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 Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.
{"title":"Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Using Mini-C(T) Specimens","authors":"M. Ickes, J. Hall, R. Carter","doi":"10.1115/PVP2018-84967","DOIUrl":"https://doi.org/10.1115/PVP2018-84967","url":null,"abstract":"The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals .\u0000 In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor.\u0000 The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"130039849","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Hajime Fukumoto, Y. Wada, H. Matsunaga, Takeru Sano, Hiroshi Kobayashi
As is well known, low alloy steels are widely used as materials for high pressure vessels because of their high tensile strength and reasonable price, but also show severe hydrogen embrittlement. Therefore, in 2016, the authors introduced a scenario for the safe use of low alloy steels in highly pressurized hydrogen gas as a “Guideline” at ASME PVP 2016 [1]. Following discussions with stakeholders and experts in recent years, we published Technical Document (TD) as an industrial standard prior to regulation, on the safe use of ground storage vessels made of low alloy steels in Hydrogen Refueling Stations (HRSs) based on performance requirements. This article presents an outline of the TD describing the required types of testing as performance requirements for confirming the good hydrogen compatibility of low alloy steels, such as controlling tensile strength in an appropriate range, confirming leak-before-break, determining the life of ground storage vessels by fatigue testing and determining the inspection term by fatigue crack growth analysis using the fatigue crack growth rate in highly pressurized hydrogen.
{"title":"Introduction of Technical Document in Japan for Safe Use of Ground Storage Vessels Made of Low Alloy Steels for Hydrogen Refueling Stations","authors":"Hajime Fukumoto, Y. Wada, H. Matsunaga, Takeru Sano, Hiroshi Kobayashi","doi":"10.1115/PVP2018-84099","DOIUrl":"https://doi.org/10.1115/PVP2018-84099","url":null,"abstract":"As is well known, low alloy steels are widely used as materials for high pressure vessels because of their high tensile strength and reasonable price, but also show severe hydrogen embrittlement. Therefore, in 2016, the authors introduced a scenario for the safe use of low alloy steels in highly pressurized hydrogen gas as a “Guideline” at ASME PVP 2016 [1]. Following discussions with stakeholders and experts in recent years, we published Technical Document (TD) as an industrial standard prior to regulation, on the safe use of ground storage vessels made of low alloy steels in Hydrogen Refueling Stations (HRSs) based on performance requirements. This article presents an outline of the TD describing the required types of testing as performance requirements for confirming the good hydrogen compatibility of low alloy steels, such as controlling tensile strength in an appropriate range, confirming leak-before-break, determining the life of ground storage vessels by fatigue testing and determining the inspection term by fatigue crack growth analysis using the fatigue crack growth rate in highly pressurized hydrogen.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"10 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128874896","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The ASME Boiler and Pressure Vessel Code (Section III and Section VIII) provides requirements to avoid a ratcheting (accumulating permanent strain) condition under cyclic thermal load application. The ratchet check in this code is based on the solutions presented by Miller in 1959. One important focus in Miller’s work was to estimate the accumulated plastic strain under cyclic loading. The existing pressure vessels and piping codes have been adopting Miller’s ratchet boundary solution where there is no cyclic plastic accumulation of strain. However, some of these codes also provide limit on accumulated plastic strain under ratcheting conditions. Since the cyclic loading also causes fatigue damage in thee component, the question how to account for the interaction of ratchet deformation, which may contribute to damage in the material, and fatigue damage arises, since the fatigue curves are obtained from tests in the absence of ratcheting. This paper investigates the solutions to calculate growth strain (incremental plastic strain) and their application in design including taking into account the interaction with fatigue. Finite element analysis is presented to validate the analytical solutions.
{"title":"Beyond Shakedown-Ratcheting Boundary","authors":"R. Adibi-Asl, W. Reinhardt","doi":"10.1115/PVP2018-85050","DOIUrl":"https://doi.org/10.1115/PVP2018-85050","url":null,"abstract":"The ASME Boiler and Pressure Vessel Code (Section III and Section VIII) provides requirements to avoid a ratcheting (accumulating permanent strain) condition under cyclic thermal load application. The ratchet check in this code is based on the solutions presented by Miller in 1959. One important focus in Miller’s work was to estimate the accumulated plastic strain under cyclic loading. The existing pressure vessels and piping codes have been adopting Miller’s ratchet boundary solution where there is no cyclic plastic accumulation of strain. However, some of these codes also provide limit on accumulated plastic strain under ratcheting conditions. Since the cyclic loading also causes fatigue damage in thee component, the question how to account for the interaction of ratchet deformation, which may contribute to damage in the material, and fatigue damage arises, since the fatigue curves are obtained from tests in the absence of ratcheting. This paper investigates the solutions to calculate growth strain (incremental plastic strain) and their application in design including taking into account the interaction with fatigue. Finite element analysis is presented to validate the analytical solutions.","PeriodicalId":128383,"journal":{"name":"Volume 1A: Codes and Standards","volume":"52 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127154212","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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}
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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