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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015

作者： Xian, Lin Jiang, Guangming Yu, Hongxing Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9781510811843

Natural convective heat transfer in enclosures is widely researched in extensive range of engineering application such as passive residual heat removal system (PRHRS) in nuclear plant, solar collectors and cooling of electronic equipment. After Fukushima incident, the application and researching of PRHRS have been gaining more and more attention. It still has no proper calculating tools to analyze natural convective heat transfer with large L/D ratio and high Rayleigh number in PRHRS. In this paper, based on some former natural convective heat transfer experiments, numerical simulations and scale analysis, the single phase natural convective heat transfer correlation from the vertical tube have been developed. Transient single phase natural convection heat transfer along vertical tube in a cylindrical enclosure has been investigated for water with widely used computational fluid dynamic (CFD) engineering program ANSYS CFX-12. The influence brought by different geometry of tube and mesh scales has been evaluated by the results of former experiments. The ratio of tube length to diameter is a dimensionless factor affects heat transfer rate. The evaluation of thermal stratification to natural convective heat transfer in enclosure has been studied. Numerical simulation covers large L/D and high Rayleigh number natural convection: 108RaL14. Based on the scale analysis of natural convection cylindrical boundary equations the new dimensionless group has been put forward contains L/D factor. The new natural convective heat transfer along the vertical single tube model has been build according to the former experiments1 results. After combining numerical simulation results, the new model covers the higher Rayleigh number range of natural convection.

关键词： Natural convection

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Large eddy simulation of turbulent flow in the rod bundle with different spacer grids 23

Large eddy simulation of turbulent flow in the rod bundle wi...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Wei, Zonglan Yu, Zhang Liu, Songtao Science and Technology on Reactor System Design Technology Laboratory NPIC Chengdu Sichuan China Nuclear Power Design and Research Sub-Institute NPIC Chengdu Sichuan China

ISBN: (纸本)9784888982566

Large eddy simulation(LES) of the swirling flow in the rod bundle subchannels with spacer grids are presented. According to the rod bundle flow benchmark experiment, the 5×5 rod bundle with a split-type spacer grid is simulated by this transient simulation with a fine hybrid mesh. Mean velocity components, RMS (root mean square) velocity components and turbulent kinetic energy profiles at different locations downstream the mixing vanes are compared between the experiment and simulation results. LES results can reproduce the mixing turbulent flow velocity field well. We also have studied the influence of the mixing vanes' bending degree on the subchannel flow structures by using LES to simulate the 3×3 rod bundle with the typical PWR spacer grid. The mixing vanes' bending degree is modified to 35°and 40°based on the original spacer grid, whose mixing vanes' bending degree is 30°. The larger bending degree produces a stronger mean coolant mixing effect, while larger pressure loss and fluctuation come together with this effect. The increased pressure loss would generate additional flow resistance, and larger fluctuation of coolant flow would make the mixing effect unstable, as well as would cause flow-induced vibration and grid-to-rod fretting phenomena. The LES method is a viable tool to optimize the spacer grid design and provide transient flow field data for mechanical analysis. Copyright © 2015 by JSME.

关键词： Mixing

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CFD simulation of single-phase flow in a 5×5 rod bundle with spacer grid 23

CFD simulation of single-phase flow in a 5×5 rod bundle wit...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Li, Quan Jiao, Yongjun Chen, Jie Yu, Junchong Department of Engineering Physics Tinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Spacer grid is an important component in PWR fuel assemblies for its significant influence on thermal-hydraulic characteristics of the reactor core. In this study, single-phase CFD technology is used to study the flow and resistance in a 5×5 rod bundle with spacer grid. CFX 12 is chosen as the tool software. The geometries of spacer grid and rod bundle are finished by UG code. To simulate the contact of springs and dimples with fuel rods, small gaskets are induced. The SST model is chosen for turbulent simulation. The results of resistance coefficients including both local loss coefficients and frictional loss coefficients under different Re numbers are conformed well to the experiment. The agreement of calculated liquid velocity distribution downstream the spacer grid with the experiment is also well but less satisfactory. We draw attention to the influence of mixing vanes and find that mixing vanes are the main components that induce the lateral flow, but take up little percents in the local loss coefficient of spacer grid. The lateral flow caused by mixing vanes will increase the frictional loss coefficient downstream the spacer grid. These results are beneficial to the optimization design of spacer grid. Copyright © 2015 by JSME.

