Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management最新文献

{"title":"Welded Joint Evaluation for Chromium Controlled Carbon Steel Piping to Improve FAC Resistance","authors":"Yoshio Uemoto, Takahiro Kawabe, H. Shibata, Shoh Tarasawa, H. Asano, J. Kaneda","doi":"10.1115/ICONE26-81913","DOIUrl":"https://doi.org/10.1115/ICONE26-81913","url":null,"abstract":"For condensate and feed water piping in nuclear power plants, it is desired to mitigate the pipe wall thinning risk due to Flow-Accelerated Corrosion (FAC). In aspect of material selection, low alloy steels are generally applied to improve FAC resistance. However, low alloy steels are inferior to carbon steels from the point of material cost and construction efficiency due to requirement of post weld heat treatment (PWHT). On the other hand, chromium is known as the most effective element to improve FAC resistance, and it is reported that a certain improvement of FAC resistance is also expected for carbon steels by increasing chromium content to over 0.10 wt%.\u0000 Such chromium controlled carbon steels are manufactured within the chemical composition range specified by material code of carbon steels, such as ASME B&PV Code Sec.II. Therefore, the amount of alloy content is lower than those for low alloy steels. The authors expect that PWHT can also be exempted for a certain thickness range of chromium controlled carbon steels, according to the exemption condition for normal carbon steels by ASME B&PV Code Sec.III. Furthermore, the chromium controlled carbon steels are generally cheaper than low alloy steels for base materials such as pipe and plate.\u0000 However, since chromium content of normal welding materials for carbon steels is generally lower than 0.05 wt%, chromium controlled carbon steel welding materials are specially-produced material. It makes the procurability worse compared to normal carbon steel welding materials. Additionally it should be confirmed if the increased chromium content affects the soundness of welded joint.\u0000 From the above reasons, it is necessary to decide the appropriate welding materials and methods for the chromium controlled carbon steel piping, considering the procurability of welding materials and the soundness of welded joint. In this study, the authors prepared the test pieces which simulate the assumed circumferential butt welded joints, then conducted the mechanical test such as tensile, impact, bend and hardness test to evaluate the soundness of welded joints. Furthermore, the authors evaluated the chromium content distribution of welded joints by using the Electro Probe Micro Analyzer (EPMA), in order to confirm if the chromium content is maintained over 0.10 wt% within the whole expected area.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"2015 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132447805","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}

{"title":"Evaluation of Local Damage to Reinforced Concrete Panels Subjected to Oblique Impact of Rigid and Soft Missiles","authors":"A. Nishida, M. Nagai, H. Tsubota, Yinsheng Li","doi":"10.1115/ICONE26-82615","DOIUrl":"https://doi.org/10.1115/ICONE26-82615","url":null,"abstract":"Many empirical formulas have been proposed for evaluating local damage to reinforced concrete (RC) structures caused by impacts of rigid missiles. Most of these formulas have been derived based on impact tests normal to the target structures. Up to now, few impact tests oblique to the target structures have been carried out. This study has been conducted with the purpose of proposing a new formula for evaluating the local damage caused by oblique impacts based on previous experimental and simulation results.\u0000 In this paper, the results of simulation analyses for evaluating the local damage to a RC panel subjected to normal and oblique impacts by rigid and soft missiles, by using the simulation method that was validated using the results of previous impact experiments. Based on the results of these simulation analyses, the effects of the rigidity of the missile as well as the impact angle on the local damage to the target structures are clarified.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"106 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128178702","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}

{"title":"Standby Equipment Reliability Data Analysis on Risk Monitor of Nuclear Power Plant","authors":"Yingfei Ma, Zhijian Zhang, He Wang, Sijuan Chen, Anqi Xu, Gang-yang Zheng","doi":"10.1115/ICONE26-82590","DOIUrl":"https://doi.org/10.1115/ICONE26-82590","url":null,"abstract":"Reliability data works as the basis of risk monitor of nuclear power plant. The failure modes of the equipment in a nuclear power plant can be divided into operation failure, demand failure and standby failure. A standby equipment is affected by the demand stress and the standby stress simultaneously, so the method of reliability data analysis must consider the two types of failure. The reliability data in online risk monitor should reflect the change of equipment reliability with time, including standby equipment. A method to deal with the reliability data of the standby equipment is presented in this paper. This model takes into account the failure of the equipment during the spare time and the failure of the starting time. Considering the characteristics of the reliability data in the nuclear power plant, the method of parameter estimation is studied. Finally, this method is applied to online risk monitor in nuclear power plant and the suggestion of reliability data application is put forward.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"124 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116829558","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}

