Response of steel-concrete composite panels to in-plane loading

Steel-Concrete Composite (SCC) panels consist of steel faceplates with welded shear studs and concrete infill. The shear studs, which perform a similar function to bond of rebars in reinforced concrete, act as springs resisting the shear slip between the faceplates and concrete. In spite of extensive research in the 1980s and 1990s due to the interest in using SCC in offshore construction and in nuclear power plants, and recently related Codes, there is no analytical method to-date to predict the shear studs response to in-plane loading. Shear connectors are sized according to various criteria unrelated to their actual forces under in-plane loads, such as prevention of faceplate buckling or out-of-plane shear resistance. This paper presents a closed analytical solution of the equilibrium and compatibility differential equations for steel and concrete displacements of SCC panels, based on distributed shear springs idealization. Analytical results presented in this paper are validated by test results of SCC panels loaded by pure shear forces and can be used as practical design formulas for the in-plane portion of the design loads.     

基于单变量求解方法的华龙一号内置换料水箱过滤器的设计和研究

为了解决华龙一号（HPR1000）事故后安全壳内置换料水箱（IRWST）过滤器设计中的压降求解问题,本文提出了一种单变量求解IRWST过滤器压降的方法,通过在过滤模块和汇流槽之间增加阻力部件,将IRWST过滤器压降求解中的多组变量转化为阻力部件的流通面积这一单组变量,实现了IRWST过滤器的压降求解。结果表明:采用单变量求解方法,可使每个过滤模块的碎渣量和流量相同,通过对IRWST过滤器的压降值计算,可确定IRWST过滤器的初步过滤面积;通过碎渣压降试验对IRWST过滤器的初步过滤面积进行了验证,其结果满足安全系统的设计要求。      

HTR-PM螺栓拉伸机无测量杆螺柱预紧力标定试验研究

为了解决高温气冷堆示范工程(HTR-PM)无测量杆螺柱预紧力的控制问题,保证反应堆一回路压力边界的法兰密封,需要对无测量杆螺柱的预紧力进行标定。以HTR-PM中M56无测量杆螺柱为例,采用液压拉伸机对其进行标定试验,找到螺栓拉伸机拉伸预紧力与螺柱残余预紧力的关系曲线;分析了螺栓拉伸机拉伸前后导致螺柱残余预紧力下降的原因,再通过材料力学本构关系,建立了螺栓拉伸机拉紧力与螺柱回弹后残余预紧力的理论关系式。结果表明,试验获得的螺柱联接体系中的残余预紧力及螺母旋紧前的预紧力关系式都与理论分析比较接近;螺栓拉伸机相同出力下,实际设备管嘴法兰螺柱的残余预紧力会比标定值大,但这更有利于法兰面的密封。     

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.     

Experimental Observation of the Edge and Divertor Plasma in HL-2A

Edge plasma characteristics were studied by a fast-scanning 4-probe array and a Much/Reynolds stress/Langmuir 10-probe movable array in the boundary region. These probes could measure the edge plasma temperature, density, poloidal electric field, radial electric field, Reynolds stress, poloidal rotation velocities and their profiles, which could be obtained by changing the radial positions of the probe array shot by shot. The measured results were used to analyse plasma confinement, turbulent fluctuations and correlations. The fixed flush 3-probe arrays were mounted on the 4-divertor neutralization plates at the same toroidal cross-section in the divertor chamber. These probes were used to measure the profiles of the electron temperature, density and float potential in the divertor chamber. Edge plasma behaviours in both limiter configuration and divertor configuration are compared. The decay lengths of the edge temperature and density were measured and is emphasized for plasma behaviours of the supersonic molecular beam injection and lower hybrid current drive. The dependence of the radial gradient of Reynolds stress on the poloidal flow and the radial gradient of the electric field on turbulent loss are discussed.     

Edge Structure of Reynolds Stress and Poloidal Flow on the HL-1M Tokamak

1. IntroductionThe determination Of electrostatic Reynolds stressand plasma poloidal flow velocity in scrape-off 18yer(SOL) and on the boundary of tokajxnak plasma havebeen of prime importance due to its potential rolein confinement and the L-H mode transition [1-5].As the plasma confinement is sensitive to the edgeconditions, various mechanisms have been theoretically proposed to explain the creation of a shearedpQloidal flow [6-8]. In brief, the theories attemptingto explain the L--H tra…     

Comparison between fluctuation of floating potential gradient and velocity of blob structure on HL-2A tokamak

Several results based on the Langmuir probes' data on the HL-2A tokamak are presented. The blob structures' radial and poloidal drift velocities, estimated by the gradient of floating potential and by time delay evaluation, are compared in different line-averaged density and electron cyclotron resonance heating conditions. A positive correlation is observed in the comparison between blobs' radial velocity estimated by the two methods mentioned above, regardless of the situation differences mentioned above. Correlation is also observed in the comparison between the blobs' poloidal velocity estimated by the two methods in different situations, while a shift due to the different line-averaged density is observed. These results imply that the radial gradient of floating potential may have some value as a reference during data analysis in low-parameter discharge.     

