SIMULATION OF MICROMIXING IN A TURBULENT TUBULAR REACTOR WITH CONCENTRIC FEED

Turbulence on the axis of a pipe is isotropic and homogeneous and the rate of turbulent energy dissipation, ε, is known as a function of the velocity, pipe diameter and viscosity. When, however, a concentric feed pipe is introduced (as in a reactor), the ε field does not seem to be known. Evidence from fast, mixing-controlled reactions indicates a disturbance near to and downstream from the feed pipe with an increase in ε. Because of the lack of experimental information, the radial and axial distribution of ε has been modeled. One parameter—ε at the outlet of the feedpipe—is undetermined in this model. It is shown how the radial dispersion and especially the micromixing and reaction of fluid emerging isokinetically from the feed pipe can then be calculated, using a one-dimensional, radially averaged formulation. An application to four coupled, mixing-controlled reactions is included. Simulated and measured product distributions compared satisfactorily at various concentration levels and two solution viscosities. Whereas the new model is plausible, more work is needed to understand how a feed pipe modifies the flow in its vicinity.     

德国艾柯夫SL-500采煤机变频器故障及分析

晋城煤业集团引进德国艾柯夫SL-500采煤机,在使用过程中,短时间内连续出现多台变频器烧坏现象,这在艾柯夫采煤机使用过程中是从未有过的事,也给本公司的原煤生产造成了很大影响。为了杜绝此类事情的再次发生,特对本次的故障排除方法进行了认真的分析总结,为今后应用与修理采煤机提供了意见。     

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

1 Introduction With respect to the inherent safety of nuclear re- actors, application of passive systems/components including natural circulation phenomena as the main mechanism is intended to simplify the safety-related systems and to improve their reliability, to reduce the effect of human errors and equipment failures, and to provide more time to enable the operators to prevent or mitigate serious accidents. Natural circulation is the main mode of heat removal for removing decay heat from t…     

反应堆堆芯围筒结构热流固耦合热变形分析

针对反应堆堆芯围筒热流固耦合问题,采用三维有限元法研究堆芯围筒的热变形.考察ANSYS的三维实体热单元SOLID 70,三维实体单元SOLID 45,三维表面热效应单元SURF 152和三维热-流耦合管单元FLUID 116等单元类型的特点和实用性.建立堆芯围筒、吊篮和冷却剂的温度分析有限元模型：堆芯围筒和吊篮采用SOLID 70,结构表面与冷却剂的对流传热表面采用SURF152,堆芯围筒与吊篮之间冷却剂采用FLUID 116.采用SOLID 45建立堆芯围筒有限元模型,根据得到的堆芯围筒、吊篮和冷却剂的温度场结果分析堆芯围筒热变形.结果表明,在考虑堆芯围筒及吊篮固体和流体的交叉耦合的基础上,采用三维有限元法能比较客观地模拟反应堆堆芯处的复杂运行环境.     

海洋条件对一回路松脱部件监测系统的影响分析

为保障海洋核动力装置反应堆一回路的运行安全,需要对一回路冷却剂内易松脱部件实施有效的监测。松脱部件监测系统是监测反应堆一回路冷却剂零部件松脱事件的专有系统。为提高海洋运行环境下系统监测有效性,分析了因海洋环境特征及反应堆结构特殊性带来的松脱部件监测本底噪声复杂、探测灵敏度提高、前端仪表性能要求提高、声传播路径变化、电气设备维修更换、报警分析困难、海洋机械环境适应性待验证等多项设计闲难,并给出了解决思路和建议。该研究对海洋环境运行下压水堆一回路松脱部件系统的设计具有指导意义。     

配电网短路故障分析及其识别方法研究

分析了在不同中性点接地方式下单相接地和两相短路接地故障的相电压及中性点电压变化特征,结合三相短路和两相短路的故障特征,提出了完整的配电网短路故障识别方法。仿真结果表明,该方法具有采集的模拟量少、判据简单、运算简便的优点,具有一定的实用价值。     

Stability of Impulsive Stochastic Neutral Partial Differential Equations With Infinite Delays

In this paper, we study the existence and asymptotic stability in the pth moment of the mild solutions to impulsive stochastic neutral partial differential equations with infinite delays. Sufficient conditions ensuring the stability of the impulsive stochastic system are established. The results are obtained via the Banach fixed point theorem.     