{"title":"Development of Fuel Route/Dropped Load PSA for UK ABWR","authors":"Daisuke Taniguchi, Naoki Hirokawa, Y. Ishiwatari","doi":"10.1115/ICONE26-82022","DOIUrl":"https://doi.org/10.1115/ICONE26-82022","url":null,"abstract":"Generic Design Assessment (GDA) of Hitachi-GE’s United Kingdom (UK) Advanced Boiling Water Reactor (ABWR) was completed in December 2017. Various Probabilistic Safety Assessment/Analyses (PSAs) are utilized to provide the risk insights for design of UK ABWR. This paper provides the characteristics of “Fuel Route/Dropped Load” PSA which is included in the scope of UK ABWR PSA and focuses on the effects of dropped load. “Fuel Route” means the all areas where the new or spent fuels are moved or stored. Shutdown PSA and Spent Fuel Storage Pool (SFP) PSA cover large parts of risk assessment for the fuel route. However, large number of lifting operation such as refueling with fuel handling machine or heavy equipment lifting operation including cask operation by reactor building crane should be considered separately from Shutdown PSA and SFP PSA. That is because dropped loads potentially caused by such lifting operation may not always cause initiating events considered in the Shutdown PSA or SFP PSA but potentially cause a mechanical failure of the spent fuels and subsequent radiological consequences. “Fuel Route/Dropped Load PSA” is additionally developed and quantified to address the potential radiological release due to the lifting operation.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128675256","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}

{"title":"ASME Conference Presenter Attendance Policy and Archival Proceedings","authors":"","doi":"10.1115/icone26-ns2","DOIUrl":"https://doi.org/10.1115/icone26-ns2","url":null,"abstract":"","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"38 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128348752","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}

{"title":"Level 1 PRA for External Vessel Storage Tank of Japan Sodium-Cooled Fast Reactor in Scheduled Refueling","authors":"H. Yamano, Kenichi Naruto, K. Kurisaka, H. Nishino, Y. Okano","doi":"10.1115/ICONE26-81079","DOIUrl":"https://doi.org/10.1115/ICONE26-81079","url":null,"abstract":"Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure by conducting probabilistic risk assessment for EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, human error probability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10−6 /year in this paper. By considering the secondary sodium freezing, the fuel damage frequency was twice increased. The dominant accident sequence resulted from the common cause failure of the damper opening and/or the human error for the switching from the stand-by to the operation mode in the three stand-by cooling circuits. The second dominant accident sequence following the secondary pump trip is sodium freezing caused by the failure of air blower trip in the air cooler due to the common cause failures of secondary sodium flowmeter failure or erroneous opening of the air cooler damper. The Fussell-Vesely importance and risk achievement worth analyses have indicated high risk contributions.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"81 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114285161","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}

{"title":"Demonstrative HEAF (High Energy Arcing Fault) Fire Tests of High and Low Voltage Switchgears of Nuclear Power Plants","authors":"K. Shirai, T. Miyagi, M. Iwata, K. Tasaka, Junghoon Ji","doi":"10.1115/ICONE26-82177","DOIUrl":"https://doi.org/10.1115/ICONE26-82177","url":null,"abstract":"High Energy Arcing Faults (HEAF) have the potential to cause extensive damage to the failed electrical components and distribution systems along with adjacent equipment and cables within the zone of influence (ZOI). Furthermore, the significant energy released during HEAF event can act as an ignition source to other components within the area of the HEAF. In Japan, during the Great East Japan Earthquake occurred in 2011, the seismic induced HEAF fire event, which induced the whole damage of the multiple high voltage switchgears, was observed in Onagawa Nuclear Power Plant (NPP).\u0000 In response, in August 2017, the NRA (Nuclear Regular Authority) in Japan amended the safety requirement for the power supply to consider the influence of the successive fire due to the HEAF event (hereinafter HEAF fire event). Therefore, it is urgently necessary to establish the design criteria to prevent the HEAF fire event, and enhance the experiment data of the HEAF fire event. In order to estimate the total arc energy during the HEAF event and obtain the threshold value to prevent the HEAF fire for the existed non-arc proof electrical cabinets, several series of three-phase internal arc tests with high (6.9kV class) and low (480V class) voltage electrical cabinets were executed. We executed internal arc tests with full scale high/low voltage metal-enclosed switchgear components (non-arc proof type, copper bus conductor), and evaluated arc energy, the mechanical damage of the cabinet and the surrounding equipment due to the impulsive pressure and the possibility of successive fire occurrence. In case of high voltage switchgear, when the arcing energy exceeded 25.3MJ, successive fire was identified. Especially, in the case where the arc flash was discharged in the circuit breaker room, a 2-second arcing duration in a three-phase short-circuit current with 18.9kA (measured arcing energy over 40MJ) caused successive fire which required extinguishment. On the other hand, in case of low voltage power center, when the arcing energy exceeded 19MJ, successive fire was identified. According to these demonstrative tests, this paper presents the evaluation method to estimate total arc discharge energy during the HEAF event for high and low voltage electrical cabinets.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133760835","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}

