Bending stresses at incipient plastic collapse for pipes with circumferential surface flaws are predicted by net-section stress approach. Appendix C-5320 of ASME B&PV Code Section XI provides a formula of bending stress at the plastic collapse, where the formula is applicable for both inner and outer surface flaws. That is, the collapse stresses for pipes with inner and outer surface flaws are the same, because of the pipe mean radius at the flawed section being entirely the same. Authors considered the separated pipe mean radii at the flawed ligament and at the un-flawed ligament. Based on the balances of axial force and bending moment, formulas of plastic collapse stresses for each inner and outer flawed pipe were obtained. It is found that, when the flaw angle and depth are the same, the collapse stress for inner flawed pipe is slightly higher than that calculated by Appendix C-5320 formula, and the collapse stress for outer flawed pipe is slightly lower than that by Appendix C-5320 formula, as can be expected. The collapse stresses derived from the three formulas are almost the same in most instances. For less common case where the flaw angle and depth are very large for thick wall pipes, the differences amongst the three collapse stresses become large.
{"title":"Prediction for Plastic Collapse Stresses for Pipes With Inner and Outer Circumferential Flaws","authors":"K. Hasegawa, Yinsheng Li, V. Mareš, V. Lacroix","doi":"10.1115/PVP2018-84951","DOIUrl":"https://doi.org/10.1115/PVP2018-84951","url":null,"abstract":"Bending stresses at incipient plastic collapse for pipes with circumferential surface flaws are predicted by net-section stress approach. Appendix C-5320 of ASME B&PV Code Section XI provides a formula of bending stress at the plastic collapse, where the formula is applicable for both inner and outer surface flaws. That is, the collapse stresses for pipes with inner and outer surface flaws are the same, because of the pipe mean radius at the flawed section being entirely the same.\u0000 Authors considered the separated pipe mean radii at the flawed ligament and at the un-flawed ligament. Based on the balances of axial force and bending moment, formulas of plastic collapse stresses for each inner and outer flawed pipe were obtained. It is found that, when the flaw angle and depth are the same, the collapse stress for inner flawed pipe is slightly higher than that calculated by Appendix C-5320 formula, and the collapse stress for outer flawed pipe is slightly lower than that by Appendix C-5320 formula, as can be expected. The collapse stresses derived from the three formulas are almost the same in most instances. For less common case where the flaw angle and depth are very large for thick wall pipes, the differences amongst the three collapse stresses become large.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"32 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":"129144774","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}
For several decades the nuclear industry has used structural weld overlays (SWOL) to repair and mitigate cracking within pressurized water reactor (PWR) components such as nozzles, pipes and elbows. There are two known primary mechanisms that have led to cracking within PWR components. One source of cracking has been primary water stress corrosion cracking (PWSCC). Numerous SWOL repairs and mitigations were installed in the early 2000s to address PWSCC in components such as pressurizer nozzles. However, nearly all of the likely candidate components for SWOL repairs have now been addressed in the industry. The other cause for cracking has been by fatigue, which usually results from thermal cycling events such as leakage caused by a faulty valve close to the component. The PWR components of most concern for fatigue cracking are mainly stainless steel. Thus, ASME Section XI Code Case N-504-4 would be a likely basis for SWOL repairs of these components, although this Code Case was originally drafted to address stress corrosion cracking (SCC) in boiling water reactors (BWR). N-504-4 includes the requirements for the SWOL design and subsequent analyses to establish the design life for the overlay based on predicted crack growth after the repair. This paper presents analysis work performed using Code Case N-504-4 to establish the design life of a SWOL repair applied to a boron injection tank (BIT) line nozzle attached to the cold leg of an operating PWR. The overlay was applied to the nozzle to address flaws found within the stainless steel base metal during inservice examination. Analyses were performed to calculate the residual stresses resulting from the original fabrication and the subsequent SWOL repair. In addition, post-SWOL operating stresses were calculated to demonstrate that the overlay does not invalidate the ASME Section III design basis for the nozzle and attached pipe. The operating and residual stresses were also used for input to a fatigue crack growth (FCG) analysis in order to establish the design life of the overlay. Lastly, the weld shrinkage from the application of overlay was evaluated for potential impact on the attached piping, restraints and valves within the BIT line. The combined analyses of the installed SWOL provide a basis for continued operation for the remaining life of the plant.
