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}
J. Solin, T. Seppänen, W. Mayinger, H. E. Karabaki
Unexpected findings on time and temperature dependent behavior have been recorded during our research on fatigue performance of niobium stabilized stainless steel. Cyclic straining at 325°C and low strain rates resulted in higher stress responses than in higher rate tests. This effect is particular strong in PWR water environment. Subsurface bulk effect in environment is in contrast to the assumption on similar responses in air and environment, which is the foundation of the ‘companion specimen’ method where the strain in environment is measured from a parallel specimen similarly tested in air. Our data shows that environmental effects caused by PWR water cannot be isolated as a separate issue. Environment, temperature and strain rate are factors, which interactively affect the cyclic response and fatigue performance of stainless steel in relevant temperatures and loading conditions. The current ASME Code Section III design curve is based on different translation of the laboratory data than that made by Langer et al. The resulting effect is not as radical as caused by replacement of the original test data to a new highly scattered data base. But also the procedural changes have detectable effects and open a door for a provocative question: “is the design curve still compatible with the code?”
{"title":"Hidden Roles of Time and Temperature in Cyclic Behavior of Stainless Nuclear Piping","authors":"J. Solin, T. Seppänen, W. Mayinger, H. E. Karabaki","doi":"10.1115/PVP2018-84936","DOIUrl":"https://doi.org/10.1115/PVP2018-84936","url":null,"abstract":"Unexpected findings on time and temperature dependent behavior have been recorded during our research on fatigue performance of niobium stabilized stainless steel.\u0000 Cyclic straining at 325°C and low strain rates resulted in higher stress responses than in higher rate tests. This effect is particular strong in PWR water environment. Subsurface bulk effect in environment is in contrast to the assumption on similar responses in air and environment, which is the foundation of the ‘companion specimen’ method where the strain in environment is measured from a parallel specimen similarly tested in air.\u0000 Our data shows that environmental effects caused by PWR water cannot be isolated as a separate issue. Environment, temperature and strain rate are factors, which interactively affect the cyclic response and fatigue performance of stainless steel in relevant temperatures and loading conditions.\u0000 The current ASME Code Section III design curve is based on different translation of the laboratory data than that made by Langer et al. The resulting effect is not as radical as caused by replacement of the original test data to a new highly scattered data base. But also the procedural changes have detectable effects and open a door for a provocative question:\u0000 “is the design curve still compatible with the code?”","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"10 9","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131687975","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}
Low alloy steels are extensively used in pressure boundary components of nuclear power plants. The structural integrity of the components made of low alloy steels can be evaluated by the procedure of flaw evaluation provided by Section XI of the ASME Boiler and Pressure Vessel Code. According to the Code, the range of stress intensity factor ΔK can be used to determine the fatigue crack growth rates of the material. However, it has been reported that crack closure behavior also strongly influence the fatigue crack growth rate under strong compressive load cycles. This paper discusses the relation between ΔK and the fatigue crack growth rate for cracks in low alloy steels exposed to air. Compressive-tensile cyclic loadings were applied to center-notched plates to obtain the fatigue crack growth curves. The test data demonstrated that effective SIF range ΔKeff more accurately described the crack growth property due to plasticity induced crack closure. Comparing the test results with the reference crack growth curves in the ASME Code Section XI, it may seem that the crack growth prediction based on the Code underestimates the crack growth rates for compressive-tensile cyclic loadings under high stress level.
