The heat exchanger tube of pressurized water reactor (PWR) steam generator is the key component between primary circuit and secondary circuit. Due to the flow induced vibration (FIV) of the heat exchanger tube bundle, fretting corrosion will occur between the heat exchanger tubes and the supports. Under the synergism between wear and corrosion, it would be accelerated that the failure of heat exchanger tubes. Therefore, in this study, the self-designed fretting corrosion experimental equipment was used to conduct fretting corrosion experimental studies on 316L stainless steel, commonly used as steam generators tube in nuclear power plants, in sodium chloride (NaCl) solutions with different concentrations (mass fraction of 1%, 3.5% and 5%), respectively. The corrosion tendency and corrosion rate of the tube were analyzed by electrochemical technology. The fretting corrosion behaviors of 316L stainless steel in NaCl solution and the effects of ion concentration on fretting corrosion behaviors were studied, and the synergism between corrosion and wear was quantitatively analyzed. The surface of the wear scars were analyzed by Scanning Electron Microscope (SEM) and white light confocal three-dimensional profilometer. Combined with these test results, the damage mechanism of fretting corrosion was analyzed. The research results reflect the synergistic mechanism between wear and corrosion, but considering the difference between the experimental setting and the operating conditions of PWR steam generator, the experimental results can’t be directly used to predict the wear of heat exchanger tubes.
{"title":"Fretting Corrosion Behavior of 316L Stainless Steel Heat Exchanger Tube in NaCl Solution","authors":"Xu Ma, Shengzan Zhang, W. Tan, Guorui Zhu","doi":"10.1115/pvp2022-84442","DOIUrl":"https://doi.org/10.1115/pvp2022-84442","url":null,"abstract":"\u0000 The heat exchanger tube of pressurized water reactor (PWR) steam generator is the key component between primary circuit and secondary circuit. Due to the flow induced vibration (FIV) of the heat exchanger tube bundle, fretting corrosion will occur between the heat exchanger tubes and the supports. Under the synergism between wear and corrosion, it would be accelerated that the failure of heat exchanger tubes. Therefore, in this study, the self-designed fretting corrosion experimental equipment was used to conduct fretting corrosion experimental studies on 316L stainless steel, commonly used as steam generators tube in nuclear power plants, in sodium chloride (NaCl) solutions with different concentrations (mass fraction of 1%, 3.5% and 5%), respectively. The corrosion tendency and corrosion rate of the tube were analyzed by electrochemical technology. The fretting corrosion behaviors of 316L stainless steel in NaCl solution and the effects of ion concentration on fretting corrosion behaviors were studied, and the synergism between corrosion and wear was quantitatively analyzed. The surface of the wear scars were analyzed by Scanning Electron Microscope (SEM) and white light confocal three-dimensional profilometer. Combined with these test results, the damage mechanism of fretting corrosion was analyzed. The research results reflect the synergistic mechanism between wear and corrosion, but considering the difference between the experimental setting and the operating conditions of PWR steam generator, the experimental results can’t be directly used to predict the wear of heat exchanger tubes.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"292 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115315922","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}
Heikki Keinänen, Juha Kuutti, N. Hytönen, P. Nevasmaa, C. Huotilainen, I. Virkkunen, S. Bhusare, Aloshious Lambai, G. Mohanty
As nuclear power plants age and their lifetimes are being extended, the possibility and need to perform repairs of safety critical and hard to replace components is ever increasing. For example, defects in the reactor pressure vessel caused by exposure to high temperature, pressure, and corrosive environment together with neutron irradiation are often repaired by different repair welding techniques. Moreover, the need for such repairs may come at short notice requiring that qualified and optimized techniques and solutions are readily available. Developments of repair welding techniques using robotized gas metal arc welding cold metal transfer to repair a linear crack like defect beneath the cladding, which extended into the reactor pressure vessel steel have been presented in previous works [8–9]). In the latest piece of research [10], the repair welding of a thermally embrittled and cladded low-alloy steel plate with two groove excavations filled using Alloy 52 was presented. In the paper, the two welds were characterized with micrographs and microhardness measurements. This work further evaluates in more detail the differences and similarities of the repair welds welded using two different welding directions, 0-degree and 45-degree, and corresponding bead patterns. Residual stresses were measured from the two repair-weld cases using the contour method. Despite significant differences in the weld bead order and consequent welding procedure, the resulting residual stresses were very similar. It was expected that the crisscross weld bead pattern would cause the subsequent weld layers to induce stresses counteracting the previous layer and thus reduce the overall residual stress field. However, this does not appear to be the case. Both weld areas showed tensile stresses around 300 MPa, which is close to the yield stress of the weld material. Balancing compressive stress is induced to the