Alexandria M. Scott, Louis W. Turicik, J. Hall, A. Udyawar, Amy E. Freed, Elliot J. Long
The low alloy steel pressurizer (PZR) vessels in pressurized water reactor (PWR) nuclear power plants are potentially susceptible to embrittlement due to thermal ageing over the life of the plant (40–80 years). This paper determines the amount of PZR thermal ageing embrittlement, which can be accommodated based on a comparison of PZR and the U.S. NRC approved reactor pressure vessel (RPV) 10 CFR 50, Appendix G Pressure-Temperature (P-T) limit curves for the current operating U.S. PWR fleet. The maximum amount of postulated thermal ageing embrittlement, in terms of a shift in nil-ductility reference temperature (ΔRTNDT), which is permissible before the generic PZR P-T limit curves exceed the NRC-approved RPV P-T limit curves is provided in this paper. The generic P-T limit curves are determined for current operating U.S. PWR representative Westinghouse and Combustion Engineering (CE) PZR designs for various levels of postulated thermal ageing embrittlement. The locations for consideration within the PZR are the cylindrical shell to bottom head girth weld (includes consideration of the adjacent shell longitudinal seam weld), lower head region in the vicinity of the heater sleeve penetrations, and the surge nozzle corner region. The methodology to calculate the PZR P-T limit curves is per 10 CFR 50, Appendix G and the 2017 Edition of ASME Section XI, Appendix G. The PZR thermal ageing ΔRTNDT values determined in this paper could be compared to estimated or empirical values of thermal ageing embrittlement to determine if or when PZR embrittlement may impact a plant’s 10 CFR 50, Appendix G heatup and cooldown P T limit curves, and any primary loop pressure boundary design fracture mechanics evaluations.
{"title":"Impact of Thermal Ageing Embrittlement on Westinghouse and Combustion Engineering-Designed Pressurized Water Reactor Pressurizers Based on Pressure-Temperature Limit Comparison","authors":"Alexandria M. Scott, Louis W. Turicik, J. Hall, A. Udyawar, Amy E. Freed, Elliot J. Long","doi":"10.1115/pvp2022-85520","DOIUrl":"https://doi.org/10.1115/pvp2022-85520","url":null,"abstract":"\u0000 The low alloy steel pressurizer (PZR) vessels in pressurized water reactor (PWR) nuclear power plants are potentially susceptible to embrittlement due to thermal ageing over the life of the plant (40–80 years). This paper determines the amount of PZR thermal ageing embrittlement, which can be accommodated based on a comparison of PZR and the U.S. NRC approved reactor pressure vessel (RPV) 10 CFR 50, Appendix G Pressure-Temperature (P-T) limit curves for the current operating U.S. PWR fleet. The maximum amount of postulated thermal ageing embrittlement, in terms of a shift in nil-ductility reference temperature (ΔRTNDT), which is permissible before the generic PZR P-T limit curves exceed the NRC-approved RPV P-T limit curves is provided in this paper. The generic P-T limit curves are determined for current operating U.S. PWR representative Westinghouse and Combustion Engineering (CE) PZR designs for various levels of postulated thermal ageing embrittlement. The locations for consideration within the PZR are the cylindrical shell to bottom head girth weld (includes consideration of the adjacent shell longitudinal seam weld), lower head region in the vicinity of the heater sleeve penetrations, and the surge nozzle corner region. The methodology to calculate the PZR P-T limit curves is per 10 CFR 50, Appendix G and the 2017 Edition of ASME Section XI, Appendix G. The PZR thermal ageing ΔRTNDT values determined in this paper could be compared to estimated or empirical values of thermal ageing embrittlement to determine if or when PZR embrittlement may impact a plant’s 10 CFR 50, Appendix G heatup and cooldown P T limit curves, and any primary loop pressure boundary design fracture mechanics evaluations.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"21 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126807678","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 impact test has historically been used in a qualitative and comparative manner to infer toughness behaviour and determine the brittle to ductile transition temperature (TBD) of low alloy ferritic steels used in reactor pressure vessels (RPVs). The simple and quick setup makes it an attractive test given the ease of data generation to assess the suitability of a given material; however, the scatter in the data produced is significant and the test does not provide a value of fracture toughness. Quasi-static tests using high-constraint geometries (e.g. single-edge notch bend (SENB) specimens) are used to determine fracture toughness properties, whilst the Charpy impact test (governed by the ASTM E23 and ISO 148 standards) gives insight into the dynamic fracture response of a material. There is significant interest, demonstrated by recent work, in utilising Charpy impact test data to predict fracture toughness properties and material behaviour, which typically require expensive and time-consuming test procedures. The ongoing digital transformation of industry and proposals of digital twins becoming ubiquitous relies intrinsically on high-quality data inputs and fully understanding the underlying mechanistic relationships governing material behaviour. This work examines the relationships between microstructure, temperature, and quasi-static and dynamic fracture behaviour of a low alloy ferritic steel (comparable in composition to SA508). The microstructures are analysed before a Charpy impact pendulum is used to determine the energy absorbed by standard V-notch samples from −196 °C to 200 °C and the fracture surfaces examined. A distinct transition zone is observed and the data is compared to historic fracture data of the material. The results are discussed in light of applicability to a digital twin and the framework for a machine learning model to predict the fracture behaviour and reduce error in transition behaviour is proposed.
