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The study on overall I&C system scheme for floating nuclear power plant ACP100S and ACP25S 10

The study on overall I&C system scheme for floating nuclear ...

10th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies, NPIC and HMIT 2017

作者： Chen, Zhi Long-Tao, Liao Dan-Rong, Song You, Kai Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9781510851160

Floating nuclear power plant (FNPP) is built on the offshore platform and it is defined as a kind of Small and Modular reactor(SMR) due to its power level. The FNPP can provide power supply, fresh water and high temperature steam and other products. So it can be used as special energy sources, such as regional power supply, district heating and offshore oil exploitation. At present, two types of FNPP, named ACP100S and ACP25S, are developed by China National Nuclear Corporation (CNNC). In this paper, the overall technical parameters and overall technical scheme of ACP100S and ACP25S are introduced firstly. On the basis of these schemes, the different arrangements of nuclear power plant which are used for meeting the different energy demands are discussed. Due to the different configuration of nuclear power plant, the requirements for the Instrument and Control (I&C) system are different, so the paper discusses some overall configurations of I&C system under the several typical arrangements models, i.e., the single reactor and the single turbine, the double reactors and double turbines, and the double reactors and the single turbine. Furthermore, the characteristics of these overall I&C configurations are analyzed. © 2017 Westinghouse Electric Company LLC. All Rights Reserved.

关键词： Nuclear power plants

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Elastic and plastic analysis of O-ring seal for reactor pressure vessel 13th

Elastic and plastic analysis of O-ring seal for reactor pres...

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13th Asia-Pacific Symposium on Engineering Plasticity and its Applications, AEPA 2016

作者： Shao, Xuejiao Zheng, Liangang Zhang, Liping Fu, Xiaolong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9783035710243

The elasto-plastic analysis for reactor Pressure Vessel seal is necessary because of nuclear safety is actively demanded. The existed work based on simplified way to simulate the seal ring, such as uniform stress, spring elements, is too conservative. Therefore, it is necessary to simulate seal ring using 3D solid model considering the elasto-plastc deformation. In this paper, the two-dimensional model is adopted to simulate the compressing and springback character of inconel 718 O-ring using nonlinear isotropic hardening constitutive model and contact algorithm. The simulation approach and constitutive model are validated according to the deformation of seal ring using ANSYS. The seal analysis for reactor pressure vessel is achieved by considering complex loads, such as bolt pre-tightening force and temperature/pressure transient with the usage of an elasto-plastic constitutive model. It is found that the deformation of seal ring obtained by finite element is close to true value. The application of 3D solid model can reduce excessive conservatism effectively and improve the precision of seal analysis compare to the simplified method. © 2017 Trans Tech Publications, Switzerland.

关键词： O rings

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The CFD calculation of a stratification break-up test of light gas in a closed vessel

The CFD calculation of a stratification break-up test of lig...

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2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017

作者： Yazhe, Lu Yu, Zhang Xu, Ran Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9784890471676

During the course of a hypothetical severe accident in a light water reactor, spray systems could be used in the containment in order to prevent overpressure in case of a steam line break and to enhance the gas mixing in case of the presence of hydrogen. Thus these two effects induced by spray might compete with each other having an influence on the hydrogen risk. Focusing on the dynamic effect of spray, the impact of spray on the breakup of a light gas layer in a closed vessel is simulated and the experimental test evaluated here is TOSQAN 113 test that belongs to the dynamic part of SARNET FP6 benchmarks. ANSYS FLUENT 14.5 is chosen for the simulation tool and spray droplets are tracked with using the Euler-Lagrange approach-Discrete Phase Model (DPM). The code validation and several sensitivity studies are done to sum up the basic method of numerical simulation of spray with CFD code and assess the influence of calculation settings and parameters, such as turbulence model and spray boundary condition. The code validation shows that the results of calculations acceptably agree with the experimental data and the sensitivity studies reveal that spray boundary condition is a rather important factor for the accuracy of simulation. Conclusion gained in from the present work can provide some reference of the numerical analysis of containment.