base material with somewhat lower magnitude, peaking around 200 MPa. This indicates that the main determinant of the residual stress field is the weld material yield behavior. The microstructural characterization of the two weld orientations included microhardness and nanohardness measurements across the low-alloy steel and Alloy 52 weld fusion boundary, where the hardness peak was at the coarsegrained heat-affected zone adjacent to the fusion boundary. The 0-degree weld gives a higher microhardness peak than the 45-degree weld, indicating a slightly higher mismatch in properties, but the nanohardness measurements could not confirm this. Also, in the microstructural analysis, no great differences are seen other than few weld defects, especially voids. The elemental analysis using energy dispersive X-ray spectrometry across the fusion boundary shows expected minor dilution of alloy elements, e.g. chromium, which affects the materials corrosion properties. Electron backscatter diffraction mapping and nanoindentation measurements were performed across the weld i
{"title":"Effect of Welding Direction and Bead Pattern in Alloy 52 / SA508 Repair Weld","authors":"Heikki Keinänen, Juha Kuutti, N. Hytönen, P. Nevasmaa, C. Huotilainen, I. Virkkunen, S. Bhusare, Aloshious Lambai, G. Mohanty","doi":"10.1115/pvp2022-84662","DOIUrl":"https://doi.org/10.1115/pvp2022-84662","url":null,"abstract":"\u0000 As nuclear power plants age and their lifetimes are being extended, the possibility and need to perform repairs of safety critical and hard to replace components is ever increasing. For example, defects in the reactor pressure vessel caused by exposure to high temperature, pressure, and corrosive environment together with neutron irradiation are often repaired by different repair welding techniques. Moreover, the need for such repairs may come at short notice requiring that qualified and optimized techniques and solutions are readily available.\u0000 Developments of repair welding techniques using robotized gas metal arc welding cold metal transfer to repair a linear crack like defect beneath the cladding, which extended into the reactor pressure vessel steel have been presented in previous works [8–9]). In the latest piece of research [10], the repair welding of a thermally embrittled and cladded low-alloy steel plate with two groove excavations filled using Alloy 52 was presented. In the paper, the two welds were characterized with micrographs and microhardness measurements. This work further evaluates in more detail the differences and similarities of the repair welds welded using two different welding directions, 0-degree and 45-degree, and corresponding bead patterns.\u0000 Residual stresses were measured from the two repair-weld cases using the contour method. Despite significant differences in the weld bead order and consequent welding procedure, the resulting residual stresses were very similar. It was expected that the crisscross weld bead pattern would cause the subsequent weld layers to induce stresses counteracting the previous layer and thus reduce the overall residual stress field. However, this does not appear to be the case. Both weld areas showed tensile stresses around 300 MPa, which is close to the yield stress of the weld material. Balancing compressive stress is induced to the base material with somewhat lower magnitude, peaking around 200 MPa. This indicates that the main determinant of the residual stress field is the weld material yield behavior.\u0000 The microstructural characterization of the two weld orientations included microhardness and nanohardness measurements across the low-alloy steel and Alloy 52 weld fusion boundary, where the hardness peak was at the coarsegrained heat-affected zone adjacent to the fusion boundary. The 0-degree weld gives a higher microhardness peak than the 45-degree weld, indicating a slightly higher mismatch in properties, but the nanohardness measurements could not confirm this. Also, in the microstructural analysis, no great differences are seen other than few weld defects, especially voids. The elemental analysis using energy dispersive X-ray spectrometry across the fusion boundary shows expected minor dilution of alloy elements, e.g. chromium, which affects the materials corrosion properties.\u0000 Electron backscatter diffraction mapping and nanoindentation measurements were performed across the weld i","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"447 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115280451","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 ferritic steels used in fabricating cost-effective cylinders for hydrogen storage are susceptible to hydrogen embrittlement. A model first proposed by Amaro et al. (ARFiDS model) for predicting the effects of hydrogen pressure on the kinetics of hydrogen assisted fatigue crack growth (HA-FCG) in X-100 pipeline steels for pressure between 1 and 20 MPa is adapted in this study and used for predicting the HA-FCG behavior of SA372 Grade J Class 70 pressure vessel steels for pressures ranging between 10 and 102 MPa. The crack growth kinetics in these steels exhibit a typical two-region behavior labelled as the transient and steady-state regions, characterized by distinct power-law exponents in the relationship between fatigue crack growth rate, da/dN, and the cyclic stress intensity parameter, ΔK. The predicted HA-FCG behavior from the model is compared with experimental data for SA 372 steel for hydrogen pressures ranging from 10 to 102 MPa and is shown to perform well for load ratios, R, of 0.2 and 0.5 over a wide range of crack growth rates. The phenomenological basis for the model is discussed.
