池式反应堆堆内流场数值模拟

以板型燃料组件池式反应堆为研究对象,采用计算流体动力学程序CFX5对堆内流场进行了数值模拟,结果表明:流过堆芯燃料组件的流速较大,在燃料组件出口位置流速达到最大值;无论是否带有围桶,堆内压降均主要集中在堆芯燃料组件上,入口流量增大,堆芯燃料组件上的压降随之增加;堆芯上部腔室和下部腔室的压力变化很小;在相同的入口流量下,带与不带围桶的堆芯进出口差压非常接近。     

板状燃料组件非线性振动特性

设计1个能模拟堆芯燃料组件工作环境和力学边界条件的台架,研究板状燃料组件的非线性振动行为。对安装在振动台上的系统进行白噪声随机激励,得到燃料组件响应-激励关系不同方向、不同介质、不同振级共42种工况的系列曲线。试验结果表明:随着激励振级加大,板状燃料组件的第一阶共振频率减小;板状燃料组件在水中的第一阶共振频率远比在空气中的小,且非线性特性仍呈软弹簧特性;板状燃料组件在环境中的振动阻尼大,能大大抑制抗震响应。     

Design of an intelligent fuzzy logic controller for a nuclear research reactor

The main objective of this paper is to design an intelligent controller system based on the concepts of fuzzy logic. This latter will be used to control the power of a nuclear reactor. The principle of this controller is based on rules established from experiments used with a classical controller and from the knowledge and the expertise of the operators of the reactor. This intelligent controller could be used in parallel with the actual system, which is semiautomatic, as a decision aided system to assist the operators in the control room.     

Safety aspects of the combined HTTR/steam reforming complex for nuclear hyrogen production

The High-Temperature Engineering Test Reactor (HTTR) in Oarai, Japan, has the potential to demonstrate the production of hydrogen by steam reforming and using nuclear process heat as primary energy input. Particular safety aspects for such a combined nuclear/chemical complex have been investigated such as fire and explosion hazard at presence of flammable gases (LNG, H₂, CO) near the reactor building. A methane vapor cloud in the open atmosphere or partially obstructed areas is highly unlikely to detonate and damage the reactor building. Theoretical assessments and experimental studies significant to the HTTR-steam reforming system, include the spreading and combustion behavior of cryogens and flammable gases providing the basis for a comprehensive safety analysis of the nuclear/chemical facility.     

Study on Pb-Bi natural circulation phenomena

For the development of 45w%Pb-55w%Bi cooled direct contact boiling water small fast reactor (PBWFR), experimental study on Pb-Bi-water direct contact boiling two-phase flow has been performed using Pb-Bi-water direct contact boiling two-phase flow loop. For stable start-up of the boiling flow operation, Pb-Bi single-phase natural circulation must be realized in a Pb-Bi flow system of the loop before water injection into Pb-Bi. The Pb-Bi flow system consists of a four-heater-pin bundle, a chimney, an upper plenum, a level meter tank, a cooler, and an electromagnetic flow meter. A stable Pb-Bi single-phase natural circulation was realized in the range of flow rate from 1.5 l/min to 4.8 l/min by heating Pb-Bi in the heater-pin bundle with a power up to 7.7 kW. The inlet and outlet temperatures of the heater bundle were in the ranges from 243°C to 278°C, and from 251°C to 278°C, respectively. The natural circulation flow was simulated analytically using one-dimensional flow model including frictional, form and drag forces. Total hydraulic head through the loop were calculated from Pb-Bi densities at measured Pb-Bi temperatures in the loop. It was found that the calculated flow rate agreed well with the measured ones, which indicated the validity of the analytical models.     

Fracture experiments on through wall cracked elbows under in-plane bending moment: Test results and theoretical/numerical analyses

Fracture assessment of pipe bends or elbows with postulated through wall crack is very essential for leak-before-break qualification of primary heat transport system piping of nuclear power plants. The methodology for fracture assessment of cracked elbows is still in developing stage. Any new development in theoretical aspect requires experimental validation. However, fracture test data on cracked elbows is not so abundant as straight pipes. The earlier experiments on cracked elbows were focused mainly on the determination of limit load. Other fracture parameters e.g. crack growth, crack initiation load or crack opening displacement were not reported in the open literature. Against this backdrop, a comprehensive experimental and theoretical program on component integrity has been initiated at Reactor Safety Division (RSD) of Bhabha Atomic Research Center (BARC), India. Under this program, a number of fracture tests have been carried out on elbows with through wall circumferential/axial cracks subjected to in-plane closing/opening bending moment. These test data are then thoroughly analysed numerically through non-linear finite element analyses, analytically through limit load comparison and also through comparison of crack initiation loads by finite element and R6 methods. These test data may be utilized in future for validation of new theoretical developments in the integrity assessment of through wall cracked elbows.     

振荡流反应器注入分散过程模拟

振荡流反应器(OFR)是一种新型的化工设备。对OFR进行深入的实验研究的同时，也在不断地对它进行数值模拟。采用计算流体力学对振荡流反应器的注入分散过程进行了三维模拟。给出了粒子分散的拉格朗日方程以及相应的边界条件和初始条件，然后根据所建模型给出了三维网格非结构化划分形式；在离散过程中，采用有限体积法和SIMPLE法求解和运用粒子浓度标准方差的处理技术，从而使模拟结果更为可靠。最后以净流量为零的情况下，St=1．0、Re0=200，注入粒子数量为：1000个OFR单腔室粒子注入分散过程为例进行了数值模拟，并对结果进行了简要的分析。     

Application of advanced core process monitoring procedures in German power reactors

The nuclear reactor core design and the nuclear fuel management have been changed remarkable during the last few years. This development was initiated by increasing costs for the fuel recycling and nuclear waste storage. The fuel material, the fuel pellet fabrication, the fuel assembly structure and the core composition have been varied to get an effective fuel exploitation. Based on advanced core process conditions the reactor power and the fuel burn-up have been increased at German plants in recent years. Improved dynamic process monitoring procedures are required to get more information about the varied core process behaviour during the reactor operation. Since several years ISTec has been performed investigations to the process monitoring based on process signal measurements in German nuclear power plants. Using the standard instrumentation of the plants process signals have been measured and analysed by means of the digital data acquisition system SIGMA. The measured time signals are influenced by core process transients, global and local process fluctuations and by signal line transfer functions. Advanced time series analysis methods have been applied to separate different process effects in the multiple signal matrix. The separation of different process influences can improve significantly the information about the process condition in the reactor core.     

中国先进研究堆冷中子导管的模拟研究

应用蒙特卡罗方法对中国先进研究堆(CARR)上的两条冷中子导管CNG1和CNG2的设计方案进行了模拟研究。在假定的冷中子源谱分布下,两条导管出口处的中子注量率均达到1×109cm-2·s-1以上。CNG1和CNG2的中子传输效率分别为50%和42%,中子束最大发散角分别为2 2°和1 9°。中子束沿水平方向的分布均匀,其最大起伏不超过3%。重力显著影响中子束沿垂直方向的分布。     

安全分析程序SAC-PREARS的开发

介绍了SAC-PREARS程序的基本数学模型。利用MISAP程序和通用程序TRAC-PFI对SAC-PREARS计算结果进行了验证。结果表明：SAC-PREARS程序能够有效地计算和分析核供热堆PRHRS的稳态和瞬态热工水力特性。