{"title":"An Investigation Into the Applicability of Charpy Dynamic Fracture Tests for a Digital Twin","authors":"Fabian S. Sorce, D. Cogswell, C. Davies","doi":"10.1115/pvp2022-84857","DOIUrl":"https://doi.org/10.1115/pvp2022-84857","url":null,"abstract":"\u0000 The Charpy impact test has historically been used in a qualitative and comparative manner to infer toughness behaviour and determine the brittle to ductile transition temperature (TBD) of low alloy ferritic steels used in reactor pressure vessels (RPVs). The simple and quick setup makes it an attractive test given the ease of data generation to assess the suitability of a given material; however, the scatter in the data produced is significant and the test does not provide a value of fracture toughness. Quasi-static tests using high-constraint geometries (e.g. single-edge notch bend (SENB) specimens) are used to determine fracture toughness properties, whilst the Charpy impact test (governed by the ASTM E23 and ISO 148 standards) gives insight into the dynamic fracture response of a material. There is significant interest, demonstrated by recent work, in utilising Charpy impact test data to predict fracture toughness properties and material behaviour, which typically require expensive and time-consuming test procedures. The ongoing digital transformation of industry and proposals of digital twins becoming ubiquitous relies intrinsically on high-quality data inputs and fully understanding the underlying mechanistic relationships governing material behaviour. This work examines the relationships between microstructure, temperature, and quasi-static and dynamic fracture behaviour of a low alloy ferritic steel (comparable in composition to SA508). The microstructures are analysed before a Charpy impact pendulum is used to determine the energy absorbed by standard V-notch samples from −196 °C to 200 °C and the fracture surfaces examined. A distinct transition zone is observed and the data is compared to historic fracture data of the material. The results are discussed in light of applicability to a digital twin and the framework for a machine learning model to predict the fracture behaviour and reduce error in transition behaviour is proposed.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"45 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132231298","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}
Laser powder bed fusion (LPBF) is an additive manufacture technique which builds components up in layers from a powder feedstock, using a scanning laser to selectively melt the powder into the required shape. The process of LPBF can often introduce defects into the structure of a part, since the powder may not fully melt and leave holes, or pores, in the sample. Excessive laser power may also cause the powder to vaporise and create pores. In whatever manner these pores are formed, they can significantly impact the properties of the finished component. Since pores and small defects already exist in LPBF components, the void growth and ductile fracture behaviour of LPBF components under multiaxial stress conditions needs to be characterised and predicted. In this work, notched bar tensile tests have been performed on samples with a range of notch acuities and hence multiaxial stress states. These tests have enabled ductile damage models to be calibrated and finite element (FE) simulations of the notched bar tests performed. The model was validated by comparison to the experimental results. The model agrees well with the results in many cases assessed in this work, but sometimes suffers from discrepancies and premature failure due to variability in material tensile properties, emphasising the need for sensitivity studies.
