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

{"title":"The Influence of Thermal Deformation on the Dynamic Characteristics of AMB Rotor of HTR-PM Helium Blower","authors":"G. Du, Jinpeng Yu, Hong Wang, Lei Zhao","doi":"10.1115/ICONE26-81132","DOIUrl":"https://doi.org/10.1115/ICONE26-81132","url":null,"abstract":"Helium blower is the core component of high temperature gas cooled reactor, which rotor is supported by active magnetic bearings (AMBs). The special advantage of AMB is that there is no contact between bearing and rotor, and this permits operation with small friction, long service life, no lubrication system, and no pollution to the helium environment. [1–3] Helium blower rotor is mainly composed of rotating shaft, impeller, motor, cooling blower and so on, which runs in an uneven temperature field that the impeller runs in a helium chamber of 250 degrees centigrade, and the motor housing’s outer surface temperature is 65 degrees centigrade. The temperature rises from standstill to stable operation will cause the thermal deformation of rotor and bearing, leading to the change of gap between rotor and bearing, which will lead to the change of electromagnetic force of AMB. The electromagnetic force determines the bearing stiffness and bearing damping of the AMB, so the change of temperature is the most important to the stiffness and damping of the AMB, which can affect the dynamic characteristics of the rotor. Through finite element method (FEM) to calculate the temperature field and displacement field of helium blower, the change of the gap of AMB and rotor is calculated. The rotor radial displacement orbits are obtained through numerical simulation, which are affected by thermal deformation. Finally, the results of numerical simulation are verified by experiments. The simulation and experiments both show that temperature rise can increase the vibration amplitude of rotor, so the influence of thermal deformation should be considered when designing the active magnetic bearings.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"75 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":"124988259","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 the Shaking Dynamic Response of Steam Generator LOCA","authors":"Qian Huang, Xiaofei Yu, Huan-huan Qi, N. Jiang, Feng-chun Cai, Zhipeng Feng","doi":"10.1115/ICONE26-81278","DOIUrl":"https://doi.org/10.1115/ICONE26-81278","url":null,"abstract":"Dynamic response research of steam generator loss of coolant accident (SG LOCA) is essential for the reliability and safety consideration. According to the differences of LOCA loading phenomena, two types of LOCA loads affect the SG: rarefaction wave travels through the primary fluid in the U-tubes, and the SG shakes due to reactor coolant loop(RCL) motions transmitted by the primary loop piping, former loading phenomena evaluation is called SG rarefaction analysis while latter is called shaking analysis. This paper place particular emphasis on shaking analysis.\u0000 At present, the published literatures about LOCA mainly focus on RCL LOCA, reactor LOCA and fuel assembly LOCA, few reports concentrate on shaking dynamic response analysis of SG LOCA. Both Westinghouse and AREVA’s methods according to their research reports are to decouple the SG from the RCL: This method results in low computational efficiency as RCL LOCA and SG LOCA are evaluated separately and the decoupling error is uncertainty, meanwhile, in the vicinity of the nodes where the displacements are imposed, distorted reaction forces are usually found.\u0000 Through reasonable simplification and equivalence, a detailed nonlinear FEM model of steam generator (SG) of a China 3rd generation nuclear power plant (NPP) is established, this model is then connected with the reactor coolant loop (RCL) to carry out the SG LOCA shaking dynamic response analysis. By calculation, the maximum absolute stresses of SG heat transfer tube bundle and its variation with tube diameter and upper supports reacting forces are obtained. In order to study the effect of SG decoupling from the RCL on shaking dynamic response, a comparative study of decoupling /coupling methods is conducted. Results shows that SG decoupling has a significant impact on the calculation result, the calculation method of coupling is more closer to the real situation and worthy to recommended. Related analytical procedures and calculation results lay the foundation for future SG shaking dynamic analysis and SG design of subsequent power plants.