A program underway at Oak Ridge National Laboratory (ORNL) to support private industry design of a system for production of 99Mo includes review of Zircaloy-4 (Zry-4) properties as a potential material for a primary vessel in the system. Because Zry-4 is not a material specified in the ASME Boiler and Pressure Vessel Code, a review has been performed to obtain the necessary data in the non-irradiated condition from the literature and from additional testing of Zry-4 to enable creation of tables of material properties required for ASME Code Section III Article ND-3000 design. A strategy for an in-vessel surveillance program is also recommended for early detection of potential corrosion and irradiation degradation issues during operational life of the vessel. Additionally, the ASME Code Section II, Part D was reviewed to ascertain the requirements for proposing a new material for Code approval. A plate of Zry-4 procured by ORNL was used to obtain tensile properties that were used as the basis for determining the maximum allowable stresses applicable to the vessel design. The materials properties of primary interest for ASME Code application are tensile properties, fracture toughness, fatigue, and five specific physical properties required by the Code for approval of a new material. The literature review revealed that all reported experimental data showed tensile strengths to substantially exceed the minimum specifications in ASTM B352. Sufficient fracture toughness and fatigue data are available to conclude that Zry-4 has sufficient fracture and fatigue resistance in the unirradiated condition for vessel design and operation. Of the physical properties reviewed, only thermal conductivity revealed a discrepancy that is recommended for further review. A surveillance program plan has been developed that is probably the maximum that would need to be implemented for the specific vessel design considered. Once adequate non-irradiated and irradiated data are available from preproduction testing of the Zry-4 and weld metal and HAZ materials, the program could be reduced to reflect that either initial properties are very good and/or the effects of embrittlement appear to be minimal. However, the overall design of a surveillance program is, of course, also dependent on the ASME Code design class for the pressure vessel.
{"title":"Material Properties of Non-Irradiated Zircaloy 4 in Support of ASME Code Acceptance for Pressure Vessel Design","authors":"R. Nanstad, W. Server, B. Kombaiah, J. Geringer","doi":"10.1115/pvp2019-93654","DOIUrl":"https://doi.org/10.1115/pvp2019-93654","url":null,"abstract":"\u0000 A program underway at Oak Ridge National Laboratory (ORNL) to support private industry design of a system for production of 99Mo includes review of Zircaloy-4 (Zry-4) properties as a potential material for a primary vessel in the system. Because Zry-4 is not a material specified in the ASME Boiler and Pressure Vessel Code, a review has been performed to obtain the necessary data in the non-irradiated condition from the literature and from additional testing of Zry-4 to enable creation of tables of material properties required for ASME Code Section III Article ND-3000 design. A strategy for an in-vessel surveillance program is also recommended for early detection of potential corrosion and irradiation degradation issues during operational life of the vessel. Additionally, the ASME Code Section II, Part D was reviewed to ascertain the requirements for proposing a new material for Code approval. A plate of Zry-4 procured by ORNL was used to obtain tensile properties that were used as the basis for determining the maximum allowable stresses applicable to the vessel design. The materials properties of primary interest for ASME Code application are tensile properties, fracture toughness, fatigue, and five specific physical properties required by the Code for approval of a new material. The literature review revealed that all reported experimental data showed tensile strengths to substantially exceed the minimum specifications in ASTM B352. Sufficient fracture toughness and fatigue data are available to conclude that Zry-4 has sufficient fracture and fatigue resistance in the unirradiated condition for vessel design and operation. Of the physical properties reviewed, only thermal conductivity revealed a discrepancy that is recommended for further review. A surveillance program plan has been developed that is probably the maximum that would need to be implemented for the specific vessel design considered. Once adequate non-irradiated and irradiated data are available from preproduction testing of the Zry-4 and weld metal and HAZ materials, the program could be reduced to reflect that either initial properties are very good and/or the effects of embrittlement appear to be minimal. However, the overall design of a surveillance program is, of course, also dependent on the ASME Code design class for the pressure vessel.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"10 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82845576","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}
R. Rebak, Shenyang Huang, M. Schuster, S. Buresh, E. Dolley
General Electric (GE) is working with the US Department of Energy (DOE) to develop advanced technology fuel (ATF) for light water reactors (LWR) that will have enhanced tolerance to failure under severe accident conditions. The development of materials for the current fuel is aimed at Generation III LWR but findings may be extended to future Generation IV reactors. One of the concepts pursued by GE is to use iron-chromium-aluminum (FeCrAl) or IronClad for the cladding due to its outstanding resistance to reaction with air and steam at temperatures higher than 1000°C. Ferritic FeCrAl alloys have been used for almost nine decades in the industry, but never in nuclear applications, therefore its fabrication and mechanical aspects for nuclear use needs to be evaluated. Results show that billets of FeCrAl can be produced via traditional melting and using powder metallurgy, and these billets can later be processed to high strength full length cladding tubes having less than half a millimeter wall thickness. The tubes can be joined to the caps via several welding processes.