几十年来，核工业一直使用结构焊接覆盖层(SWOL)来修复和减轻压水堆(PWR)部件(如喷嘴、管道和弯头)的裂缝。有两种已知的主要机制导致了压水堆组件的开裂。裂缝的来源之一是原发性水应力腐蚀裂缝(PWSCC)。21世纪初，为了解决稳压器喷嘴等部件的PWSCC问题，安装了大量的SWOL维修和缓解措施。然而，几乎所有可能用于SWOL维修的候选部件现在都已经在行业中得到了解决。开裂的另一个原因是疲劳，这通常是由热循环事件引起的，例如由靠近组件的故障阀门引起的泄漏。压水堆中最容易发生疲劳开裂的部件主要是不锈钢。N-504-4包括对SWOL设计的要求，以及根据修复后预测的裂纹扩展来确定覆盖层设计寿命的后续分析。本文介绍了使用Code Case N-504-4进行的分析工作，以确定SWOL修复的设计寿命，该修复应用于连接在运行中的压水堆冷腿上的硼注入罐(BIT)管线喷嘴。覆盖层应用于喷嘴，以解决在使用检查中发现的不锈钢母材内部的缺陷。分析计算了由原始制造和随后的SWOL修复产生的残余应力。此外，对swol后的工作应力进行了计算，以证明覆盖不会使喷嘴和附加管道的ASME Section III设计依据失效。工作应力和残余应力也被用于疲劳裂纹扩展(FCG)分析，以确定覆盖层的设计寿命。最后，评估了覆盖层对管道、约束装置和阀门的潜在影响。对安装的SWOL进行综合分析，为工厂剩余寿命的持续运行提供了基础。
{"title":"Analysis of a Weld Overlay to Address Fatigue Cracking in a Stainless Steel Nozzle","authors":"S. Marlette, A. Udyawar, J. Broussard","doi":"10.1115/PVP2018-84722","DOIUrl":"https://doi.org/10.1115/PVP2018-84722","url":null,"abstract":"For several decades the nuclear industry has used structural weld overlays (SWOL) to repair and mitigate cracking within pressurized water reactor (PWR) components such as nozzles, pipes and elbows. There are two known primary mechanisms that have led to cracking within PWR components. One source of cracking has been primary water stress corrosion cracking (PWSCC). Numerous SWOL repairs and mitigations were installed in the early 2000s to address PWSCC in components such as pressurizer nozzles. However, nearly all of the likely candidate components for SWOL repairs have now been addressed in the industry. The other cause for cracking has been by fatigue, which usually results from thermal cycling events such as leakage caused by a faulty valve close to the component. The PWR components of most concern for fatigue cracking are mainly stainless steel. Thus, ASME Section XI Code Case N-504-4 would be a likely basis for SWOL repairs of these components, although this Code Case was originally drafted to address stress corrosion cracking (SCC) in boiling water reactors (BWR). N-504-4 includes the requirements for the SWOL design and subsequent analyses to establish the design life for the overlay based on predicted crack growth after the repair.\u0000 This paper presents analysis work performed using Code Case N-504-4 to establish the design life of a SWOL repair applied to a boron injection tank (BIT) line nozzle attached to the cold leg of an operating PWR. The overlay was applied to the nozzle to address flaws found within the stainless steel base metal during inservice examination. Analyses were performed to calculate the residual stresses resulting from the original fabrication and the subsequent SWOL repair. In addition, post-SWOL operating stresses were calculated to demonstrate that the overlay does not invalidate the ASME Section III design basis for the nozzle and attached pipe. The operating and residual stresses were also used for input to a fatigue crack growth (FCG) analysis in order to establish the design life of the overlay. Lastly, the weld shrinkage from the application of overlay was evaluated for potential impact on the attached piping, restraints and valves within the BIT line. The combined analyses of the installed SWOL provide a basis for continued operation for the remaining life of the plant.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"26 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":"132662463","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
This work analyzes the applicability of the ASME Boiler and Pressure Vessel Code procedures to calculate fatigue crack initiation under multiaxial stresses and/or strains, in particular when caused by non-proportional loads that lead the principal directions at the critical point to vary with time, e.g. due to out-of-phase bending and torsion loads induced by independent sources. Classic uniaxial fatigue damage models are usually inappropriate for analyzing multiaxial loads, since they can generate highly inaccurate predictions. Moreover, it is shown that the ASME procedures can lead to non-conservative results for non-proportional load histories.