Tokamak edge plasma rotation in the presence of the biased electrode

Electrode biasing system was designed, constructed, and installed on the IR-T1 tokamak, and then biasing experiments were carried out. Also, using a Mach probes the effects of radial electric field (produced by biased electrode) on the poloidal and toroidal components of the edge plasma velocity were investigated. The results showed an increase in both toroidal and poloidal components of the edge plasma velocity during biasing regime. Results compared and discussed. During positive biasing, increased E_r tends to slow the poloidal rotation in the electron diamagnetic drift direction, i.e., to speed up rotation in the ion diamagnetic drift direction. An increased toroidal rotation velocity has the opposite effect on the poloidal rotation.     

Mechanical design of the second ICRF antenna for EAST

In order to satisfy the requirements of heating plasma on EAST project, 3 MW ion cyclotron range of frequency (ICRF) heating system will be available at the second stage. Based on this requirement, the second ICRF antenna, has been designed for EAST. The antenna which is planned to operate with a frequency ranging from 30 MHz to 110 MHz, comprises four poloidal current straps. The antenna has many cooling channels inside the current straps, faraday shield and baffle to remove the dissipated RF loss power and incoming plasma heat loads. The antenna is supported via a cantilever support box to the external support structure. Its assembly is plugged in the port and fixed on the support box. External slideway and bellows allow the antenna to be able to move in the radial direction. The key components of the second ICRF antenna has been designed together with structural and thermal analysis presented.     

The effect of fuel rod loading speed on spacer grid spring force

The burnup-dependent grid-to-rod gap combined with the fluid-induced vibration may generate grid-to-rod fretting wear-induced fuel failure for some fuel assemblies in a certain burnup range. The grid-to-rod gap is dependent on initial spacer grid spring force, spring force relaxation and cladding creepdown. It is found that the initial spring force is reduced during the fuel rod loading into the fuel assembly skeleton. The extent of the initial spring force loss is strongly dependent on the fuel rod loading speed. Based on the initial spring force loss data obtained from two kinds of fuel rod loading speeds of 0.18 and 0.33 m/s, it can be said that the higher rod loading speed generates the larger initial spring force loss. This is because the higher speed generates the larger overshooting of spring deflection during the fuel rod loading. The extent of overshooting may be affected by axial misalignment of SG cells, spring-to-fuel rod end plug contact angle, ballooning of FR end plug weld region and the extent of gravity-induced FR bowing, combining with the fuel rod loading speed. The rod loading speed of 0.33 m/s is found to produce some spacer grid cells less than a minimum initial spring force requirement of 12 N against the grid-to-rod fretting wear-induced failure. In order to produce initial spacer grid spring force meeting the minimum spring force requirement, it is recommended that the lower rod loading speed be used, combined with axially aligned spacer grid cells and lower contact angle of spring-to-fuel rod end plug.     

Evaluation of the probability of a leak in the flange joint of a steam-generator collector cover as a result of stud failure

A two-stage approach to evaluating the probability of leak appearing in the flange joint as a result of a failure of the securing studs is described. First the probability of only one stud failing is calculated. This probability depends on, specifically, the integrity of all other studs. The probabilities of different combinations of the relative arrangements of whole and failed studs are calculated at the second stage. The method is illustrated for the calculation of the probability of a leak in the region of the collector cover for a PGV-1000 steam generator. Normative data on the strength and mechanical characteristics of the structural materials and loads corresponding to the nominal operating regimes are used in the calculations. __________ Translated from Atomnaya énergiya, Vol. 102, No. 6, pp. 344–347, June, 2007.     

The plastic buckling of a steel liner encased in a concrete cylinder

The inelastic buckling and postbuckling performances of a steel liner encased in a rigid concrete containment vessel are studied by taking a strip of unit width from a pattern — either rectangular or diamond — along the circumferential direction in such a way that the strip will have the maximum deflection of a buckled panel. The complete load-deflection curves are obtained and the effect of initial imperfections is also included in the curves. From the discussion of failure modes, a design criteria can be obtained for the liner to maintain its integrity under accident conditions. A simplified spring model is used to calculate the maximum shear displacement and the corresponding shear force of studs at buckling. A design analysis procedure is developed from the limit and the ultimate design conditions and can be used to determine the liner thickness, studsizes and stud spacings in both the axial and circumferential directions.     