关键词： Boundary conditions

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SMR built-in pressurizer operation transients analysis

SMR built-in pressurizer operation transients analysis

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2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017

作者： Songwei, Li Yu, Liu Xi, Chen Zhongchun, Li Danrong, Song Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9784890471676

Three primary design principles, meanwhile the design features, are carried through the R&D process of Chinese Small Module reactor (ACP100): integral layout, compact design and modularized configuration. As a substantial demonstration of the above design conceptions, an upper plenum built-in pressurizer is proposed for the pressure control system design for Chinese SMR. In this design, branches with large diameters connected to the primary loop (i.e. surge line, etc.) are eliminated, and this will significantly reduce the LOCA possibility due to the pressure boundary breach. This study mainly focuses on the feature of responses of the built-in pressure control system during the system transients, such as step load increase/decrease, ramp load increase/decrease and partial loss of electrical load. Initial status of the plant, assumptions, postulated transient condition and methodology of this analysis are described in the first section. Important thermal hydraulic parameters that can picture the transient characteristics of the coolant system are given in the second section. It shows that, during those transients, no reactor safety related signals or safety guard features are triggered or actived. It also shows that Nuclear/thermal power response rapidly and stably, and average coolant temperature and system pressure are welled controlled within the range of normal deviation of the set point values of safety related system. This study results get a good agreement with the operating experiment study of the build-in pressure control system.

关键词： Transient analysis

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Computation of fluid thrust force from the pressure release tank of nuclear power plant pressurizer 17

Computation of fluid thrust force from the pressure release ...

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17th International Topical Meeting on Nuclear reactor Thermal Hydraulics, NURETH 2017

作者： Xiong, Fu-Rui Jiang, Nai-Bin Wu, Ge Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

Pressure release tank is a key component of nuclear power plant safeguard. Evaluating the fluid thrust force from high pressure and high temperature flow jet under postulated accidents is an essential task for nuclear system design and safety estimation. In this work, flow jet impingement from pressure release tank to nearby building roof is computed via a two-stage strategy. First, the RELAP5 code is applied to estimate thermal-hydraulic conditions at pressure release tank outlet. The thermal-hydraulic computational model includes key components of pressure release tank and its connection to the pressurizer. Based upon the thermal-hydraulic estimation within the pressure release tank, detailed computational fluid dynamics (CFD) simulation is performed with full considerations of how flow thrust is applied on nearby structures. The two-stage computational strategy provides a practical paradigm for fluid thrust force estimation. © 2016 Association for Computing Machinery Inc. All Rights Reserved.

关键词： Computational fluid dynamics

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A study on the sealing behaviors of the laterally one side restrained C-ring under internal pressure by numerical simulation 25

A study on the sealing behaviors of the laterally one side r...

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2017 25th International Conference on Nuclear Engineering, ICONE 2017

作者： Yuanyuan, Dong Ying, Luo Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9784888982566

C-ring, which is one of the main sealing elements of reactor Pressure Vessels, is composed of a close-wound helical spring surrounded by two metal linings. C-ring has better resilience and sealing characteristics. Just because of the complex structure of C-ring, it was studied by experiments almost. There were also some numerical simulations on C-ring. With these methods, the compression-decompression process of C-ring under ideal condition can be simulated. But them cannot be used to study the sealing behavior of laterally one side restrained C-ring under internal pressure, which is fit for practical condition. This paper will establish a 3D numerical simulation method to study the C-ring which is mentioned as above. In this paper, it can be obtained that with internal pressure forcing and sealing groove squeezing, sealing performance of C-ring will reduce;when p>20MPa, the ratio of reducing becomes bigger;when p>140MPa, it will generate leakage;inner lining and outer lining will generate biggish plastic deformation, while the stress distribution in spring changes much. © Copyright 2017 ASME.

关键词： Numerical methods

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Study of control rod device mechanism missile impact on protection plate

Study of control rod device mechanism missile impact on prot...