{"title":"Fabrication and Mechanical Aspects of Using FeCrAl for Light Water Reactor Fuel Cladding","authors":"R. Rebak, Shenyang Huang, M. Schuster, S. Buresh, E. Dolley","doi":"10.1115/pvp2019-93128","DOIUrl":"https://doi.org/10.1115/pvp2019-93128","url":null,"abstract":"\u0000 General Electric (GE) is working with the US Department of Energy (DOE) to develop advanced technology fuel (ATF) for light water reactors (LWR) that will have enhanced tolerance to failure under severe accident conditions. The development of materials for the current fuel is aimed at Generation III LWR but findings may be extended to future Generation IV reactors. One of the concepts pursued by GE is to use iron-chromium-aluminum (FeCrAl) or IronClad for the cladding due to its outstanding resistance to reaction with air and steam at temperatures higher than 1000°C. Ferritic FeCrAl alloys have been used for almost nine decades in the industry, but never in nuclear applications, therefore its fabrication and mechanical aspects for nuclear use needs to be evaluated. Results show that billets of FeCrAl can be produced via traditional melting and using powder metallurgy, and these billets can later be processed to high strength full length cladding tubes having less than half a millimeter wall thickness. The tubes can be joined to the caps via several welding processes.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"19 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82912594","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}
Analysis of a generic dissimilar metal weld (DMW) susceptible to primary water stress corrosion cracking (PWSCC) in a pressurized water reactor (PWR) is used to compare the newly developed probabilistic models (xLPR code) to the previously performed deterministic leak before break (LBB) analyses. The objective of this scoping analysis is to develop a generic reactor loop composed of representative welds and to investigate the safety margins in the presence of PWSCC at the Alloy 82/182 locations. These locations have been previously studied and approved for LBB, however not in the presence of active degradation such as PWSCC. The purpose of this study is to investigate potential increase in risk due to this mechanism. Comparisons of the individual weld probabilistic results to the deterministic LBB analysis are made as the first results of this study. Additionally the individual welds are combined into a configuration representative of the primary loop. This configuration is then tested against the criterion recommended by the xLPR acceptance group. This xLPR criterion is then compared to the existing LBB criterion to assess the change, if any, in risk due to PWSCC.
