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

{"title":"Creep Deformation Analysis of a Pipe Specimen Based on Creep Damage Evaluation Method","authors":"J. Katsuyama, Yoshihito Yamaguchi, Yinsheng Li","doi":"10.1115/ICONE26-82568","DOIUrl":"https://doi.org/10.1115/ICONE26-82568","url":null,"abstract":"After the Fukushima Daiichi nuclear power station (1F) accident, it has become more important to develop methods for evaluating the failure behavior of the nuclear components under severe conditions. We are, therefore, researching prediction methods for creep deformation and failure behavior of nuclear components under elevated temperature conditions based on finite element analysis. In this study, as part of a project called Components and Structures under Severe Accident Loading (COSSAL), we perform failure analysis of a large-scale pipe experiment to validate our prediction methods based on a creep damage evaluation method. The analytical results obtained from the conventional Norton and modified Garofalo constitutive laws, as well as the Kachanov-Rabotnov law considering material damage due to creep deformation, were compared with experimental results. We conclude that the Kachanov-Rabotnov law allows for the most accurate analysis among the three creep constitutive laws considered herein.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"18 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":"134214986","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":"Design of the Sampling Measurement and Radiochemistry Lab in the Nuclear Island of HTR-PM","authors":"Mengqi Lou, Wenqian Li, F. Xie, Jian-zhu Cao, Liqiang Wei, J. Tong, J. Kong","doi":"10.1115/ICONE26-81701","DOIUrl":"https://doi.org/10.1115/ICONE26-81701","url":null,"abstract":"Radiation monitoring system is very important to the safe operation of the nuclear power plant. The radiation monitor, including the on-line type and the off-line type, can provide the radioactive level for a given system, certain areas, or unique substance promptly. However, as a supplementary method, the sampling measurement can supply more accurate information about the source term. In this paper, we present the design of the sampling measurement in the nuclear island of HTR-PM. The sampling measurement contains the helium sampling from the primary coolant, the radioactive dust sampling from the primary loop, the liquid sampling from the tritiated water and some process systems, the gas sampling from certain areas, etc. The frequency of the sampling measurement depends on regulatory requirements and actual operational demands. The control values of the samples are settled on the basis of the source term analysis and regulatory requirements. The radiochemistry lab has been designed to meet the requirement of the sampling measurement, in which the main instruments include a NaI γ spectrometer, two high-purity germanium γ spectrometers, a four-channel ultralow background α/β analyzer, and a low background liquid scintillation counter. The characteristics of these instruments will be described in details in the paper.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"41 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":"133116328","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":"WATCH Loop Development and Commissioning Tests","authors":"W. Byers, G. Wang","doi":"10.1115/ICONE26-82626","DOIUrl":"https://doi.org/10.1115/ICONE26-82626","url":null,"abstract":"It is essential to understand how crud build-up leads to elevated cladding temperatures and abnormal cladding corrosion for Boiling Water Reactor (BWR) fuel. With a heated rod test loop configuration, the influence of water chemistry on crud build-up and cladding corrosion behavior can be assessed.\u0000 A new test loop named the Westinghouse Annular Thermal Crud Hydraulic (WATCH) Loop, has been designed to grow simulated BWR crud under normal BWR operating conditions. WATCH simulates the bottom meter of a BWR core in terms of temperature, flow, steam quality and heat flux. Test results from this loop can be used to benchmark risk assessment tools, which in turn are used to make predictions and minimize or avoid BWR fuel rod failures, consistent with guidelines from the Institute of Nuclear Power Operations (INPO) and the Electric Power Research Institute (EPRI).\u0000 In this paper, the Westinghouse Annular Thermal Crud Hydraulic (WATCH) Loop design and development are introduced. The results of a commissioning test and a run producing zinc-rich deposits are also presented.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"46 9","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"113987449","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":"Design and Development of the Platform for Significance Determination Process System for Nuclear Power Plant","authors":"Qinfang Zhang, Guoxu Zhang, Zilong Wang, Qi Dong, Guofeng Tang","doi":"10.1115/ICONE26-81399","DOIUrl":"https://doi.org/10.1115/ICONE26-81399","url":null,"abstract":"Significance determination process (SDP) is one of important application tools in the scheme of risk-informed and performance based regulation. It is used to help the analyst evaluate the risk significance of a particular safety concerned event. By referring to the design and development of SDP system, this paper gives a brief introduction on the design and development of the system. An autonomously developed PSA quantification engine is incorporated into the SDP system, and it helps to resolve quantification issue for event risk evaluation. From qualified screening to risk significance quantification, SDP system is designed to realize the whole process for event risk analysis. SDP system could potentially help analysts from both regulatory agency and NPP licensees to quickly evaluate the risk significance of a particular event.