Computational-experimental modeling of processes in the protective shell with a passive condenser present in the passive heat-removal system

The modeling of the operation of a passive condenser based on a numerical solution of three-dimensional equations of hydrodynamics is examined. Questions associated with correct modeling of turbulent transport under free-convection conditions are examined. A model taking account of the dynamics of a condensate film and the conditions of heat and mass transfer on its surface is proposed for surface condensation. The results are used to develop recommendations for closure relations used in point codes with whose help design validation of a passive system removing heat from beneath the protective shell is performed.     

静态随机存取存储器辐射效应测试系统的研制

由于空间辐射效应会导致SRAM器件单粒子翻转、单粒子锁定等现象的产生。文中介绍了SRAM辐射效应测试装置的硬件、软件构成及有关测试技术。通过对SRAM芯片电流的检测、断电保护,解决了在SRAM实验过程中SRAM芯片的损坏问题,利用该装置在不同的辐射实验源上对SRAM进行辐射效应实验研究,获得了预期的实验数据。为星载计算机系统存储器的运行寿命评估及加固设计提供重要参考依据。     

Demonstrative testing of honeycomb passive autocatalytic recombiner for nuclear power plant

Passive autocatalytic recombiners (PAR) are widely being used as hydrogen control device in the current and advanced light water reactors (ALWRs). The PARs lend themselves to very effective means of circumventing buildup of combustible or detonable hydrogen gas mixtures in the reactor containment. Korea Nuclear Technology Inc. has recently developed a new PAR system with high porous catalyst material in the shape of honeycomb. The honeycomb PAR catalyst has a design characteristic of improved hydrogen removal performance by increasing the surface area and enhancing the flow rate through the catalyst at the same time, without increasing PAR size compared to the conventional PARs. The experimental study was focused on the development of the hydrogen depletion rate correlation of the honeycomb PAR. Two different sizes of PARs, KPAR-40 and KPAR-T2, have been employed in the tailor-made Integral Test Facility and Performance Test Facility. Multiple tests were conducted in various conditions of pressure, temperature, and hydrogen concentration. The hydrogen depletion rate correlation and the PAR performance constant were determined from the experimental results, which can be applied to the honeycomb PAR system. Also determined was the scale effect due to the PAR size, i.e., the number of catalysts in a PAR.     

HTR fuel coating separate effect test PYCASSO

The “analytical” PYCASSO (PYrocarbon irradiation for Creep and Swelling/Shrinkage of Objects) irradiations focus on determining the effects of neutron irradiation in the temperature range of 900-1100 °C, excluding effects due to the presence of fuel, such as pressurization or chemical attack by fission products. These irradiations can therefore be considered separate effect tests, where only the influence of neutron fluence and temperature on coatings and coating combinations is investigated.For this purpose dedicated particles have been manufactured consisting of surrogate kernels (ZrO₂ and Al₂O₃) with different types of PyC/SiC/ZrC coatings and coating combinations. All specimens delivered have been extensively characterized, such that even potentially small changes due to the irradiation in dimensions, microstructure and density can be determined accurately after irradiation.Partners involved in this irradiation are CEA (France), JAEA (Japan) and KAERI (South Korea). The PYCASSO irradiations take place in the High Flux Reactor (HFR) in Petten, and are coordinated by NRG (The Netherlands). The partnership for PYCASSO was initiated by the RAPHAEL (V)HTR European 6th Framework Program and is integrated in the Generation IV International Forum VHTR Fuel and Fuel Cycle project.     

Analysis of large-scale tests for AP-600 passive containment cooling system

All next-generation light water reactors utilize passive systems to remove heat via natural circulation and are significantly different from past and current nuclear plant designs. One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 h without action by the reactor operator. During a design-basis accident (DBA), i.e., either a loss-of-coolant or a main-steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annular space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-1D code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single-phase flow, transport equations for the k − two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-1D results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized.     

Scaling laws for thermal-hydraulic system under single phase and two-phase natural circulation

Scaling criteria for a natural circulation loop under single phase and two-phase flow conditions have been derived. For a single phase case the continuity, integral momentum, and energy equations in one-dimensional area average forms have been used. From this, the geometrical similarity groups, friction number, Richardson number, characteristic time constant ratio, Biot number, and heat source number are obtained. The Biot number involves the heat transfer coefficient which may cause some difficulties in simulating the turbulent flow regime. For a two-phase flow case, the similarity groups obtained from a perturbation analysis based on the one-dimensional drift-flux model have been used. The physical significance of the phase change number, subcooling number, drift-flux number, friction number are discussed and conditions imposed by these groups are evaluated. In the two-phase flow case, the critical heat flux is one of the most important transients which should be simulated in a scale model. The above results are applied to the LOFT facility in case of a natural circulation simulation. Some preliminary conclusions on the feasibility of the facility have been obtained.     