{"title":"Modelling the Effects of Hydrogen Pressure on Fatigue Crack Growth Behavior in SA372 Pressure Vessel Steels","authors":"A. Saxena, K. Findley","doi":"10.1115/pvp2022-83958","DOIUrl":"https://doi.org/10.1115/pvp2022-83958","url":null,"abstract":"\u0000 Low alloy ferritic steels used in fabricating cost-effective cylinders for hydrogen storage are susceptible to hydrogen embrittlement. A model first proposed by Amaro et al. (ARFiDS model) for predicting the effects of hydrogen pressure on the kinetics of hydrogen assisted fatigue crack growth (HA-FCG) in X-100 pipeline steels for pressure between 1 and 20 MPa is adapted in this study and used for predicting the HA-FCG behavior of SA372 Grade J Class 70 pressure vessel steels for pressures ranging between 10 and 102 MPa. The crack growth kinetics in these steels exhibit a typical two-region behavior labelled as the transient and steady-state regions, characterized by distinct power-law exponents in the relationship between fatigue crack growth rate, da/dN, and the cyclic stress intensity parameter, ΔK. The predicted HA-FCG behavior from the model is compared with experimental data for SA 372 steel for hydrogen pressures ranging from 10 to 102 MPa and is shown to perform well for load ratios, R, of 0.2 and 0.5 over a wide range of crack growth rates. The phenomenological basis for the model is discussed.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"75 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126178197","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. Rensman, Davide Frittitta, F. Fusari, Nicola Ronchi
Minimum Pressurization Temperature (MPT) is the lowest temperature at which a hydroprocessing reactor can start pressurizing safely. A reliable MPT evaluation is necessary due to the susceptibility of hydroprocessing reactor materials to the combined effects of temper embrittlement and hydrogen embrittlement. Accurate estimation of the MPT is important for owners, who are looking for the lowest MPT values to reduce start-up time. Several methods to develop an MPT have been adopted in the past mainly based upon experimental data and/or acquired from material exposed in reactors over time. Most of these methods are based on historical concepts of hydrogen embrittlement in conventional 2¼Cr-1Mo alloys. For 2¼Cr-1Mo-¼V low alloy steels, hydrogen diffusion and trapping capacity are different and thus, potential for internal cracking is different. In addition, experiments to assess the influence of external hydrogen environment on stable cracking have been reported and provide more insight into stable crack growth potential due to a hydrogen atmosphere. The traditional methods could lead to an overly conservative approach. The result could lead to an over-estimation (i.e., a shift to higher temperature envelope) of the MPT curve. In this paper the authors describe a methodology for deriving the MPT for new built 2¼Cr-1Mo-¼V low alloy steel reactors. It is based on API TR-934 F parts 3 and 4 [1],[2] combined with some practices from WRC Bulletin 562 [3]. A case study will be described and a comparison with results according to historical calculations will be presented.
