The result of a wall failure in-pile experiment under the EAGLE project

The WF (wall failure) test of the EAGLE program, in which 2 kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR (Impulse Graphite Reactor) of NNC/Kazakhstan. In this test, a 3 mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10 mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 s after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events. A preliminary interpretation on the WF test results is presented in this paper.     

Analysis of low load transients for the Vandellòs-II NPP: Application to operation and control support

The Technical University of Catalonia (UPC) has been jointly working with the Asociación Nuclear Ascó-Vandellòs (ANAV) for a number of years in order to establish, qualify and use best estimate (BE) models for the reactors under the control of ANAV. ANAV is the consortium that is responsible for operation of the Ascó and Vandellòs-II reactors. The reactors are Westinghouse-design three-loop PWRs with an approximate electrical power of 1000 MW. The existing integral plant models for each plant are currently used for many different purposes among which are support of plant operation and control. Quite a number of studies have been done in order to improve both safety and plant competitiveness. Most of these dynamic analyses were carried out in relation to transients starting at nominal full power or at least, very close to full power. This paper develops a specific use of the Vandellòs-II plant model for operation and control support at low power involving new ranges of system actuation parameters. It also examines scenarios that are somewhat different from those typically analysed. The study starts showing the results of an assessment case, which is a start-up test and provides some additional qualification, and subsequently attempts to establish calculations to support both an improvement in feed water controls and to set up operating recommendations for low-load manual operation of feed water turbo-pump. Both results hopefully, will produce an outcome, which leads to an improvement in safety and reduces reactor trip probability.     

我国核裂变能可持续发展战略研究

本文从科学技术的角度,分析我国国民经济增长对核能发展的需求,探讨我国核能的近中期发展战略构想,指出为实现其发展战略目标而应重点研究的关键科学技术与问题,并提出一些相关的政策建议,为国家制定核能中长期发展规划提供参考。     

中空纤维酶膜反应器及其动力学初步探讨──Ⅰ．单酶系统中空纤维酶膜反应器

单酶系统中空纤维酶膜反应器是用天津纺织工学院研制的ＴＦＣ－００３型聚砜中空纤维超滤器固定β－淀粉酶，在一定条件下催化可溶性淀粉制取麦芽糖的反应器。试验中，对温度、ｐＨ、酶及底物浓度变化动力学参数进行了初步探讨，试验重复性好。     

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。     

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.