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

Study on PRA Procedure Considering Combination of Multiple Events Using DQFM Methodology 基于DQFM方法考虑多事件组合的PRA程序研究

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-82086

H. Yamakawa, H. Muta

引用次数: 1

A PSA Case Study: Promoting the Reliability of the CRSS on the CMRR Through the ATWS Mitigation System PSA案例研究:通过ATWS缓解系统提高CMRR上CRSS的可靠性

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-81755

Heng Yu, Guangze Wang, Da-zhi Qian, Yuchuan Guo, B. Hu

{"title":"A PSA Case Study: Promoting the Reliability of the CRSS on the CMRR Through the ATWS Mitigation System","authors":"Heng Yu, Guangze Wang, Da-zhi Qian, Yuchuan Guo, B. Hu","doi":"10.1115/ICONE26-81755","DOIUrl":"https://doi.org/10.1115/ICONE26-81755","url":null,"abstract":"An increasing number of PSA programs concerning research reactors have been launched across the world. As with many other reactors, the CMRR (China Mianyang Research Reactor), a typical pool-type research reactor, regards the control rod shutdown system (CRSS) as its primary shutdown system which enables the reactor subcritical by dropping control rods into the core after a specific initiating event is detected. As a result, the CRSS is an essential ingredient of engineered safety features. It is necessary to enhance the reliability of the CRSS, ensuring the reactor can be successfully shut down when the ATWS — the anticipated transients without scram occurs. Therefore, additional facilities should be designed to cope with the extremely severe circumstance. Accordingly, the purpose of this paper is to evaluate the promotion of the CMRR’s safety degree and the reliability of its CRSS from the PSA’s perspective with an ATWS mitigation system installed. Results indicate that, by introducing the ATWS mitigation system, the failure probability of the CRSS can decrease from 1.52e−05 per demand to 3.35e−06 per demand, while the aggregate CDF (core damage frequency) induced by all IE (initiating event) groups, is able to decrease to a relatively low value 1.17e−05/y from its previous value 3.11e−06/y. It is apparent that the reliability of the CRSS as well as the safety degree of the overall reactor can be enhanced effectively by adding the ATWS mitigation system to the elementary design of the normal CRSS.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"115 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":"124032065","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}

引用次数: 0

Epistemic Uncertainty Quantification of Floor Responses for a Nuclear Reactor Building 核反应堆建筑楼板响应的认知不确定性量化

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-82034

Byunghyun Choi, A. Nishida, Yinsheng Li, K. Muramatsu, T. Takada

{"title":"Epistemic Uncertainty Quantification of Floor Responses for a Nuclear Reactor Building","authors":"Byunghyun Choi, A. Nishida, Yinsheng Li, K. Muramatsu, T. Takada","doi":"10.1115/ICONE26-82034","DOIUrl":"https://doi.org/10.1115/ICONE26-82034","url":null,"abstract":"After the 2011 Fukushima accident, engineers of nuclear power plants are looking beyond the basic design requirements and ensuring that countermeasures are built in to avert possible nuclear accidents. In seismic probabilistic risk assessment (SPRA), uncertainties can be classified in two ways as aleatory uncertainties or epistemic uncertainties. To improve the reliability of SPRA, the difference in seismic response due to difference of building modelings related to epistemic uncertainty was focused on. Two modeling methods were used for a seismic response analysis: a three-dimensional finite-element model and a conventional sway-rocking stick model. Simulated input ground motions related to aleatory uncertainty were generated for the input waves. Then, the seismic floor response results of the various input ground motions of the two modeling methods were quantified. For the uncertainty quantification related to the different building modelings, a statistical analysis of the floor response results of the nuclear reactor building were further performed. Finally, for the quantification of the uncertainty in the fragility analysis for SPRA, the way to use of these results were discussed.","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":"132442566","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}

引用次数: 1

Research on Structural Design and Analysis of S-CO2 Turbine Impeller S-CO2涡轮叶轮结构设计与分析研究

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-81267

Jun Wu, can ma, C. Dai, Zhenxing Zhao, L. Dai, Zhouyang Liu

{"title":"Research on Structural Design and Analysis of S-CO2 Turbine Impeller","authors":"Jun Wu, can ma, C. Dai, Zhenxing Zhao, L. Dai, Zhouyang Liu","doi":"10.1115/ICONE26-81267","DOIUrl":"https://doi.org/10.1115/ICONE26-81267","url":null,"abstract":"The Brayton cycle with supercritical carbon (S-CO2) as working medium is one of the most promising new nuclear power systems. Turbine is the key device during the working process in the Brayton power cycle. The turbine structural presents small size and extremely high rotational speed for the special physical properties of S-CO2, which increase the difficulty for the structural design and strength safety significantly. According to the aerodynamic design and optimization results of 200 kW S-CO2 radial inflow turbine, this paper proposes a detail structural design and analysis method for turbine impeller. Based on the three-dimensional blade profile data and meridional planes data, key structural design parameters are chosen and the parametric geometry model is established by CAD tools. On this basis, numerical simulation models of turbine are established to analyze the structural strength in detail. Then the influence of parameters on the turbine impeller strength is studied by a series of finite element numerical procedures. The influence mechanisms of key structural design parameters on impeller strength are discussed. Moreover, the final model of turbine impeller is obtained by parameter comparison and selection. The results show that for the initial model, the maximum von-Mises equivalent stress is 400.10 MPa, the maximum radial deformation is 0.0333 mm and the maximum axial deformation is 0.0770 mm. For the final model, the maximum von-Mises equivalent stress is 294.26 MPa, the maximum radial deformation is 0.0279 mm and the maximum axial deformation is 0.0769 mm. The maximum von-Mises equivalent stress and maximum radial deformation of structural decreases 26.45 % and 16.22 % respectively compared with the initial model. As a result, the impeller structural strength safety margin is obviously improved by the parameter analysis.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"201 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":"134242048","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}

