Venturi tube application in high power test of the SRF module

In the superconducting RF module, the dissipation power of the niobium cavity is an important parameter. In the Superconducting radio frequency (SRF) module’s acceptance test at Shanghai Synchrotron Radiation Facility (SSRF), the Venturi tube is used to measure the quality factor of SRF cavity at 4.2 K. During the test, the venturi tube is be calibrated by increasing heat load with internal heater. In this paper, the horizontal test principle and venturi effect are briefly introduced. The authors find out a...     

HXMT中时间模块的设计与检验

介绍的模块设计方案将时间信息分为两部分,秒以上时间信息采用GPS秒脉冲产生并能被HXMT控制系统同步,秒以下时间信息由系统高精度晶振产生并用GPS秒信号进行校正。针对时间模块正确性检验的难点,提出了时变分析的新方法,并用它完成了时间模块的正确性检验。     

Ripple effect and impact of test blanket module by using RAFM steels on the plasma of ITER

The losses of high-energy particles from the plasma depend on the toroidal field (TF) ripple in Tokomak machine. TBM (test blanket module), using RAFM (reduced activation ferritic/martensitic) steels as structure material, impacts on TF ripple in International Thermonuclear Experimental Reactor (ITER). The aim in this paper was to investigate the impact of TBM on TF ripple in ITER. It was analyzed based on ANSYS code and the Chinese DFLL (Dual Function Lithium Lead)-TBM as instances of analysis. The results indicated the TF ripple was still beyond the acceptable level of ITER (δ_TF < 0.3%) while considering several kinds of configurations (different masses, different dimensions, and different distances to plasma) of the DFLL-TBM. The correction coil might be one way to further reduce the effect on ripple of TF, and the ferromagnetic inserts under TF coil need to continue optimized.     

Thermal power estimation by fouling phenomena compensation using wavelet and principal component analysis

A small percentage of reactor thermal power can be overestimated because of fouling phenomena in a secondary feedwater flowmeter. This study proposes a signal processing technique for the compensation of a degraded flowmeter such a secondary feedwater flowmeter in nuclear power plants. The technique proposed is mainly focused on noise classification and step-by-step noise reduction. The noises focused are classified into the rapid distortion caused by environmental interference, the flow fluctuation according to plant state transition and the degradation by fouling phenomena qualitatively. The multi-step de-noising technique reduces each noise by three techniques step-by-step. The wavelet analysis as a low frequency pass filter to remove the rapid distortion, the linear principal component analysis (PCA) to predict a steady-state value from the fluctuation, and the non-linear PCA implemented as an autoassociative neural network (AANN) to predict an original value from the signal including fouling phenomena are developed. The main purpose of this approach is to make an AANN concentrate on compensating the degradation by fouling phenomena itself. For the demonstration the signals from a simulator and signal modeling were used so that the role and the performance of each noise removal step was represented. In addition a thermal power deviation estimator is proposed to recognize the degradation effect of each operating parameter for reactor thermal power calculation.     

Model of cladding failure estimation for a cycling nuclear unit

Nuclear reactor operating modes under multiple cyclic power changes have been promoted recently, and fuel element cladding behavior under the multiple cyclic power changes has been widely known as a key issue in terms of rod design and reliability. A model of nuclear reactor fuel rod cladding failure estimation under multiple cyclic power changes is proposed. The model is built on the basis of the following admissions of the energy version of creep theory: processes of cladding creep and destruction proceed together and affect each other, intensity of creep process is estimated by specific dispersion power W(τ), while intensity of destruction—by specific dispersion energy A(τ) accumulated during time τ. Having calculated the equivalent stress and the rate of equivalent creep strain, the condition of fuel rod cladding failure used on the basis of the energy version of the theory of creep gives us a criterion to decide if a multiple cyclic power change operating mode is permissible for a given variant of power history and coolant conditions.     