{"title":"A Methodology for Calculating the Minimum Pressurization Temperature of New Built Hydroprocessing Reactors in 2¼Cr-1Mo-¼V Low Alloy Steel","authors":"J. Rensman, Davide Frittitta, F. Fusari, Nicola Ronchi","doi":"10.1115/pvp2022-84640","DOIUrl":"https://doi.org/10.1115/pvp2022-84640","url":null,"abstract":"\u0000 Minimum Pressurization Temperature (MPT) is the lowest temperature at which a hydroprocessing reactor can start pressurizing safely. A reliable MPT evaluation is necessary due to the susceptibility of hydroprocessing reactor materials to the combined effects of temper embrittlement and hydrogen embrittlement. Accurate estimation of the MPT is important for owners, who are looking for the lowest MPT values to reduce start-up time.\u0000 Several methods to develop an MPT have been adopted in the past mainly based upon experimental data and/or acquired from material exposed in reactors over time. Most of these methods are based on historical concepts of hydrogen embrittlement in conventional 2¼Cr-1Mo alloys. For 2¼Cr-1Mo-¼V low alloy steels, hydrogen diffusion and trapping capacity are different and thus, potential for internal cracking is different. In addition, experiments to assess the influence of external hydrogen environment on stable cracking have been reported and provide more insight into stable crack growth potential due to a hydrogen atmosphere.\u0000 The traditional methods could lead to an overly conservative approach. The result could lead to an over-estimation (i.e., a shift to higher temperature envelope) of the MPT curve.\u0000 In this paper the authors describe a methodology for deriving the MPT for new built 2¼Cr-1Mo-¼V low alloy steel reactors. It is based on API TR-934 F parts 3 and 4 [1],[2] combined with some practices from WRC Bulletin 562 [3]. A case study will be described and a comparison with results according to historical calculations will be presented.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"134355391","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}
Jae-Yoon Jeong, Yun‐Jae Kim, P. Lam, Seunghyun Kim, G. Kim
Spent nuclear fuels (SNFs) are stored in stainless steel canisters at Independent Spent Fuel Storage Installations (ISFSIs) typically near the seashore. During long-term storage of these canisters in the dry cask storage system (DCSS), chloride-induced stress corrosion cracking (CISCC) could occur due to the deliquescence of concentrated salt deposits on the canister surface. To evaluate such flaws on the accessible exterior metallic portions of containment systems while in service, the ASME Section XI Code Case N-860 provides inservice inspection requirements for aging management of canisters manufactured with welded austenitic stainless steels. It is noteworthy that CISCC crack growth rate (CGR) model in Code Case N-860 consists of the constitutive equations with temperatures (canister surface temperatures, storage site yearly mean temperature, and ambient temperature measured at overpack inlet) and is independent of stress intensity factor or other environmental factors. In this work, the mean temperature effect of the local storage site on the CGR is analyzed based on the CISCC CGR model in Code Case N-860. The specified mean temperature in the Code Case is calculated yearly however, the crack growth by CISCC can be evaluated differently if the mean temperature of storage site with a large annual range of temperature such as South Korea is applied. In that case, the monthly mean temperature is adjusted as the yearly mean temperature so the effect of averaging range for calculating the mean temperature is analyzed. Firstly, climate data of some candidate sites for the storage in South Korea are measured from Korea Meteorological Administration (KMA). The climate data of the Diablo Canyon Power Plant located in California (United States) is obtained from National Weather Service (NWS) for comparison. Yearly data from 2012 to 2020 are applied and the crack growth is estimated for sites of different annual ranges.
乏燃料(snf)储存在独立乏燃料储存装置(ISFSIs)的不锈钢罐中，通常靠近海岸。在干桶储存系统(dcs)的长期储存过程中，由于罐表面的浓盐沉积潮解，可能发生氯化物诱发的应力腐蚀开裂(CISCC)。值得注意的是，Code Case N-860中的CISCC裂纹扩展速率(CGR)模型由含温度的本构方程(罐面温度、储存地年平均温度和包口实测环境温度)组成，不受应力强度因素和其他环境因素的影响。本文基于Code Case N-860中的CISCC CGR模型，分析了局部储存点的平均温度对CGR的影响。规范案例中规定的平均温度是按年计算的，但如果采用韩国等年温差较大的储存地的平均温度，则CISCC对裂缝扩展的估计会有所不同。在这种情况下，将月平均温度调整为年平均温度，从而分析平均范围对计算平均温度的影响。首先，利用韩国气象厅(KMA)对韩国部分候选地的气候数据进行了测量。位于美国加利福尼亚州的迪亚波罗峡谷电厂的气候数据是从美国国家气象局(NWS)获得的，用于比较。采用2012 ~ 2020年的年际数据，对不同年际范围的站点进行了裂缝扩展估算。