{"title":"Ductile Damage Model Development and Validation of 316L Laser Powder Bed Fusion Steel Under Multiaxial Stress Conditions","authors":"Theo Hales, T. Ronneberg, P. Hooper, C. Davies","doi":"10.1115/pvp2022-84785","DOIUrl":"https://doi.org/10.1115/pvp2022-84785","url":null,"abstract":"\u0000 Laser powder bed fusion (LPBF) is an additive manufacture technique which builds components up in layers from a powder feedstock, using a scanning laser to selectively melt the powder into the required shape. The process of LPBF can often introduce defects into the structure of a part, since the powder may not fully melt and leave holes, or pores, in the sample. Excessive laser power may also cause the powder to vaporise and create pores. In whatever manner these pores are formed, they can significantly impact the properties of the finished component. Since pores and small defects already exist in LPBF components, the void growth and ductile fracture behaviour of LPBF components under multiaxial stress conditions needs to be characterised and predicted. In this work, notched bar tensile tests have been performed on samples with a range of notch acuities and hence multiaxial stress states. These tests have enabled ductile damage models to be calibrated and finite element (FE) simulations of the notched bar tests performed. The model was validated by comparison to the experimental results. The model agrees well with the results in many cases assessed in this work, but sometimes suffers from discrepancies and premature failure due to variability in material tensile properties, emphasising the need for sensitivity studies.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"16 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124190437","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}
Due to the complex working environment, the defects were easily produced on the surface of pressure vessel plate, such as scratch, wear and corrosion pit, which reduced the service life of equipment. As a new surface coating preparation technology, cold spraying technology can be used to effectively repair and protect the substrate. With the development of cold spraying technology, more and more metal powders are used to prepare functional coatings, but aluminum powders are used most frequently due to significant plasticity and corrosion resistance. However, pure aluminum coating have a obvious shortcoming with lower bonding strength, and most of bonding strength values of coatings and steels are about 14.6 MPa. With the aim to improve the mechanical properties more of pure Al, the most representative one is the inclusion of ceramic particles as reinforcement to produce dense coatings. Thus, the Al-Al2O3 composite coatings of the different Al2O3 weight fractions were deposited on the surface of pressure vessel plate by cold spraying technology. The strength and corrosion resistance of the coatings were evaluated by tensile test, corrosion weight loss measurement and electrochemical test. The results show that the bonding strength of pure Al coating is the lowest among the four cold spray coatings, and the bonding strength sharply increases while the coating increased Al2O3 particles, and the largest values with bonding strength of Al-Al2O3 is as high as 45.4 MPa. The tensile test was carried out under the allowable stress of pressure vessel. Observations under optical microscope (OM) were also done, and the coating has excellent quality and no new cracks and holes. The corrosion weight-loss of the substrate and composite coatings were measured, and the corrosion weight loss rate of Al-Al2O3 was 5 times lower than that of pressure vessel plate. In addition, for all the coatings, the values of weight loss had little changes. It is observed that the values of the composite coating was exhibited a peak with an increasing of the Al2O3 content. The Al-20wt.%Al2O3 was shown the best corrosion resistance and the value of weight loss was 0.11(g/cm2*h), which probably was attributed to the effect of the lower porosity. Potentiodynamic polarization curves were shown the corrosion current density of composite coatings were one order of magnitude lower than that of the substrate. Therefore, we concluded that the corrosion resistance is obviously better than that of the substrate, which can effectively protect the substrate and delay the service life of the pressure vessel plate.
{"title":"Study on the Bonding Strength and Corrosion Resistance of Low-Pressure Cold Sprayed Al/Al2O3 Composite Coatings on Pressure Vessel Steel Substrate","authors":"Yonggang Wang, Xin Liu, Liang Sun","doi":"10.1115/pvp2022-84489","DOIUrl":"https://doi.org/10.1115/pvp2022-84489","url":null,"abstract":"\u0000 Due to the complex working environment, the defects were easily produced on the surface of pressure vessel plate, such as scratch, wear and corrosion pit, which reduced the service life of equipment. As a new surface coating preparation technology, cold spraying technology can be used to effectively repair and protect the substrate. With the development of cold spraying technology, more and more metal powders are used to prepare functional coatings, but aluminum powders are used most frequently due to significant plasticity and corrosion resistance. However, pure aluminum coating have a obvious shortcoming with lower bonding strength, and most of bonding strength values of coatings and steels are about 14.6 MPa. With the aim to improve the mechanical properties more of pure Al, the most representative one is the inclusion of ceramic particles as reinforcement to produce dense coatings. Thus, the Al-Al2O3 composite coatings of the different Al2O3 weight fractions were deposited on the surface of pressure vessel plate by cold spraying technology. The strength and corrosion resistance of the coatings were evaluated by tensile test, corrosion weight loss measurement and electrochemical test. The results show that the bonding strength of pure Al coating is the lowest among the four cold spray coatings, and the bonding strength sharply increases while the coating increased Al2O3 particles, and the largest values with bonding strength of Al-Al2O3 is as high as 45.4 MPa. The tensile test was carried out under the allowable stress of pressure vessel. Observations under optical microscope (OM) were also done, and the coating has excellent quality and no new cracks and holes. The corrosion weight-loss of the substrate and composite coatings were measured, and the corrosion weight loss rate of Al-Al2O3 was 5 times lower than that of pressure vessel plate. In addition, for all the coatings, the values of weight loss had little changes. It is observed that the values of the composite coating was exhibited a peak with an increasing of the Al2O3 content. The Al-20wt.