反应堆压力容器主螺栓预紧工艺优化仿真分析

应用商业软件ANSYS模拟了反应堆压力容器主螺栓十字拉伸3级预紧过程,对其预紧工艺进行了仿真优化分析。分析结果表明:3级加载方式中第1级采用对称加载、第2级采用间隔加载、第3级采用顺序加载得到的主螺栓不均匀度和离散度最小;第3级预紧采用变载荷能明显减小主螺栓的不均匀度和离散度,提高反应堆压力容器的密封性能。      

Thermal analysis of the Mirnov coils of Wendelstein 7-X

Mirnov coils are used to measure fluctuations of the magnetic field which are in particular generated by magnetohydrodynamic (MHD) modes. The underlying plasma currents have a multipolar structure in a poloidal cross-section. Therefore the amplitude of the magnetic fluctuations decays quickly with increasing distance from the plasma edge. It is hence important to place the Mirnov coils as close to the plasma edge as possible where they are exposed to high thermal loads. Two types of Mirnov coils are proposed to be used in Wendelstein 7-X (W7-X). Type 1 (44 Mirnov coils) should be mounted on the plasma side of wall protection panels with a graphite cap to shield them from direct plasma exposure. Type 2 (137 Mirnov coils) will be located behind the tiles of the heat shields. An important issue concerning the design of these Mirnov coils is to verify their suitability for steady state operation from the thermal point of view. Both steady state and transient finite element thermal analyses were performed for the Mirnov coils under different conditions and with different designs. The paper presents detailed thermal analyses of the Mirnov coils.     

Neutronic analysis of ITER radial x-ray camera

The radial x-ray camera(RXC) is designed to measure the poloidal profile of plasma x-ray emission with high spatial and temporal resolution. The RXC diagnostic system consists of internal camera module and external camera module that view the core region and outer region through the vertical slots of the diagnostic first wall and diagnostics shield module of the equatorial port plug. To ensure the normal performance of the silicon photodiode array detectors of the cameras in the hard neutron irradiation environment in ITER tokamak, it is necessary to calculate neutron flux, radiation damage and the nuclear heating of the silicon photodiode array detectors and simulate the radiation maps of the range of RXC. This work estimated the nuclear environment of RXC based on Monte Carlo N-particle transport code, plasma scenarios of ITER tokamak and the RXC-integrated ITER CLITE model. The neutron flux of silicon photodiode array detectors and the lifetime of the silicon photodiode detector in the camera were calculated. The neutronic analysis results show that the shielding design has achieved the effect as expected and is able to guarantee the normal work of the detector during the ITER deuterium–deuterium phase without replacement, three detectors of the external camera can be operated during the whole deuterium–tritium phase without replacement.     

Status of ITER blanket attachment design and related R&D

Main function of the ITER blanket system [1], [2], [3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R&D is also briefly described.     

Progress in engineering design of Indian LLCB TBM set for testing in ITER

The Indian Lead–Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) is the Indian DEMO relevant blanket module, as a part of the TBM program in ITER. The LLCB TBM will be tested from the first phase of ITER operation in one-half of an ITER port no. 2. LLCB TBM-set consists of LLCB TBM module and shield block, which are attached with the help of attachment systems. This LLCB TBM set is inserted in a water-cooled stainless steel frame called ‘TBM frame’, which also provides the separation between the neighboring TBM-sets (Chinese TBM set) in port no. 2. In LLCB TBM, high-pressure helium gas is used to cool the first wall (FW) structure and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder (CB) pebble bed to cool the TBM internals which are heated due to the volumetric neutron heating during plasma operation. Low-pressure helium is purged inside the CB zones to extract the bred tritium. Thermal-structural analyses have been performed independently on LLCB TBM and shield block for TBM set using ANSYS. This paper will also describe the performance analysis of individual components of LLCB TBM set and their different configurations to optimize their performances.     

Design considerations for access to Tokamak fusion reactors

A concept providing for access to the outside of a Tokamak fusion reactor to perform contact maintenance and for access to the interior of the reactor using locally controlled reactor-mounted remote handling machines is presented. By heavily shielding the reactor blanket and field-shaping coils, the activation of the toroidal field coils and other external components can be limited to acceptable doses. The use of low activity shielding materials, such as lead and boron carbide, result in minimal shield activity. An in-chamber remote handling machine for removing and replacing the linear and blanket modules is provided. Additional remote handling machines mounted on the outside of the reactor are conceived for gaining access to the field-shaping coils and outside of the blanket and to the poloidal bore. A preliminary assessment of the savings in the down time through the use of the proposed concept is made, and the potential impact of these savings identified.     

Conceptual design of the blanket mechanical attachment for the helium-cooled lithium-lead reactor

The conceptual design of a new type of fusion reactor based on the helium-cooled lithium-lead (HCLL) blanket has been performed within the European Power Plant Conceptual Studies. As part of this activity, a new attachment system suitable for the HCLL blanket modules had to be developed. This attachment is composed of two parts. The first one is the connection between module and the first part of a shield, called high temperature shield, which operates at a temperature around 500 °C, close to that of the blanket module. This connection must be made at the lateral walls, in order to avoid openings through the first wall and breeding zone thus avoiding complex design and fabrication issues of the module. The second connection is the one between the high temperature shield and a second shield called low temperature shield, which has a temperature during reactor operation around 150 °C. The design of this connection is complex because it must allow the large differential thermal expansion (up to 30 mm) between the two components. Design proposals for both connections are presented, together with the results of finite element mechanical analyses which demonstrate the feasibility to support the blanket and shield modules during normal and accidental operation conditions.