{"title":"Probabilistic Fracture Mechanics Analyses Comparison to LBB Assessments","authors":"R. Kurth, C. Sallaberry, E. Kurth, F. Brust","doi":"10.1115/pvp2019-93413","DOIUrl":"https://doi.org/10.1115/pvp2019-93413","url":null,"abstract":"\u0000 Analysis of a generic dissimilar metal weld (DMW) susceptible to primary water stress corrosion cracking (PWSCC) in a pressurized water reactor (PWR) is used to compare the newly developed probabilistic models (xLPR code) to the previously performed deterministic leak before break (LBB) analyses. The objective of this scoping analysis is to develop a generic reactor loop composed of representative welds and to investigate the safety margins in the presence of PWSCC at the Alloy 82/182 locations. These locations have been previously studied and approved for LBB, however not in the presence of active degradation such as PWSCC. The purpose of this study is to investigate potential increase in risk due to this mechanism. Comparisons of the individual weld probabilistic results to the deterministic LBB analysis are made as the first results of this study. Additionally the individual welds are combined into a configuration representative of the primary loop. This configuration is then tested against the criterion recommended by the xLPR acceptance group. This xLPR criterion is then compared to the existing LBB criterion to assess the change, if any, in risk due to PWSCC.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"19 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79514940","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}
Additive manufacturing (AM) offers the potential for increased design flexibility in the low volume production of complex engineering components for hydrogen service. However, the suitability of AM materials for such extreme service environments remains to be evaluated. This work examines the effects of internal and external hydrogen on AM type 304L austenitic stainless steels fabricated via directed-energy deposition (DED) and powder bed fusion (PBF) processes. Under ambient test conditions, AM materials with minimal manufacturing defects exhibit excellent combinations of tensile strength, tensile ductility, and fatigue resistance. To probe the effects of extreme hydrogen environments on the AM materials, tensile and fatigue tests were performed after thermal-precharging in high pressure gaseous hydrogen (internal H) or in high pressure gaseous hydrogen (external H). Hydrogen appears to have a comparable influence on the AM 304L as in wrought materials, although the micromechanisms of tensile fracture and fatigue crack growth appear distinct. Specifically, microstructural characterization implicates the unique solidification microstructure of AM materials in the propagation of cracks under conditions of tensile fracture with hydrogen. These results highlight the need to establish comprehensive microstructure-property relationships for AM materials to ensure their suitability for use in extreme hydrogen environments.
{"title":"Effects of Extreme Hydrogen Environments on the Fracture and Fatigue Behavior of Additively Manufactured Stainless Steels","authors":"Thale R. Smith, C. S. Marchi, J. Sugar, D. Balch","doi":"10.1115/pvp2019-93903","DOIUrl":"https://doi.org/10.1115/pvp2019-93903","url":null,"abstract":"\u0000 Additive manufacturing (AM) offers the potential for increased design flexibility in the low volume production of complex engineering components for hydrogen service. However, the suitability of AM materials for such extreme service environments remains to be evaluated. This work examines the effects of internal and external hydrogen on AM type 304L austenitic stainless steels fabricated via directed-energy deposition (DED) and powder bed fusion (PBF) processes. Under ambient test conditions, AM materials with minimal manufacturing defects exhibit excellent combinations of tensile strength, tensile ductility, and fatigue resistance. To probe the effects of extreme hydrogen environments on the AM materials, tensile and fatigue tests were performed after thermal-precharging in high pressure gaseous hydrogen (internal H) or in high pressure gaseous hydrogen (external H). Hydrogen appears to have a comparable influence on the AM 304L as in wrought materials, although the micromechanisms of tensile fracture and fatigue crack growth appear distinct. Specifically, microstructural characterization implicates the unique solidification microstructure of AM materials in the propagation of cracks under conditions of tensile fracture with hydrogen. These results highlight the need to establish comprehensive microstructure-property relationships for AM materials to ensure their suitability for use in extreme hydrogen environments.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83128672","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. Lam, A. Duncan, L. Ward, R. Sindelar, Yun‐Jae Kim, Jae-Yoon Jeong, Hyun Jae Lee, M. Lee
Stress corrosion cracking may occur when chloride-bearing salts deposit and deliquesce on the external surface of stainless steel spent nuclear fuel storage canisters at weld regions with high residual stresses. Although it has not yet been observed, this phenomenon leads to a confinement concern for these canisters due to its potential for radioactive materials breaching through the containment system boundary provided by the canister wall during extended storage. The tests for crack growth rate have been conducted on bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The test and characterization protocols are performed to provide bounding conditions in which cracking will occur. The results after 2- and 6-month exposure are examined in relation to previous studies in condensed brine and compared with other experimental data in the open literature. The knowledge gained from bolt-load compact tension testing is being applied to a large plate cut from a mockup commercial spent nuclear fuel canister to demonstrate the crack growth behavior induced from starter cracks machined in regions where the welding residual stress is expected. All these tests are conducted to support the technical basis for ASME Boiler and Pressure Vessel Section XI Code Case N-860.