Advantages of high-field tokamaks for fusion reactor development

High-field designs could reduce the cost and complexity of tokamak reactors. Moreover, the certainty of achieving required plasma performance could be increased. Strong Ohmic heating could eliminate or significantly decrease auxiliary heating power requirements and high values of n_E could be obtained in modest-size plasmas. Other potential advantages are reactor operation at modest values of , capability of higher power density and wall loading, and possibility of operation with advanced fuel mixtures. Present experimental results and basic scaling relations imply that the parameterB ²a, where B is the magnetic field and a is the minor radius, may be of special importance. A superhigh-field compact ignition experiment with very high values ofB ²a (e.g.,B ²a=150 T² m) has the potential of Ohmically heating to ignition. This short-pulse device would use inertially cooled copper plate magnets. Compact engineering test reactor and/or experimental hybrid reactor designs would use steady-state, water-cooled copper magnets and provide long-pulse operation. Design concepts are also described for demonstration/commercial reactors. These devices could use high-field superconducting magnets with 7–10 T at the plasma axis.     

The need and prospects for improved fusion reactors

Conceptual fusion reactor studies over the past 10–15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100–200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed.Nomenclature a Plasma minor radius at outboard equatorial plane (m) - A Plasma aspect ratioR _T /a - AC Annual charges ($/yr) - b Plasma minor radius in vertical direction (m) - B Magentic field at plasma or blanket (T) - B _c Magnetic field at the coil (T) - B Toroidal magnetic field (T) - B Poloidal magnetic field (T) - BOP Balance of plant - C Coil - COE Cost of electricity (mills/kWeh) - CRFPR Compact RFP reactor - CT Compact torus (FRC or spheromak) - c _FPC Unit cost of fusion power core ($/kg) - DC Direct cost ($) - DZP Dense Z-pinch - E Escalation rate (1/yr) - EDC Escalation during construction ($) - ET Elongated tokamak - F Annual fuel charges ($/yr) - FC Component of UDC not strongly dependent or FPC size ($/kWe) - FW First wall - FPC Fusion power core - f _Aux Fraction of gross electric power recirculated to BOP - f ₁ (IC+IDC+EDC)/DC - f ₂ (O&M + SCR + F)/AC - IC Indirect cost ($) - IDC Interest during construction ($) - I _w Neutron first-wall loading (MW/m²) - i Toroidal plasma current (MA) - j Plasma current density, I/a² - k _B Boltzmann constant, 1.602(10)^–16 (J/keV) - LWR Light-water (fission) reactor - MPD Mass power density 1000P_E/M_FPC (kWe/tonne) - M _N Blanket energy multiplication of 14.1-MeV neutron energy - M _FPC Mass of fusion power core (tonne) - n Plasma density (m^–3) or toroidal MHD mode number - O&M Annual operating and maintenance cost ($/yr) - p _f Plant availability factor - PFD Poloidal field dominated (CTs, RFP, DZP) - P Construction time (yr) - P_TH Thermal power (MWt) - P _E Net electric power (1-)P _ET (MWe) - P_ET Total gross electric power (MWe) - p_f Fusion power (MW) - q Tokamak safety factor (B /B _gq )(a/R _T ) - q _e EngineeringQ value, 1/e - R _T Major toroidal radius (m) - RFP Reversed-field pinch - RPE Reactor plant equipment (Account 22) - S Shield - SCR Annual spare component cost ($/yr) - SSR Second stability region for the tokamak - S/T/H Stellarator/torsatron/heliotron - ST Spherical tokamak or spherical torus - T Plasma temperature (keV) - TDC Total direct cost ($) - TOC Total overnight cost ($) - UDC Unit direct cost,TDC/10 ³ P _E ($/kWe) - V _p Plasma volume (m³) - W _p Plasma energy (GJ) - W _B Magnetic field energy (GJ) - Magnetic utilization efficiency, 2nk_BT/(B ²/2₀) - ₀ Permeability of free space, 4(10)^–7 H/m - XE Plasma confinement efficiency, a²/4_E - _e Plasma energy confinement time - _p Overall plant efficiency, _TH(1-) - _TH Thermal conversion efficiency - _FPC AverageFPC mass density (tonne/m³) - Plasma vertical elongation factor,b/a - Thickness of allFPC engineering structure surround plasma (m) - Total recirculating power fraction, (P _ET-P _E)/P _ET, or inverse aspect ratioa/R _T This work was performed under the auspices of USDOE, Office of Fusion Energy.     

CVD反应器传输过程的三维数学模型

提出了一个同时表示CVD过程的气体流动、温度分布和物质传输的三维数学模型。应用这个模型预报了在含有SiCl_4的氢气流中沉积出Si的锥台式反应器中的速度场、温度场和浓度场。所得的结果有助于增进对这类反应器中的传输过程的认识,模型亦可用于设计参数的最优化,诸如入口流量,锥台倾角等。