosen to quantify the sizing uncertainty of the UT method. The DEX approach is also time-consuming and costly, but has the advantage that it can be tailored to a specific defect-detection and defect-sizing problem. The third approach using an expert panel is the most efficient and least costly approach. Using the crack length results of the second approach, we introduce in this paper how the expert panel approach can be implemented with the application of a software package named the Sheffield Elicitation Framework (SHELF). The crack length estimation with uncertainty results of the three approaches are compared and discussed. Significance and limitations of the three uncertainty quantification approaches to risk assessment of NDE-based engineering decisions are presented and discussed.
{"title":"Three Approaches to Quantification of NDE Uncertainty and a Detailed Exposition of the Expert Panel Approach Using the Sheffield Elicitation Framework","authors":"J. Fong, N. Heckert, J. Filliben, S. Doctor","doi":"10.1115/PVP2018-84771","DOIUrl":"https://doi.org/10.1115/PVP2018-84771","url":null,"abstract":"The ASME Boiler & Pressure Vessel Code Section XI Committee is currently developing a new Division 2 nuclear code entitled the “Reliability and Integrity Management (RIM) program,” with which one is able to arrive at a risk-informed, NDE-based engineering maintenance decision by estimating and managing all uncertainties for the entire life cycle including design, material selection, degradation processes, operation and non-destructive examination (NDE). This paper focuses on the uncertainty of the NDE methods employed for preservice and inservice inspections due to a large number of factors such as the NDE equipment type and age, the operator’s level and years of experience, the angle of probe, the flaw type, etc. In this paper, we describe three approaches with which uncertainty in NDE-risk-informed decision making can be quantified: (1) A regression model approach in analyzing round-robin experimental data such as the 1981–82 Piping Inspection Round Robin (PIRR), the 1986 Mini-Round Robin (MRR) on intergranular stress corrosion cracking (IGSCC) detection and sizing, and the 1989–90 international Programme for the Inspection of Steel Components III-Austenitic Steel Testing (PISC-AST). (2) A statistical design of experiments approach. (3) An expert knowledge elicitation approach. Based on a 2003 Pacific Northwest National Laboratory (PNNL) report by Heasler and Doctor (NUREG/CR-6795), we observe that the first approach utilized round robin studies that gave NDE uncertainty information on the state of the art of the NDE technology employed from the early 1980s to the early 1990s. This approach is very time-consuming and expensive to implement. The second approach is based on a design-of-experiments (DEX) of eight field inspection exercises for finding the length of a subsurface crack in a pressure vessel head using ultrasonic testing (UT), where five factors (operator’s service experience, UT machine age, cable length, probe angle, and plastic shim thickness), were chosen to quantify the sizing uncertainty of the UT method. The DEX approach is also time-consuming and costly, but has the advantage that it can be tailored to a specific defect-detection and defect-sizing problem. The third approach using an expert panel is the most efficient and least costly approach. Using the crack length results of the second approach, we introduce in this paper how the expert panel approach can be implemented with the application of a software package named the Sheffield Elicitation Framework (SHELF). The crack length estimation with uncertainty results of the three approaches are compared and discussed. Significance and limitations of the three uncertainty quantification approaches to risk assessment of NDE-based engineering decisions are presented and discussed.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"250 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128486406","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}
In order to conduct effective and rational maintenance activity of components in nuclear power plants, it is proposed to manage fatigue degradation based on crack size corresponding to an extent of cumulative fatigue damage. The purpose of this study focuses on the influence of strain rate in simulated reactor coolant environment for fatigue crack initiation and growth. 3-dimensional replica observations were conducted for environmental fatigue test specimens in different strain rates. Crack initiation and growth were observed in the experiments. It is clarified that low strain rate influences crack propagation and coalescence and increases crack growth rate that finally decrease fatigue life.