引用次数: 0

Analytical Study on Fragility Evaluation With Uncertainty Against Fault Displacement for Nuclear Power Plant Buildings 核电厂建筑物断层位移不确定脆性评价的分析研究

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-81072

K. Ishiki, H. Tsuji, M. Kanechika, Y. Mihara

{"title":"Analytical Study on Fragility Evaluation With Uncertainty Against Fault Displacement for Nuclear Power Plant Buildings","authors":"K. Ishiki, H. Tsuji, M. Kanechika, Y. Mihara","doi":"10.1115/ICONE26-81072","DOIUrl":"https://doi.org/10.1115/ICONE26-81072","url":null,"abstract":"Japan Nuclear Safety Institute had recently reported the pioneering deterministic evaluation approach for nuclear power plant under seismic induced fault displacement. But the uncertainty of fault displacement based on probabilistic hazard analysis is described to be greater than that of other natural phenomena e.g. earthquake ground motions or seismic acceleration vibration in the report. Furthermore, for plant-wide risk assessment against fault displacement hazards beyond design basis displacement level, it is seriously necessary to promote a series of fundamental studies and develop the standard procedures regarding not only accident sequence analysis but also fragility analysis of buildings and structures as well as components and piping systems.\u0000 Based on the above background, the objective of this study is focusing to obtain basic fragility data for the aleatory and epistemic uncertainties of structural responses for nuclear power plant buildings against fault displacement. A number of nonlinear soil-structure finite element analyses against relatively large fault displacement are performed with the randomness of soil and building material properties, the uncertainly of contact parameters relating to friction between soil and building, and also the uncertainty of fault hazards such as fault types and geometries. Their quantitative results for fragility data are shown in this paper.","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":"123710442","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}

引用次数: 0

Finite Element Analysis of AMB Eddy-Current Loss in HTR-PM Primary Helium Circulator HTR-PM一次氦循环器AMB涡流损耗的有限元分析

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-81375

Jinpeng Yu, Yan Zhou, Mo Ni, Guojun Yang, Lei Zhao

引用次数: 0

Influence of Gap Size on Added Mass for Spent Fuel Storage Rack 间隙尺寸对乏燃料贮存架附加质量的影响

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-82595

D. Lu, Yu Liu, Shu Zheng

{"title":"Influence of Gap Size on Added Mass for Spent Fuel Storage Rack","authors":"D. Lu, Yu Liu, Shu Zheng","doi":"10.1115/ICONE26-82595","DOIUrl":"https://doi.org/10.1115/ICONE26-82595","url":null,"abstract":"Free standing spent fuel storage racks are submerged in water contained with spent fuel pool. During a postulated earthquake, the water surrounding the racks is accelerated and the so-called fluid-structure interaction (FSI) is significantly induced between water, racks and the pool walls[1].\u0000 The added mass is an important input parameter for the dynamic structural analysis of the spent fuel storage rack under earthquake[2]. The spent fuel storage rack is different even for the same vendors. Some rack are designed as the honeycomb construction, others are designed as the end-tube-connection construction. Therefore, the added mass for those racks have to be measured for the new rack’s design. More importantly, the added mass is influenced by the layout of the rack in the spent fuel pool.\u0000 In this paper, an experiment is carried out to measure the added mass by free vibration test. The measured fluid force of the rack is analyzed by Fourier analysis to derive its vibration frequency. The added mass is then evaluated by the vibration frequency in the air and water. Moreover, a two dimensional CFD model of the spent fuel rack immersed in the water tank is built. The fluid force is obtained by a transient analysis with the help of dynamics mesh method.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"84 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":"123835058","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}

引用次数: 0

Study on Shutdown Fire PRA for Nuclear Power Plant 核电站停堆火灾PRA研究

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-81272

Meiru Liu, Q. Zhao, Wei Deng, Jinyan Du, Lin Sun

引用次数: 0

Calibration of Inspection Strategies in Support of Aging Management Programs: A Probabilistic Approach 支持老化管理计划的检验策略校准:一种概率方法

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-81115

M. Pandey, M. Jyrkama

引用次数: 0

Overview of PSA for the UK ABWR Generic Design Assessment 英国ABWR通用设计评估PSA概述

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management Pub Date : 2018-07-22 DOI: 10.1115/ICONE26-82553

Naoki Hirokawa, Y. Ishiwatari, Daisuke Taniguchi, K. Hisamochi

引用次数: 0