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1.
Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.  相似文献   

2.
我国的快堆技术发展和实验快堆   总被引:5,自引:1,他引:4  
徐銤 《核动力工程》2000,21(1):34-38
随着我国核电技术的发展,自主研制钠冷快中子增殖堆十分必要。本文介绍了我国在研究开发快堆技术方面的历史和实验快堆的设计原则、设计简介和安全特性。  相似文献   

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Presently discernible limitations to the extent of low-cost uranium resources provide an obvious incentive for seeking better utilization of fissile material, which may well become dominating by the end of the century. The high degree of world interest in breeder reactors, which could produce more fissile material than they consume, is thus very understandable, for the substantial conversion of fertile into fissile material thereby offered would greatly increase the effective energy yield of the mined fuel. However, the large-scale use of fast breeders which would be necessary for substantial impact on resource conservation would also require that these reactors be entirely acceptable also in other ways. Particularly they would have to compete with other routes to nuclear power in areas of capital cost, maintainability, and siting flexibility. Such considerations have stimulated the evolution of the gas-cooled fast breeder and the contemplation of combinations of fast breeders and advanced high temperature converter reactors aimed at taking best advantage of the special merits of each type. Predominant influences here are the enhanced importance of breeding ratio to the effectiveness of such a combination, the special worth of U233 as a thermal reactor fuel and the value of the high temperature capability, both directly and as a factor broadening the options available to power plant design.  相似文献   

6.
铅冷快堆固有安全性的分析   总被引:2,自引:0,他引:2  
为了研究铅冷快堆的固有安全性。本文完成了25MW铅冷快堆物理和热工水力初步设计,并进行了铅的充排放实验和铅的自封性实验。在此基础上,依据核反应堆固有安全性的理论,详细地分析和比较了铅冷快堆所具有的固有安全性。分析结果表明,铅冷快堆是一种很有发展的先进核动力堆堆型。  相似文献   

7.
The Prototype Fast Breeder Reactor (PFBR) which is under construction at Kalpakkam, India, is a 500 MWe sodium cooled pool type reactor. The core of the PFBR consists of 1758 free standing subassemblies supported on the grid plate. The entire core is divided into 15 different flow zones and the flow rate required through each zone is calculated based on the fission heat generation. The coolant sodium flows from the bottom of the subassembly to top and the design of the subassembly for each flow zone is quite complex. There are 181 fuel subassemblies in PFBR core with 217 fuel pins in each subassembly, vertically held in the form of bundle within a hexagonal wrapper tube. The pins are separated by spacer wires wound around the pins helically. Analytical prediction of subassembly pressure drop, vibration and determination of inception of cavitation for this complex geometry is very difficult. So experiments were conducted extensively to get a more accurate evaluation of the design and for its qualification for the use in PFBR, which is designed for 40 years of operation.Pressure drop and cavitation experiments were carried out in water on full scale (1:1) subassemblies of all flow zones. The overall pressure drop of the subassembly determines the ratings of the pump. Cavitation of the pressure drop devices lead to erosion damage of fuelpins and may also result in reactivity fluctuation due to sodium-void effect. So it is essential to confirm that the subassembly is not cavitating in the operating regime of the reactor. Subassembly can vibrate in cantilever mode due to the turbulence in the flow and can result in reactivity fluctuation, reactor control problem and can even lead to the failure of the fuel pins. So vibration measurements were carried out in water on the maximum rated subassembly. This paper discusses various experiments carried out on PFBR subassembly, the similarity criteria followed, instrumentation, results and conclusion.  相似文献   

8.
Briefly reviewed were recent R&D activities and achievements in Japan in the area of neutronics, core design and shielding studies on fast breeder reactor.  相似文献   

9.
This paper reports some irradiation effects and recovery behavior of neutron irradiated boron carbide pellets that were used as control rod elements in the Enrico Fermi Fast Breeder Reactor. Measurements were carried out on changes in lattice parameters, thermal expansion, helium release, elastic moduli and microstructure observations by annealing the irradiated pellets at elevated temperatures. The increase in unit cell volume of B4C upon irradiation was found to be 0.22%. The recovery in lattice parameter began at around 500°C and completed at 1,000°C. It was found that the pellet showed a sharp increase in a dimensional change at about 700 to 800°C with a large amount of helium release, and the pellet which showed larger swelling released smaller amount of helium.  相似文献   

10.
池式快堆主容器地震响应分析   总被引:2,自引:1,他引:1  
翁智远  钱江  徐礼存 《核动力工程》2000,21(4):328-331,338
试图对池式快堆结构作较大简化,使计算简图既能反应堆结构的动力特性,又能使计算简便可行。从而把一个复杂的结构用一个简单的弹簧-质量体系来近似地等效代替。引入容器的变形假设和确定的液动压力假设,应用虚位移原理可获得体系在水平地震作用下的运动方程,而后考察其地震动响应。  相似文献   

11.
为了估计和预测钠火事故的后果,构建了以“有火焰薄层”为理论基础的燃烧模型和热传输模型,给出了程序计算结果与试验值的比较。比较结果证实,该计算结果可信、模型合理。程序可用来分析和预测钠池火事故。  相似文献   