{"title":"Effect Of The Mean Temperature of Storage Site on Chloride-Induced Stress Corrosion Cracking Rate in ASME Code Case N-860: Case Study","authors":"Jae-Yoon Jeong, Yun‐Jae Kim, P. Lam, Seunghyun Kim, G. Kim","doi":"10.1115/pvp2022-83766","DOIUrl":"https://doi.org/10.1115/pvp2022-83766","url":null,"abstract":"\u0000 Spent nuclear fuels (SNFs) are stored in stainless steel canisters at Independent Spent Fuel Storage Installations (ISFSIs) typically near the seashore. During long-term storage of these canisters in the dry cask storage system (DCSS), chloride-induced stress corrosion cracking (CISCC) could occur due to the deliquescence of concentrated salt deposits on the canister surface. To evaluate such flaws on the accessible exterior metallic portions of containment systems while in service, the ASME Section XI Code Case N-860 provides inservice inspection requirements for aging management of canisters manufactured with welded austenitic stainless steels. It is noteworthy that CISCC crack growth rate (CGR) model in Code Case N-860 consists of the constitutive equations with temperatures (canister surface temperatures, storage site yearly mean temperature, and ambient temperature measured at overpack inlet) and is independent of stress intensity factor or other environmental factors.\u0000 In this work, the mean temperature effect of the local storage site on the CGR is analyzed based on the CISCC CGR model in Code Case N-860. The specified mean temperature in the Code Case is calculated yearly however, the crack growth by CISCC can be evaluated differently if the mean temperature of storage site with a large annual range of temperature such as South Korea is applied. In that case, the monthly mean temperature is adjusted as the yearly mean temperature so the effect of averaging range for calculating the mean temperature is analyzed. Firstly, climate data of some candidate sites for the storage in South Korea are measured from Korea Meteorological Administration (KMA). The climate data of the Diablo Canyon Power Plant located in California (United States) is obtained from National Weather Service (NWS) for comparison. Yearly data from 2012 to 2020 are applied and the crack growth is estimated for sites of different annual ranges.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"244 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132621244","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 paper provides an overview of the Leak-before-Break best practice guidance document that has been produced under the Horizon 2020 project ATLAS+.
本文概述了在地平线2020项目ATLAS+下生成的泄漏前破坏最佳实践指导文件。
{"title":"Overview of Leak-Before-Break Best Practice Document Developed Under the ATLAS+ Project","authors":"J. Sharples, Peter Gill, Brian Daniels","doi":"10.1115/pvp2022-82129","DOIUrl":"https://doi.org/10.1115/pvp2022-82129","url":null,"abstract":"\u0000 This paper provides an overview of the Leak-before-Break best practice guidance document that has been produced under the Horizon 2020 project ATLAS+.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"3 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126232562","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In order to evaluate the hydrogen embrittlement sensitivity of 4130X material, the effect of curvature radius of the high-steel ring on material failure mode and blasting pressure was firstly carried out, and then the hydrogen embrittlement sensitivity index was evaluated for disc samples with different surface roughness. The test results show that when the curvature radius of the high-steel ring is 0.5 mm, 1.5 mm, 2.0 mm, the material is mostly destroyed by shear. When the curvature radius is greater or equal to 4.0 mm, the material will be destroyed in the form of blasting. The blasting strength tested in hydrogen environment of 4130X materials decreases when the surface roughness increase from 0.1 to 4.1 μm. For the HE sensitivity evaluation of 4130X by DPT method, it is recommended that the curvature radius should be greater or equal to 4.0 mm, and the surface roughness between 0.1 and 1.5 μm can meet the DPT requirements.