{"title":"Fatigue Crack Growth in Low Alloy Steels Under Tension-Compression Loading in Air","authors":"Kisaburo Azuma, Y. Yamazaki","doi":"10.1115/PVP2018-84467","DOIUrl":"https://doi.org/10.1115/PVP2018-84467","url":null,"abstract":"Low alloy steels are extensively used in pressure boundary components of nuclear power plants. The structural integrity of the components made of low alloy steels can be evaluated by the procedure of flaw evaluation provided by Section XI of the ASME Boiler and Pressure Vessel Code. According to the Code, the range of stress intensity factor ΔK can be used to determine the fatigue crack growth rates of the material. However, it has been reported that crack closure behavior also strongly influence the fatigue crack growth rate under strong compressive load cycles.\u0000 This paper discusses the relation between ΔK and the fatigue crack growth rate for cracks in low alloy steels exposed to air. Compressive-tensile cyclic loadings were applied to center-notched plates to obtain the fatigue crack growth curves. The test data demonstrated that effective SIF range ΔKeff more accurately described the crack growth property due to plasticity induced crack closure. Comparing the test results with the reference crack growth curves in the ASME Code Section XI, it may seem that the crack growth prediction based on the Code underestimates the crack growth rates for compressive-tensile cyclic loadings under high stress level.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"12 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":"131831745","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 Master Curve (MC) method can be used to directly determine fracture toughness of ferritic reactor pressure vessel (RPV) materials. CRIEPI has been working on the development of a testing technique to apply very small C(T) (called Mini-C(T)) specimens for the MC method. The appropriateness of using Mini-C(T) specimens for several materials including un-irradiated plate, forging, weld metal and irradiated plate has been demonstrated. Through a series of investigations, it was determined that more invalid data, due to ductile crack growth (DCG), can occur when using small size specimens. Linde-80 weld metal, used in the fabrication of some RPVs, is known as low upper shelf material, which tends to exhibit more DCG than high upper shelf materials. In the present study, two sets of 15 Mini-C(T) specimens were machined and pre-cracked from irradiated Linde-80 weld metal. Each set of specimens were provided to two different laboratories (A and B). The laboratories separately conducted the MC tests. DCG occurred even in the lower test temperature condition. About half of specimens for lab A showed excessive DCG and were subjected to the censoring. Some of specimens were rejected since the test temperature is outside of the specified range (T-To < −50°C). As a result, lab A could not obtain valid To with 15 specimens. Lab B also experienced DCG, however were able to obtain a sufficient number of valid KJc data points to determine a valid To. The obtained ToQ (lab A) and To (lab B) are sufficiently close to each other and suggests that Mini-C(T) can be used even for the low upper shelf material if the number of available specimens are sufficiently large. The combined dataset from labs A and B estimated To = 31.5°C, which is in the scatter band of To obtained by pre-cracked Chapry (PCCv), 0.5TC(T) or 1TC(T) specimens in a past Heavy-Section Steel Irradiation (HSSI) project. The overall result suggests that To can be estimated using Mini-C(T) specimens for the lower upper shelf weld material, but 15 is a marginal number of specimens for a valid estimation.
{"title":"Trial Study of the Master Curve Fracture Toughness Evaluation by Mini-C(T) Specimens for Low Upper Shelf Weld Metal Linde-80","authors":"Masato Yamamoto","doi":"10.1115/PVP2018-84906","DOIUrl":"https://doi.org/10.1115/PVP2018-84906","url":null,"abstract":"The Master Curve (MC) method can be used to directly determine fracture toughness of ferritic reactor pressure vessel (RPV) materials. CRIEPI has been working on the development of a testing technique to apply very small C(T) (called Mini-C(T)) specimens for the MC method. The appropriateness of using Mini-C(T) specimens for several materials including un-irradiated plate, forging, weld metal and irradiated plate has been demonstrated. Through a series of investigations, it was determined that more invalid data, due to ductile crack growth (DCG), can occur when using small size specimens.\u0000 Linde-80 weld metal, used in the fabrication of some RPVs, is known as low upper shelf material, which tends to exhibit more DCG than high upper shelf materials. In the present study, two sets of 15 Mini-C(T) specimens were machined and pre-cracked from irradiated Linde-80 weld metal. Each set of specimens were provided to two different laboratories (A and B). The laboratories separately conducted the MC tests. DCG occurred even in the lower test temperature condition. About half of specimens for lab A showed excessive DCG and were subjected to the censoring. Some of specimens were rejected since the test temperature is outside of the specified range (T-To < −50°C). As a result, lab A could not obtain valid To with 15 specimens. Lab B also experienced DCG, however were able to obtain a sufficient number of valid KJc data points to determine a valid To. The obtained ToQ (lab A) and To (lab B) are sufficiently close to each other and suggests that Mini-C(T) can be used even for the low upper shelf material if the number of available specimens are sufficiently large. The combined dataset from labs A and B estimated To = 31.5°C, which is in the scatter band of To obtained by pre-cracked Chapry (PCCv), 0.5TC(T) or 1TC(T) specimens in a past Heavy-Section Steel Irradiation (HSSI) project. The overall result suggests that To can be estimated using Mini-C(T) specimens for the lower upper shelf weld material, but 15 is a marginal number of specimens for a valid estimation.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"28 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":"132912374","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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