{"title":"Study on the Influence of Strain Rate on Crack Initiation and Growth in Simulated Reactor Coolant Environment of Type 316 Stainless Steel","authors":"Takahisa Nose, Takao Nakamura, T. Kitada","doi":"10.1115/PVP2018-84490","DOIUrl":"https://doi.org/10.1115/PVP2018-84490","url":null,"abstract":"In order to conduct effective and rational maintenance activity of components in nuclear power plants, it is proposed to manage fatigue degradation based on crack size corresponding to an extent of cumulative fatigue damage. The purpose of this study focuses on the influence of strain rate in simulated reactor coolant environment for fatigue crack initiation and growth. 3-dimensional replica observations were conducted for environmental fatigue test specimens in different strain rates. Crack initiation and growth were observed in the experiments. It is clarified that low strain rate influences crack propagation and coalescence and increases crack growth rate that finally decrease fatigue life.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"38 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132501126","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}
T. Ohmi, A. Yokobori, G. Ozeki, T. Kasuya, N. Ishikawa, S. Minamoto, M. Enoki
Hydrogen induced cracking occurs at the welded position of the structure due to concentration of hydrogen during cooling process of welding. A square groove weld joint is one of typical one in engineering field. Hydrogen embrittlement cracking is sometimes caused during cooling process of a weld joint. For such case, hydrogen diffusion and concentration behaviour is a significant factor. One of authors has been proposed α multiplication method which magnifies the hydrogen driving term in the diffusion equation to find out detailed behaviours of hydrogen concentration around a local stress field. In this paper, to clarify hydrogen diffusion behaviour in the square groove weld joint, a coupled analysis of heat transfer – thermal stress – hydrogen diffusion combining with α multiplication method was conducted. From these results, it was found out that for the case of a square groove weld joint, since thermal stress was not highly localized for the case of using usual value of thermal expansion coefficient of steel, hydrogen concentration behaviour is not typical. However, if thermal stress is highly localized, hydrogen was found to be localized in the side of HAZ (heat affected zone) at the interface of WM (weld metal) and HAZ and is much more typical near the outer surface side of weld joint. Hydrogen diffusion and concentration behaviours were also found to be dominated not only by local thermal stress gradient, ∇σ but also by diffusion coefficient gradient, ∇D caused by temperature difference during cooling process. In this paper, effects of these factors on hydrogen concentration were investigated based on a coupled analysis of heat transfer – thermal stress – hydrogen diffusion combining with α multiplication method.
{"title":"Hydrogen Diffusion Concentration Behaviors for Square Groove Weld Joint","authors":"T. Ohmi, A. Yokobori, G. Ozeki, T. Kasuya, N. Ishikawa, S. Minamoto, M. Enoki","doi":"10.1115/PVP2018-84192","DOIUrl":"https://doi.org/10.1115/PVP2018-84192","url":null,"abstract":"Hydrogen induced cracking occurs at the welded position of the structure due to concentration of hydrogen during cooling process of welding. A square groove weld joint is one of typical one in engineering field. Hydrogen embrittlement cracking is sometimes caused during cooling process of a weld joint. For such case, hydrogen diffusion and concentration behaviour is a significant factor. One of authors has been proposed α multiplication method which magnifies the hydrogen driving term in the diffusion equation to find out detailed behaviours of hydrogen concentration around a local stress field. In this paper, to clarify hydrogen diffusion behaviour in the square groove weld joint, a coupled analysis of heat transfer – thermal stress – hydrogen diffusion combining with α multiplication method was conducted. From these results, it was found out that for the case of a square groove weld joint, since thermal stress was not highly localized for the case of using usual value of thermal expansion coefficient of steel, hydrogen concentration behaviour is not typical. However, if thermal stress is highly localized, hydrogen was found to be localized in the side of HAZ (heat affected zone) at the interface of WM (weld metal) and HAZ and is much more typical near the outer surface side of weld joint. Hydrogen diffusion and concentration behaviours were also found to be dominated not only by local thermal stress gradient, ∇σ but also by diffusion coefficient gradient, ∇D caused by temperature difference during cooling process. In this paper, effects of these factors on hydrogen concentration were investigated based on a coupled analysis of heat transfer – thermal stress – hydrogen diffusion combining with α multiplication method.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"289 ","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114096508","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}
Pressurized water reactor environments are known to reduce the fatigue life of austenitic stainless steel components when compared to air environments. Laboratory testing has provided a means of quantifying this, allowing conservative plant assessments to be made. The majority of this testing has been isothermal and carried out on membrane loaded hollow or solid specimens. The geometry and loading of laboratory test specimens is significantly different to that experienced on plant, where complex strain waveforms are generally out of phase with temperature changes, and significant through wall strain gradients may be present. To address the issue of realistic loading, a test facility has been developed which can simulate thermal shock loading on a tubular specimen. The capability of the test facility was presented at the PVP2016 conference [PVP2016-63161]. Since then the facility has evolved, with modifications made to the rig configuration and specimen geometry in order to maximize the strain amplitude from the thermal shock, including the adoption of an annular flow geometry. These modifications were designed to optimize both the heat transfer coefficient and the speed of cycling between hot and cold water in order to induce a thermal strain that can cause mechanical failure within practicable test durations. In order to calculate the magnitude of the thermal strain, detailed calculations were required both in terms of thermal hydraulics as well as stress analyses. The latest stress analysis has been combined with state of the art life prediction models to estimate the time for crack initiation. This paper presents the results of the latest stress analysis and life prediction, including the derivation of the heat transfer coefficient for an annular flow region. The life prediction method uses best estimate strain-temperature histories from elastic-plastic finite element analysis (FEA). Heat-specific material properties have been developed during accompanying tests within the same experimental programme, and have been applied to enable cyclic hardening to be taken into account. The comparison of the prediction to an on-going test is also discussed.