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2017 Transactions of the American Nuclear Society, ANS 2017

作者： Ye, Xianhui Xiong, Furui Zhen, Bin Jiang, Naibin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan610041 China

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An investigation of numerical performance enhancement of RELAP5: Numerical stability, higher resolution and an alternative constitutive relation

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Nuclear Engineering and design 2018年 328卷 309-320页

作者： Wei Li Xiaoli Wu Koroush Shirvan Guanghui Su Science Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610213 China Massachusetts Institute of Technology 77 Massachusetts Ave Cambridge MA 02139 USA School of Nuclear Science and Technology Xi’an Jiaotong University No. 28 Xianning West Road Xi’an Shanxi 710049 China

The nuclear reactor safety system code RELAP5 decomposes complex flow system of a nuclear reactor into a series of one-dimensional control volumes connected by flow junctions, and solves a set of two-phase two-fluid equations to predict the nuclear reactor system behavior. In spite of its extensive applications, there indeed exist many numerical shortcomings in RELAP5 and it is desirable to constantly improve its numerical performance. In the present work, the numerical performance improvement to RELAP5/MOD3 is carried out from the aspects of numerical stability, high-resolution and alternative constitutive relations. For the enhancement of numerical stability, the virtual mass term is replaced and an interfacial pressure term is added in the phase momentum equations of RELAP5 to make all the characteristic roots real, thus improve the model’s hyperbolicity. In addition, the second-order Minmod TVD flux-limiter scheme replaces the original first-order upwind scheme for advection terms to reduce the numerical diffusion. Furthermore, an alternative interphase friction relation is substituted for the built-in model in RELAP5 to calculate the interphase drag for bubbly/slug flow in the vertical bundle channels. The performance improvement measures work reasonably well, as indicated by the simulation of selected numerical examples and the Bethsy 6.2TC integral effect experiment which corresponds to an intermediate break Pressurized Water reactor Loss of Coolant Accident (PWR LOCA).

关键词： Virtual mass force Interfacial pressure Flux-limiter Interphase friction RELAP5

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Research on Hydrodynamic Interference of a DSRV and a Wrecked Submarine

Research on Hydrodynamic Interference of a DSRV and a Wrecke...

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IEEE International Conference on Underwater system technology: Theory and Applications (USYS)

作者： Hao TAN Duanfeng HAN Lei JU Haitao ZHU Yiran MIAO College of Shipbuilding Engineering Harbin Engineering University Harbin 150000 China Science and Technology on Reactor System Design Technology Laboratory Unclear Power Institute of China Chengdu 610041 China

ISBN: (纸本)9781538694060

In this paper, the viscous flow method based on Computational Fluid Dynamics (CFD) is used to simulate the hydrodynamic interaction of the Deep Submergence Rescue Vehicle (DSRV) involved in the rescue of a wrecked submarine. As the object of study, the drag of the DSRV moving in unbounded flow field is numerically calculated and the calculation results are compared with the experimental values to validate the accuracy of the numerical method. When the DSRV approaches the wrecked submarine in a positive flotation in quasi-static way, the effects of different inflow velocities and the distance between the two objects on the DSRV's drag, lift and pitching moment are simulated and monitored. The results show that when the DSRV is far away from the wrecked submarine, it is subject to the repulsive force from the submarine; when the distance is gradually reduced, the repulsive force is reduced until becoming suction; when the two objects are docking, the DSRV is subjected to the repulsive force from the submarine. This study will provide a reference for the operation of DSRV drivers.

关键词： Underwater vehicles Mathematical model Hydrodynamics Force Numerical models Load modeling Drag

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Preliminary study on the radiation heat removal capacity in SMR core under high temperature condition

Preliminary study on the radiation heat removal capacity in ...

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2017 International Congress on Advances in Nuclear Power Plants: A New Paradigm in Nuclear Power Safety, ICAPP 2017

作者： Li, Zhongchun Ding, Shuhua Li, Songwei Deng, Jian Nulcear Power Institute of China Chengdu610041 China Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China

ISBN: (纸本)9784890471676

The advanced fuel design, such as accident tolerant fuel (ATF) was design concept to withstand the high temperatures for extreme condition. When it was equipped in reactor core, the heat removal by radiation would be quite considerable. The radiation heat removal in a simplified SMR core was studied in the present study. The calculation was assumed to be 2D and at steady state. The calculation models were established from fuel assembly to fuel rod. The radiation heat flux was calculated based on the heat balance model when the limitation temperature in the core was reached. The results showed that the radiation heat flux contribute large proportion when the limitation temperature on fuel clad was high. The radiation heat flux was increased with fuel clad temperature limitation value. The higher the emissivity was, the more radiation heat flux removed.

关键词： Temperature

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