非能动系统可靠性评价方法的研究

本文首先详细解释了非能动系统可靠性概念,分析各种非能动系统可靠性评价方法的特点,对比各种方法之间的区别,并指出这些可靠性评价方法共同存在的不足:没有一种方法可同时兼顾非能动系统设备可靠性与功能可靠性,不能科学地整合两者的可靠性,并且未将非能动系统整体可靠性融合进概率安全评价(PSA)模型;针对各种方法存在的不足,本文在国内外研究基础上提出研究问题与思路,而且展望了非能动系统可靠性评价方法未来的发展方向。     

UPTF test results: First three separate effect tests

The first separate effect tests were run in the Upper Plenum Test Facility — a 1:1 representation of a PWR primary system. These tests were focusing the simultaneous steam up- and water down flow phenomena at the upper tie plate, the fluid-fluid mixing in the cold leg and downcomer and the countercurrent flow conditions of steam and saturated water in a PWR-hot leg.     

60Co工业DR无损检测的计算机控制系统

介绍了计算机控制在^60Co工业DR无损检测系统中的应用，涉及步进电机控制、电气保护、计算机控制程序等方面的内容。该系统性能优良、控制方便、可靠、运行与维护费用低，应用前景良好。     

Research and development on passive cooling system

Experiments are carried out to investigate the effects of the natural convection of superheated gas as well as those of the standpipes on the temperature distributions of the components and the heat removal performance in the water-cooling panel system for the MHTGR for decay heat removal, and to verify reliability of the design and evaluation methods. The numerical results of the code THANPACST2 are compared with the experimental data to verify the numerical methods and axi-symmetric model proposed, which can simulate the three-dimensional configuration of the standpipes on the upper head of the pressure vessel by using porous body cells. The experiments revealed that temperatures increased with elevation on the upper head, because the standpipes restrict radiation heat transfer to the upper cooling panel and reduce the heat transfer area on the upper head, which was superheated by natural convection of helium gas in the pressure vessel. In the presumed accident condition in which thermal radiative heat transfer is responsible for the majority of the total heat transfer, the numerical methods were able to closely duplicate the pattern of the rising temperature profile with elevation around the top of the upper head as observed in the experiments.     

Verification of the two-phase stratified-flow model in by separate effect tests

The code which is being developed by the Gesellschaft für Anlagen- und Rcaktorsicherheit (GRS) mbH is intended to cover, by means of a single code, the entire spectrum of loss-of-coolant and transient accidents in pressurized and boiling water reactors. The actual version Mod 1.1-Cycle A has a five-equation two-phase model based on the conservation laws for liquid mass, liquid energy, vapor energy and overall momentum. The relative velocity between liquid and vapor is determined by a full-range drift-flux model for two-phase flow in horizontal and vertical pipes. The verification of this drift-flux model is carried out by both large-scale experiments and single-effect tests. The single-effect test ECTHOR investigates stratified flow during the clearance of a water-filled loop seal by a forced air flow through the loop. ECTHOR is a French test for the consideration of two-phase flow regimes in pipes for the development of the codes. The experiments are dedicated to investigating typical two-phase flow during small break loss of coolant accidents (LOCA) in pressurized water reactors (PWR).As a measure, the remaining water level in the loop is determined as a function of the air flow rate. For the verification, a comparison between and computations, on the one hand, and experiments on the other hand is carried out. The results compare very well to each other. Test runs on different numerical grids show convergence to an asymptotic limit with increasing grid refinement.     

IPWRs非能动余热排出系统运行特性分析

由于结构紧凑和采用模块化及非能动安全技术,一体化压水堆(IPWRs)特别适合于舰船核动力装置的应用。本文研究对象为基于固有安全一体化动力堆UZrHx和俄罗斯一体化压水堆ABV-6M的运行特点而概念设计的一体化压水堆。堆芯采用弧形板状燃料元件,直流蒸汽发生器形式为套管式,利用3个回路的自然循环排出堆芯余热的非能动余热排出系统以及一套能动的停堆冷却系统。运用RE-LAP5/MOD3.4程序对该反应堆在全船断电事故工况下反应堆停堆,非能动余热排出系统和能动停堆冷却系统分别投入运行进行仿真计算,分析其热工水力动态特性,保证堆芯安全。     