关键词： Turbulence models

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Application of STPA to the digital reactor protection system in NPP for system safety analysis 23

Application of STPA to the digital reactor protection system...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Liu, Zhaohui Wu, Zhiqiang Yang, Xiaohua School of Computer Science and Technology University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9784888982566

In NPP, the digital control system which integrated software and hardware are increasingly used to improve dependability and introduce new functionality. Traditional safety analysis can get a good result when handling accidents caused by component failures, but software does not fail in this way. STPA is a new hazard analysis technique based on systems theory rather than reliability theory. It considers the system as a whole (include the hardware and software) to analyze failure and causality of systems and treats safety as a control problem rather than a failure problem. Being a safety-critical system, RPS in NPP needs to be considered carefully in system safety. So, we adopt this new approach to analyze the design process. From the analysis results, we found that causal factors leading to safety accidents identified by STPA included all the hazards identified by the fault tree analysis. Furthermore, there are some causal factors that were identified by STPA only. We utilize these results of the analysis on causation factor to refine the safety requirements and reduce the occurrences of the hazardous scenarios. Copyright © 2015 by JSME.

关键词： Fault tree analysis

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The Software safety analysis based on sfta for reactor power regulating system in nuclear power plant 23

The Software safety analysis based on sfta for reactor power...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Liu, Zhaohui Liao, Longtao Wu, Zhiqiang Yang, Xiaohua School of Computer Science and Technology University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9784888982566

The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide many advantages. However, digital control systems induce new failure modes that differ from those of analog control systems. While the cost effectiveness and flexibility of software is widely recognized, it is very difficult to achieve and prove high levels of dependability and safety assurance for the functions performed by process control software, due to the very flexibility and potential complexity of the software itself. Software safety analysis (SSA) was one way to improve the software safety by identify the system hazards caused by software failure. This paper describes the application of a software fault tree analysis (SFTA) at the software design phase. At first, we evaluate all the software modules of the reactor power regulating system in nuclear power plant and identify various hazards. The SFTA was applied to some critical modules selected from the previous step. At last, we get some new hazards that had not been identified in the prior processes of the document evaluation which were helpful for our design. Copyright © 2015 by JSME.

关键词： Nuclear power plants

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Research of the bundle CHF prediction based on the minimum DNBR point and the BO point methods 16

Research of the bundle CHF prediction based on the minimum D...

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16th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2015

作者： Liu, Wei Shan, Jianqiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China School of Nuclear Science and Technology Xi'an Jiao Tong University Xi'an710049 China

ISBN: (纸本)9781510811843

There are two ways in the development of CHF correlation in rod bundles: one is based on round tube CHF correlations, considering the effects of spacers and CHF promoters, effects of cold walls, etc. The other one is to develop the correlation directly using the rod bundles CHF experimental data which includes two methods: me minimum DNBR point method and the BO point method. In this study, the local thermal-hydraulic parameters at DNB occurrence point of each set of bundle CHF data were obtained by using the subchannel code ATHAS. Both of the minimum DNBR point and the BO point methods were then applied to develop a new CHF correlation for rod bundles. The so-called "three-step method" was used to determine the shape of me correlation and the non-linear regression analysis method was applied to determine the coefficients of the correlation. Based on the large database of CHF tests, a new CHF correlation named the ACC correlation has been developed. The analysis and assessment results indicate that me ACC correlation can fit the experimental data well with high prediction accuracy and correct parametric trends. Compared with the minimum DNBR point method, me BO point method is recommended to predict me risk of DNB because it is more reasonable and conservative with better predication rate. Coupled with subchannel code ATHAS, this correlation can simulate the thermalhydraulics performances of rod bundles exactly.

关键词： Forecasting

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Thermal hydraulic analysis on the success criteria of steam generator tube rupture accident 23

Thermal hydraulic analysis on the success criteria of steam ...

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23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015

作者： Ran, Fu Dan, Wu Ding, Shuhua Qian, Peng Libo, Qian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888982566

The SGTR studies correspond to the double-ended guillotine rupture of a steam generator tube, which allows unimped blowdown from both ends of the severed tube. Occurrence of the SGTR a ccident leads to an inc rease in contamination of t he secondary system as well as activ ity release to atmosphe re due to leakage of radioactive coolant from the RCS. The refore the au tomatic or operator actions should be taken in tim e to terminate the leakage. For ACP1000, the ti me window of the operator action is prolonged by adding new protection signals and equipment. In present work, the maximal time interval of th e operator action after t he SGTR is studied with these new m easures. And after the operator takes actions, in accordance with the procedure, no SG overflow takes place till the pressure of the primary side is equal to that of the secondary side. Copyright © 2015 by JSME.