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"11 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":"122360975","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":"Optimal Design and Performance Simulation of a Novel Semi-Active Vibration Absorber for Pipeline System of NPP","authors":"Zhiguo Wei, J. Gou, M. Liao, Shaodan Li, L. Dai","doi":"10.1115/ICONE26-81741","DOIUrl":"https://doi.org/10.1115/ICONE26-81741","url":null,"abstract":"The vibration of pipeline system induced by pump is one of the most important factors affecting the safety and stability of nuclear power plant (NPP) operation, and is also the most prominent vibration and noise transmission channel in the plant area. Analysis results show that various support structures such as pipe clamps, hangers and brackets are the main transmission routes of pipeline vibration. However, it is usually difficult to obtain a satisfactory absorption effect by using traditional passive vibration isolation device, especially for a variety of low-frequency vibration. Besides, the effective isolation frequency is fixed which cannot adapt to the variation of the operation condition excitation source. Therefore, it is of great practical significance to propose a method and scheme which can effectively restrain the low-frequency vibration of pipeline system of NPP. In order to meet the above requirements, an electromagnetic semi-active vibration absorber is designed in this paper. Its electromagnetic stiffness can be adjusted by changing the excitation current, so as to achieve the purpose of changing the natural frequency online and improving the vibration attenuation effect. In order to verify the rationality of its electromagnetic structure, the magnetic field distribution of vibration absorber is simulated and analyzed. Results show that it is close to magnetic saturation inside without leakage. On this basis, a platform of simulated condensate water pipeline system for vibration experiment is set up, on which the frequency response together with time & frequency-domain vibration characteristics of typical support points are tested, and the stiffness, damping and other parameters of the vibration absorber are optimized according to the test results. Finally, the vibration of pipeline system is simulated and compared. The results show that under the condition of the optimal combination of parameters, the vibration spectrum of the pipeline is obviously reduced after the vibration absorber is used, which indicates a significant vibration control effect.","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":"116727357","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 Structural Integrity Assessment of a Nuclear Boiler Superheater Bifurcation at High Temperature","authors":"B. Nadri, Robert X. Wang","doi":"10.1115/ICONE26-81167","DOIUrl":"https://doi.org/10.1115/ICONE26-81167","url":null,"abstract":"Steam generating boilers in gas cooled nuclear reactors in the UK operate at high temperatures and some of them have been in service for more than 30 years and are now facing the challenges from long term operation extension demand. The tubular components experience surface metal losses due to exposure to oxidation and corrosive environment and as a result, some tubes suffer from restricted flow which may lead to an increased creep-fatigue crack initiation damage. To maintain or recover boiler heat transfer efficiency, internal chemical cleaning of selected boiler tubes is carried out, which introduces additional metal loss in the tube wall, weakening its load bearing capacity.\u0000 Some boiler components are subject to high temperature, pressure and mechanical loadings in large number of operating cycles through life, introducing creep in addition to cyclic fatigue damage.\u0000 In support of an operational safety case and plant long term operation extension requirements, structural integrity assessments have been carried out on a critical boiler component — bifurcation, taking into account tube wall metal loss for extended long term services, including the effects of possible future chemical cleaning operations.\u0000 This paper presents the finite element analyses and R5 Volume 2/3 assessment work carried out for the structural integrity substantiation of a stainless steel boiler tube bifurcation. The bifurcation is a tubular component subject to significant applied displacement due to long range thermal expansion of the neighbouring components. The initial study following normal industry practice using a decoupled analysis approach showed that the strain ranges obtained would exceed the creep-fatigue crack initiation capacity and plastic ratchetting would occur which would lead to short term, incremental plastic collapse, hence a safety case could not be made.\u0000 To meet the challenge, the analysis and assessment processes have been examined. A coupled FE analysis approach was used to remove the pessimism associated with the decoupled analysis approach. This approach captures the displacement-controlled nature of the system loads and allows a more realistic assessment. In addition, the plant life has been divided into a number of assessment periods such that the more realistic metal loss appropriate for each period could be used. Furthermore, segregated temperature zones have been considered in the assessment, leading to a significant reduction in the creep-fatigue crack initiation damage and a satisfactory extended long term operation safety case.