{"title":"Crack Growth Rate Testing and Large Plate Demonstration Under Chloride-Induced Stress Corrosion Cracking Conditions in Stainless Steel Canisters for Storage of Spent Nuclear Fuel","authors":"P. Lam, A. Duncan, L. Ward, R. Sindelar, Yun‐Jae Kim, Jae-Yoon Jeong, Hyun Jae Lee, M. Lee","doi":"10.1115/pvp2019-94031","DOIUrl":"https://doi.org/10.1115/pvp2019-94031","url":null,"abstract":"\u0000 Stress corrosion cracking may occur when chloride-bearing salts deposit and deliquesce on the external surface of stainless steel spent nuclear fuel storage canisters at weld regions with high residual stresses. Although it has not yet been observed, this phenomenon leads to a confinement concern for these canisters due to its potential for radioactive materials breaching through the containment system boundary provided by the canister wall during extended storage. The tests for crack growth rate have been conducted on bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The test and characterization protocols are performed to provide bounding conditions in which cracking will occur. The results after 2- and 6-month exposure are examined in relation to previous studies in condensed brine and compared with other experimental data in the open literature. The knowledge gained from bolt-load compact tension testing is being applied to a large plate cut from a mockup commercial spent nuclear fuel canister to demonstrate the crack growth behavior induced from starter cracks machined in regions where the welding residual stress is expected. All these tests are conducted to support the technical basis for ASME Boiler and Pressure Vessel Section XI Code Case N-860.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"7 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74709174","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}
Jia Xiao, Zhijun Li, Li Jiang, L. Ye, Kun Yu, Jianjun Liang, Shuangjian Chen, Zezhong Chen
Two Alloy N/316H bimetallic plates have been fabricated by explosive welding and rolling technologies respectively. Metallographic observations indicate that the rolled bimetallic plate has a straight bond interface, in which some cavities and precipitates exist. While the explosive welded plate shows a wavy bond interfaces. The interface thermal expansion mismatch between the two alloys were evaluated in the two plates at high temperature. Results show that the thermal expansion coefficient of 316H is larger than that of Alloy N. The thermal expansion coefficient of the substrate plates depends on the thickness ratio between Alloy N and 316H, which reaches the maximum when the ratio is 1:4.
{"title":"Interface Microstructure and Thermal Expansion Mismatch in Alloy N/316H Bimetallic Plates","authors":"Jia Xiao, Zhijun Li, Li Jiang, L. Ye, Kun Yu, Jianjun Liang, Shuangjian Chen, Zezhong Chen","doi":"10.1115/pvp2019-93585","DOIUrl":"https://doi.org/10.1115/pvp2019-93585","url":null,"abstract":"\u0000 Two Alloy N/316H bimetallic plates have been fabricated by explosive welding and rolling technologies respectively. Metallographic observations indicate that the rolled bimetallic plate has a straight bond interface, in which some cavities and precipitates exist. While the explosive welded plate shows a wavy bond interfaces. The interface thermal expansion mismatch between the two alloys were evaluated in the two plates at high temperature. Results show that the thermal expansion coefficient of 316H is larger than that of Alloy N. The thermal expansion coefficient of the substrate plates depends on the thickness ratio between Alloy N and 316H, which reaches the maximum when the ratio is 1:4.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"104 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75737945","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}
M. Dalal, J. Penso, David J. Dewees, Robert G. Brown
Creep is progressive deformation of material over an extended period when exposed to elevated temperature and stresses below the yield strength. Poor Creep ductility and cracking can be a problem above 900 °F (482°C) in the HAZ of low alloy (Cr-Mo) steel. High stress areas, including supports, hangers and fittings are more vulnerable to cracking. Creep cracking has occurred in longitudinal pipe welds with excessive peaking or welds with poor quality. Numerous incidents of cracking in low alloy (Cr-Mo) steel have been reported in the power industry and in refineries with major concern in longitudinal seam welds as well as highly stressed welds in reactors-heaters interconnecting piping. This paper presents the results of an assessment performed on reactors-heaters interconnecting piping in a catalytic reformer unit with a maximum operating temperature of about 950 °F (510 °C) at 250 psig (1.7 MPa) (> 40 years in-service). Comprehensive inspection including visual, dye penetrant testing, thickness measurements and peaking measurements have been performed. Phased Array Ultrasonic Testing (PAUT) was utilized to detect crack-like defects and flaws. Detailed pipe stress analysis and finite element analyses (FEA) were also performed.