%Al2O3 was shown the best corrosion resistance and the value of weight loss was 0.11(g/cm2*h), which probably was attributed to the effect of the lower porosity. Potentiodynamic polarization curves were shown the corrosion current density of composite coatings were one order of magnitude lower than that of the substrate. Therefore, we concluded that the corrosion resistance is obviously better than that of the substrate, which can effectively protect the substrate and delay the service life of the pressure vessel plate.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"80 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130618341","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. Dutilleul, Robert Widdison, John Crossley, W. Kyffin, M. Albert, D. Gandy
As part of a Department of Energy (DOE) funded programme assessing advanced manufacturing techniques for SMR applications, the Nuclear AMRC and EPRI have been developing Electron Beam Welding (EBW) parameters and procedures based upon SA508 Grade 3 Class 1 base material. For linear EB welds, the start and stop regions can be managed by using sacrificial run on/off blocks. However, for circumferential welds, such as joining shell to flange components, the use of sacrificial material is not possible. As such an effective method of closing the keyhole must be developed to ensure that weld defects are not entrained. This process is termed ‘slope out’ welding. This paper presents results from the steady state welding in the 80–90 mm material thickness range, showing that weld properties meet specification requirements. Subsequently the paper describes the steps in developing an effective slope out welding procedure for circumferential welds. Weld quality was assured by Phased Array Ultrasonic Testing (PAUT) in conjunction with weld sectioning. All welds were assessed against ASME V requirements. The results presented in this study clearly indicate that defect free and repeatable electron beam welds, including the slope out region, can be produced on thick section pressure vessel steel.
{"title":"Slope Out Welding Development for Thick Section Electron Beam Welding for Pressure Vessel Applications","authors":"T. Dutilleul, Robert Widdison, John Crossley, W. Kyffin, M. Albert, D. Gandy","doi":"10.1115/pvp2022-85478","DOIUrl":"https://doi.org/10.1115/pvp2022-85478","url":null,"abstract":"\u0000 As part of a Department of Energy (DOE) funded programme assessing advanced manufacturing techniques for SMR applications, the Nuclear AMRC and EPRI have been developing Electron Beam Welding (EBW) parameters and procedures based upon SA508 Grade 3 Class 1 base material.\u0000 For linear EB welds, the start and stop regions can be managed by using sacrificial run on/off blocks. However, for circumferential welds, such as joining shell to flange components, the use of sacrificial material is not possible. As such an effective method of closing the keyhole must be developed to ensure that weld defects are not entrained. This process is termed ‘slope out’ welding.\u0000 This paper presents results from the steady state welding in the 80–90 mm material thickness range, showing that weld properties meet specification requirements. Subsequently the paper describes the steps in developing an effective slope out welding procedure for circumferential welds. Weld quality was assured by Phased Array Ultrasonic Testing (PAUT) in conjunction with weld sectioning. All welds were assessed against ASME V requirements.\u0000 The results presented in this study clearly indicate that defect free and repeatable electron beam welds, including the slope out region, can be produced on thick section pressure vessel steel.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"27 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132032006","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}
A. Wisbey, David Coon, Mark Chatterton, Josh Barras, D. Guo, Kun Yan, M. Callaghan, W. Mirihanage
Additive manufacturing (AM) offers the potential for significantly reducing the time and cost of new nuclear components. This process may also permit unique design features, for example internal geometries. However, the limitations of the technology need to be better understood to enable implementation and accreditation. Here a “blown powder” and laser melting process, within a helium shielded environment, was used to fabricate austenitic stainless steel 316L walls of ∼2.4 mm thickness, with the deposition parameters minimizing the surface roughness. A key aim was to evaluate the effect of the as-deposited surface finish and the bulk material on the tensile and fatigue properties. In addition, the effect of material orientation was also considered to be important. Microstructural characterization demonstrated the complex nature of the grain morphology arising from the as-manufactured AM process, including elongated grains following the thermal gradients. However, areas of equiaxed grains were also observed at the sample surfaces. Si-Mn-O particles, up to ∼20 μm in diameter, were noted throughout the samples produced. Residual strains have also been measured and correlated with microstructural features. The tensile performance was generally similar to wrought 316L material but exhibited some anisotropy. The fatigue endurance of as-deposited AM 316L was significantly lower than wrought material. However, surface grinding of the AM 316L was shown to be beneficial. It was noted that in all cases examined, fatigue crack initiation was found to occur at the Si-Mn-O particles, in both surface finishes — clearly a performance limitation.