{"title":"Issues With Multiaxial Fatigue Assessment in the ASME Boiler and Pressure Vessel Code","authors":"M. Meggiolaro, J. Castro, Hao Wu","doi":"10.1115/PVP2018-84979","DOIUrl":"https://doi.org/10.1115/PVP2018-84979","url":null,"abstract":"This work analyzes the applicability of the ASME Boiler and Pressure Vessel Code procedures to calculate fatigue crack initiation under multiaxial stresses and/or strains, in particular when caused by non-proportional loads that lead the principal directions at the critical point to vary with time, e.g. due to out-of-phase bending and torsion loads induced by independent sources. Classic uniaxial fatigue damage models are usually inappropriate for analyzing multiaxial loads, since they can generate highly inaccurate predictions. Moreover, it is shown that the ASME procedures can lead to non-conservative results for non-proportional load histories.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"6 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":"114888176","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}
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}
Wataru Nishi, T. Hirota, M. Ochi, Daiki Takagoshi, Kiminobu Hojo
In order to determine the target flaw depths for performance demonstration (PD) of nondestructive testing of cast austenitic stainless steel (CASS) pipes, probabilistic fracture mechanics (PFM) code, “PREFACE”, was developed to evaluate ductile fracture and plastic collapse considering change in mechanical properties due to thermal aging and uncertainties of the mechanical properties. In the previous study[1], the tables of the target flaw depths for the highest ferrite content material of the Japanese CASS pipes at the fully saturated thermal aging condition were calculated for several flaw lengths and several stress levels by the PREFACE code. However, the sensitivity study of the ferrite content on the target flaw depth revealed that the fully saturated thermal aging condition may not be the most severe condition for the target flaw depth. In this study, the PREFACE code was modified to improve the dependency of ferrite content on true stress-strain curve and to apply limit load method directly at the failure mode of plastic collapse. To confirm of the correct relation between ferrite content and aging time, deterministic sensitivity analyses were performed. For validation of the PFM function, the results of the deterministic analysis and PFM analysis were compared.
{"title":"Improvement of Target Flaw Sizes of CASS Pipe for PD Approval Using PFM Code Preface","authors":"Wataru Nishi, T. Hirota, M. Ochi, Daiki Takagoshi, Kiminobu Hojo","doi":"10.1115/PVP2018-85015","DOIUrl":"https://doi.org/10.1115/PVP2018-85015","url":null,"abstract":"In order to determine the target flaw depths for performance demonstration (PD) of nondestructive testing of cast austenitic stainless steel (CASS) pipes, probabilistic fracture mechanics (PFM) code, “PREFACE”, was developed to evaluate ductile fracture and plastic collapse considering change in mechanical properties due to thermal aging and uncertainties of the mechanical properties. In the previous study[1], the tables of the target flaw depths for the highest ferrite content material of the Japanese CASS pipes at the fully saturated thermal aging condition were calculated for several flaw lengths and several stress levels by the PREFACE code. However, the sensitivity study of the ferrite content on the target flaw depth revealed that the fully saturated thermal aging condition may not be the most severe condition for the target flaw depth. In this study, the PREFACE code was modified to improve the dependency of ferrite content on true stress-strain curve and to apply limit load method directly at the failure mode of plastic collapse. To confirm of the correct relation between ferrite content and aging time, deterministic sensitivity analyses were performed. For validation of the PFM function, the results of the deterministic analysis and PFM analysis were compared.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"10 2","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"113964432","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 the past 10 years, different laboratory test results lead the International Standard Development Organizations (SDO) to review their fatigue design rules in different directions, in particular to consider consequences of environmental effects on existing design rules. The key document that ask different questions to Code developers is the USNRC NUREG 6909 report: “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials” that confirms some environmental effects on S-N fatigue tests on small specimen. The open question is: how to transfer these results to Fatigue Design Rules of plant components? This paper will review existing codified rules in major nuclear Codes; in particular USA ASME Boiler and Pressure Vessel Code Section III and French AFCEN RCC-M Code. The paper will make a first historical background of these Codes and analyze background of these rules by consideration of field experience and existing tests results. To conclude, the paper will summarize to-day “fatigue road maps” to evaluate margins and screening criteria to assure reliable and safe codified design fatigue life evaluation.