CPLD在HIRFL-CSR电源控制模块中的应用

介绍了一种采用复杂可编程逻辑器件（CPLD）实现HIRFL—CSR电源控制模块中单片机外围电路的方法，将低位地址锁存、地址译码、数据总线、分频电路、比较、计数以及逻辑电路集成于一片CPLD．大大缩小了印制板的面积并提高了系统可靠性．同时，由于CPLD的现场可编程性，使整个系统的灵活性显著增强。     

Status of design of the HTR test module China

Within the frame of German-Chinese R&D cooperation, initiated at the end of 1988, the HTR-Module technology will be implemented in the People's Republic of China as an element of future energy supply. Furthermore, the partners agreed a common further development of this technology. As a first step, the design of a 10-MW(th) experimental reactor was begun, which is characterized by the main features of the commercial-sized HTR-Module. With this so-called Test Module to be errected close to Beijing, it is planned, as one main topic, to demonstrate the fuel element integrity in the temperature range of 1600°C.     

Increasing the power of the high temperature reactor module

To alleviate the economic problems of the modular pebble bed high temperature reactor, its design was modified in such a way that the power output was increased from 200 to 350 MWth. The core geometry was changed from cylindrical to annular, and the pressure vessel diameter was increased to 6.7 m. Control rods are placed in both the outer reflector and the graphite central column. In a safety analysis, loss of heat sink, loss of coolant and water ingress accident were examined. Reactor shutdown and decay heat removal take place passively, and the maximum fuel temperature stays theoreticallybelow 1600 °C, implying full retention of the fission products in the fuel elements. The central column has a diminishing effect on the positive reactivity effect of water ingress. A cost analysis shows that the specific investment costs of a four-module plant would decrease by 26% and the electricity generating costs would reduce by 19%.     

Design plan and requirement of test module and testing items in IFMIF

The main purpose of present phase of IFMIF/EVEDA (International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities) is to produce a detailed report of IFMIF engineering design with the integrated design of all facilities in IFMIF. The main function of the IFMIF is to give the demanded design database for the licensing of DEMO reactors and further reactors, and it is achieved from the materials data set obtained from the high, medium, and low flux test modules (HFTM, MFTM and LFTM) of IFMIF. In the evaluation using small specimens, developments and guidelines of small specimen test technique or technology (SSTT) are also demanded for the achievements. This paper is summarized and also analyzed about the design plan status and requirements in these test modules from users, and testing items and test methodology in IFMIF.     

Modelling the loss and recovery of electric power

General analytical models are developed to quantify the frequencies and durations of transients involving loss of off-site power or total loss of alternating current. Such transients are often important initiating events in probabilistic safety studies of nuclear power plants. Recursive equations are developed for the frequencies of production loss events when there is one or more standby systems available and a grace period to start a reserve unit. The methodology is illustrated with numerical applications.     

New thermal optimization scheme of power module in solid-state amplifier

The new 1 kW power module for ADS project needs the optimization of cooling design including water flow and tunnel layout, and the water flow of three tons per hour was chosen to be a goal for a 20 kW power source.According to analysis from the insertion and integrated loss, about 24 modules were integrated into the rated power. Thus, every module has a cooling flow of 2.1 L/min for RF heat load and power supply loss, which is very hard to achieve if no special consideration and techniques. A new thermal simulation method was introduced for thermal analysis of cooling plate through CST multi-physics suite,especially for temperature of power LDMOS transistor.Some specific measures carried out for the higher heat transfer were also presented in this paper.     

Development of water-cooled solid breeder test blanket module in JAEA

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as test blanket module (TBM). JAEA has been extensively developing a water-cooled solid breeder test blanket module (WCSB TBM) for ITER. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by hot isostatic press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 °C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m², 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.     

ITER中国液态锂铅实验包层模块结构热应力数值模拟

使用有限元程序对中国向国际热核实验堆ITER实验包层工作组提交的双功能锂铅实验包层模块(DFLL-TBM)的两种结构设计方案即双冷LiPb包层DLL和单冷准静态LiPb包层SLL进行热应力数值模拟,在包层结构设计、热工水力学设计和中子学计算基础上,给出包层结构温度场和应力场分布,依据ITER高温结构设计标准,进一步对包层高温部件进行力学性能分析.根据这些模拟结果,分析两种结构基本设计方案的合理性和可行性,并作为进一步优化分析的基础.     