{"title":"The Mechanical Performance of Additively Manufactured 316L Austenitic Stainless Steel","authors":"A. Wisbey, David Coon, Mark Chatterton, Josh Barras, D. Guo, Kun Yan, M. Callaghan, W. Mirihanage","doi":"10.1115/pvp2022-84543","DOIUrl":"https://doi.org/10.1115/pvp2022-84543","url":null,"abstract":"\u0000 Additive manufacturing (AM) offers the potential for significantly reducing the time and cost of new nuclear components. This process may also permit unique design features, for example internal geometries. However, the limitations of the technology need to be better understood to enable implementation and accreditation.\u0000 Here a “blown powder” and laser melting process, within a helium shielded environment, was used to fabricate austenitic stainless steel 316L walls of ∼2.4 mm thickness, with the deposition parameters minimizing the surface roughness.\u0000 A key aim was to evaluate the effect of the as-deposited surface finish and the bulk material on the tensile and fatigue properties. In addition, the effect of material orientation was also considered to be important.\u0000 Microstructural characterization demonstrated the complex nature of the grain morphology arising from the as-manufactured AM process, including elongated grains following the thermal gradients. However, areas of equiaxed grains were also observed at the sample surfaces. Si-Mn-O particles, up to ∼20 μm in diameter, were noted throughout the samples produced. Residual strains have also been measured and correlated with microstructural features.\u0000 The tensile performance was generally similar to wrought 316L material but exhibited some anisotropy. The fatigue endurance of as-deposited AM 316L was significantly lower than wrought material. However, surface grinding of the AM 316L was shown to be beneficial. It was noted that in all cases examined, fatigue crack initiation was found to occur at the Si-Mn-O particles, in both surface finishes — clearly a performance limitation.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"11 3 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123682066","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. Katsuyama, Yoshihito Yamaguchi, Y. Nemoto, T. Furuta, Y. Kaji
To assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi, we have been developing an analysis method based on coupled analysis of three-dimensional multi-physics simulations composed of radiation transport, thermal-hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this simulation, Monte Carlo radiation transport calculation is firstly performed by using PHITS code to compute proton dose distribution considering molten conditions of core materials. Then the deposit energies at each location is imported into TH analysis code ANSYS Fluent with the same geometry and temperature distribution is simulated by thermal-fluid dynamics. Finally, temperature distribution obtained from TH analysis is applied to thermal-elastic-plastic-creep analyses using FINAS-STAR and then damage evaluation is carried out based on several criterions such as Kachanov, Larson-Miller-parameter, melting point. To conduct such analyses, we also have continued to obtain experimental data on creep deformation in high temperature range. In this study, to predict time and location of reactor pressure vessel (RPV) lower head rupture of boiling water reactors (BWRs) considering creep damage mechanisms, we performed creep damage evaluations based on developing analysis method by using detailed three-dimensional model of RPV lower head with control rod guide tubes, stub tubes and welds. From the detailed analysis results, it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.