关键词： Nuclear fuels

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Research on Post Nonlinear Geometric Algorithm for BSS of Mixed Vibration Simulation Signals from the Impeller Pumps

Research on Post Nonlinear Geometric Algorithm for BSS of Mi...

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The second International Symposium on Socially and Technically Symbiotic systems (STSS2015);The 7th International Symposium on Symbiotic Nuclear Power systems (ISSNP2015)

作者： Sun Shi-yan XIA Hong CHEN Zhi-hui YANG Bo LIU Mei-ru Fundamental Science on Nuclear Safety and Simulation Technology Laboratory College of Nuclear Science and TechnologyHarbin Engineering University Science and Technology on Reactor System Design Technology Laboratory Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China

In nuclear power plants,for rotating equipment,a normal method for monitoring the modes and diagnosing its fault is monitoring and analyzing its vibration or noise ***,the detected signals are usually the non-linear mixed signals produced by multiple pieces of equipment or their different *** order to distinguish equipment's status,Blind Source Separation technique need to be applied to separate the mixed signals,then equipment's mode can be distinguished *** existing mature BSS algorithms are based on the assumption of the linear *** it to the nonlinear mixed problem directly may cause failure or result in totally wrong *** this paper,we apply geometric algorithm to separate the mixed vibration simulation *** method follows the post-nonlinear mixing model assumption,and develops its separating measure exploiting the difference between a linear mixture and a nonlinear one from their distributions in a three-dimensional *** impeller pump,which is a typical rotating machinery in nuclear power plants,is chosen as the study *** build a three-dimensional model of a kind of impeller pump,use Finite Element Analysis method to create some correlative stress and mode analysis,decorate appropriate monitoring points to get the most intense vibration simulation signal in the pump,and then construct the mixed *** using the post-nonlinear geometric algorithm,the mixed vibration simulation signals of the pump are *** with correlation coefficient,the algorithm performs *** can obtain the same vibration peak frequency from the result as that from the source signals.

关键词： Blind Source Separation post-nonlinear model geometric algorithm the impeller pump vibration simulation signal

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Neutron transport study based on assembly modular ray tracing MOC method

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Hedongli Gongcheng/Nuclear Power Engineering 2015年第3期36卷 157-161页

作者： Tian, Chao Zheng, Youqi Li, Yunzhao Li, Shuo Chai, Xiaoming School of Nuclear Science and Technology Xi'an Jiaotong University Xi'an710049 China State Nuclear Power Software Development Center Beijing102200 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

It is difficulty for the MOC method based on Cell Modular Ray Tracing to deal with the irregular geometry such as the water gap between the PWR lattices. Hence, the neutron transport code NECP-Medlar based on Assembly Modular Ray Tracing is developed. CMFD method is used to accelerate the transport calculation. The numerical results of the 2D C5G7 benchmark and typical PWR lattice prove that NECP-Medlar has an excellent performance in terms of accuracy and efficiency. Besides, NECP-Medlar can describe clearly the flux distribution of the lattice with water gap. ©, 2015, Yuan Zi Neng Chuban She. All right reserved.

关键词： Ray tracing

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Analysis of RCS depressurization effects on containment temperature for Qinshan phase II nuclear power plant 22

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2014 22nd International Conference on Nuclear Engineering, ICONE 2014

作者： Lili, Liu Ming, Zhang Jian, Deng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Chengdu China

ISBN: (纸本)9780791845936

A severe accident code was applied for modeling of a typical pressurized water reactor (PWR) nuclear power plant, and the effects of RCS depressurization on the gas temperature of the relief tank cell in the containment during a station blackout (SBO) induced accident was analyzed. The sensitivity calculation indicated that the hydrogen generation rate obviously increased due to RCS depressurization in a critical stage. The results show that RCS depressurization can play an important role in hydrogen generation rate and total accumulation, and the temperature of the containment atmosphere is highly influenced by hydrogen combustion. High temperature induced by hydrogen combustion may degrade the equipment and instruments capabilities. Based on this analysis, a feasible strategy of RCS depressurization for mitigating the accident consequence is provided for developing the capacity of the SBO treatment of Qinshan Phase II Nuclear Power Plant (QSP-II NPP). Copyright © 2014 by ASME.

关键词： Nuclear power plants

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