{"title":"Nuclear Fatigue Codified Design Rules: Development Status, Margins and Screening Criteria","authors":"C. Faidy","doi":"10.1115/PVP2018-84698","DOIUrl":"https://doi.org/10.1115/PVP2018-84698","url":null,"abstract":"In the past 10 years, different laboratory test results lead the International Standard Development Organizations (SDO) to review their fatigue design rules in different directions, in particular to consider consequences of environmental effects on existing design rules.\u0000 The key document that ask different questions to Code developers is the USNRC NUREG 6909 report: “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials” that confirms some environmental effects on S-N fatigue tests on small specimen. The open question is: how to transfer these results to Fatigue Design Rules of plant components?\u0000 This paper will review existing codified rules in major nuclear Codes; in particular USA ASME Boiler and Pressure Vessel Code Section III and French AFCEN RCC-M Code.\u0000 The paper will make a first historical background of these Codes and analyze background of these rules by consideration of field experience and existing tests results.\u0000 To conclude, the paper will summarize to-day “fatigue road maps” to evaluate margins and screening criteria to assure reliable and safe codified design fatigue life evaluation.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"29 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":"122983554","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}
Mengyu Chai, Yan Song, Zao-xiao Zhang, Q. Duan, G. Cheng
2.25Cr-1Mo-0.25V steel, which is a high strength low alloy (HSLA) steel, has been widely used for structural material of hydrogenation reactor due to its excellent combination of mechanical properties and resistance to hydrogen embrittlement (HE). However, it still suffers serious hydrogen damage during the aggressive service environment. When sufficient hydrogen concentrates in the reactor steel, the ductility and strength of the steel will be greatly decreased. Such a phenomenon of reduction of toughness is known as HE, and it can significantly weaken the safety and reliability of equipment. Therefore, the aim of this investigation is to focus on the effect of hydrogen on fracture toughness behavior of 2.25Cr-1Mo-0.25V steel. The fracture mechanics specimens in geometry of single edge notch bending were used. The immersion charging method was used to pre-charge hydrogen inside the specimens. Moreover, the fracture toughness of specimens with and without hydrogen pre-charging were measured following the ASTM E1820 standard. Finally, the fracture morphology was observed by scanning electron microscopy (SEM) to identify the HE mechanisms. The results of the present investigation showed that the pre-charged hydrogen resulted in significant reduction of fracture toughness of 2.25Cr-1Mo-0.25V steel, indicating a reduced crack growth resistance of specimens in the presence of hydrogen. Furthermore, the uncharged specimens failed in a ductile manner, whereas the fracture of pre-charged specimens is a mixed ductile and brittle fracture mode. It was believed that the hydrogen-induced decohesion (HEDE) mechanism contributed to the HE in hydrogen pre-charged specimens.