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"82 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":"130238827","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":"Transient Analysis of Fire Protection System at a Nuclear Power Plant Using Computer Code USLAM","authors":"A. Arastu, E. Tom","doi":"10.1115/ICONE26-82622","DOIUrl":"https://doi.org/10.1115/ICONE26-82622","url":null,"abstract":"A new Method Of Characteristics (MOC) based computer code, USLAM, has been used to analyze the Fire Protection System (FPS) of a Nuclear Power Plant. Some unique features of this system are that it has a complex geometry, has many operating modes, is connected to an adjacent nuclear power plant FPS, and has a significant number of high elevation standpipes or risers. In a typical FPS, firefighting water is pumped from a low elevation reservoir at atmospheric pressure to the whole system including higher elevations regions where water hammer due to water column separation & rejoining may occur. A low capacity keep-full system is employed to keep the system pressurized during standby conditions. A loss of system pressure due to the opening of the pre-action or deluge valves can lead to void formation at high elevations whose collapse can result in severe water hammer. A catastrophic valve failure due to a water hammer event at a nuclear power plant (Arastu, et al, 1999) disabled the entire FPS.\u0000 The analysis presented in this paper is to evaluate the possibility of such a water hammer event and mitigate it in an effective manner. The mitigation method employed is the use of multiple vacuum breakers. As many as 11 vacuum breakers are used for this purpose. The paper discusses the analysis results without and with vacuum breakers for all operating modes. The basic methodology of the USLAM code is also discussed.","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":"130142210","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":"Design of the Important Area Airborne Radioactivity Monitoring System and Calculation of the Alarm Threshold of HTR-PM","authors":"Wenqian Li, Chuan Li, Jian-zhu Cao, F. Xie","doi":"10.1115/ICONE26-81232","DOIUrl":"https://doi.org/10.1115/ICONE26-81232","url":null,"abstract":"Unlike the conventional pressurized water reactor (PWR), helium is adopted as the coolant in the high-temperature gas-cooled reactor pebble-bed module (HTR-PM). Due to the leakage of the primary coolant, in certain process rooms and some important areas, the airborne radioactivity should be monitored for the purpose of personnel radiation protection and emergency control. The system executing above function in HTR-PM is called important area airborne radioactivity monitoring system. This system consists of five separate monitoring channels, in which online and/or sampling monitoring devices will be included. For radiation monitoring system of a reactor, the calculation and setting of alarm thresholds for online devices is an important issue. The set of alarm thresholds requires not only accurate calculation but also feedback from actual operational experience. The design of the high-temperature reactor is different from the PWR, so the set of the alarm thresholds need special considerations. Based on the calculation of source terms and production-removal model, the theoretical radioactive levels of the sampled gases of all the online monitors were calculated and the alarm thresholds adopted in the current design are presented in this paper. The set of the alarm thresholds takes the influences of the background and other factors into account. This paper provides a method for calculating and determining the alarm thresholds of the radiation monitoring system, which can provide references for other nuclear power plants.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"195 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":"122527772","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":"Reliability Prediction for the Squib Valve of Advanced Passive PWR by Hardened Test Method","authors":"Shao-hua Ge, Zhang Qinfang","doi":"10.1115/ICONE26-81822","DOIUrl":"https://doi.org/10.1115/ICONE26-81822","url":null,"abstract":"The Squib valve is never used in the Pressurized Water Reactor (PWR) before, and it plays an important role in the advanced passive PWR accident mitigation. Thus the reliability of squib valve has received wide attention and high consideration. As a pyrotechnic mechanical device, the currently accepted approach for predicting the reliability of the squib valve requires hundreds or thousands of consecutive, successful tests of very similar components. The hardened test method for predicting the function reliability of the squib valve is presented in this paper. The squib valves are tested under hardened conditions. If they can successfully be actuated under hardened conditions, then the functional reliability under normal condition can be predicted based on the stress-strength model and one-shot product reliability assessment theory. In this way, high reliability of the squib valve can be assessed with only a limited number of tests. The hardened test method can provide considerable cost savings in testing pyrotechnical devices.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"4 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":"116996179","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 of γ-Ray 3D Scanning Detection and Analysis System","authors":"Chen Chen, S. Hou, Quan-hu Zhang, X. Su, Wenming Zuo","doi":"10.1115/ICONE26-81929","DOIUrl":"https://doi.org/10.1115/ICONE26-81929","url":null,"abstract":"With the development of nuclear industry, nuclear facilities have accumulated a large amount of pending nuclear waste during operation. Usually, they are stored in a closed container, and their physical and chemical forms are complicated and unevenly distributed. γ-ray 3D scanning detection and analysis system using non-destructive analysis of γ-ray analysis method to achieve a closed container of radionuclide recognition, positioning, quantitative measurement and analysis. The system has broad application prospects in the fields of nuclear safety, nuclear material control and accounting, radioactive recyclables and non-destructive measurement of nuclear waste. γ-ray 3D scanning detection and analysis technology is a comprehensive high-tech integrated nuclear science, nuclear radiation detection, radiation imaging, image processing, mechanical design, automatic control, computer technology and other disciplines.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"287 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":"124570287","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}