{"title":"Evaluation of Hydrogen Embrittlement Sensitivity of 4130X Material Based on the Disc Method","authors":"Jian-ming Zhai","doi":"10.1115/pvp2022-84745","DOIUrl":"https://doi.org/10.1115/pvp2022-84745","url":null,"abstract":"\u0000 In order to evaluate the hydrogen embrittlement sensitivity of 4130X material, the effect of curvature radius of the high-steel ring on material failure mode and blasting pressure was firstly carried out, and then the hydrogen embrittlement sensitivity index was evaluated for disc samples with different surface roughness. The test results show that when the curvature radius of the high-steel ring is 0.5 mm, 1.5 mm, 2.0 mm, the material is mostly destroyed by shear. When the curvature radius is greater or equal to 4.0 mm, the material will be destroyed in the form of blasting. The blasting strength tested in hydrogen environment of 4130X materials decreases when the surface roughness increase from 0.1 to 4.1 μm.\u0000 For the HE sensitivity evaluation of 4130X by DPT method, it is recommended that the curvature radius should be greater or equal to 4.0 mm, and the surface roughness between 0.1 and 1.5 μm can meet the DPT requirements.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"94 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126237109","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}
N. Cofie, D. Dedhia, Gary Dominguez, M. Uddin, C. Harrington, N. Glunt, D. Shim
To address the dynamic effects of pipe break required by the General Design Criteria 4 (GDC-4), Appendix A of 10CFR50, deterministic leak-before-break (LBB) analysis (based on margins on leak rate and critical flaw size) as described in Standard Review Plan (SRP) 3.6.3 has been utilized by the nuclear industry. The LBB deterministic margins are generally met for large diameter piping (NPS 10 and greater) using a traditional LRD of 1.0 gpm. For smaller diameter piping, it is typically difficult to meet the required margins using LRD of 1.0 gpm necessitating a lower LRD. An alternate approach is to employ a probabilistic method to determine the probability of rupture and compare it to an appropriate probabilistic acceptance criterion. Recent release of the Extremely Low Probability of Rupture (xLPR) software has made the use of such a probabilistic approach possible. In this paper, a feasibility study was performed using xLPR to investigate if low probability of rupture can be demonstrated for a typical small diameter piping nozzle with a dissimilar metal (DM) butt weld that is susceptible to primary water stress corrosion cracking (PWSCC). This study assumes a small surface crack and grows this crack to determine if rupture occurs by a through-wall crack or a surface crack. Several sensitivity studies were performed to investigate the effects of key input variables on the rupture probabilities. The study has shown that it is feasible to justify low probability of rupture for small diameter DM welds susceptible to PWSCC. However, the WRS profile used in the evaluation has strong influence on the probability of rupture. The study has also shown the limitations in applying the deterministic LBB approach on a broader basis to small diameter DM welds in the presence of PWSCC, since rupture by surface cracks instead of through-wall cacks cannot be summarily dismissed and is a fundamental assumption in deterministic LBB with SRP 3.6.3.
{"title":"Application of Leak-Before-Break to Small Diameter Piping Nozzles With Dissimilar Metal Butt Welds Susceptible to PWSCC Using xLPR","authors":"N. Cofie, D. Dedhia, Gary Dominguez, M. Uddin, C. Harrington, N. Glunt, D. Shim","doi":"10.1115/pvp2022-86180","DOIUrl":"https://doi.org/10.1115/pvp2022-86180","url":null,"abstract":"\u0000 To address the dynamic effects of pipe break required by the General Design Criteria 4 (GDC-4), Appendix A of 10CFR50, deterministic leak-before-break (LBB) analysis (based on margins on leak rate and critical flaw size) as described in Standard Review Plan (SRP) 3.6.3 has been utilized by the nuclear industry. The LBB deterministic margins are generally met for large diameter piping (NPS 10 and greater) using a traditional LRD of 1.0 gpm. For smaller diameter piping, it is typically difficult to meet the required margins using LRD of 1.0 gpm necessitating a lower LRD. An alternate approach is to employ a probabilistic method to determine the probability of rupture and compare it to an appropriate probabilistic acceptance criterion. Recent release of the Extremely Low Probability of Rupture (xLPR) software has made the use of such a probabilistic approach possible.\u0000 In this paper, a feasibility study was performed using xLPR to investigate if low probability of rupture can be demonstrated for a typical small diameter piping nozzle with a dissimilar metal (DM) butt weld that is susceptible to primary water stress corrosion cracking (PWSCC). This study assumes a small surface crack and grows this crack to determine if rupture occurs by a through-wall crack or a surface crack. Several sensitivity studies were performed to investigate the effects of key input variables on the rupture probabilities.\u0000 The study has shown that it is feasible to justify low probability of rupture for small diameter DM welds susceptible to PWSCC. However, the WRS profile used in the evaluation has strong influence on the probability of rupture. The study has also shown the limitations in applying the deterministic LBB approach on a broader basis to small diameter DM welds in the presence of PWSCC, since rupture by surface cracks instead of through-wall cacks cannot be summarily dismissed and is a fundamental assumption in deterministic LBB with SRP 3.6.3.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"2014 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123831271","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 demonstrated an experimental testing process to quantitatively test for the effects of pressurized hydrogen on polymeric materials. It was shown that some o-ring materials are susceptible to deleterious property changes. A high-pressure hydrogen testing system with a window was fabricated and was used to soak elastomeric o-rings in hydrogen at temperatures from 22 to 40°C and 13.8 to 16.5 MPa for up to 96 hours. Next, the o-rings were observed through the window during the depressurization. Finally, the o-rings were characterized before and after hydrogen exposure for hardness, weight, tensile strength, and ultimate elongation. The data showed the following four quantitative trends 1) durometer decreased up to 14%, 2) weight was mostly unchanged, 3) tensile strength decreased up to 35%, 4) ultimate elongation decreased up to 55%.