12.
钠冷快中子增殖堆(钠冷快堆)是一种最成熟、最具商业化前途的快堆堆型。但由于其材料、冷却剂安全性及经济竞争力等方面的原因,国内仍处于实验堆运行阶段。由于缺乏钠冷快堆安全监管方面的法规、标准、技术及经验,监管工作面临巨大挑战。本文将钠冷快堆与压水堆进行比较,并将钠冷快堆的特点与监管工作的特点相结合,从堆芯、系统及设备等方面提出15个监管重要关注点,并给出一系列相关建议。  相似文献   

13.
建立改进型快谱超临界水冷堆(SCFR-M)堆芯模型,探讨点火区燃料棒直径和增殖区水棒直径对堆芯转换比的影响,得到合理的燃料组件设计形式。设计并计算6种不同堆芯布置的反应堆增殖特性和空泡反应性,并分析燃料中235U和239Pu成分对堆芯转换比和空泡系数的影响,提高了转换比;研究燃料成分对堆芯转换比的影响。结果表明:减小氢原子数与重金属原子数之比(H/HM),增加堆芯增殖燃料组件数目并采用合理布置可满足堆芯负空泡反应系数,且可以提高堆芯转换比;降低燃料中Pu同位素质量分数可以使堆芯转换比大幅增加,同时使堆芯的空泡反应性系数负值更大;当点火燃料组件采用Pu同位素质量分数为20.8%的MOX燃料,增殖燃料组件采用0.2%富集度235U的贫铀燃料,6号设计方案可以使堆芯的初始转换比达到1.03128,且空泡反应性系数为负,初步达到超临界水冷快堆的增殖要求。进一步对堆芯的缓发中子有效份额、能谱、中子注量率、功率分布进行计算,分析研究增殖堆芯的物理特性。  相似文献   

14.
Experimental and Computational Fluid Dynamics (CFD) investigations have been carried out on a 1/5th scale model of the inlet plenum of steam generator (SG) used in the Fast Breeder Reactor (FBR) technology. The distribution of liquid sodium in the inlet plenum of the steam generator strongly affects the thermal as well as mechanical performance of the steam generator. In the present work, flow distribution in a scaled down model has been investigated. Various strategies adopted for obtaining uniform flow distribution have been evaluated. Experiments have been conducted to measure the axial and radial velocity distributions using Ultrasonic Velocity Profiler (UVP) under a variety of geometries. Computational Fluid Dynamics (CFD) studies have been carried out for various geometries. On the basis of these experiments and CFD simulations, various flow distribution devices have been compared.  相似文献   

15.
This work explores the use of Real-parameter Genetic Algorithm and analyses its performance in the steam condenser (or Circulating Water System) optimization study of a 500 MW fast breeder nuclear reactor. Choice of optimum design parameters for condenser for a power plant from among a large number of technically viable combination is a complex task. This is primarily due to the conflicting nature of the economic implications of the different system parameters for maximizing the capitalized profit. In order to find the optimum design parameters a Real-parameter Genetic Algorithm model is developed and applied. The results obtained are validated with the reference study results.  相似文献   

16.
浸没在液态钠中的快堆堆芯组件在地震作用下发生振动,可能导致组件结构损坏或堆芯结构变形,从而影响反应堆结构完整性和安全.流体使该振动表现为强烈的非线性,因此,研究地震引起的流固耦合效应对快堆抗震分析十分重要.本文主要研究流固耦合问题中附加质量的计算方法,该方法由Westergaard首先提出,是一种考虑水体对结构作用的简化动力学计算方法,它将动水压力等效成质量附加在结构上,质量等效原则自提出在各行业得到广泛应用,但缺乏详细理论推导.本文首先推导出附加质量公式,并对该公式进行有效性分析;接着对单根和两根组件用CASTEM在空气和水中进行建模;最后将频率、碰撞力分别与试验值比较.结果表明,计算值和试验值吻合.  相似文献   

17.
The radial propagation of sodium boiling by thermal process in a 1,500 MWe FBR is analyzed with the BOIP-T code. Since the model used embodies some uncertainty, several modes of molten fuel behavior are considered. The time required for boiling to propagate from one channel to its neighbor is calculated. Even on the most conservative basis, the boiling propagation requires at least 10 sec, which indicates that the radial propagation of sodium boiling could be prevented by a suitable core protection system.  相似文献   

18.
This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55°C; piping surface, 80°C) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB).  相似文献   

19.
The prediction accuracies of key neutronic characteristics including burnup properties evaluated with use of the sensitivity-based methodology have been reviewed for a fast breeder reactor. The bias factor method, the cross section adjustment method and the combined method are used to evaluate the prediction accuracies. The calculation method of sensitivity coefficients used in the uncertainty analysis is discussed. The three methods are compared from the theoretical and numerical points. For the numerical comparison, they are applied to a 1,000 MWe fast breeder reactor. The prediction uncertainties are within the range of 0.7~1.0% for keff , 3~5% for control rod worth, 1~2% for 239Pu fission rate distribution, 12% for burnup reactivity loss and 1.5% for breeding ratio. These values are much smaller than those predicted without any integral data.  相似文献   

20.
The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large storage capacity because the number of waste packages produced is significantly reduced by a factor of 5 from that of the glass waste package in the FBR fuel cycle without PT.  相似文献   

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