{"title":"A Thermomechanical PWR Test Facility to Investigate Thermal Shock Loading on a Small Scale Tubular Specimen","authors":"Peter Gill, N. Platts, C. Currie, E. Grieveson","doi":"10.1115/PVP2018-84923","DOIUrl":"https://doi.org/10.1115/PVP2018-84923","url":null,"abstract":"Pressurized water reactor environments are known to reduce the fatigue life of austenitic stainless steel components when compared to air environments. Laboratory testing has provided a means of quantifying this, allowing conservative plant assessments to be made. The majority of this testing has been isothermal and carried out on membrane loaded hollow or solid specimens. The geometry and loading of laboratory test specimens is significantly different to that experienced on plant, where complex strain waveforms are generally out of phase with temperature changes, and significant through wall strain gradients may be present. To address the issue of realistic loading, a test facility has been developed which can simulate thermal shock loading on a tubular specimen.\u0000 The capability of the test facility was presented at the PVP2016 conference [PVP2016-63161]. Since then the facility has evolved, with modifications made to the rig configuration and specimen geometry in order to maximize the strain amplitude from the thermal shock, including the adoption of an annular flow geometry. These modifications were designed to optimize both the heat transfer coefficient and the speed of cycling between hot and cold water in order to induce a thermal strain that can cause mechanical failure within practicable test durations. In order to calculate the magnitude of the thermal strain, detailed calculations were required both in terms of thermal hydraulics as well as stress analyses. The latest stress analysis has been combined with state of the art life prediction models to estimate the time for crack initiation.\u0000 This paper presents the results of the latest stress analysis and life prediction, including the derivation of the heat transfer coefficient for an annular flow region. The life prediction method uses best estimate strain-temperature histories from elastic-plastic finite element analysis (FEA). Heat-specific material properties have been developed during accompanying tests within the same experimental programme, and have been applied to enable cyclic hardening to be taken into account. The comparison of the prediction to an on-going test is also discussed.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"1969 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128037614","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}
X. Chen, M. Sokolov, Y. Katoh, M. Rieth, L. Clowers
Eurofer97 is one of leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, high neutron irradiation damage on first wall materials can cause irradiation embrittlement and reduce the fracture toughness of RAFM steels. Therefore, it is critical to select proper testing techniques to characterize the fracture toughness of RAFM steels with high fidelity. In this manuscript, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) × 3.3mm (width) × 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 steel based on the ASTM E1921 Master Curve method. The testing yielded a provisional Master Curve reference temperature ToQ of −89°C of unirradiated Eurofer97 steel heat J362A in the normalized and tempered condition. The results are within the normal scatter range of Master Curve reference temperature T0 for Eurofer97 steel, indicating suitability of applying M4CVN specimens for characterizing the transition fracture toughness of Eurofer97 steel.