{"title":"Effect of Hydrogen on Fracture Toughness Behavior of 2.25Cr-1Mo-0.25V Steel","authors":"Mengyu Chai, Yan Song, Zao-xiao Zhang, Q. Duan, G. Cheng","doi":"10.1115/PVP2018-84486","DOIUrl":"https://doi.org/10.1115/PVP2018-84486","url":null,"abstract":"2.25Cr-1Mo-0.25V steel, which is a high strength low alloy (HSLA) steel, has been widely used for structural material of hydrogenation reactor due to its excellent combination of mechanical properties and resistance to hydrogen embrittlement (HE). However, it still suffers serious hydrogen damage during the aggressive service environment. When sufficient hydrogen concentrates in the reactor steel, the ductility and strength of the steel will be greatly decreased. Such a phenomenon of reduction of toughness is known as HE, and it can significantly weaken the safety and reliability of equipment. Therefore, the aim of this investigation is to focus on the effect of hydrogen on fracture toughness behavior of 2.25Cr-1Mo-0.25V steel. The fracture mechanics specimens in geometry of single edge notch bending were used. The immersion charging method was used to pre-charge hydrogen inside the specimens. Moreover, the fracture toughness of specimens with and without hydrogen pre-charging were measured following the ASTM E1820 standard. Finally, the fracture morphology was observed by scanning electron microscopy (SEM) to identify the HE mechanisms. The results of the present investigation showed that the pre-charged hydrogen resulted in significant reduction of fracture toughness of 2.25Cr-1Mo-0.25V steel, indicating a reduced crack growth resistance of specimens in the presence of hydrogen. Furthermore, the uncharged specimens failed in a ductile manner, whereas the fracture of pre-charged specimens is a mixed ductile and brittle fracture mode. It was believed that the hydrogen-induced decohesion (HEDE) mechanism contributed to the HE in hydrogen pre-charged specimens.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"35 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":"123169789","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
C. Pétesch, T. Lebarbé, O. Gelineau, D. Vallot, M. Blat-Yrieix
The 2018 edition of the RCC-MRx Code (1) will be issued, by the end of the 2018, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up rules applicable to research reactor components (coming from the RCC-MX 2008 developed in the context of the Jules Horowitz Reactor project), to reactor components operating at high temperature (coming from the RCC-MR 2007) and to the Vacuum Vessel of ITER (also coming from the RCC-MR 2007). This edition takes the benefits of an important feedback of the users, such as Jules Horowitz Reactor or ASTRID project, but also from ITER and MYRRHA projects. In parallel, in compliance with the EC’s objectives and its own policy of openness, AFCEN proposes to make its codes evolve, taking into account the needs and expectations of European stakeholders (operators, designers, constructors, suppliers ...) through a workshop called CEN Workshop 64 phase 2. The end of the workshop, planned for 2018, will allow to integrate recommendations issued from this work in the code. This paper gives an overview of the performed work and also identifies the work to be done for a development of a standard such as RCC-MRx code.
{"title":"2018 RCC-MRx Code Edition: Context, Overview, On-Going Developments","authors":"C. Pétesch, T. Lebarbé, O. Gelineau, D. Vallot, M. Blat-Yrieix","doi":"10.1115/PVP2018-84706","DOIUrl":"https://doi.org/10.1115/PVP2018-84706","url":null,"abstract":"The 2018 edition of the RCC-MRx Code (1) will be issued, by the end of the 2018, in French and English versions by AFCEN (Association Française pour les règles de Conception et de Construction des Matériels des Chaudières Electro-nucléaires). This Code set up rules applicable to research reactor components (coming from the RCC-MX 2008 developed in the context of the Jules Horowitz Reactor project), to reactor components operating at high temperature (coming from the RCC-MR 2007) and to the Vacuum Vessel of ITER (also coming from the RCC-MR 2007).\u0000 This edition takes the benefits of an important feedback of the users, such as Jules Horowitz Reactor or ASTRID project, but also from ITER and MYRRHA projects.\u0000 In parallel, in compliance with the EC’s objectives and its own policy of openness, AFCEN proposes to make its codes evolve, taking into account the needs and expectations of European stakeholders (operators, designers, constructors, suppliers ...) through a workshop called CEN Workshop 64 phase 2. The end of the workshop, planned for 2018, will allow to integrate recommendations issued from this work in the code.\u0000 This paper gives an overview of the performed work and also identifies the work to be done for a development of a standard such as RCC-MRx code.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"74 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":"130798369","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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