{"title":"Testing for the Effects of Pressurized Hydrogen on Polymeric Elastomers","authors":"J. Ellis, Jessica Whitman, L. Zoller","doi":"10.1115/pvp2022-81859","DOIUrl":"https://doi.org/10.1115/pvp2022-81859","url":null,"abstract":"\u0000 This work demonstrated an experimental testing process to quantitatively test for the effects of pressurized hydrogen on polymeric materials. It was shown that some o-ring materials are susceptible to deleterious property changes. A high-pressure hydrogen testing system with a window was fabricated and was used to soak elastomeric o-rings in hydrogen at temperatures from 22 to 40°C and 13.8 to 16.5 MPa for up to 96 hours. Next, the o-rings were observed through the window during the depressurization. Finally, the o-rings were characterized before and after hydrogen exposure for hardness, weight, tensile strength, and ultimate elongation. The data showed the following four quantitative trends 1) durometer decreased up to 14%, 2) weight was mostly unchanged, 3) tensile strength decreased up to 35%, 4) ultimate elongation decreased up to 55%.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"52 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121438429","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}
Yufeng Feng, Ying-zhe Wu, J. Kuang, C. Gu, Jinyang Zheng, Z. Hua, Ruizhe Gao
Many materials exhibit a ductile-brittle transition as the temperature decreases. Liquefied gas and cryogenic equipment become more and more widely used in the industry, especially the accelerated commercialization of the liquid hydrogen industry. However, there is a lack of cryogenic mechanical property testing platform and published material mechanical property data below the liquid nitrogen temperature zone (77K). The development of cryogenic mechanical property testing platform is the basis for obtaining cryogenic mechanical property data of materials. Therefore, based on the existing cryogenic mechanical testing platform of the team, a system-level design of the modified cryogenic mechanical testing platform was designed and implemented. The modified platform is named Mechanical Performance Testing Platform for Liquid Hydrogen Temperature Zone and it is using a Gifford-Mcmahon refrigerator to provide cryogenic cooling and a refrigerated helium circulation loop to transfer the cooling to the specimen effectively. The lowest test temperature can reach the liquid hydrogen temperature range 20K. Finally, tensile tests of S30408 (Equivalent to UNS S30400) austenitic stainless steel at 20K and 77K are carried out by the Testing Platform.
{"title":"Development of Material Mechanical Properties Testing Platform for Liquid Hydrogen Temperature Zone","authors":"Yufeng Feng, Ying-zhe Wu, J. Kuang, C. Gu, Jinyang Zheng, Z. Hua, Ruizhe Gao","doi":"10.1115/pvp2022-84452","DOIUrl":"https://doi.org/10.1115/pvp2022-84452","url":null,"abstract":"\u0000 Many materials exhibit a ductile-brittle transition as the temperature decreases. Liquefied gas and cryogenic equipment become more and more widely used in the industry, especially the accelerated commercialization of the liquid hydrogen industry. However, there is a lack of cryogenic mechanical property testing platform and published material mechanical property data below the liquid nitrogen temperature zone (77K). The development of cryogenic mechanical property testing platform is the basis for obtaining cryogenic mechanical property data of materials.\u0000 Therefore, based on the existing cryogenic mechanical testing platform of the team, a system-level design of the modified cryogenic mechanical testing platform was designed and implemented. The modified platform is named Mechanical Performance Testing Platform for Liquid Hydrogen Temperature Zone and it is using a Gifford-Mcmahon refrigerator to provide cryogenic cooling and a refrigerated helium circulation loop to transfer the cooling to the specimen effectively. The lowest test temperature can reach the liquid hydrogen temperature range 20K. Finally, tensile tests of S30408 (Equivalent to UNS S30400) austenitic stainless steel at 20K and 77K are carried out by the Testing Platform.","PeriodicalId":434862,"journal":{"name":"Volume 4B: Materials and Fabrication","volume":"4 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115339975","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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