{"title":"Master Curve Fracture Toughness Characterization of Eurofer97 Using Miniature Multi-Notch Bend Bar Specimens for Fusion Applications","authors":"X. Chen, M. Sokolov, Y. Katoh, M. Rieth, L. Clowers","doi":"10.1115/PVP2018-85065","DOIUrl":"https://doi.org/10.1115/PVP2018-85065","url":null,"abstract":"Eurofer97 is one of leading candidates of reduced activation ferritic martensitic (RAFM) steels for first wall structural materials of early demonstration fusion power plants. During fusion plant operation, high neutron irradiation damage on first wall materials can cause irradiation embrittlement and reduce the fracture toughness of RAFM steels. Therefore, it is critical to select proper testing techniques to characterize the fracture toughness of RAFM steels with high fidelity. In this manuscript, we present the feasibility study of using pre-cracked miniature multi-notch bend bar specimens (M4CVN) with a dimension of 45mm (length) × 3.3mm (width) × 1.65mm (thickness) to characterize the transition fracture toughness of Eurofer97 steel based on the ASTM E1921 Master Curve method. The testing yielded a provisional Master Curve reference temperature ToQ of −89°C of unirradiated Eurofer97 steel heat J362A in the normalized and tempered condition. The results are within the normal scatter range of Master Curve reference temperature T0 for Eurofer97 steel, indicating suitability of applying M4CVN specimens for characterizing the transition fracture toughness of Eurofer97 steel.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"88 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131988807","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}
During design of Class I components in Nuclear Power Plants, cumulative usage factors (CUF) are conservatively calculated to estimate fatigue damage, and results must be below the limits of the applicable codes. Nevertheless, when these results are used to evaluate the possibility of using these components for an extended life, the results are frequently above code limits. Many Nuclear Power Plants have installed commercial fatigue monitoring systems at critical components in order to assess transient severity and cycle count for life extension fatigue calculations among other reasons. Since the commissioning of the system, unexpected operation modes and thermal stratification was discovered and evaluations needed to be done. Findings, interpretations and solving are presented in this paper.
{"title":"Fatigue Monitoring: Case Studies in Nuclear Power Plant","authors":"G. Bourguigne, F. Schroeter","doi":"10.1115/PVP2018-84007","DOIUrl":"https://doi.org/10.1115/PVP2018-84007","url":null,"abstract":"During design of Class I components in Nuclear Power Plants, cumulative usage factors (CUF) are conservatively calculated to estimate fatigue damage, and results must be below the limits of the applicable codes. Nevertheless, when these results are used to evaluate the possibility of using these components for an extended life, the results are frequently above code limits. Many Nuclear Power Plants have installed commercial fatigue monitoring systems at critical components in order to assess transient severity and cycle count for life extension fatigue calculations among other reasons. Since the commissioning of the system, unexpected operation modes and thermal stratification was discovered and evaluations needed to be done. Findings, interpretations and solving are presented in this paper.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134292940","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}
D. Andrés, R. Lacalle, S. Cicero, J. Álvarez, M. Pinzón
The extension of the operation period of nuclear plants requires an accurate characterisation of the vessel materials, in order to monitor their embrittlement due to neutron irradiation. This need poses a challenge, since the availability of specimens inside the vessels to characterise their evolution is rather scarce. Therefore, innovative techniques have to be applied, in order to reduce the number of tests and the volume of the specimens. In this paper, the Master Curve approach has been employed, combined with the use of small punch notched specimens. The Master Curve methodology allows to evaluate the embrittlement of the material using a single parameter: the reference temperature, T0. This parameter has been estimated for several steels by means of modified small punch specimens, which are characterised by their reduced dimensions: only 10 × 10 × 0.5 mm. The obtained results have been compared with those obtained by means of conventional testing and a methodology to estimate T0 by means of small punch tests together with the Master Curve has been proposed.
{"title":"Estimation of the Reference Temperature, T0, by Means of the Small Punch Testing Technique","authors":"D. Andrés, R. Lacalle, S. Cicero, J. Álvarez, M. Pinzón","doi":"10.1115/PVP2018-84250","DOIUrl":"https://doi.org/10.1115/PVP2018-84250","url":null,"abstract":"The extension of the operation period of nuclear plants requires an accurate characterisation of the vessel materials, in order to monitor their embrittlement due to neutron irradiation. This need poses a challenge, since the availability of specimens inside the vessels to characterise their evolution is rather scarce. Therefore, innovative techniques have to be applied, in order to reduce the number of tests and the volume of the specimens. In this paper, the Master Curve approach has been employed, combined with the use of small punch notched specimens. The Master Curve methodology allows to evaluate the embrittlement of the material using a single parameter: the reference temperature, T0. This parameter has been estimated for several steels by means of modified small punch specimens, which are characterised by their reduced dimensions: only 10 × 10 × 0.5 mm. The obtained results have been compared with those obtained by means of conventional testing and a methodology to estimate T0 by means of small punch tests together with the Master Curve has been proposed.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"18 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124215152","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}
The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.