{"title":"Longitudinal Seam Welded Piping Assessment in Refinery Reformer Units","authors":"M. Dalal, J. Penso, David J. Dewees, Robert G. Brown","doi":"10.1115/pvp2019-93706","DOIUrl":"https://doi.org/10.1115/pvp2019-93706","url":null,"abstract":"\u0000 Creep is progressive deformation of material over an extended period when exposed to elevated temperature and stresses below the yield strength. Poor Creep ductility and cracking can be a problem above 900 °F (482°C) in the HAZ of low alloy (Cr-Mo) steel. High stress areas, including supports, hangers and fittings are more vulnerable to cracking. Creep cracking has occurred in longitudinal pipe welds with excessive peaking or welds with poor quality. Numerous incidents of cracking in low alloy (Cr-Mo) steel have been reported in the power industry and in refineries with major concern in longitudinal seam welds as well as highly stressed welds in reactors-heaters interconnecting piping. This paper presents the results of an assessment performed on reactors-heaters interconnecting piping in a catalytic reformer unit with a maximum operating temperature of about 950 °F (510 °C) at 250 psig (1.7 MPa) (> 40 years in-service). Comprehensive inspection including visual, dye penetrant testing, thickness measurements and peaking measurements have been performed. Phased Array Ultrasonic Testing (PAUT) was utilized to detect crack-like defects and flaws. Detailed pipe stress analysis and finite element analyses (FEA) were also performed.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"74 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90434989","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}
B. Arroyo, P. González, L. Andrea, J. Álvarez, R. Lacalle
In this work, the incremental step loading technique to measure hydrogen embrittlement threshold in steels from ASTM F1624 standard is applied to the Small Punch test technique. For the experimental program, a medium strength steel is employed, simulating the hydrogen embrittlement environment by a cathodic polarization of 5 mA/cm2 in an acid electrolyte mainly consisting of 1N H2SO4 in H2O. Regular standard tests on cylindrical tensile specimens were carried out under the same environment following the ASTM F1624 standard, observing the same trends in both cases, which validates the methodology proposed. In order to adapt the aforementioned standard to small punch testing, the duration of the loading steps had to be modified, proposing much shorter ones according to the Small Punch specimen dimensions and the punch rate taking place in these scenarios, which is pointed in bibliography. This proposal allows to obtain a threshold load by using at least 3 specimens in a total time of around a week.
{"title":"Application of the Incremental Step Loading Technique to Small Punch Tests in Hydrogen Embrittlement","authors":"B. Arroyo, P. González, L. Andrea, J. Álvarez, R. Lacalle","doi":"10.1115/pvp2019-93550","DOIUrl":"https://doi.org/10.1115/pvp2019-93550","url":null,"abstract":"\u0000 In this work, the incremental step loading technique to measure hydrogen embrittlement threshold in steels from ASTM F1624 standard is applied to the Small Punch test technique. For the experimental program, a medium strength steel is employed, simulating the hydrogen embrittlement environment by a cathodic polarization of 5 mA/cm2 in an acid electrolyte mainly consisting of 1N H2SO4 in H2O. Regular standard tests on cylindrical tensile specimens were carried out under the same environment following the ASTM F1624 standard, observing the same trends in both cases, which validates the methodology proposed.\u0000 In order to adapt the aforementioned standard to small punch testing, the duration of the loading steps had to be modified, proposing much shorter ones according to the Small Punch specimen dimensions and the punch rate taking place in these scenarios, which is pointed in bibliography. This proposal allows to obtain a threshold load by using at least 3 specimens in a total time of around a week.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"29 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90938013","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. Hayashi, Shigeaki Tanaka, Abe Tomonori, Seiji Sakuraya, S. Ooki, Takayuki Kaminaga
Continuous improvement of the structural integrity evaluation methodology in the plant life management (PLM) evaluations is of increasing importance for aged light water reactors. In PLM evaluations, structural integrity evaluations are required for degradation mechanisms considered in the subject equipment and components. Austenitic stainless steels used in reactor internal components are known to show decreases in ductility and fracture toughness due to accumulated neutron irradiation damage. In Japan, “Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code)” provides fracture evaluation method and criterion, based on the linear elastic fracture mechanics, for irradiated stainless steels of boiling water reactor (BWR) internal components. The fracture toughness criterion, however, was developed with limited materials testing data and knowledge available at that time and it has not been revised since the code originally established. In this study, fracture toughness criteria for structural integrity evaluation were discussed and developed with the latest database on fracture toughness of irradiated austenitic stainless steels, including additional material testing data obtained in this study for the neutron fluence range of interest from 1 to 3 dpa. First, the fracture toughness data of austenitic stainless steels irradiated in BWR conditions were compiled to evaluate the correlation between fracture toughness and neutron fluence. Material characteristics potentially affecting fracture toughness, such as chemical composition and specimen orientation, were also considered and discussed in the development of the fracture toughness criteria. Based on the results, the fracture toughness criteria for irradiated austenitic stainless steels were proposed for fracture evaluation of the BWR internal components.