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"43 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":"125962487","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":"Probabilistic Analysis of Creep-Induced SGTR for NPP","authors":"Li Wenjing, Z. Wentao, Yu Xinli, Wei Wei","doi":"10.1115/ICONE26-81032","DOIUrl":"https://doi.org/10.1115/ICONE26-81032","url":null,"abstract":"Severe accident analysis is the technical basis of severe accident prevention and mitigation, and severe accident phenomenon analysis can provide important technical support for Level-2 Probabilistic Safety Analysis (L2 PSA) and severe accident management of NPP. This paper evaluates the probability of creep-induced steam generator tubes rupture (SGTR) which might lead to direct release of radioactive material to environment based on Risk-Oriented Accident Analysis Methodology (ROAAM). According to Larson-Miller creep failure model, the material parameters are provided first, then a two-loop NPP model is established with MAAP4 code to assess the temperature and pressure of the pipes for typical sequences, and eventually the probability of creep-induced SGTR phenomenon for typical sequences is calculated.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"34 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":"129744886","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":"Eulerian Two-Fluid Model for Aerosol Removal in Filtered Containment Venting Scrubbers","authors":"Ji-Su Kim, J. W. Park, Minkyun Kim","doi":"10.1115/ICONE26-81506","DOIUrl":"https://doi.org/10.1115/ICONE26-81506","url":null,"abstract":"Eulerian steady one-dimensional two-fluid mass and momentum conservation equations for gas and water droplet coupled with the particle removal rate are formulated. The major flow variables are phasic velocities and pressure and they are obtained by integrating the four conservation equations and one particle removal rate along the horizontal throats of the venturi scrubbers by using the forth-order Runge-Kutta method. Popular correlations are used for the impaction parameter, the gas-to-water droplet drag coefficient and the droplet size. All the equations are integrated into one computer code named REMAERO based on the MATLAB. It is found that most particles are removed at the throat entrance as already known from the previous experiments and this is due to the particle removal rate proportional to the particle concentration. Particle removal efficiencies from the present method shows better agreement with existing experimental data than previous one-dimensional models.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"73 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":"114622316","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":"Residual Stress Measurement of Sealing Glass Based on Optical Fiber Sensing Technology","authors":"L. Mingze, Fan Zhichun, Diao Xingzhong, Yan He","doi":"10.1115/ICONE26-82135","DOIUrl":"https://doi.org/10.1115/ICONE26-82135","url":null,"abstract":"Metal-to-glass electrical penetration assemblies (EPA) are highly sophisticated equipment and have been used for electrical connection in containment structures or pressure vessels in nuclear plants because of their high temperature resistance and good hermeticity. One important factor to keep hermeticity and reliability can be attributed to the initial residual stress in sealing glass of metal-to-glass EPA. If the residual stress is too high, small defects easily take place in the sealing materials. An insufficient prestress also cannot meet the requirement of high pressure application. To study the influence of residual stress on hermeticity, we developed a novel method of residual stress measurement in metal-glass sealing based on an embedded optical fiber sensor. The fiber Bragg grating (FBG) sensor was embedded in the glass material during the EPA manufacturing, and the residual stress along the grating could be retrieved via optical fiber sensing technique. Basing on our existing metal-glass sealing technique, the initial residual stress could be modulated by changing the sealing process, then the change of different initial residual stress was measured by the embedded FBG, through which the impact of residual stress on metal-glass sealing hermeticity could be finally revealed. A finite element model was established basing on linear elastic theory, then the localized stress along the FBG and the global stress distribution had been investigated theoretically. Taking the stress measuring by FBG as a breakthrough point, the effect of initial residual stress on sealing hermeticity was studied experimentally. The results showed that the residual stress should be an important assessment indicator to metal-to-glass sealing. This research also provided a new approach to optimize EPA manufacture.