{"title":"Damage Evaluations for BWR Lower Head in Severe Accident Based on Multi-Physics Simulations","authors":"J. Katsuyama, Yoshihito Yamaguchi, Y. Nemoto, T. Furuta, Y. Kaji","doi":"10.1115/pvp2022-84609","DOIUrl":"https://doi.org/10.1115/pvp2022-84609","url":null,"abstract":"\u0000 To assess rupture behavior of the lower head of reactor pressure vessel in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi, we have been developing an analysis method based on coupled analysis of three-dimensional multi-physics simulations composed of radiation transport, thermal-hydraulics (TH) and thermal-elastic-plastic-creep analyses. In this simulation, Monte Carlo radiation transport calculation is firstly performed by using PHITS code to compute proton dose distribution considering molten conditions of core materials. Then the deposit energies at each location is imported into TH analysis code ANSYS Fluent with the same geometry and temperature distribution is simulated by thermal-fluid dynamics. Finally, temperature distribution obtained from TH analysis is applied to thermal-elastic-plastic-creep analyses using FINAS-STAR and then damage evaluation is carried out based on several criterions such as Kachanov, Larson-Miller-parameter, melting point. To conduct such analyses, we also have continued to obtain experimental data on creep deformation in high temperature range. In this study, to predict time and location of reactor pressure vessel (RPV) lower head rupture of boiling water reactors (BWRs) considering creep damage mechanisms, we performed creep damage evaluations based on developing analysis method by using detailed three-dimensional model of RPV lower head with control rod guide tubes, stub tubes and welds. From the detailed analysis results, it was concluded that failure regions of BWR lower head are only the control rod guide tubes or stub tubes under simulated conditions.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121335324","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}
P. Arffman, J. Lydman, N. Hytönen, Z. Que, S. Lindqvist
Project BRUTE has investigated weld materials extracted from the decommissioned Barsebäck 2 reactor pressure vessel. The materials investigated originate from the pressure vessel head (RPVH) and beltline regions. The performed mechanical testing include tensile, Charpy impact and fracture toughness testing. Tensile testing with miniature specimens demonstrates a difference of over 50 MPa in the yield and ultimate tensile strengths of the RPVH and beltline materials. Beltline specimens tested at the operating temperature exhibit discontinuity past the yield region, possibly indicating dynamic strain aging. Charpy impact tests were performed around the transition region of the material. Transition curves were fitted, and reference temperatures T28J of −85 °C and −106 °C were determined for RPVH and beltline materials, respectively. This indicates better material properties at beltline compared to the RPVH, in agreement with tensile results. The reference temperatures T28J were further utilized to estimate brittle fracture initiation toughness reference temperatures T0. Fracture toughness testing follows the Master Curve methodology defined in ASTM standard E1921. Reference temperatures T0 were determined at −115.1 °C and −101.1 °C for the RPVH and beltline, respectively, but the tests indicate inhomogeneity in both materials. The mean reference temperatures of the multimodal models TM were determined at −110.0 °C and −96.5 °C, and the associated, margin adjusted lower confidence bounds TM5%, MA at −13.3 °C and −53.0 °C for the RPVH and beltline materials, respectively. The latter values indicate that the inhomogeneity is more extensive in the RPVH. The estimates obtained from the Charpy impact toughness results do not correlate consistently with the fracture toughness-based transition temperature, possibly due to the inhomogeneity of the materials. The results show that the safety of the materials can be assessed reliably, provided that contemporary methods, equipment and analyses are used.
{"title":"BRUTE: Evaluation of Mechanical Properties of True Reactor Pressure Vessel Material From Barsebäck 2","authors":"P. Arffman, J. Lydman, N. Hytönen, Z. Que, S. Lindqvist","doi":"10.1115/pvp2022-83819","DOIUrl":"https://doi.org/10.1115/pvp2022-83819","url":null,"abstract":"\u0000 Project BRUTE has investigated weld materials extracted from the decommissioned Barsebäck 2 reactor pressure vessel. The materials investigated originate from the pressure vessel head (RPVH) and beltline regions. The performed mechanical testing include tensile, Charpy impact and fracture toughness testing.\u0000 Tensile testing with miniature specimens demonstrates a difference of over 50 MPa in the yield and ultimate tensile strengths of the RPVH and beltline materials. Beltline specimens tested at the operating temperature exhibit discontinuity past the yield region, possibly indicating dynamic strain aging.\u0000 Charpy impact tests were performed around the transition region of the material. Transition curves were fitted, and reference temperatures T28J of −85 °C and −106 °C were determined for RPVH and beltline materials, respectively. This indicates better material properties at beltline compared to the RPVH, in agreement with tensile results. The reference temperatures T28J were further utilized to estimate brittle fracture initiation toughness reference temperatures T0.