{"title":"Research on Auxiliary Bearing Structure With Buffer Shim Based on LS-DYNA for Helium Circulator of HTR-10","authors":"Guojun Yang, Zhe Sun, Xingnan Liu, Zhengang Shi","doi":"10.1115/ICONE26-81304","DOIUrl":"https://doi.org/10.1115/ICONE26-81304","url":null,"abstract":"The active magnetic bearing (AMB) is introduced to the helium circulator of the 10 MW high temperature gas-cooled reactor (HTR-10). Auxiliary bearing bears huge instantaneous impact load when the AMB fails to work in the helium circulator. The huge load may cause serious damage to the auxiliary bearing. In order to reduce the impact load and protect the AMB and rotor system, a new type of auxiliary bearing with axial buffer shim is presented in this paper. There are two different structures of auxiliary bearing to be studied in the paper. One is the auxiliary bearing with the buffer shim and another one is without the buffer shim. The finite element method and LS-DYNA software are applied to analyze the structure characteristic of the auxiliary bearing. The rotor dropping trajectory and the stress distribution of the auxiliary bearing are analyzed by comparing two different auxiliary bearing structures during the rotor drop. The stress change of the auxiliary bearing in each impact course is mainly studied, and the law of stress variation in impact course is analyzed. Besides, the stress distribution and deformation of two auxiliary bearing is the focus research when impact force is maximum in the paper. Finally, the research shows the auxiliary bearing structure without the buffer shim can bear the huge impact of the rotor drop, and the addition of buffer shim can also reduce the damage of the rotor drop to the auxiliary bearing structure. These researches’ result provides an important reference for the experiment of rotor drop, and has laid a theoretical foundation for the practical application of this structure.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"37 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133803138","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}

{"title":"A Systematic Study of the Material Performance of Hot Isostatically Pressed Type 316L Stainless Steel Powder for the Civil Nuclear Sector","authors":"W. Kyffin, D. Gandy, Barry Burdett","doi":"10.1115/ICONE26-81438","DOIUrl":"https://doi.org/10.1115/ICONE26-81438","url":null,"abstract":"Hot Isostatic Pressing (HIP) of type 316L stainless steel powder has been an established manufacturing practice for more than twenty-five years in the oil and gas sector and more recently in the naval defence sector. To demonstrate the capability of the powder metallurgy HIP (PM/HIP) for nuclear power applications a systematic study of 316L commercial powder production, encapsulation/consolidation providers and selected HIP parameters was undertaken by the Nuclear AMRC in collaboration with the Electric Power Research Institute (EPRI). In the study, the 316L powder specification limited the oxygen content of the powder to under 130 parts per million (ppm), which reflects the improvements that commercial powder suppliers have been making over the past decade to ensure greater powder cleanliness.\u0000 The test programme assessed powder supply, HIP service provider and HIP sustain time. Excellent test results were achieved across the full range of variables studied with all billets meeting the specification requirements of ASTM A988 and additional requirements imposed based on nuclear manufacturing standards.\u0000 Significantly, the study demonstrated the robustness of the PM/HIP supply chain, as material produced via differing HIP service providers resulted in very consistent material properties across the destructive test programme. Furthermore, no significant difference in material properties were noted for material HIP’ed between 2–8 hours hold time, suggesting that the HIP process window is large. Both these results are significant from an end-user standpoint as they highlight the uniformity of the process through the full manufacturing cycle from powder procurement to destructive testing.\u0000 Despite all material passing specification requirements, some property variation was noted for differing powder suppliers. Considering the systematic approach, this was attributed to powder composition, with both low oxygen and high nitrogen contents contributing to improvements in Charpy impact strength and tensile strength respectively.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"24 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115634020","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}

{"title":"Study on Methods for HCLPF Value of Nonlinear Supports System of Steam Generator","authors":"Feng-chun Cai, Xian-hui Ye, Qian Huang, Wen-zhong Zhang","doi":"10.1115/ICONE26-81279","DOIUrl":"https://doi.org/10.1115/ICONE26-81279","url":null,"abstract":"High confidence of low probability of failure (HCLPF) values of equipment, representing the seismic capacities of the equipment, are the fundamental ingredient in seismic probability safety assessment (SPSA) and seismic margin analyses (SMA). In this paper, two methods for calculating the HCLPF values of equipment were investigated, fragility analysis, and conservative deterministic failure margin (CDFM). These methods are linear methods. Based on these methods, HCLPF value of equipment can be computed conveniently by scaling the results of the existing seismic analysis. For a nonlinear systems, the HCLPF values based on these linear scaling methods are unrealistic. For a complicated nonlinear equipment or structure, a detail nonlinear model was used to derive the seismic capacity. The results by this method are realistic, but cost calculation time. In this paper, a nonlinear model of reactor coolant system coupled reactor building was built. This model includes the steam generator and considers the nonlinear factors of steam generator such as gap in the supports, plasticity of hot leg and cold leg. Forced motion was applied to the base of reactor building. And seismic response of the steam generator was calculated iteratively by scaling the ground motion level step by step. Based on these calculations, a curve of load on the supports VS peak ground acceleration (PGA) can be obtained. Then based on these curves and allowable load of supports of steam generator, which derived from stress analysis on support of steam generator, seismic capacity of the supports of steam generator was determined. Then the HCLPF Value of the supports of steam generator was obtained by this nonlinear time history analysis and was compared with the results based on the CDFM. The two results were different. Therefore, the HCLPF seismic capacity of equipment with nonlinearity, such as gap nonlinearity, should be calculated by nonlinear time history method.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"14 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114801952","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}