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"29 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":"132477281","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":"Discussion of Issues for Small Modular Reactor PSA","authors":"Liu Tao","doi":"10.1115/ICONE26-82157","DOIUrl":"https://doi.org/10.1115/ICONE26-82157","url":null,"abstract":"Small module reactors (SMRs) are in ascendant in the world due to their shorter design & construction duration and higher safety level. Probabilistic safety assessment (PSA) will play an irreplaceable role in small module reactor design demonstration and other application stages. A mature PSA methodology framework is mature and has successfully used in commercial nuclear power plant application. While several issues are worthy of discussion when PSA performed for SMRs project. Based on traditional PSA framework and its fundamental elements, potential technical issues put forward for discussion for SMRs project PSA, which will provide references for the combination of PSA and SMRs project.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"176 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":"132496078","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":"Structural Dynamic Transient Analysis of Fire Protection System at a Nuclear Power Plant","authors":"M. Dong, E. Tom","doi":"10.1115/ICONE26-82627","DOIUrl":"https://doi.org/10.1115/ICONE26-82627","url":null,"abstract":"NRC Information Notice 98-31[1] describes a water hammer event that occurred at Columbia Generating Station. In this event, actuation of a Fire Protection System (FPS) led to voiding in a tall dead-ended riser with subsequent water hammer in the riser when the main fire pumps were activated. Due to the severity of the event, NRC requires that FPS with similar configurations at other nuclear power plants be analyzed for similar events. This paper describes an evaluation of two selected riser piping systems performed to quantify the susceptibility of the potential water hammer events in the FPS at a nuclear power plant. The FPS was modeled in a proprietary computer program, USLAM (Hydraulic Analysis Code), and selected operational scenarios (analysis cases) were analyzed to quantify the potential waterhammer loads (fluid forces). A tall riser piping system in the Reactor Building was chosen as a sample system for structural dynamic analysis. Based on the results of the sample system, it was concluded that the fire protection piping systems could experience piping stress far exceed the faulted allowable loads as allowed by the ASME/ANSI piping code. A cost-effective mitigation scheme was then proposed plan in paper ICONE65-82622[2] as Part 1 of this study with consideration only hydraulic loading where 11 vacuum breakers are to be installed at various locations of the FPS. The structural analyses discussed only piping stress in this paper and demonstrate for the effectiveness of the proposed mitigation scheme from the revised hydrodynamic loads. As the continuation of Ref. 2, the results from this study validate the acceptance of these two piping systems.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"8 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":"121787794","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":"The Compatibility of Nuclear Graphite With Molten Salt in the Molten Salt Reactor","authors":"Zhoutong He, Hui Tang, Can Zhang, Yantao Gao, H. Xia, Xingtai Zhou","doi":"10.1115/ICONE26-82065","DOIUrl":"https://doi.org/10.1115/ICONE26-82065","url":null,"abstract":"In thermal Molten Salt Reactors, the nuclear graphite core is in direct contact with the molten salt coolant. Due to the porous nature of nuclear graphite, the molten salt may infiltrate the nuclear graphite, which may affect the mechanical strength and irradiation behavior of the nuclear graphite. In order to evaluate the infiltration behavior of molten salt in nuclear graphite, both FLiNaK and FLiBe salts were used to infiltrate two typical nuclear graphite grades: IG110 and NBG18. The pressure dependence of the infiltration weight gain ratio was measured. The influence of molten salt infiltration on the thermal properties of these two graphite grades, such as their thermal expansion behavior and thermal conductivity, was also measured. The mechanical strength of the FLiNaK-infiltrated graphite was measured at room temperature and elevated temperature, and showed that the mechanical strength of the nuclear graphite was enhanced at room temperature and weakened at elevated temperature by molten salt infiltration. Finally, the thermal expansion coefficient and the fracture surface analysis measured after FLiNaK infiltration indicated that the stress induced by molten salt infiltration could be one of the reasons for the graphite property changes.","PeriodicalId":237355,"journal":{"name":"Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management","volume":"234 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":"114543224","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 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}

{"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}