\u0000 Fracture toughness testing follows the Master Curve methodology defined in ASTM standard E1921. Reference temperatures T0 were determined at −115.1 °C and −101.1 °C for the RPVH and beltline, respectively, but the tests indicate inhomogeneity in both materials. The mean reference temperatures of the multimodal models TM were determined at −110.0 °C and −96.5 °C, and the associated, margin adjusted lower confidence bounds TM5%, MA at −13.3 °C and −53.0 °C for the RPVH and beltline materials, respectively. The latter values indicate that the inhomogeneity is more extensive in the RPVH. The estimates obtained from the Charpy impact toughness results do not correlate consistently with the fracture toughness-based transition temperature, possibly due to the inhomogeneity of the materials.\u0000 The results show that the safety of the materials can be assessed reliably, provided that contemporary methods, equipment and analyses are used.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129467344","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 fatigue analyses included in nuclear rules of KTA [1]–[2] and ASME [3] are based on defined loads (specified or measured loads and frequencies). It is assumed that highly cyclic loadings or resonance vibrations are avoided by appropriate design. Often these loads are recorded by measurements during commissioning or during operation. In particular, in pressure vessel and reactor internals such vibrational excitations cannot be excluded, so that fatigue loadings in the HCF regime and even in the VHCF regime can occur. Since these kinds of loading situations also appear in combination with fatigue loadings in the LCF regime, load collectives are to be considered, as they are not explicitly taken into account in the current analysis of the nuclear regulations. Furthermore, no generally validated method, especially a consolidated damage accumulation model is available. Furthermore, design fatigue curves for austenitic steels in the applicable international design codes were extended by extrapolation from originally 106 up to 1011 load cycles [1]–[3]. However, the existing database for load cycles equal to or above 107 is still insufficient. Therefore, international efforts are currently ongoing in order to expand the database through international co-operations and compile a safe high cycle fatigue (HCF) database [4]. This is particularly important in combination with the influence of the cooling medium and its consideration according to established international standards as the database of the Argonne National Laboratory ANL [5] for fatigue behavior under medium conditions. For the range from HCF to VHCF and for their combination with LCF loads (collective effect) and the currently discussed limit values above which the cooling medium has an effective influence on the fatigue strength are not sufficiently consolidated. These aspects gain in importance particularly in the long-term operation context. A recently finished cooperative research project aims at contributing to closing these mentioned gaps by generation of a data and assessment basis for the fatigue behavior of welded austenitic stainless steels at high numbers of load cycles [6]. The following topics will be discussed in detail in the paper: • Fatigue behavior at variable amplitude loading (combination of LCF / HCF and LCF / VHCF) • Development of a fatigue assessment methodology under consideration of the transient endurance limit and damage accumulation effects including assessment and adaptation of appropriate fatigue damage parameters
KTA[1] -[2]和ASME[3]核规则中包含的疲劳分析基于定义载荷(指定或测量的载荷和频率)。假设通过适当的设计可以避免高循环载荷或共振振动。通常这些负载是在调试或运行期间通过测量记录下来的。特别是在压力容器和反应堆内部，不能排除这种振动激励，因此在HCF状态甚至在VHCF状态下都可能发生疲劳载荷。由于这些类型的载荷情况也与LCF制度中的疲劳载荷结合出现，因此需要考虑载荷集体，因为它们在当前的核法规分析中没有明确考虑。而且，目前还没有一个普遍有效的方法，特别是一个统一的损伤累积模型。此外，在适用的国际设计规范中，奥氏体钢的设计疲劳曲线通过外推从原来的106个荷载循环扩展到1011个荷载循环[1]-[3]。但是，负载周期等于或大于107的现有数据库仍然不够。因此，目前国际上正在努力通过国际合作扩大数据库，并编制安全高循环疲劳(HCF)数据库[4]。考虑到冷却介质的影响，并根据既定的国际标准(如Argonne National Laboratory ANL[5]的介质条件下疲劳行为数据库)来考虑，这一点尤为重要。对于从HCF到VHCF的范围，以及它们与LCF载荷的组合(集体效应)，以及目前讨论的冷却介质对疲劳强度有有效影响的限值没有得到充分的整合。这些方面在长期操作环境中尤为重要。最近完成的一项合作研究项目旨在通过生成焊接奥氏体不锈钢在高载荷循环次数下的疲劳行为的数据和评估基础来弥补上述差距[6]。本文将详细讨论以下主题:•可变振幅载荷下的疲劳行为(LCF / HCF和LCF / VHCF的组合)•考虑瞬态耐久性极限和损伤累积效应的疲劳评估方法的开发，包括评估和适应适当的疲劳损伤参数
{"title":"Investigations on Multi-Stage Tests and Transient Endurance Limit Behavior Under Low-, High- and Very High Cycle Fatigue Loads","authors":"T. Schopf, S. Weihe, J. Rudolph","doi":"10.1115/pvp2022-84718","DOIUrl":"https://doi.org/10.1115/pvp2022-84718","url":null,"abstract":"\u0000 The fatigue analyses included in nuclear rules of KTA [1]–[2] and ASME [3] are based on defined loads (specified or measured loads and frequencies). It is assumed that highly cyclic loadings or resonance vibrations are avoided by appropriate design. Often these loads are recorded by measurements during commissioning or during operation. In particular, in pressure vessel and reactor internals such vibrational excitations cannot be excluded, so that fatigue loadings in the HCF regime and even in the VHCF regime can occur. Since these kinds of loading situations also appear in combination with fatigue loadings in the LCF regime, load collectives are to be considered, as they are not explicitly taken into account in the current analysis of the nuclear regulations. Furthermore, no