功率MOS、IGBT单粒子烧毁、栅穿效应模拟实验研究

建立了利用^252Gf裂片源,模拟空间重离子引起的单粒子烧毁、栅穿效应的实验方法和测试装置,开展了功率MOS器件、IGBT的单粒子烧毁、栅穿效应的模拟试验研究,给出了被试器件单粒子烧毁、栅穿效应的损伤阈值,以及随器件偏置的变化规律。     

New proposal of reactivity coefficient estimation method using a gray-box model in nuclear power plants

A new method for estimating reactivity parameters, such as moderator temperature coefficient (MTC) and void reactivity coefficient (VRC), is proposed using steady-state noise data. In order to solve the ill-posed problem of reactivity parameter estimation, a concept of a gray box model is newly introduced. The gray box model includes a first principle based model and a black-box fitting model. The former model acts as a priori knowledge based constraints in a parameter estimation problem. After establishing the gray box and noise source models, the maximum likelihood estimation method based on Kalman filter is applied. Furthermore, it is shown that the frequency domain approach of the gray box model is useful in the case of VRC estimation. The effectiveness of the proposed algorithms is shown through numerical simulation and actual plant data analysis.     

HIRFL-CSR主环磁铁电源数据采集控制模块设计

介绍了一种基于微处理器TMS320VC5402的磁铁电源电流数据采集控制模块,给出了模块的硬件组成和软件设计.经过测试表明该模块具有非常高的控制精度,处理能力强,速度快,而且集成度高,便于组成网络控制系统,现已应用于HIRFL-CSR主环六极磁铁电源控制系统中.     

Seismic response prediction for cabinets of nuclear power plants by using impact hammer test

An effective method to predict the seismic response of electrical cabinets of nuclear power plants is developed. This method consists of three steps: (1) identification of the earthquake-equivalent force based on the idealized lumped-mass system of the cabinet, (2) identification of the state-space equation (SSE) model of the system using input-output measurements from impact hammer tests, and (3) seismic response prediction by calculating the output of the identified SSE model under the identified earthquake-equivalent force. A three-dimensional plate model of cabinet structures is presented for the numerical verification of the proposed method. Experimental validation of the proposed method is carried out on a three-story frame which represents the structure of a cabinet. The SSE model of the frame is accurately identified by impact hammer tests with high fitness values over 85% of the actual frame characteristics. Shaking table tests are performed using El Centro, Kobe, and Northridge earthquakes as input motions and the acceleration responses are measured. The responses of the model under the three earthquakes are predicted and then compared with the measured responses. The predicted and measured responses agree well with each other with fitness values of 65-75%. The proposed method is more advantageous over other methods that are based on finite element (FE) model updating since it is free from FE modeling errors. It will be especially effective for cabinet structures in nuclear power plants where conducting shaking table tests may not be feasible. Limitations of the proposed method are also discussed.     

ITER中国液态锂铅实验包层模块结构设计与加工

根据ITER实验包层的发展目标,实验要求,限制条件,结合聚变发电反应堆FDS-Ⅱ DLL/SLL包层方案设计了DFLL-TBM原型结构,给出了加工工艺和装配序列方案.该实验模块特点是极向LiPb流道易于布置FCI流道插件,"]"型隔板和"盒形"背板式联箱简化冷却方案和结构.这种简单的结构易于加工制造,易于派生出在ITER不同运行阶段实验的系列模块,符合在ITER进行SLL-TBM和DLL-TBM两种包层模块实验的策略.     

Fire-induced loss of nuclear power plant safety functions

A methodology is developed for evaluating the probability for loss of nuclear power plant safety functions due to fire. A framework for the investigation of fire scenarios involving safety-related equipment is established which models fire development as an event tree consisting of a series of ignition, detection, suppression, and propagation steps. The methodology has been applied to a representative BWR. Variations in the methodology are discussed for application to specific plants. Conservative estimates of core-damage probabilities due to fire were obtained; application of the methodology to a particular BWR including specific knowledge of cable locations, fire-retardants, detectors, etc. would result in considerably lower probabilities.