Scaling laws for contained explosions

Assessment of structural integrity under postulated accident conditions in fast reactors has been based, in the past, on results of scale model tests conducted with chemical explosives. Though emphasis has currently shifted to the development and use of elaborate computational models to determine structural response in the postulated accident, idealised scale model experiments still serve the useful purpose of providing by extrapolation estimates of pressure, impulse and deformation without much expense or loss of time. However, the choice of appropriate scaling laws is important in performing the extrapolation. The results of experiments carried out for the case of a charge of high explosives set off in open waterfilled cylindrical vessels are presented and compared with earlier work by others. Measured shock overpressures at the vessel wall are reported as also impulse values derived by determining the area under the pressure-time traces. Deviations from free-field scaling laws have been observed which are significant for wall pressures and less so for impulse received by the vessel wall.     

面向对象测试方法在观测控制系统中的应用

介绍了面向对象测试方法在观测控制系统（OCS）中的应用。与OCS开发过程和模型相对应,介绍了迭代的面向对象的测试过程,在过程中相应于面向对象的基于组件的开发方法,应用面向对象的测试方法,提出适合OCS的方法和内容以及OCS的测试重点,包括各开发模型的测试,与子系统的接口测试,应用组件的测试,体系结构的测试。     

Mechanical testing of a FW panel attachment system for ITER

An objective of experiments and finite element simulations was to check the stiffness, the strength and the fatigue resistance of the attachment of the First Wall panels onto a shield block of blanket modules according to the ITER 2001 design. The panel has a poloidal key at the rear side (in so-called option A with the rear access bolting) and it is attached by means of special studs located on a key-way in the shield block. Special device for a test of stud tensile pre-load relaxation during a thermal cycling was developed. True-to-scale panels, the shield block mock-up and simplified studs were fabricated and the assembly was loaded alternatively by radial moment, poloidal force or poloidal moment simulating the loading during off-normal plasma operations. Thermal cycling led to an acceptable stud pre-load relaxation. Mechanical cycling caused neither the pre-load relaxation nor the loss of the contact in the key-way nor a damage of the attachment system. The combination of poloidal moment and radial force during vertical displacement events (VDEs) seems to be a most dangerous case because it could lead to the loss of the key–key-way contact.     

A reactor system for radiation research and materials testing

Translated from Atomnaya Énergiya, Vol. 68, No. 6, pp. 438–441, June, 1990.     

Cyclic testing of shear keys for the ITER magnet system

Shear keys are to be used to support the out-of-plane loading of the toroidal field (TF) coils during a plasma pulse in ITER. At the inner intercoil structures (IIS) a set of poloidal shear keys is used to take the shear load at each connection between adjacent TF coils. Solid circular keys have been selected as reference. At the outer intercoil structures (OIS) adjustable conical shear keys and friction joint based shear panels are used to take the shear load. Low voltage electrical insulation is required at the flanges of the IIS and OIS, plus for all the bolts, poloidal keys and adjustable keys. This electrical insulation has to withstand large compression associated with some shear or slippage. A ceramic coating was selected for this purpose. The main scope of the experimental campaign was the mechanical testing of the shear keys and the electrical insulation in operational conditions relevant to ITER. Both keys were made of Inconel 718, provided with a ceramic alumina coating and inserted into flanges made of cast AISI 316 LN. The adjustable conical shear key was pre-loaded at room temperature and subject to cyclic shear loads of 2.5 MN for a large number of cycles (about 30,000) at cryogenic temperature (77 K). The conical key and the alumina coating remained undamaged after the test. Another test campaign was then performed with higher shear loads (up to 3 MN) to reach a sufficient safety margin even with the friction effect due to the pre-load. A set of 15,000 cycles were completed followed by some cycles at higher loads to reach the ultimate limit, which is the shear load to be experienced by the key in case of a poloidal field (PF) coil short.     

AP1000非能动堆芯冷却系统及系统设计瞬态研究

AP1000在标准设计中革新性重大改进之一就是采用了独特的非能动堆芯冷却系统(PXS)。目前世界上在役核电厂和在建核电工程中,AP1000非能动堆芯冷却系统是第一个完全采用非能动手段来达到堆芯冷却、冷却剂补充以及限制放射性释放等安全功能的安全相关系统。文章结合AP1000非能动堆芯冷却系统设计与运行,应用包络方法对一些重要的设计瞬态进行研究分析,从而得出系统设计的合理性和系统功能实现的可行性,为自主研发ACP100、ACP600、ACP1000等第三代核电技术提供借鉴和参考。