generally validated method, especially a consolidated damage accumulation model is available. Furthermore, design fatigue curves for austenitic steels in the applicable international design codes were extended by extrapolation from originally 106 up to 1011 load cycles [1]–[3]. However, the existing database for load cycles equal to or above 107 is still insufficient. Therefore, international efforts are currently ongoing in order to expand the database through international co-operations and compile a safe high cycle fatigue (HCF) database [4]. This is particularly important in combination with the influence of the cooling medium and its consideration according to established international standards as the database of the Argonne National Laboratory ANL [5] for fatigue behavior under medium conditions. For the range from HCF to VHCF and for their combination with LCF loads (collective effect) and the currently discussed limit values above which the cooling medium has an effective influence on the fatigue strength are not sufficiently consolidated. These aspects gain in importance particularly in the long-term operation context. A recently finished cooperative research project aims at contributing to closing these mentioned gaps by generation of a data and assessment basis for the fatigue behavior of welded austenitic stainless steels at high numbers of load cycles [6].\u0000 The following topics will be discussed in detail in the paper:\u0000 • Fatigue behavior at variable amplitude loading (combination of LCF / HCF and LCF / VHCF)\u0000 • Development of a fatigue assessment methodology under consideration of the transient endurance limit and damage accumulation effects including assessment and adaptation of appropriate fatigue damage parameters","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129746579","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}
Satoshi Kumagai, Kiminobu Hojo, T. Hirota, Keiji Kawanishi
The European project “Advanced Structural Integrity Assessment Tools for Safe Long Term Operation (ATLAS+)” is aiming to achieve a reasonable balance between safety and long term operation of nuclear reactor pressure coolant boundary systems. In the project, the Work Package 3 deals with development and validation of finite element analysis (FEA) using the Gurson-Tvergaard-Needleman (GTN) model to predict a ductile crack growth behavior of a pipe structure from those of laboratory specimens such as SE(T) and C(T) specimens. By using the parameter sets of the GTN model obtained by other members in the project, the authors predicted the fracture behaviors of three pipes with a circumferential through-wall crack or a circumferential outer surface crack by the four-point bending test conducted by a member of the project. As a result, the predicted maximum loads agreed with those of experiments within the error of 2.2 %, and the predicted ductile crack growth amounts were nearly equal to those on the fractured surface. In addition, the J-resistance of the pipe with a though-wall crack was calculated by the FEA by the node release technique with a crack growth criterion according to the crack growth predicted by the GTN model. Consequently, the J-resistance of the pipe is 3.1 to 3.6 times larger than that of C(T) specimens, which means that a prediction by the conventional fracture mechanics has a lot of margin to the actual fracture behavior.
{"title":"Ductile Tearing Prediction of Ferritic Pipes by GTN Model for ATLAS+ European Project (Report 2)","authors":"Satoshi Kumagai, Kiminobu Hojo, T. Hirota, Keiji Kawanishi","doi":"10.1115/pvp2022-84554","DOIUrl":"https://doi.org/10.1115/pvp2022-84554","url":null,"abstract":"\u0000 The European project “Advanced Structural Integrity Assessment Tools for Safe Long Term Operation (ATLAS+)” is aiming to achieve a reasonable balance between safety and long term operation of nuclear reactor pressure coolant boundary systems. In the project, the Work Package 3 deals with development and validation of finite element analysis (FEA) using the Gurson-Tvergaard-Needleman (GTN) model to predict a ductile crack growth behavior of a pipe structure from those of laboratory specimens such as SE(T) and C(T) specimens. By using the parameter sets of the GTN model obtained by other members in the project, the authors predicted the fracture behaviors of three pipes with a circumferential through-wall crack or a circumferential outer surface crack by the four-point bending test conducted by a member of the project. As a result, the predicted maximum loads agreed with those of experiments within the error of 2.2 %, and the predicted ductile crack growth amounts were nearly equal to those on the fractured surface. In addition, the J-resistance of the pipe with a though-wall crack was calculated by the FEA by the node release technique with a crack growth criterion according to the crack growth predicted by the GTN model. Consequently, the J-resistance of the pipe is 3.1 to 3.6 times larger than that of C(T) specimens, which means that a prediction by the conventional fracture mechanics has a lot of margin to the actual fracture behavior.","PeriodicalId":434925,"journal":{"name":"Volume 4A: Materials and Fabrication","volume":"43 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128176620","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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