{"title":"Fracture Toughness Criteria of Irradiated Austenitic Stainless Steels for Structural Integrity Evaluation of BWR Internal Components","authors":"T. Hayashi, Shigeaki Tanaka, Abe Tomonori, Seiji Sakuraya, S. Ooki, Takayuki Kaminaga","doi":"10.1115/pvp2019-93441","DOIUrl":"https://doi.org/10.1115/pvp2019-93441","url":null,"abstract":"\u0000 Continuous improvement of the structural integrity evaluation methodology in the plant life management (PLM) evaluations is of increasing importance for aged light water reactors. In PLM evaluations, structural integrity evaluations are required for degradation mechanisms considered in the subject equipment and components. Austenitic stainless steels used in reactor internal components are known to show decreases in ductility and fracture toughness due to accumulated neutron irradiation damage. In Japan, “Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code)” provides fracture evaluation method and criterion, based on the linear elastic fracture mechanics, for irradiated stainless steels of boiling water reactor (BWR) internal components. The fracture toughness criterion, however, was developed with limited materials testing data and knowledge available at that time and it has not been revised since the code originally established.\u0000 In this study, fracture toughness criteria for structural integrity evaluation were discussed and developed with the latest database on fracture toughness of irradiated austenitic stainless steels, including additional material testing data obtained in this study for the neutron fluence range of interest from 1 to 3 dpa. First, the fracture toughness data of austenitic stainless steels irradiated in BWR conditions were compiled to evaluate the correlation between fracture toughness and neutron fluence. Material characteristics potentially affecting fracture toughness, such as chemical composition and specimen orientation, were also considered and discussed in the development of the fracture toughness criteria. Based on the results, the fracture toughness criteria for irradiated austenitic stainless steels were proposed for fracture evaluation of the BWR internal components.","PeriodicalId":23651,"journal":{"name":"Volume 6B: Materials and Fabrication","volume":"45 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82794884","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}
S. Matsuoka, T. Iijima, Satoko Yoshida, J. Yamabe, H. Matsunaga
Three types of strength tests, slow strain rate tensile (SSRT), fatigue life, and fatigue crack growth (FCG) tests, were performed using six types of aluminum alloys, 5083-O, 6061-T6, 6066-T6, 7N01-T5, 7N01-T6, and 7075-T6, in air and 115 MPa hydrogen gas at room temperature. All the strength properties of every material were not deteriorated in 115 MPa hydrogen gas. In all the materials, FCG rates were lower in 115 MPa hydrogen gas than in air. This was considered to be due to a lack of water- or oxygen-adsorbed film at crack tip in hydrogen gas. In 5083-O, 6061-T6 and 6066-T6, relative reduction in area (RRA) were remarkably higher in 115 MPa hydrogen gas than in air. These differences were attributed to a hydrostatic pressure produced in 115 MPa hydrogen gas. In contrast, in 7N01-T5, 7N01-T6 and 7075-T6, the values of RRA in 115 MPa hydrogen gas were nearly the same as those in air. Observation of fractured specimens inferred that the degree of such a hydrogen-induced improvement was determined by the fracture mode (e.g. cup-and-cone or shear fracture), which is dominated by the microstructure morphology.
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