{"title":"Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Using Mini-C(T) Specimens","authors":"M. Ickes, J. Hall, R. Carter","doi":"10.1115/PVP2018-84967","DOIUrl":"https://doi.org/10.1115/PVP2018-84967","url":null,"abstract":"The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals .\u0000 In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor.\u0000 The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130039849","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}
Hajime Fukumoto, Y. Wada, H. Matsunaga, Takeru Sano, Hiroshi Kobayashi
As is well known, low alloy steels are widely used as materials for high pressure vessels because of their high tensile strength and reasonable price, but also show severe hydrogen embrittlement. Therefore, in 2016, the authors introduced a scenario for the safe use of low alloy steels in highly pressurized hydrogen gas as a “Guideline” at ASME PVP 2016 [1]. Following discussions with stakeholders and experts in recent years, we published Technical Document (TD) as an industrial standard prior to regulation, on the safe use of ground storage vessels made of low alloy steels in Hydrogen Refueling Stations (HRSs) based on performance requirements. This article presents an outline of the TD describing the required types of testing as performance requirements for confirming the good hydrogen compatibility of low alloy steels, such as controlling tensile strength in an appropriate range, confirming leak-before-break, determining the life of ground storage vessels by fatigue testing and determining the inspection term by fatigue crack growth analysis using the fatigue crack growth rate in highly pressurized hydrogen.
{"title":"Introduction of Technical Document in Japan for Safe Use of Ground Storage Vessels Made of Low Alloy Steels for Hydrogen Refueling Stations","authors":"Hajime Fukumoto, Y. Wada, H. Matsunaga, Takeru Sano, Hiroshi Kobayashi","doi":"10.1115/PVP2018-84099","DOIUrl":"https://doi.org/10.1115/PVP2018-84099","url":null,"abstract":"As is well known, low alloy steels are widely used as materials for high pressure vessels because of their high tensile strength and reasonable price, but also show severe hydrogen embrittlement. Therefore, in 2016, the authors introduced a scenario for the safe use of low alloy steels in highly pressurized hydrogen gas as a “Guideline” at ASME PVP 2016 [1]. Following discussions with stakeholders and experts in recent years, we published Technical Document (TD) as an industrial standard prior to regulation, on the safe use of ground storage vessels made of low alloy steels in Hydrogen Refueling Stations (HRSs) based on performance requirements. This article presents an outline of the TD describing the required types of testing as performance requirements for confirming the good hydrogen compatibility of low alloy steels, such as controlling tensile strength in an appropriate range, confirming leak-before-break, determining the life of ground storage vessels by fatigue testing and determining the inspection term by fatigue crack growth analysis using the fatigue crack growth rate in highly pressurized hydrogen.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128874896","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}
The ASME Boiler and Pressure Vessel Code (Section III and Section VIII) provides requirements to avoid a ratcheting (accumulating permanent strain) condition under cyclic thermal load application. The ratchet check in this code is based on the solutions presented by Miller in 1959. One important focus in Miller’s work was to estimate the accumulated plastic strain under cyclic loading. The existing pressure vessels and piping codes have been adopting Miller’s ratchet boundary solution where there is no cyclic plastic accumulation of strain. However, some of these codes also provide limit on accumulated plastic strain under ratcheting conditions. Since the cyclic loading also causes fatigue damage in thee component, the question how to account for the interaction of ratchet deformation, which may contribute to damage in the material, and fatigue damage arises, since the fatigue curves are obtained from tests in the absence of ratcheting. This paper investigates the solutions to calculate growth strain (incremental plastic strain) and their application in design including taking into account the interaction with fatigue. Finite element analysis is presented to validate the analytical solutions.
{"title":"Beyond Shakedown-Ratcheting Boundary","authors":"R. Adibi-Asl, W. Reinhardt","doi":"10.1115/PVP2018-85050","DOIUrl":"https://doi.org/10.1115/PVP2018-85050","url":null,"abstract":"The ASME Boiler and Pressure Vessel Code (Section III and Section VIII) provides requirements to avoid a ratcheting (accumulating permanent strain) condition under cyclic thermal load application. The ratchet check in this code is based on the solutions presented by Miller in 1959. One important focus in Miller’s work was to estimate the accumulated plastic strain under cyclic loading. The existing pressure vessels and piping codes have been adopting Miller’s ratchet boundary solution where there is no cyclic plastic accumulation of strain. However, some of these codes also provide limit on accumulated plastic strain under ratcheting conditions. Since the cyclic loading also causes fatigue damage in thee component, the question how to account for the interaction of ratchet deformation, which may contribute to damage in the material, and fatigue damage arises, since the fatigue curves are obtained from tests in the absence of ratcheting. This paper investigates the solutions to calculate growth strain (incremental plastic strain) and their application in design including taking into account the interaction with fatigue. Finite element analysis is presented to validate the analytical solutions.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"52 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127154212","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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